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Measurements of radioactivity
CHAPTER
Measurements of radioactivity
4
Chapter outline 4.1 Radiation interaction with matter .....................................................................124 4.1.1 Heavy-charged particles .................................................................125 4.1.1.1 Energy loss due to ionization ................................................ 126 4.1.1.2 Ranges of charged particles ................................................. 129 4.1.1.3 Rutherford scattering .......................................................... 132 4.1.2 Electrons ......................................................................................133 4.1.3 γ- and X-rays .................................................................................134 4.1.3.1 Absorption coefficients ........................................................ 137 4.1.4 Neutrons .......................................................................................138 4.1.5 Penetrating powers of ionizing radiations .........................................141 4.2 Radiation detectors .........................................................................................143 4.2.1 Charged particle detection ..............................................................144 4.2.2 γ- and X-ray detection ....................................................................145 4.2.2.1 Gas-filled detectors ............................................................. 147 4.2.2.2 Scintillation detector ........................................................... 152 4.2.2.3 Semiconductor detector ...................................................... 155 4.2.2.4 Thermoluminescent detectors .............................................. 167 4.2.2.5 Nuclear track detectors ....................................................... 168 4.2.2.6 Photographic film as a radiation detector ............................... 168 4.2.2.7 Neutron detection ............................................................... 168 4.2.3 New developments .........................................................................169 4.2.4 Personal dosimetry detectors ..........................................................171 4.3 Radiometric methods .......................................................................................172 4.3.1 Counting .......................................................................................172 4.3.1.1 2π α-counting method ........................................................ 175 4.3.1.2 4π β-counting with a 4π gas-flow counter .............................. 176 4.3.1.3 4π β-counting with a liquid scintillation spectrometer .............. 177 4.3.1.4 4π α-counting with a liquid scintillation spectrometer .............. 178 4.3.1.5 4π β–γ coincidence counting method .................................... 178 4.3.2 γ-Spectrometry ..............................................................................181 4.3.3 β-Particle spectrometry ..................................................................188 4.3.4 α-Particle spectrometry ..................................................................191 4.3.5 Liquid scintillation measurement method ........................................195 4.3.5.1 External standard method .................................................... 199 Radioactivity in the Environment. https://doi.org/10.1016/B978-0-444-64146-5.00004-5 # 2019 Elsevier B.V. All rights reserved.
123
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CHAPTER 4 Measurements of radioactivity
4.3.5.2 Sample channel ratio method .............................................. 200 4.3.5.3 Automatic efficiency tracing method ..................................... 200 4.3.6 Radiochemical analysis ..................................................................203 4.3.6.1 Introduction ....................................................................... 203 4.3.6.2 Analysis of strontium ........................................................... 204 4.3.6.3 Analysis of tritium ............................................................... 206 4.3.6.4 Caesium analysis ................................................................ 207 4.3.6.5 Determination of actinides ................................................... 210 4.3.7 Rapid methods ..............................................................................216 4.3.7.1 Rapid determination of transuranic elements and plutonium .... 217 4.3.7.2 Rapid determination of 90Sr ................................................. 217 4.4 Nonradiometric methods ..................................................................................218 4.4.1 Methods based on X-ray spectrometry ..............................................220 4.4.2 Methods based on ultraviolet-visible spectroscopy ............................222 4.4.2.1 Inductively coupled-plasma-optical emission spectrometry (ICP-OES) .......................................................................... 223 4.4.2.2 Laser-excited resonance-ionization spectroscopy (LERIS) ........ 224 4.4.3 Methods based on mass spectrometry ..............................................226 4.4.3.1 Inductively coupled-plasma mass spectrometry ...................... 228 4.4.3.2 Accelerator mass spectrometry ............................................ 234 4.4.4 Laser-induced photoacoustic spectroscopy .......................................241 4.5 QA/QC procedures ...........................................................................................243 4.5.1 Intercomparison ............................................................................246 4.5.2 Reference materials .......................................................................251 References ............................................................................................................266 Further reading ......................................................................................................274
4.1 Radiation interaction with matter The detection of radiation is based on the interactions of the various types of ionizing radiations with matter. The differences between the interactions and the penetrating abilities of the various radiations are very relevant to radiation detection and measurement—e.g., they partly explain the variety of detector types and designs. Radiations can be grouped into directly ionizing radiations and indirectly ionizing radiations. Directly ionizing radiations include all charged particles such as α-particles and heavier ions and β-particles. All charged particle radiations lose energy interaction with the orbital electrons or nuclei of atoms in the materials they traverse. There are two main processes involving the orbital electrons: 1. Atomic or molecular excitation, with the emission of light resulting from subsequent deexcitation. 2. Ionization, which involves the ejection of an orbital electron, resulting in the creation of an ion pair. Indirectly ionizing radiations include some types of electromagnetic radiations and neutrons. These radiations interact with matter by giving rise to secondary radiation
4.1 Radiation interaction with matter
which is ionizing. Indirectly ionizing radiations lose energy by collisions with electrons, or atomic nuclei, and the charged particles thus set in motion interact in turn with the orbital electrons or nuclei.
4.1.1 Heavy-charged particles Heavy-charged particles lose energy by Coulomb interaction with the electrons and the nuclei of the absorbing materials. The collision of heavy-charged particles with free and bound electrons results in the ionization or excitation of the absorbing atom, whereas the interaction with nuclei leads only to a Rutherford scattering between two types of nuclei. Thus the energy spent by the particle in electric collisions results in the creation of electron–hole pairs, whereas the energy spent in nuclear collisions is lost to the detection process. The interactions of charged particles with matter are generally expressed in terms of the energy loss per unit path length (also known as stopping power) and the total range of the particle. The range is most important in detector selection, although some applications require the stopping power for particle identification. For heavy-charged particles in which the path is a straight line, the theoretical energy loss while undergoing collisions with electrons is given by dE 14πZ1 Z2 e2 N 2mv2 ln ¼ I dz mv2
(4.1)
where m, e is the mass, charge of electron; Z1 is the atomic number of moving particle; Z2 is the atomic number of stopping material; v is the ion velocity; N is the number of atoms per unit volume and I is the average ionization potential of stopping atom. The logarithmic term varies slowly with energy so that it is approximately correct to write the energy loss as dE Constant ¼ dx E
(4.2a)
dE Z1 Constant ¼ dx E
(4.2b)
or as
to identify the charged particle atomic number. The range is obtained by integrating the stopping power dE/dx over the energy: range ¼
ð0
dE ð dE=dx Þ Emax
(4.3)
Nuclear collisions can become an important part of the energy loss process, especially in the case of heavy ions and fission fragments. The theory describing this process is too complicated for a brief summary. Finally, it should be mentioned that channelling effects (the steering of charged particles in open regions in the lattice) could reduce the specific ionization loss.
125
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CHAPTER 4 Measurements of radioactivity
4.1.1.1 Energy loss due to ionization A charged particle moving through matter loses energy as a consequence of collisions with atomic electrons. Let us first consider the effect of a fast-charged particle on an unbound electron. As the particle passes by, the electron “sees” the rapidly changing electric field as an impulse
Δp ¼ FΔt
Zi e2 b2
2b v
(4.4)
where Zi is the atomic number of the incident-charged particle, v is the centre-ofmass velocity and b is the distance of closest approach, called the impact parameter. The energy transferred to the electron is then ðΔpÞ2 2Zi2 e4 1 EðbÞ ¼ ¼ 2me me v2 b2
(4.5)
In order to find the total loss of the kinetic energy by the charged particle to the atomic electrons in the medium, the number of electrons, dn, affected by the charged particle must be known. It follows
dE ¼ EðbÞdn ¼
2Zi2 e2 m e v2
1 ðNZi 2πdbdxÞ b2
(4.6)
where N is the number density of the target atoms and Zt is the atomic number of the target material. The above expression includes the estimate of the number of electrons within a cylinder having its axis along the path of the charged particle. The energy loss per distance traversed is then dE 4πZi2 Zt e4 N ¼ dx m e v2
ð bmax bmin
db 4πZi2 Zt e4 N bmax log ¼ bmin b me v2
(4.7)
We shall now discuss the values of bmax and bmin. For bmax we shall consider electrons bound in atomic orbits. In order to perturb their orbit the duration of perturbation, b/v, must be smaller than the periods, τ, of the system under consideration: b/v < τ ¼ 1/ν. This determines bmax as bmax ¼
v hνi
(4.8)
where hνi is the appropriate average of the electron frequencies. As a result, the electron binding energy becomes important, it impedes energy absorption by the bound electron. Taking the relativistic corrections into account, the duration of the perturbation is shortened by a factor γ ¼ (1 β2)1/2; therefore, bmax ¼ γv=hνi
(4.9)
The lower limit on b, bmin, is determined by quantum mechanics since the electron can be localized with respect to the heavy ion only to the accuracy of its de Broglie wavelength: h ħ bmin > ¼ p γme v
(4.10)
4.1 Radiation interaction with matter
By taking these values for bmin and bmax, one obtains dE 4πZi2 Zt e4 N m e v2 γ 2 log ¼ ħhνi dx m e v2
(4.11)
The quantity 2πħhi is a special average of the excitation and ionization potentials of the atoms in the stopping material. 2πħhνi ¼ I
(4.12)
A more complete theory along these lines was developed by Bethe (1930) using Born’s approximation. He obtained the following expression: dE 4πZi2 Zt e4 ¼ m e v2 dx
"
! # wme ðγvÞ2 2 β Ck ln I
(4.13)
where Ck is a correction term for the “nonparticipating” K-shell electrons that are strongly bound and are partially screened by outer electrons. The mean ionization-excitation potential, I, can be experimentally determined. Its relation to Z is given by the semiempirical formula: I ¼ 9:1Z 1 + 1:9Z2=3 eV
(4.14)
The general form for the energy loss formula allows us to draw some conclusions. Energy loss of a charged particle is proportional to its mass, square of its charge and inversely proportional to its energy, dE Zi2 Zi2 M ∝ ¼ dx v2 2E
(4.15)
Therefore, simultaneous measurement of kinetic energy, E, and energy loss, dE/dx, allows the determination of the mass of the charged particle, since dE E∝ Zi2 M dx
(4.16)
Fig. 4.1 shows the rate of energy loss for protons in some materials (C, Al, Fe, Pb) commonly used as targets or stopping foils; the unit of dE/dx is keV/(mg/cm2). Fig. 4.2 shows energy loss curves for hydrogen and helium isotopes in silicon. These additional curves are given for convenience because silicon is a common detector material, but, in general, these curves can be obtained from the proton curves by using the following relations: dE dE ðdeuteronÞ ¼ ðprotonÞ1 dx dx E E 2
(4.17a)
dE dE ðtritonÞ ¼ ðprotonÞ1 dx dx E E 3
(4.17b)
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CHAPTER 4 Measurements of radioactivity
1000 500
Al Fe
200
dE/dx (keV/mg/cm2)
100
C
50 20
Pb
10 5
2 0.1
0.2
0.5
1.0
2
5
10
20
50
Proton energy (MeV) FIG. 4.1 The rate of energy loss for protons in C, Al, Fe and Pb.
103
Energy loss (keV μm–1)
128
10
2
4
He
101
3
He
3
H 2
H H
1
100
1
10 Energy (MeV)
100
FIG. 4.2 Energy vs energy loss curves for hydrogen and helium isotopes in silicon.
100 200
Stopping power (MeV / g cm–2)
4.1 Radiation interaction with matter
1000
2000
500
1000
200
500
100
200
50
100
20
α d
50
10
20
5
10
p
5 2 2
1
1 Si Ge
0.05 0.1 0.2 0.05 0.1 0.2
0.5 0.5
1 1
5
2 2
5
10 20 10 20
50 100 200 50 100 200
500 500
Energy (MeV)
FIG. 4.3 Stopping power vs energy for protons, deuterons and α-particles in Si and Ge.
dE 3 dE He ¼ 4 ðprotonÞ1 dx dx E E 3
(4.17c)
dE dE ðaÞ ¼ 4 ðprotonÞ1 dx dx E E 4
(4.17d)
The factor 4 appearing in Eqs (4.17c) and (4.17d) is valid only when the equilibrium charge of the helium ion is essentially 2. The specific ionization loss measures the amount of energy lost by the particle per unit-length of its track; the range indicates how deeply the particle penetrates the absorbing material. Silicon and germanium are the two most common materials in semiconductor industry and especially in radiation detector manufacturing. Therefore, we present here the stopping power of Ge and Si for p, d and α-particles as a function of energy, as shown in Fig. 4.3.
4.1.1.2 Ranges of charged particles The range of a charged particle of incident energy Ei in a material in which its rate of energy loss is dE/dx is given by RðEi Þ ¼
ð Ei 0
dE dE=dx
(4.18)
If dE/dx is known for 0 E Ei, then the range can easily be calculated. For example, Fig. 4.4 shows the range (μm) of protons in Fe for proton energies up to 3 MeV,
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CHAPTER 4 Measurements of radioactivity
40
30
Range (μm)
130
20
10
0 0
1.0
2.0
3.0
Proton energy (MeV) FIG. 4.4 Range of protons in Fe.
while the range as a function of the energy in silicon and germanium for α-particles, protons and deuterons is shown in Fig. 4.5. Unfortunately, stopping cross sections have not been measured for very low energies nor can they be calculated with reliability. Therefore, computed range–energy relations are subject to considerable uncertainty at low energies. On the other hand, range differences from, say, 1 MeV to Ei can be calculated with confidence. The following curves give such range differences, i.e., Rdif ðEi Þ ¼
ð Ei
dE
1 MeV dE=dx
(4.19)
The total range is given by Rdif(Ei) + R(1 MeV). Table 4.1 lists some estimates of R (1 MeV) for the materials considered here. With better estimates (or actual measurements) this table can be corrected. Fig. 4.6 shows range differences of p, d, t, 3He and 4 He for carbon. All of the range differences in Table 4.1 (except for carbon) are given in mm. The density of carbon, however, depends on its manufacture and it is recommended that the density be measured for the sample used before converting the range differences in mg/cm2 to range differences in mm. The following equations are frequently useful for obtaining ranges for particles in terms of proton ranges: Re ðEÞ ¼ 2Rp
1 E 2
(4.20a)
4.1 Radiation interaction with matter
100 50
p (Ge)
20 10
p (Si)
5
d (Ge)
2
Range (g cm–2)
1
d (Si)
0.5 0.2 0.1
α (Ge)
0.05 0.02
α (Si)
0.01 0.005 0.002 0.001 0.0005
0.0002 0.05 0.1 0.2 0.5 1 2 5 10 20 50 100 200 500 Energy (MeV)
FIG. 4.5 Proton, deuteron and α-particle ranges in Si and Ge.
Table 4.1 Approximate ranges for 1 MeV Particle Material
Unit
p
d
t
He3
He4
C Al Si Fe Ge Pb NaI
mg/cm2 mm mm mm mm mm mm
2.7 0.0146 0.0170 0.0075 0.0130 0.0116 0.0218
1.9 0.011 0.013 0.0061 0.0108 0.0095 0.014
1.7 0.010 0.012 0.0057 0.0100 0.0087 0.011
0.55 0.0035 0.0041 0.0019 0.0033 0.0027 0.0029
0.59 0.0037 0.0043 0.0020 0.0034 0.0028 0.0025
1 E 3
(4.20b)
1 E + 0:25mg=cm2 4
(4.20c)
Rt ðEÞ ¼ 3Rp Rα ðEÞ Rp
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CHAPTER 4 Measurements of radioactivity
500
d
t
200
α
100
Energy (MeV)
132
3
He
50
p 20 10 5
2 1 1
2
5
10
5
2
2
10
2
5
10
3
2
5
10
4
2
Range difference (1 MeV mg/cm2)
FIG. 4.6 Range differences for carbon of 1 MeV particles.
RHe
3
3 4 ðEÞ ¼ Rα E 4 3
(4.20d)
where the subscripts p, d, t, α and He3 refer to protons, deuterons, tritons, α-particles and He3 ions, respectively. Of course, when range differences are considered, the approximate additive factor for helium ion ranges does not contribute. When considering the range of charged particles it is useful to look at the property of the beam of charged particles. Fig. 4.7 shows the range curve for a beam of particles penetrating to a given depth.
4.1.1.3 Rutherford scattering In addition to interacting with atomic electrons, charged particles interact with nuclei when passing through matter. When approaching the nucleus the charged particle feels a potential, U ðr Þ ¼
Zi Zt e2 ‘ð‘ + 1Þħ2 + r 2mr 2
(4.21)
The first term is due to the Coulomb repulsion, while the second term is the result of angular momentum of relative motion. The approaching charged particle will undergo scattering on the above potential. As early as 1906, Rutherford detected anomalously large α-particle scattering by thin sheets of mica, gold and some other materials. He has explained this scattering in one of the most cited papers ever (Rutherford, 1911). Rutherford’s most significant assumption was that the scattering
4.1 Radiation interaction with matter
N
R0
s
R1
x
FIG. 4.7 Range curve showing the number of particles in a beam penetrating to a given depth.
centre of the target atom was the atomic charge concentrated into a nucleus of 1012 cm. Assuming a Coulomb potential (U(r) ¼ ZiZte2/r) between the α-particle and the target nuclei, the differential scattering cross section is classically derived as dσ Zi Zt e2 ¼ 2μv dΩ
1 θ sin 4 2
(4.22)
where Zie is the electronic charge of the impinging particle (in this case an α-particle), Zte is that of the target nucleus, μ is the reduced mass of the two particles, v is the velocity of the centre of mass and θ is the centre-of-mass scattering angle.
4.1.2 Electrons The interaction of electrons with matter is similar to the interaction of heavy particles, with the following differences: 1. Nuclear collisions are not part of the interaction because of the very light electron mass. 2. At energies higher than a few MeV, radioactive processes (bremsstrahlung) must be considered in addition to the inelastic electron collision. 3. Again because of their light mass, electrons are so intensely scattered that their trajectory in the material is a jagged line; therefore, the concept of range as previously used cannot be applied. Rather, the concept of zero-transmission range is introduced. This is done by means of absorption experiments, which permit definition of the absorber thickness resulting in zero-electron transmission at a given energy. Fig. 4.8 shows the zero-transmission range as a function of energy in silicon and germanium.
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CHAPTER 4 Measurements of radioactivity
10
Energy (MeV)
134
1
100 Ge Si 10–1
10
–2 0
10
1
10
2
10 Range (μm)
3
10
4
10
FIG. 4.8 Zero-transmission range vs energy for electrons in Si and Ge.
4.1.3 γ- and X-rays The interaction of ionizing electromagnetic radiation with matter is different from the processes previously mentioned, and the concept of ranges and specific ionization loss cannot be applied. Only the three most important absorption processes are considered: the photoelectric effect, the Compton effect and the pair-production effect. The corpuscular description of electromagnetic radiation is the most appropriate for these effects, as one photon in a well-collimated beam of N0 photons disappears at each interaction. The attenuation of the photon beam can be described by a simple exponential law N ¼ N0 exp ðμxÞ
(4.23)
where N is the remaining photons in the beam after traversing distance x, and the absorption coefficient μ is the sum of three terms due to the three above-mentioned processes. These processes are strongly dependent on the energy of effect, such as Rayleigh scattering. Thomson scattering and others are much less important and can be ignored in detection processes (see Fig. 4.9). In the photoelectric interaction, the photon ejects a bound electron from an atom. All of the photon energy, hv, is given to the atom, which ejects the electron with an energy hv – E1, where E1 is the binding energy of the electron. The excited atom then releases energy E1 by decaying to its ground state. In this process, the atom releases one or more photons (and possibly an electron, called an auger electron). The cross section of the photoelectric effect increases rapidly with the atomic number Z and decreases with increasing energy. The Compton effect is essentially an elastic collision between a photon and an electron; during this interaction, the photon gives a fraction of its energy to the electrons, and its frequency ν is therefore decreased. The cross section for this effect decreases with increasing energy, but the decrease is less rapid than for the photoelectric effect.
4.1 Radiation interaction with matter
1.0 0.8 Compton
Fraction of total attenuation
0.6
Compton
0.4 Photoelectric
0.2
Pair
(A) 1.0 Photoelectric
Pair
0.8 0.6 0.4 Compton 0.2
0.01
0.1
1
(B)
10
100
hϖ (MeV)
FIG. 4.9 Relative contributions of various photon interactions to the total attenuation coefficient for (A) carbon and (B) lead.
The energy of electron and the scattered photon are given by
1 Eγ ¼ E0 1 + E0 =mc2 ð1 cos θÞ E0 ¼ E0 Eγ ¼ E0
ðE0 =mc2 Þð1 cos θÞ 1 + ðE0 =mc2 Þð1 cos θÞ
(4.24) (4.25)
where E0 is the incident photon energy; Eγ is the scattered photon energy; Ee is the electron energy; m is the electron mass; c is the velocity of light and θ is the angle between incident and scattered γ-ray directions. The maximum energy loss by the photon is for a head-on collision (θ ¼ 180 degree) and is equal to: E0 Eγ ¼
E0 1 + mc2 =2E0
(4.26)
The fractional energy loss of low-energy photons is small since the scattering is nearly elastic, but becomes appreciable at higher energies. The probability of scattering to a particular angle is a complicated function of energy and angle, but can be generally described as becoming increasingly peaked at small angles as the photon energy increases. The total cross section depends on the
135
CHAPTER 4 Measurements of radioactivity
number of electrons available, or as the atomic number Z of the material. The energy distribution of Compton electrons for several γ-ray energies is shown in Fig. 4.10. In the pair-production effect, a high-energy photon near a nucleus gives up its energy to produce an electron–positron pair. The photon energy goes into the rest-mass energy and the kinetic energy of the electron–positron pair. The minimum energy necessary for this effect is set by elementary relativistic considerations at the value of 1.022 MeV, an amount equivalent to two electron rest masses. The cross section P for pair production increases with energy. Up to energies of 10 MeV, the P/Z ratio remains constant with energy. At higher energies the cross section starts to decrease. Fig. 4.11 summarizes values of the linear absorption coefficients of the above-mentioned effects as a function of γ-ray energy for silicon and germanium. In the pulse-height distributions of Compton interactions of γ-rays in scintillation detectors there are two prominent features usually present: (1) the Compton edge, which corresponds to the maximum energy that can be transferred to an electron by the γ-ray and (2) the backscatter peak, which corresponds to the absorption of a photon which has been scattered through 180 degree in the material surrounding the detector. The energy of the Compton edge is given by Ec ¼
Eγ
(4.27)
ð1 + m0 c2 Þ=2 Eγ
where Eγ is the energy of the incident γ-ray. The energy of the backscatter peak is given by Eb ¼ Eγ Ec ¼ m0 c2 = 2 + m0 c2 =Eγ :
(4.28)
Eγ = 0.5 MeV
Differenntial cross section
136
Eγ = 1.0 MeV Eγ = 2.0 MeV
0.5
1.0
Electron energy (MeV) FIG. 4.10 Compton scattered electron energy distribution.
1.5
2.0
4.1 Radiation interaction with matter
1
10
Linear absorption coefficients (cm-1)
Ge(PE) 0
10
Si(PE) Ge(C)
–1
10
Ge(PP)
Si(C)
Si(PP) 10–2
10–3 10–2
10–1
100 Energy (mev)
101
102
FIG. 4.11 Linear absorption coefficients vs γ-ray energy for Si and Ge (PE ¼ photoelectric, C ¼ Compton, PP ¼ pair production).
The quantities Ec and Eb are shown in Figs. 4.12 and 4.13 as functions of Eγ for both low- (<0.7 MeV) and high-energy (0.5–4 MeV) regions.
4.1.3.1 Absorption coefficients The probability of interacting with matter in one of these three processes can be expressed as a cross section or as an absorption coefficient. The absorption coefficient contains the cross section and is therefore more practical in calculating absorption fractions (Hubbel, 1982). The attenuation coefficient for a beam of γ-rays is related to the number of γ-rays removed from the beam, either by absorption or scattering. For the Compton effect, the absorption cross section is determined by the energy absorbed by the electron, which is the total collision energy minus the average scattered photon energy. For all three processes, the total attenuation coefficient μ is the sum of the three partial attenuation coefficients: μtotal ¼ μphotoelectric + μCompton + μpair production
(4.29)
The attenuation coefficients themselves are defined in terms of thickness of material or surface weight of material. This is just using a thickness x (cm) or a surface weight ρx (g/cm2) where ρ is the density (in g/cm3). The number of primary photons n removed from a beam of n photons is
137
CHAPTER 4 Measurements of radioactivity
0.5
Energy of compton edge or scattered γ-ray (mev)
138
0.3 0.2 Back-scater peak 0.1 0.07 0.05 0.03
Compton edge
0.02
0.01 0
0.1
0.2 0.3 0.4 0.5 Incident γ-ray energy (mev)
0.6
0.7
FIG. 4.12 Energy of Compton edge and backscattered peak for γ-rays of energy <0.7 MeV.
dn ¼ μdx n
(4.30)
n ¼ n0 eμdx
(4.31)
which integrates to for an initial beam intensity n0. For surface density ρx the equation is n ¼ n0 e
μ ðρxÞ ρ
(4.32)
where μ is known as the linear attenuation coefficient and μ/ρ as the mass attenuation coefficient. Graphs of the values for NaI are shown in Fig. 4.14, showing the total μ/ρ value and Compton, photoelectric and pair-production components. Fig. 4.15 shows total μ/ρ values for lead and copper. The K and L edges in the photoelectric absorption are at the energy where the photon can eject a K or L shell electron, thus providing an additional absorption mechanism.
4.1.4 Neutrons The interaction of neutrons with matter is quite different from that of either charged particles or γ-rays. Depending on their energy, neutrons interact with matter by various processes.
4.1 Radiation interaction with matter
Energy of comption edge or scattered γ-ray (mev)
5 3 compton edge
2
1.0 0.7 0.5 0.3
Back-scatter peak
0.2
0.1 0.5
1.0
3 2 Incident γ-ray energy (mev)
4
FIG. 4.13 Energy of Compton edge and backscattered peak for high-energy γ-rays (0.5–4 MeV).
1. Elastic scattering: The neutron shares its initial kinetic energy with the nucleus, which suffers recoil only and is not left in an excited state. The smaller the mass of the nucleus, the greater the fraction of the kinetic energy taken by it. The average fraction of the neutron energy transferred per collision to a medium of atomic weight A is given by 2A/(1 + A)2. A 2-MeV neutron gets thermalized in about 18 collisions in water and in about 420 collisions in lead. 2. Inelastic scattering: Inelastic scattering is possible only with fast neutrons: the scattered neutron carries less energy than the incident neutron and the nucleus goes into an excited state. The excited nucleus either emits a γ-ray or remains in a metastable state. 3. Capture: The incident neutron is captured by the target nucleus forming a compound nucleus which may be excited and emitted γ-radiation. This reaction is probably the most common, since thermal neutrons can induce this reaction in nearly all nuclides. The excitation energy of the target nucleus may be emitted in a single photon or in several. Every such capture results in energy emission amounting to about 6–10 MeV. Hence, materials in which neutron capture is allowed to take place for purposes of attenuation are so chosen that, as a result of the capture, charged particles or photons are emitted that can be easily absorbed. Cadmium and boron are commonly used for capturing thermal neutrons. 4. Particle emission: In this type of reaction, the interaction of the incident neutron with the target nucleus may lead to the emission of particles such as protons,
139
CHAPTER 4 Measurements of radioactivity
102
Iodine K edge
101
NaI
100
(m/r) (cm2/g)
140
–1
10
–2
10
Compton
Total
Photoelectric Pair production
–3
10
–4
10
0.1
1.0
10.0
Gamma ray energy (mev)
FIG. 4.14 Absorption coefficients for NaI.
neutrons and αs. Since the charged particles will have to overcome the Coulomb barrier before escaping the nucleus, this type of reaction is most probable for light nuclides and fast neutrons. 5. Fission: In this process the compound nucleus splits into two fission fragments with the emission of one or more neutrons. Fission reactions take place with thermal neutrons in 235U, 239Pu and 233U and with fast neutrons in many heavy nuclides. Essentially, the absorption of neutrons occurs in two distinct stages. Fast neutrons are slowed down by elastic and inelastic scattering processes with nuclei, particularly light nuclei like carbon and hydrogen. The slowed-down neutrons are then captured, as the capture cross section for low-energy neutrons is high for most elements. It should be noted that neutron capture and certain nuclear reactions are the only interactions that can make the receiving medium radioactive, α-, β-, γ- and X-rays cannot make a medium radioactive. They can ionize a medium, but that does not make the medium radioactive. An ionized stable atom is not radioactive, because ionization alters only the electron structure of an atom, not the nuclear structure, which determines whether an atom is radioactive or stable.
4.1 Radiation interaction with matter
L edges 2
10
1
(m/r) (cm2/g)
10
K edge
Lead
100
Copper –1
10
10–2
102
102
102
Gamma ray energy (mev) FIG. 4.15 Absorption coefficients for Pb and Cu.
4.1.5 Penetrating powers of ionizing radiations Ionizing radiation gives up some or all its energy for each interaction it undergoes. This means that radiation gradually loses its energy as it passes through a medium, and that at best, only some of the incident amount of radiation can pass entirely through any medium. The radiation that does not emerge from the medium is not trapped inside it—the radiation energy has been transferred to the medium, resulting in ionization and excitation. Some or all of the radiation in a beam entering a medium may cease to exist, some may exit from the medium with diminished energy and some may emerge without having undergone any interactions, depending on the penetrating power of the specific radiation. In general, the penetrating power of ionizing radiation is determined by • • •
the type of radiation, the energy of the radiation, the medium the radiation passes through.
141
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CHAPTER 4 Measurements of radioactivity
The penetrating power of: • • • •
α-particles are the heaviest of the four most important ionizing radiations and they move the slowest; an α has two positive charges; an α therefore has a high chance for interactions, rapidly loses its energy and so only travels for short distances, especially in dense media and it is possible to stop α-particles completely—after having lost all its energy, an α captures two free electrons and becomes an ordinary stable 4He atom and is then no longer an ionizing particle.
The penetrating power of: • • • •
β-particles are much lighter than α and travel much faster; a β has only one negative charge; a β therefore has a much smaller chance for interactions and loses its energy slower than an α, and so travels longer distances in media and it is possible to stop β-particles completely—after having lost all its energy, a β may be captured by a positive ion, becoming an ordinary orbital electron, and so ceases to be an ionizing particle.
The penetrating power of X- and γ-radiation of: • • •
photons travel at the speed of light and are not electrically charged; a photon therefore has a much smaller chance for interactions than either α or β and γ-rays give up their energy only a little at a time and it is almost impossible to stop all photons in a beam completely; the beam is only weakened or attenuated by a medium.
The penetrating power of: • •
• •
Neutrons have a very small chance for interactions, because they are not electrically charged. Neutrons travel long distances through dense media and shorter distances in less dense media. Media containing lots of hydrogen atoms slow down neutrons the quickest. When a neutron collides with an atom that has about the same mass as a neutron (like a hydrogen atom), it loses the most energy (think of a marble colliding with another marble, as opposed to a marble colliding with a soccer ball). Slower neutrons travel shorter distances in materials that have a greater chance of capturing (absorbing) them. Eventually a neutron is either absorbed by a nucleus or it decays into a proton and an electron; these are eventually neutralized much like α and β. It is very difficult to completely stop neutrons.
4.2 Radiation detectors
Table 4.2 Ranges of ionizing radiations with (maximum energy of 1 MeV) Radiation
Electrical charge
Range in air
Range in tissue
α β γ n
+2 –1 0 0
6 mm 3m >500 m >500 m
0.008 mm 4 mm >65 cm >65 cm
Data taken from US Bureau of Radiological Health. 1970. Radiological Health Handbook, Edition 1970. U.S. Dept. of Health, Education, and Welfare, Public Health Service, Food and Drug Administration, Bureau of Radiological Health, Rockville, MD and from Cember, H. 1983. Introduction to Health Physics. Pergamon Press, Oxford.
Alpha particles
Beta particles
Gamma rays
Paper
Aluminum
Lead
FIG. 4.16 Penetrating powers of ionizing radiations: α-particles can be stopped by paper, β-particles can be stopped by aluminium, γ-radiation is weakened by lead, but is never greatly totally blocked; neutrons will pass through lead, but will be stopped by thick wax or concrete.
The penetrating abilities of the various radiations are compared in Table 4.2 and Figs. 4.16 and 4.17. The ranges given in Table 4.2 are approximate and are only valid for the specified energy, since radiation ranges are energy dependent.
4.2 Radiation detectors Ionizing radiation cannot be detected by any of the human senses. Nor can we perceive the cause of radiation, namely radioactivity (“glowing in the dark” is, unfortunately, a myth). Man has to use special instruments to detect radiation. These instruments are based on the interactions of ionizing radiation with matter, in particular • •
ionization in gases and ionization and excitation in certain solids.
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CHAPTER 4 Measurements of radioactivity
Concrete
Alpha particles Beta particles Gamma rays
FIG. 4.17 Penetrating powers of ionizing radiation in tissue: γ and n radiation penetrates easily; βparticles penetrate a few centimetres at max. α-Particles do not penetrate the dead layer of the skin.
A variety of instruments based on these principles exists for the assessment of radiological hazards in the environment. The operating principles and practical applications of these instruments are discussed further. Measurement techniques often also require calculations using the instrument reading, in order to obtain the desired quantity.
4.2.1 Charged particle detection Charged particle detection is a process in which the particles interact directly with the material in the region of ionization, and the path of the particle is clearly defined. The efficiency for detecting a particle once it has entered the detector is nearly unity, so that efficiencies for standard geometries are generally calculated from the solid angle alone. For a small detector at sufficiently large distance from a point source (sourceto-detector distance greater than about three times detector diameter), the solid angle is A/x2 steradians, where A is the detector area and x is the source-to-detector distance. At closer distances the following formula should be used where θ1 is the maximum angle subtended by a circular detector solid angle ¼ 2π
ð θ1
sinθdθ ¼ 2π ð1 cosθ1 Þ
(4.33)
0
For extended sources at close distances experimental calibration should be made. Many types of detectors, such as Geiger-M€uller counters, proportional counters and scintillation detectors, are used for charged particle detection. The selection is made on the basis of resolution and range of particle in the gas or scintillator. In some cases, the particles are not completely stopped within the detector for an energy measurement, but deposit only a portion of their energy. This is related to the relative ionization of the particle and can be used to identify different kinds of particles.
4.2 Radiation detectors
Semiconductor detectors are also used for charged particle detection, but differ from those used for γ-rays in being very thin and operating at room temperature. They have very good resolution and are used for total energy, or partial energy loss, known as stopping power or dE/dx measurements. Silicon-charged particle detectors have a p-i-n structure in which a depletion region is formed by applying reverse bias, with the resultant electric field collecting the electron–hole pairs produced by an incident-charged particle. The resistivity of the silicon must be high enough to allow a large enough depletion region at moderate bias voltages. A traditional example of this type of detector is the gold–silicon detector. In this detector, the n-type silicon has a Schottky barrier contact as the positive contact, and deposited aluminium is used at the back of the detector as the ohmic contact. A modern version of the charged particle detector is called PIPS, an acronym for passivated implanted planar silicon. This detector employs implanted rather than surface-barrier contacts and is therefore more rugged and reliable than the silicon surface-barrier (SSB) detector it replaces. At the junction there is a repulsion of majority carrier (electrons in the n-type and holes in p-type) so that a depleted region exists. An applied reverse bias widens this depleted region which is the sensitive detector volume, and can be extended to the limit of breakdown voltage. Detectors are generally available with depletion depths of 100 and 1000 μm, with the cost approximately proportional to the depletion depth. Detectors are specified in terms of surface area and α- or β-particle resolution as well as depletion depth. The resolution depends largely on detector size, being best for small area detectors. Α resolution of 12–35 keV and β-resolution of 6–30 keV are typical. Areas of 25–3000 mm2 are available as standard, with larger detectors available in various geometries for custom applications. Additionally, PIPS detectors are available with fully depleted silicon wafers so that a dE/dx energy loss measurement can be made, with the particle exiting into another detector to measure the remaining energy.
4.2.2 γ- and X-ray detection The kinds of detectors commonly used for γ- and X-ray detection can be categorized as follows: a. gas-filled detectors, b. scintillation detectors and c. semiconductor detectors. Gas-filled detectors are used for X-rays or low-energy γ-rays. These include ionization chambers, proportional counters and Geiger-M€uller counters. Scintillation detectors are used in conjunction with a photomultiplier tube to convert the scintillation light pulse into an electric pulse. Solid crystal scintillators such as CsI or NaI are commonly used, as well as plastics and various liquids.
145
CHAPTER 4 Measurements of radioactivity
Scintillator Gas proportional Counts
146
Si(Li)
0
100
200 300 Channel number
FIG. 4.18 The pulse resolving ability of three types of X-ray detection: scintillator, gas proportional and Si(Li).
Semiconductor detectors, made from single crystals of very pure germanium or silicon, are the highest performance detector type. The superior resolution of these detectors has revolutionized data-gathering for X-ray and γ-ray measurements. The comparison of the pulse resolving ability of the three types of X-ray detectors: scintillator, gas proportional and Si(Li) is shown in Fig. 4.18. The choice of a particular detector type for an application depends on the γ-energy range of interest and the application’s resolution and efficiency requirements. The detector must have sufficient material to absorb a large fraction of the γ-ray energy. Thus, a gas-filled proportional counter is suitable for 14.4 keV γ-rays or for X-rays, but would not “see” 1 MeV γ-rays because the probability of absorbing the γ-ray energy is too low. Further, the higher γ-ray energies are more effectively absorbed by higher Z materials. Other considerations are count rate capability, resolution and if timing applications are involved, pulse rise time. The efficiency of a detector is a measure of how many pulses occur for a given number of γ-rays. Various kinds of efficiency definitions are in common use for γ-ray detectors: 1. Absolute efficiency: The ratio of the number of counts produced by the detector to the number of γ-rays emitted by the source (in all directions). 2. Intrinsic efficiency: The ratio of the number of pulses produced by the detector to the number of γ-rays striking the detector. 3. Relative efficiency: Efficiency of one detector relative to another; commonly that of a germanium detector relative to a 3 in diameter by 3 in long NaI crystal, each at 25 cm from a point source, and specified at 1.33 MeV only.
4.2 Radiation detectors
10–3 8 6
57
Co
109
Cd 139
Ce
4
203
Efficiency
2
Hg
113
Sn
85
Sr 137
–4
10 8 6
Cs
88
Y
60
Co
4
88
Y
2 –5
10
4
6 8 100
2
4 6
1000
2
Energy (kev)
FIG. 4.19 Efficiency calibration for Ge detector.
4. Full-energy peak (or photopeak) efficiency: The efficiency for producing fullenergy peak pulses only, rather than a pulse of any size for the γ-ray. An example of a full-energy peak efficiency curve for a germanium detector is shown in Fig. 4.19.
4.2.2.1 Gas-filled detectors A gas-filled detector is basically a metal chamber filled with gas and containing a positively biased anode wire. A photon passing through the gas produces free electrons and positive ions by the interactions previously described. The electrons are attracted to the anode wire and collected to produce an electric pulse (see Fig. 4.20). At low anode voltages, the electrons may recombine with the ions. Recombination may also occur for a high density of ions. At a sufficiently high voltage nearly all electrons are collected, and the detector is known as an ionization chamber. At higher voltages the electrons are accelerated towards the anode at energies high enough to Positive electrode
Meter Negative electrode Fill gas
FIG. 4.20 Basic elements of gas-filled detector.
– + Battery
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CHAPTER 4 Measurements of radioactivity
10
10
Continuous discharge
12
GeigerMueller counter
10
8
Ions collected
148
10
106 Ionization chamber 104
10
2
Proportional counter
500 Anode voltage (V)
1000
FIG. 4.21 Gas detector output vs anode voltage.
ionize other atoms, thus creating a larger number of electrons. The detector is known as a proportional counter. At higher voltages the electron multiplication is even greater, and the number of electrons collected is independent of the initial ionization. € This detector is the Geiger-Muller counter, in which the large output pulse is the same for all photons. At still higher voltages continuous discharge occurs. The different voltage regions are indicated schematically in Fig. 4.21. The actual voltages can vary widely from one detector to the next, depending on the detector geometry and the gas type and pressure. The ionization chamber is the simplest form of gas-filled detector. It consists of a chamber provided with two electrodes coupled to an electric potential. Gas ions created by the radiation are attracted to the respective electrodes of opposite charge, causing an electric current (called the ion current) to flow between the electrodes. The ion current is electrically amplified and is measured with a micro-ampmeter calibrated to read in dose rate units. The electric current is related to the dose rate as follows: the electric current is proportional to the ion current, which is a measure of the rate of ionization in the gasfilled chamber, which in turn is a measure of the rate at which the gas in the chamber absorbs energy from the radiation passing through it, i.e., of the dose rate. Ions created inside the chamber walls may enter the chamber causing an incorrect reading. This can be avoided by manufacturing the walls from a material with similar ionization properties to the gas. For an air-filled chamber, such chamber walls are
4.2 Radiation detectors
Ionizing particle
-
+
+
+
Amplifier -
+
Ion pair
+ –
Gas-equivalent chamber wall
mA meter –
+
FIG. 4.22 Working principle of an ionization chamber.
called air-equivalent walls or simply air walls. If the detector is required to respond to β-radiation, the chamber must have thin walls or a thin window to allow the βs to enter (normally structural materials will stop the βs before they can enter the detector chamber). The working principle of an ionization chamber is illustrated in Fig. 4.22.
4.2.2.1.1 Proportional counter If the voltage applied to a gas-filled detector is increased beyond a certain point, it ceases to function like an ionization chamber, and an effect known as gas amplification occurs. Gas amplification occurs when electrons produced by the radiation through ionization of the gas in the chamber are accelerated by the applied voltage to such an extent that they gain enough kinetic energy to cause further ionization, resulting in a cascade of ionization that causes an electric pulse. The original amount of ionization caused by the radiation is thus amplified to create a much larger electric signal. The principle of gas amplification is illustrated in Fig. 4.23. High voltage
- - -
+ - - - - -
-
-
-
-
Initial ion pair +
– Gas amplification.
-
-
-
FIG. 4.23
-
149
150
CHAPTER 4 Measurements of radioactivity
If the applied voltage is not too high, the size of the output pulse is proportional to the amount of energy deposited in the detector by the incoming radiation particle/ photon. (This is why the detector is called a proportional counter.) It is called a counter because the number of output pulses is counted by a counting system. The readout can be either a total number of counts or a count rate (in cpm or cps). The count rate is a measure of the rate at which individual radiation particles/photons cause pulses in the detector. For this reason, proportional counters can be used to determine the amount of radioactivity in a sample. The proportionality of the detector can be used to distinguish between different radiation energies of types of radiation, based on pulse size. A specific type of proportional counter that is used for accurate counting of α- and β-activity on smear samples, is the gas-flow proportional counter. In some types of gas-flow proportional counters, the sample is put inside the detector for greater sensitivity—there is no structural material to absorb the radiation before it can be detected. Fig. 4.24 shows a diagram of the detector of such a gas-flow proportional counter. Other gas-flow proportional counters have very thin Mylar windows sealing off the detector’s gas chamber. During counting, samples are positioned very close to the window.
€ ller counter 4.2.2.1.2 Geiger-Mu If a very high electric potential is applied across the electrodes of a gas-filled detector, the gas amplification achieves a maximum value; this means that all pulses are amplified to the same maximum pulse size. All output pulses then have the same size and are no longer proportional to the energy deposited by radiation; an output pulse is only a sign of some radiation particle/photon having been detected—it cannot be said what type or energy. This type of detector is called a Geiger-M€uller counter (also generally known as a Geiger counter or GM tube). A GM tube is usually of the general construction shown in Fig. 4.20. If the detector is to be sensitive to β-radiation, it is fitted with a thin end-window (see Fig. 4.25) that allows the β-particles to penetrate into the detector’s sensitive volume. GM tubes Positive electrode
Gas inlet
Gas outlet
FIG. 4.24 Diagram of windowless gas-flow proportional counter used for α and β counting. The sample is inserted into the detector on a sliding tray.
4.2 Radiation detectors
Positive electrode
Thin end-window
Gas tight insulator
Negative electrode
FIG. 4.25 A typical thin end-window GM tube for β-detection.
are widely used because they are quite sensitive to radiation, but simple and rugged in construction. Gas-filled detectors do not respond instantaneously; their metres take some time to reach the eventual reading. This is caused by the time it takes to collect all the electric charges and the characteristics of the electric circuit. The time constant is used to quantify the response time. The time constant is the time an instrument takes to indicate 63% of its eventual reading. An instrument with a large time constant responds slowly, and one with a small time constant responds quickly. The significance of the response time is that in practice one must wait for the instrument reading to reach its full response and stabilize before taking a reading; this is especially important for portable instruments. A rule of thumb is that one must wait for at least three times the time constant before taking a precise reading. A time constant for an ionization chamber is in the order of 10 s; for Geiger counters it can vary from a few seconds to more than 20 s. The resolving or dead time of a detector is the minimum period of time that must elapse after the detector has detected a particle/photon, before it is able to detect the next particle/photon. Dead time is caused by the time the detector takes to collect all the charges created by an ionization avalanche, and recover for the start of the next impulse. While collecting the charge from one pulse, the detector is “dead,” that is it cannot register another pulse. A particle/photon that arrives in the detection volume while the detector is still “dead” will not be detected. The higher the activity/dose rate, the more particles/photons will be “missed.” In other words, detector dead time causes the instrument to increasingly under-respond at high dose rates/source activities. Dead time is particularly significant for Geiger counters; they have the largest dead times (up to 200 μs) because they have the most charge to collect (maximum gas multiplication, maximum size ionization avalanche), which takes time. At very high dose rates, a Geiger counter with large dead time may become completely “paralyzed”; its reading drops right down and stays there, even if the dose rate or activity is further increased. Geiger counters can be designed to electronically compensate for counts lost due to dead time. However, many are not. It is important to know whether a particular
151
152
CHAPTER 4 Measurements of radioactivity
(A)
(B)
FIG. 4.26 Two generations of Geiger counters: (a) An old Geiger counter shown together with iPhone having an attached Geiger probe; (b) iPhone with its attachment and old Geiger counter probe.
Geiger counter is dead time compensated, so that if not, its drawback can be kept in mind. There has been a significant advance in miniaturization of Geiger counters. Today, one can acquire a “smart Geiger nuclear radiation detector” for iOS iPhone or Android phones, some of them of questionable quality. One such product is shown in Fig. 4.26 in comparison with classical instrument. There are number of sites on web offering the sale of such devices.
4.2.2.2 Scintillation detector Scintillation means the production of small flashes of light. Some crystals, e.g., sodium iodide (NaI) convert the ionization and excitation produced by radiation into a light pulse or scintillation. The amount of light that is produced is proportional to the energy deposited by the radiation particle/photon. The small light pulse is converted into an electric pulse by an electric component called a photomultiplier tube. The size of the amplified electric pulse is proportional to the energy deposited by the radiation photon/particle (for details, see Birks, 1964; Knoll, 1979). The basic components and operating principle of a scintillation detector are illustrated in Fig. 4.27. The proportionality of the output pulses of a scintillation counter to the deposited radiation energy enables the output to be used
4.2 Radiation detectors
Gamma ray in
NaI crystal
Optical pipe
Housing
Photomultiplier
Magnetic shield
Tube base
Preamplifier
HV in
Signal out
Preamplifier power supply
FIG. 4.27 Basic components and operating principle of a scintillator detector.
• •
to discriminate between different types and energies of radiation, e.g., with portable instruments and as input pulses for a γ-spectrometer.
There are also liquid scintillation counters that use a scintillation liquid instead of a crystal. The sample to be counted is dissolved in the scintillating liquid. This method is suitable for α- and β-counting.
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CHAPTER 4 Measurements of radioactivity
The properties of a scintillation material required for good detectors are transparency, availability in large size and large light output proportional to γ-ray energy. Relatively few materials have good properties for detectors. Thallium-activated NaI and CsI crystals are commonly used, as well as a wide variety of plastics. Both NaI and CsI require an activator such as Thallium for proper operation. NaI is the dominant material for γ-detection because it provides good γ-ray resolution and is economical. However, plastics have much faster pulse light decay and find use in timing applications, even though they often offer little or no energy resolution. The actual process by which light is produced is very complex. The high Z of iodine in NaI gives good efficiency for γ-ray detection. Energy dependence of absorption coefficient for NaI is shown in Fig. 4.14. A small amount of Tl is added in order to activate the crystal, so that the designation is usually NaI(Tl) for the crystal. The best resolution achievable is about 8.0% for the 662 keV γ-ray from 137Cs for 2 in diameter by 2 in long crystal, and is slightly better for larger sizes. Typical spectra are shown in Fig. 4.28 for 137Cs and 22Na γ-sources. NaI is slightly nonlinear (about 5%) at low energies (below 200 keV) because of light output variations with γ-energy. The light decay time constant in NaI is about 0.25 μs. Typical charge-sensitive preamplifiers translate this into an output pulse rise time of about 0.5 μs. Fast coincidence measurements cannot achieve the very short resolving times that are possible with plastic, especially at low γ-ray energies. 1200
800 0.662 MeV
137
Cs
400
Counts
154
0 1200
20
800
40
60
80
100
120
140
0.511 MeV 22
400
Na
1.274 MeV
0 20
40
60
80 100 Channel number
FIG. 4.28 NaI(Tl) spectra for
137
Cs and
22
Na γ sources.
120
140
160
4.2 Radiation detectors
Many configurations of NaI detectors are commercially available, ranging from very thin crystals for X-ray measurements to large crystals with multiple phototubes. Crystals built with a well to allow nearly spherical (4π) geometry counting of weak samples are also a standard configuration. Many types of plastic scintillators are commercially available and find applications in fast timing, charged particle or neutron detection, as well as in cases where the rugged nature of the plastic (compared to NaI), or very large detector sizes, are appropriate. Subnanosecond rise times are achieved with plastic detectors coupled to fast photomultiplier tubes, and these assemblies are ideal for fast timing work. Photomultiplier tubes are extremely sensitive light detectors, which multiply the current produced by incident photons by up to 108 times. Since their invention in the 1930s they have seen huge developments that have increased their performance significantly (see Wright, 2017).
4.2.2.3 Semiconductor detector Semiconductors are materials that do not normally conduct electricity because their crystals do not contain enough free-charged particles to carry the current, but that do become conducting when atoms in the crystal become ionized (Knoll, 1979; Bertolini and Coche, 1968). When a relatively small voltage (25–300 V) is applied across the crystal, and it is exposed to ionizing radiation, the electric field sweeps the free-charged particles formed by the radiation out of the crystal. This creates an electric pulse in the external circuit. The size of the pulse is proportional to the radiation energy deposited in the semiconductor. The number of pulses per size range can be counted, and the count rate can be used to determine the activity of the radiation source. Like scintillation detectors, semiconductor detectors are usually used in γ-spectrometer set-ups to identify radionuclides and determine their activities in a sample. A semiconductor detector is much more expensive and somewhat more troublesome to operate than a scintillation detector, but it can distinguish much better between different radiation energies and is better for nuclide identification. The group IV elements silicon and germanium are by far the most widely used semiconductors, although some compound semiconductor materials are finding use in special applications as development work on them continues. Semiconductor detectors have a P-I-N diode structure in which the intrinsic (I) region is created by depletion of charge carriers when a reverse bias is applied across the diode. When photons interact within the depletion region, charge carriers (holes and electrons) are freed and are swept to their respective collecting electrode by the electric field. The resultant charge is integrated by a charge-sensitive preamplifier and converted to a voltage pulse with amplitude proportional to the original photon energy. Since the depletion depth is inversely proportional to net electrical impurity concentration, and counting efficiency is also dependent on the purity of the material, large volumes of very pure material are needed to ensure high counting efficiency for high-energy photons.
155
156
CHAPTER 4 Measurements of radioactivity
Table 4.3 Some of the key characteristics of various semiconductors used as detector materials Material
Z
Band gap (eV)
Energy/e–h pair (eV)
Si Ge CdTe HgI2 GaAs
14 32 48–52 80–53 31–33
1.12 0.74 1.47 2.13 1.43
3.61 2.98 4.43 6.5 5.2
Prior to the mid-1970s, the required purity levels of Si and Ge could be achieved only by counter-doping P-type crystals with the N-type impurity, lithium, in a process known as lithium-ion drifting. Although this process is still widely used in the production of Si(Li) X-ray detectors, it is no longer required for germanium detectors since sufficiently pure crystals have been available since 1976. The band gap figures in Table 4.3 signify the temperature sensitivity of the materials and the practical ways in which these materials can be used as detectors. Just as Ge transistors have much lower maximum operating temperatures than Si devices, so do Ge detectors. As a practical matter both Ge and Si photon detectors must be cooled in order to reduce the thermal charge carrier generation (noise) to an acceptable level. This requirement is quite aside from the lithium precipitation problem which made the old Ge(Li), and to some degree Si(Li) detectors, perishable at room temperature. The most common medium for detector cooling is liquid nitrogen; however, recent advances in electrical cooling systems have made electrically refrigerated cryostats a viable alternative for many detector applications. In liquid nitrogen (LN2)-cooled detectors, the detector element (and in some cases preamplifier components) is housed in a clean vacuum chamber which is attached to or inserted in a LN2 Dewar (see Fig. 4.29 as an illustration of the often used configuration). The detector is in thermal contact with the liquid nitrogen which cools it to around 77°C. At these temperatures, reverse leakage currents are in the range of 109 to 1012 A. In electrically refrigerated detectors, both closed-cycle Freon and helium refrigeration systems have been developed to eliminate the need for liquid nitrogen. Besides the obvious advantage of being able to operate where liquid nitrogen is unavailable or supply is uncertain, refrigerated detectors are ideal for applications requiring long-term unattended operation, or applications such as undersea operation, where it is impractical to vent LN2 gas from a conventional cryostat to its surroundings. There are various types and configurations of semiconductor detectors. One type, the surface-barrier detector, is especially useful for α-spectrometry. This type of
4.2 Radiation detectors
Detector holder
Electrical feedthroughs
End cap
Preamp housing
Tailstock
Fill/vent tubes
LN2 transfer collar
Necktube
Molecular sieves
Dewar
Insulation
Coldfinger
FIG. 4.29 Vertical dipstick cryostat.
detector uses a semiconductor wafer, and the sensitive volume of the detector is the surface layer of the wafer. It detects a radiation very effectively, because the α-particles are not absorbed before they reach the sensitive volume (the usual problem with other detectors). The energy lost by ionizing radiation in semiconductor detectors ultimately results in the creation of electron–hole pairs. Details of the processes through which incoming radiation creates electron–hole pairs are not well known, but the average energy ε necessary to create an electron–hole pair in a given semiconductor at a given temperature is independent of the type and the energy of the ionizing radiation. The values of ε are 3.62 eV in silicon at room temperature, 3.72 eV in silicon at 80 K and 2.95 eV in germanium at 80 K. Since the forbidden bandgap value is 1.115 eV for silicon at room temperature and is 0.73 eV for germanium at 80 K, it is clear that not all the energy of the ionizing radiation is spent in breaking covalent bonds. Some of it is ultimately released to the lattice in the form of phonons. The importance of these interactions for detectors is in how they relate to incident γ-ray energy deposited in the crystal. An ideal detector converts all of the energy of
157
CHAPTER 4 Measurements of radioactivity
Double escape peak Counts
158
0.5
Full energy peak Single escape peak
1.0 1.5 Detected energy (mev)
2.0
FIG. 4.30 Idealized spectrum.
the γ-ray into an electric pulse that is directly proportional to the γ-ray energy, that is, linear. For γ-rays the Compton scattering often results in only a fraction of the energy being deposited because the γ-ray can scatter and then escape from the crystal without further interaction. The full-energy peak can be produced by a photoelectric absorption, or one or more Compton scatterings followed by photoelectric absorption. If pair production occurs, the positron slows down in the material and then annihilates, producing two 511 keV γ-rays. Each of these may escape from the detector totally, or leave part of their energy by Compton scattering. If one or both totally escapes, the deposited energy is the full energy 511, or 1022 keV, leading to designation of these peaks as “single escape” and “double escape” peaks. An idealized spectrum from a source of γ-rays at 2 MeV is shown in Fig. 4.30. The constant value of ε for different types of radiation and for different energies contributes to the versatility and flexibility of semiconductor detectors for use in nuclear spectroscopy. The low value of ε compared with the average energy necessary to create an electron–ion pair in a gas (typically 15–30 eV) results in the superior spectroscopic performance of semiconductor detectors. If all of the energy lost by ionizing radiation in a semiconductor was spent breaking covalent bonds in the detector’s sensitive volume, no fluctuations would occur in the number of electron–hole pairs produced by ionizing radiation of a given energy. At the other extreme, if that energy entering the semiconductor detector that is partitioned between breaking covalent bonds and lattice vibrations or phonon production were completely uncorrelated, Poisson statistics would apply. The variance in the number of electron–hole pairs n would then be hni2 ¼ n. In fact, neither of these suppositions simulates reality. As the incoming ionizing radiation gives up energy, a large shower of hot electrons is created. After many generations, the energy of these hot electrons gets close to the ionization energy necessary to create an electron–hole pair in the semiconductor detector, so that there are several
4.2 Radiation detectors
possible competing mechanisms for energy loss. Thus the Fano factor F is introduced to modify the more familiar Poisson relation for this case. The equation for the variance can be written as hn0 i2 ¼ Fhni2 ¼ Fn:
In the case where there are no fluctuations in the number of electron–hole pairs, F would be zero; in the case where Poisson statistics apply, F would be equal to 1. Since the energy necessary to create electron–hole pairs in semiconductor detectors is much smaller than that of the incoming ionizing radiation, it can be concluded that F is closer to zero than to 1. The true value of F for silicon and germanium is still unknown; the conflicting theories on the subject do not lead to experimentally distinguishable results. However, by assuming a value of 0.1 for F in both silicon and germanium, satisfactory agreement with measured results is found in most cases. By assuming a value of 0.1 for the Fano factor, the following formula gives the germanium detector resolution at LN2 temperature: pffiffiffi ΔE ¼ 1:27 E
(4.34)
with E measured in eV. ΔE must be summed in quadrature with the FWHM keV noise ΔN in order to obtain the measured energy resolution ΔEs: ΔEs ¼
qffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi ðΔEÞ2 + ðΔN Þ2
(4.35)
The value of K for silicon at room temperature is of little interest because, in such conditions, other factors such as fundamental statistics dominate energy resolution values. These simple formulas show that, as expected from the better statistics due to the lower value of ε, when the energy resolution is dominated by the detector contribution, germanium detectors have an advantage over silicon detectors. The equivalent circuit of a semiconductor detector operated as a spectrometer is shown in Fig. 4.31. In most cases, effects of high resistance of the reverse-biased junction are negligible. If a zero-electric-field radiation-insensitive region is present
Z
I(t)
CD
RD
FIG. 4.31 Equivalent circuit of semiconductor detector: I(t) ¼ current generator; CD ¼ capacitance of the depletion region; Z ¼ series impedance; RD is the resistance of the depletion region.
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in the detector, its impedance (a parallel RC combination) appears in series with the circuit and is indicated in Fig. 4.31 by the impedance Z. The impedance also accounts for any resistance (or resistance-capacitance combination) appearing in series with the contacts. When semiconductor detectors are used as spectrometers, they are invariably connected to a charge-sensitive (integrating) preamplifier with a high-dynamic input capacitance. The charge-sensitive preamplifier integrates on its feedback capacitance the current signal delivered by the detector and feeds the resulting voltage signal to the filter amplifier (main amplifier). The time behaviour of the current signal at the input of the charge-sensitive preamplifier is determined by the current signal’s shape and by the effect of the equivalent circuit shown in Fig. 4.31. The effect of the equivalent circuit is usually either negligible or easily calculated, whereas detailed considerations on the charge collection process in the detector are needed to calculate the induced current signal I(t). The current delivered by the signal generator I(t) is induced on the contacts of the detector by the motion of the charge carriers created by the ionizing radiation. Therefore, the first problem in determining I(t) is calculating the motion of the charge carriers in the detector’s electric field. When this problem is solved, the induced charge can be calculated by electrostatic considerations. The charge carriers created by the ionizing radiation drift to the contacts of opposite polarity, following the lines of force of the electric field established by the applied voltage. In the case of heavily ionizing particles such as fission fragments, the drift process does not begin immediately due to the creation of the charge cloud. The electric field E(r) in the detector can be calculated from known quantities: applied bias voltage, detector geometry and resistivity of the bulk material. Once the electric field is known, the motion of a charge carrier created at a given point r0 or the detector volume can be calculated by using the values for the drift velocity Vd as a function of the electric field E. Thus the differential equation dr ¼ Vd ½Eðr Þ dt
(4.36)
can be written for every charge carrier and can be solved if the initial positions r0 are known only when the charge carriers are created along a well-defined track (heavycharged particles). In the case of β, x and γ-radiation, the only information on r0 values is of a statistical nature. The integration of Eq. (4.36) leads to r(t) for every charge carrier created. The charge induced by every carrier can then be calculated by electrostatic considerations. For instance, in the case of a detector with parallel contacts and a field E(x) across a distance W, the charge induced by a carrier q moving along a length Δx in the direction of the field is given by Δq ¼ q
Δx W
(4.37)
independently of the shape of E(x). Eqs (4.36) and (4.37) (or the appropriate induction equation) yield the contribution to I(t) of every single charge carrier and, by integration over all the created charge carriers, the total I(t) function.
4.2 Radiation detectors
The rise time Tt of the pulse generated by a semiconductor detector can be measured at the output of a charge-sensitive preamplifier. If the preamplifier is sufficiency fast, Tt is determined by the following factors: 1. the charge collection time TR, 2. the rise time of the detector equivalent circuit, in most cases a negligible quantity and 3. the plasma time. In most cases TR is the dominant factor. Although a precise calculation of TR can be quite complex, the order of magnitude of TR can be easily obtained by the following formulas: TR ffi W 107 s
(4.38a)
for silicon detectors at room temperature and TR ffi W 108 s
(4.38b)
for germanium detectors at LN2 temperature. In these formulas, W is the thickness of the depletion region measured in mm. For silicon detectors and for planar HPGe detectors, the value of W is provided with each detector. For coaxial Ge detectors, W is the radius of the cylinder. The formulas given earlier are indicative only of orders of magnitude and do not give exact values. The previous discussion did not consider trapping effects, which result in a loss of charge to the collection process and consequent distortion of the shape of the peak as observed with a multichannel analyser. Trapping of a charge carrier in a semiconductor occurs when the carrier is captured by an impurity or imperfection centre and is temporarily lost to any charge transport process. In semiconductor detectors, it is useful to introduce the quantity τ+ (mean free drift time): τ + ¼ ðNt , σVth Þ1
(4.39)
where Nt is the density of trapping centres, σ is the trapping cross section and Vth is the thermal velocity. Note that τ+ does not ordinarily coincide with the classical lifetime in photoconductivity theories. This is because in photoconductivity the traps are generally filled, while in a depleted detector, the traps are generally empty. The trapped charge carrier can be reemitted in the relevant band and take part again in the charge transport process. The average time spent by a carrier in a trap is called the mean detrapping time τD and is strongly temperature dependent Eτ τD ¼ Cexp kT
(4.40)
where C is the constant, Eτ is the activation energy of the trap, k is the Boltzmann’s constant and T is the absolute temperature. If the mean detrapping time is of the same order of magnitude as, or larger than, the electronic clipping-time constants, the charge carrier is lost to the charge
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collection process or is collected with significantly reduced efficiency. The result is poor energy resolution and peak tailing. On the other hand, if the mean detrapping time is orders of magnitude shorter than the charge collection time due to drift of the carriers, then the trap has no effect on the charge collection process. For this reason, normally used dopants such as Li, P, B and Ga, which are shallow donors or acceptors, do not act as traps. It can be shown that, to the first-order approximation, the efficiency of collection of a charge carrier subjected to trapping with a mean free drift time τ+ is given by
TR η¼1 2τ +
(4.41)
where η is the collected fraction of the created charge. In a modern germanium γ-ray spectrometer the charge collection efficiency is of the order of 0.999 and as TR is of the order of 107 s, then τ+ according to Eq. (4.41), is of the order of 104 s. As typical values of Vth and σ are 107 cm s1 and 1013 cm2 respectively, the maximum concentration of trapping centres permissible in the detector is of the order of 1010 cm3 corresponding to approximately 1 for every 1012 atoms of germanium.
4.2.2.3.1 Germanium detector Germanium detectors are semiconductor diodes having a P-I-N structure in which the intrinsic (I) region is sensitive to ionizing radiation, particularly X-rays and γ-rays. Under reverse bias, an electric field extends across the intrinsic or depleted region. When photons interact with the material within the depleted volume of a detector, charge carriers (holes and electrons) are produced and are swept by the electric field to the P and N electrodes. This charge, which is in proportion to the energy deposited in the detector by the incoming photon, is converted into a voltage pulse by an integral charge-sensitive preamplifier. Because germanium has relatively low band gap, these detectors must be cooled in order to reduce the thermal generation of charge carriers (thus reverse leakage current) to an acceptable level. Otherwise, leakage current-induced noise destroys the energy resolution of the detector. Liquid nitrogen, which has a temperature of 77°K is the common cooling medium for such detectors. The detector is mounted in a vacuum chamber which is attached to or inserted into an LN2 Dewar. The sensitive detector surfaces are thus protected from moisture and condensable contaminants. A modern Ge detector is a suitably shaped cylinder of highly purified germanium; it is rather hard to imagine that on 1010 atoms of germanium we have only one atom of an impurity. The modern metallurgical methods, zone refining, allow us to obtain such extreme purity of material. There are very few laboratories where germanium can be refined to this purity, and a single crystal can be grown; there are only three that produce such crystal for commercial uses. And even these three have difficulties in producing really big crystals. The technology of making a γ-ray detector from a piece of pure germanium is in principle simple but in practice very demanding. The Ge crystal must be
4.2 Radiation detectors
• • •
•
properly shaped, equipped with suitable electrical contacts, mounted in a cryostat (two important criteria: it must have excellent contact with the cold finger, and it should be mounted in such a way as to have minimum microphonics) and connected to the first stage of the preamplifier (which is usually placed inside the cryostat).
After a good vacuum has been created in the cryostat, and the detector is cooled by placing the assembly in liquid nitrogen, the characteristics of the detectors are measured. If the detector shows the resolution of 1.68 keV, it will obtain a high price tag. With the resolution of 2.2 keV, it will cost only half as much. And if it has resolution of 3 keV, it will be thrown away. The production of the detector is only to a smaller extent science. Mainly, it is good workmanship, applying modern technologies—and hoping for good results. There are different types of germanium detectors. Their geometry and construction features are illustrated in Fig. 4.32. The “classical” coaxial Ge detector is made of p-type germanium, and is used for spectroscopy of γ-rays. It covers the energy range from 100 keV to several MeV. On the low-energy side, its efficiency is limited by the fact that low-energy γ-rays cannot penetrate the wall of the cryostat. High-energy γ-rays might not be detected because they just pass the volume of the detector without creating a signal that would reflect all the energy of the ray.
(C) (A)
(B)
(D)
FIG. 4.32 (A) p-Type germanium detector; (B) n-type germanium detector; (C) planar germanium detector; and (D) low-energy germanium detector.
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If the detector crystal is made of n-type germanium, the outer electrode can be made rather thin. Include a thin aluminium or a beryllium window at the face of the detector, so that low-energy electromagnetic radiation will be able to enter the volume of the detector, and you get a γ-ray detector that is good also for soft γ-rays, and X-rays. Instead of a cylindrically shaped detector, one can take only a slice of the monocrystal. This will make a planar Ge detector. With a beryllium entrance window, it can be used for X-rays. The last addition to the world of Ge detectors is the lowenergy germanium detector. If made by Canberra, it will be called LEGE. It is excellent for low-energy γ-rays; its energy range extends from 10 to 300 keV. This is indicated in Fig. 4.33. The detector geometries also result in different energy resolutions especially for lower γ-ray energies. This is shown in Fig. 4.34. Typical absolute efficiency curves for various Ge detectors in the Marinelli-beaker configuration are shown in Fig. 4.35, while Fig. 4.36 shows typical absolute efficiency curves for various Ge detectors with 2.5 cm source to the end-cap spacing. Detector type Ultra LEGe LEGe Coaxial Ge REGe and XtRa well 0
10 100 Energy (keV)
1
1000
FIG. 4.33 γ-Ray detection energy interval for different types of detection.
3 keV
Resolution
164
Well 1 keV Coax.
ReGe XtRa Large LeGe Small LeGe 100 eV
5.9 10
100
Energy
FIG. 4.34 Typical resolution vs energy.
122
1332
10,000
4.2 Radiation detectors
Efficiency
0.1
0.01
0.001 10
100
1000
10,000
Energy (keV) FIG. 4.35 Detection efficiency of the HPGe detector in the Marinelli-beaker configuration.
10.0 1
2
4
Efficiency (%)
3 1.0
0.1
0.01
5
10
20
50 100 200 Energy (keV)
500
1000
2000
FIG. 4.36 Typical absolute efficiency curves for various Ge detectors with 2.5 cm source to end-cap spacing: (1) REGe, 15% relative efficiency; (2) LEGe, 100 cm2 15 mm thick; (3) LEGe, 200 mm2 10 mm thick; and (4) coaxial Ge, 10% relative efficiency.
4.2.2.3.2 Silicon detector Occasionally, high-resolution X-ray detectors made of very pure silicon can be found on the market. It seems, however, that the purification process of silicon does not yield the same excellent results as in the case of germanium. It is more of a lucky
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chance that sometimes the metallurgists succeed in producing a batch of silicon with an extremely low amount of impurities. Then, intrinsic silicon detectors are produced. Most of the time, the silicon crystal with little leakage current is obtained by the lithium drifting process. It should be noted that this process in the case of silicon results in a very stable product. At room temperature (without high voltage connected), a Si(Li) detector can be stored for years, without losing its properties. When it is cooled again, it will display the same characteristics as when it was new. Therefore, the search for ultrapure silicon material does not have the same significance as in the case of germanium. The most frequently used silicon detector is a small pellet of lithium-drifted silicon, some 6 mm in diameter and 3–4 mm thick. It is good only for X-rays, in the energy range between 4 and 50 keV. In fact, there are some standard sizes of Si (Li) detectors, areas of 30 and 100 mm2 are considered normal, anything else is a detector made by specifications of the user. The energy deposited in a silicon detector by an X-ray is small, accordingly the electrical signal produced by collecting the charges from the detector is also small. Obviously, these signals will be correctly detected and analysed only if they are clearly above the electronic noise. Therefore, the problem of signal-to-noise ratio is particularly critical with silicon detectors. As a rule, the electronics contributes more than half to the FWHM of an X-ray peak. This was probably one of the reasons that a novel type of electronic gadget was developed: the optically reset preamplifier. The planar germanium and the silicon detector are used for spectroscopy of X-rays. In one aspect, the Si(Li) detector is better: its escape peaks are very small, and do not interfere with the spectrum proper. In the planar germanium detector, as much as 15% of the X-rays coming into the detector will be registered as escape peaks, making the spectrum which is complex anyway, even more difficult to analyse. The planar germanium detector can be recommended if high-energy X-rays; or very low-energy γ-rays are to be analysed. The efficiency of detection for X-rays at very low energies, say at 4 eV, strongly depends on the thickness of the entrance windows at the top of the detector. This is usually a beryllium foil, strong enough to hold vacuum, but thin enough to let X-rays enter the detector. Its thickness is specified in thousands of an inch; 1-mL-thick Be windows are frequently used, thinner are available. Thicker windows are recommended for countries with high humidity. For X-rays detected by Si(Li) crystal, the efficiency also depends on energy (see Fig. 4.37). At the low-energy end, the efficiency decreases towards zero: no wonder, very soft X-rays cannot penetrate into the detector and if they do not come in they cannot be registered. Some manufacturers are replacing Be windows with a window made of lower Z material which improves low-energy response of the detector (see Fig. 4.38). At the high-energy end, the X-ray might travel through the detector without an interaction; the efficiency will decrease with the energy. Note that X-rays interact with the silicon mainly by the photoeffect; this is good—there is very little of the Compton tail, the spectrum looks cleaner.
4.2 Radiation detectors
Full energy detection efficienty (%)
100 80 0.0075 mm (0.3 mil)
60
0.0075 mm (0.3 mil) 3 mm thick
40
0.0075 mm (0.3 mil)
5 mm thick
20 0 C
N
O
0.1
F Ne Na MgAl Si
1
10
100
Energy (keV)
FIG. 4.37 Efficiency of Si(Li) detector as a function of X-ray energy for different thicknesses of Be window and detector sizes.
Intensity
O
C Quantum window spectrum Be window spectrum Ca
0
1
2
3 Energy (keV)
4
5
FIG. 4.38 X-ray spectrum of a detector with low-Z material window.
4.2.2.4 Thermoluminescent detectors When radiation energy is absorbed by crystals of certain materials (e.g. lithium fluoride, lithium borate and calcium fluoride), the absorbed energy is trapped (stored) as displaced electrons within the crystal structure. If the material is heated later after the exposure, the trapped electrons are released and the stored absorbed energy is released in the form of visible light. This process is called thermoluminescence. Materials having this characteristic are called thermoluminescent.
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After its exposure to ionizing radiation, a thermoluminescent detector (TLD) is read-out in special apparatus (the TLD reader) as follows: The TLD material is heated to a suitable temperature (about 200°C) with a heating element or lamp; the trapped electrons in the TLD return to their normal energy state, releasing their extra energy as a light pulse. The light output is converted to an electric pulse by a photomultiplier tube (the working principle of which is illustrated in Fig. 4.27 for the scintillation counter). The size of the electric pulse is measured, and is proportional to the light output from the TLD material, which in turn is proportional to the total radiation energy absorbed, i.e., to the total radiation dose accumulated over the time that the TLD material was exposed. The TLD material can be used again after read-out. Thermoluminescent materials are commonly used in personal dose metres, in so-called TLD badges.
4.2.2.5 Nuclear track detectors When charged particles, e.g., α-particles, impinge on certain types of plastic materials like polycarbonate or cellulose nitrate, they cause radiation damage tracks in the material. The tracks can be made visually detectable through chemical or electrochemical etching procedures. The visible tracks can be counted using a microscope, microfilm reader or automatic image analysers. The number of tracks is used to calculate the total amount of radiation to which the detector material was exposed. Nuclear track detectors can be used to determine concentrations of α-emitting radon and thoron; the number of tracks is related to the radon or thoron concentration. Nuclear track detectors can also be used to indirectly detect fast neutrons. Fast neutrons interact with the base material of a special film and cause recoil protons to be released. These protons then cause damage tracks in the film which can be made visible and counted as described earlier. The number of tracks can be used to determine the neutron dose.
4.2.2.6 Photographic film as a radiation detector Ionizing radiation affects photographic film in the same way as light does. Films used for radiation measurement are sealed in light-tight packets. When the film is exposed, the radiation causes a latent image which becomes visible as blackening when the film is developed. The optical density (blackness) of the film is measured by passing a beam of light through it. The radiation dose that the film received can then be inferred from the optical density.
4.2.2.7 Neutron detection Neutrons have mass but no electrical charge. Because of this they cannot directly produce ionization in a detector, and therefore cannot be directly detected. This means that neutron detectors must rely on a conversion process where an incident neutron interacts with a nucleus to produce a secondary-charged particle. These
4.2 Radiation detectors
charged particles are then directly detected and from them the presence of neutrons is deduced. The most common reaction used in neutron detection today is n + 3 He ! p + 3 H + 765 keV
where both the proton and the 3H are detected by gas-filled proportional counters using 3He fill gas. In addition, the following reactions are used to detect neutrons: •
•
•
•
B(n, α) reaction: A neutron collides with a 10B atom and an α-particle is released in the process. The ionization caused by the α-particle can then be measured. This reaction is used in the BF3 gas-filled proportional tube. Boron-lined proportional counters and boron-loaded scintillators are other examples of neutron detectors using this reaction. Nuclear fission: A gas-filled detector, typically an ionization chamber, is coated inside with a thin film of fissile material like uranium. Absorption of neutrons by this material causes nuclear fission that produces highly ionizing fission fragments, which can be readily detected by the ionization chamber. Proton recoil: Fast neutrons can knock hydrogen nuclei right out of their atoms. The resulting so-called recoil protons can then cause ionization that can be detected and measured. This reaction can be utilized in gas-filled detectors, nuclear track films and scintillators. Neutron activation: This is the term used for neutron reactions that result in the formation of radioactive nuclides. This type of reaction is used in solid-state use-once devices containing one or more materials that are activated by neutron radiation. The induced activity of each material can be measured and the neutron exposure can then be calculated from these activities. 10
4.2.3 New developments In this section we mention some of the new developments in this very active field. Most of the effort is directed towards searching for new materials which would improve the detector resolution and make detectors more available to the researcher. The efforts are usually problem oriented trying to provide better detection capabilities for the solution of particular problems. For example, Sabet et al. (2016) have demonstrated the feasibility of a novel technique for fabrication of high-resolution CsI(Tl) scintillation detectors dedicated to single photon emission computed tomography (SPECT). The scintillators are fabricated using laser-induced optical barriers technique to create optical microstructures; or optical barriers, inside the CsI(Tl) crystal bulk. The laser-processed CsI(Tl) crystals were 3, 5 and 10 mm thickness. In their work, the authors focus on the simplest pattern of optical barriers in such a way that the barriers were created inside the crystal to form pixel-like patterns. The monolithic CsI(Tl) scintillator samples were fabricated with optical barrier patterns with 1.0 1.0 mm2 and 0.625 0.625 mm2
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pixels. A 4 4 array was exposed to 122 keV γ-rays and peak-to-valley (P/V) ratios were calculated. The P/V values greater than 2.3 were calculated suggesting that pixels can be resolved. The results obtained show that optical barriers can be considered as a robust alternative to mechanically pixelated arrays and can provide high spatial resolution while maintaining the sensitivity and providing cost-effective solution (see also Nagarkar et al., 2004; Sabet et al., 2012). Increasing interest in energetic particle (cosmic rays) effects on weather and climate has motivated the development of miniature scintillator-based detectors intended for deployment on meteorological radiosondes or unmanned airborne vehicles. Aplin et al. (2017) have reported the measurement of ionizing radiation in the atmosphere with a new balloon-borne detector. Their detector is made by a 1.0 1.0 0.8 cm3 caesium iodide thellium activated, CsI(Tl), scintillator with an integrated 1 cm2 silicon PIN photodiode, operated under reverse bias of 12 V. The CsI(Tl) scintillators are more mechanically robust with a greater intrinsic efficiency per unit volume than similar scintillators such as NaI(Tl) and in addition the wavelength of their optical output is optimal for photodiode detection (see Fioretto et al., 2000; Saint-Gobain, 2016). The principle of operation is that ionizing radiation creates a flash of light in the scintillator, which is detected by the photodiode and converted into a charge pulse, proportional to the energy of the incoming particle. The associated current pulses are converted to voltage pulses by fast rise time current to voltage convertor. Coupled with additional electronics units this detector can be run from low bias voltage (12 V) and low current (20 mA) and also has the important advantage of being able to resolve particle energy. Difficult challenges in the areas of nuclear proliferation, homeland security and defence have focused attention on nuclear detection having a goal of enabling the rapid and effective detection, location and characterization of threat materials. The science of radiation detection materials intended for nuclear detection has seen greatly increased attention in recent years. A number of researchers have pursued formalized discovery of novel radiation detection materials for nuclear detection applications. It is assumed that formalized discovery can play a role in finding radiation detection materials for nuclear detection similar to the role it has played in other areas of science. Within the domain of electro-optical materials there are several examples of materials that have been discovered as a result of a detailed understanding of material performance like negative index refraction materials (Shelby et al., 2001), photonic materials (Cregan et al., 1999) and some lasing materials (Faist et al., 1994). Within the domain of radiation detection there has been significant number of research activities on the formalized discovery of materials for applications in highenergy research and medical diagnostics (Peurrung, 2008). Formalized discovery involves the use of one or more methods intended to accelerate the pace and effectiveness of discovery. This could be first principles performance prediction, highthroughput screening, combinational methods, informatic methods, material design, etc. Application of this methodology to search for new detector materials, especially scintillators is described in the work of Deng et al. (2007), van Loef et al. (2007) and Awater (2017).
4.2 Radiation detectors
The review paper by Milbrath et al. (2008) describes the current state of radiation detection material science with particular emphasis on national security needs. In this chapter, the radiation detector materials physics is reviewed with the aim of setting the stage for performance metrics that determine the relative merit of existing and new materials. Semiconductors and scintillator are the two primary classes that are evaluated and the state of art and limitations for each are presented.
4.2.4 Personal dosimetry detectors Personal radiation detectors (PRDs) detect and localize radiation sources generated by man-made devices, they are worn on the person and provide radiation detection in the immediate area around the wearer. The instrument can quickly pinpoint the location of radioactive source, allowing the wearer to respond to the exact location of the threat. The wearer should be able to use PRD effectively while performing other tasks. The PRD should not require the attention of the wearer until it alarms (Stephens, 2018). PRDs offer the functionality of traditional Geiger counter as well as an improved range of detection of multiple types of ionizing radiation. Variations include units that can monitor dose rate, stay time and peak values; options are available for user that may operate in explosive environments. They can be conveniently grouped into following groups: • • • •
basic survey metres, γ-surveys, surface contamination and postaccident response instruments and portable survey metres with full range of modular probes (Stephens, 2018).
Researchers at the national Space Biomedical Research Institute and the U.S. Naval Academy have developed a new radiation detector to be used during space missions where energy levels are dangerous and approximate doses are estimated. The device, called a microdosimeter, is small and low powered and it can measure atmospheric radiation levels in real time (Sauser, 2007). NASA is using dosimeters on every mission, measure the total accumulated amount of radiation exposure and can take form of a badge, pen size tube or a digital readout. However, the drawback of the procedure used is that the doses of radiation astronauts received while in space are not known until they return to Earth. The microdosimeter described here contains an array of cells made from silicon, each the size of a red blood cell and arranged on an electronic board like the squares of chess board, total size of a package of cigarettes. Each cell continuously measures the deposited energy spectrum. This system takes real-time measurements of radiation levels and can alert astronauts immediately if they are at risk. While extremely important to manned space missions, the microdosimeter will have Earth-based application also, especially in areas of nuclear power, different medical and industrial applications. At this point we refer reader to a key dosimeter reference: Radiation Dosimeters for Response and Recovery, Market Survey Report (DHS/OSTP/NUSTL) June 2016
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Table 4.4 U.S. Radiation Dosimeter Standards Standard no. Pub. year ANSI N42.49A 2011 ANSI NJ42.49B 2013 ANSI N13.11 2009 ANSI N42.20 2003 ANSI N322 2003
Standard title American National Standard for Performance Criteria for Alarming Electronic Personal Emergency Radiation Detectors (PERDs) for Exposure Control American National Standard for Performance Criteria for Nonalarming Personal Emergency Radiation Detectors (PERDs) for Exposure Control Personnel Dosimetry Performance—Criteria for Testing (2009) Performance Criteria for Active Personnel Radiation Monitors Inspection, Construction, and Performance Requirements for Direct Reading Electrostatic/Electroscope Type Dosimeters
(see NUSTL, 2016). One should keep in mind that the choice of radiation dosimeters should be governed by the standards they have to satisfy. The appropriate American National Standards Institute (ANSI) standards that cover radiation dosimetry in various applications standards are listed in Table 4.4. Dosimeters differ from other radiation detection devices that are designed for the purpose of preventing a radiological release by alerting a responder to the presence of radiation. Commercially available dosimeters range from low cost, passive devices that store personnel dose information for later readout, to more expensive, battery operated devices that display immediate dose and dose rate information. Readout method, dose measurement range and precision, size, weight, ruggedness and price are important selection factors. In the report NUSTL (2016), instruments have been organized into three main categories: electronic personal dosimeters, EPD, selfreading dosimeters and processed dosimeters. The market survey (NUSTL, 2016) identified 53 radiation dosimeters including 42 EPDs, 2 self-reading dosimeters, 5 processed dosimetry systems designed for user readout (4 of which are portable) and 4 processed dosimeters offered by dosimetry provider services.
4.3 Radiometric methods 4.3.1 Counting The fundamental statistical treatment by Currie (1968) shows that the (low) limit of detection is proportional to the square root of the number of background continuum counts under the peak region of interest, where the proportionality factor varies with the confidence level chosen. Since the detection limit is expressed in counts, a more
4.3 Radiometric methods
interesting parameter is the minimum detectable activity (MDA), expressed in becquerels and defined as the smallest quantity of radioactive nuclide which can be determined reliably, given the prevailing conditions of the specific spectral measurement. The MDA is inversely proportional to the absolute detection efficiency at the full-energy photopeak. Smaller MDAs can be obtained by lowering the background and increasing the detection efficiency (see also Watt and Ramsden, 1964; Friedlander et al., 1964). Therefore, for the measurement of the concentrations of radionuclides in environmental samples and especially for low concentration levels, it is essential to reduce the background rate as well as improving the counting efficiency. When considering the background rate reduction, it is important to pay attention to the origins of the background. These are 1. 2. 3. 4.
natural and artificial radioactivity in the environmental circumstances, radioactivity in the detector and/or shielding material, cosmic radiation and instrumental noise.
In the surrounding environmental materials, e.g., a concrete wall of a building, the soil, air or water, there are various kinds of natural and artificial radionuclides which give rise to a background. The most important are 40K, 226Ra and its decay products and 232Th and its decay products. As a result, many lines originating from these nuclides are observed even in a heavy shielding box, during the spectrometry with a Ge detector. The background due to these radiations can be reduced considerably by shielding with heavy materials such as lead and iron. Typically, such background can be reduced to a hundredth by a 10 cm thickness of lead or 30 cm of iron. In making a shielding box, it is important that the detector is surrounded entirely (4π direction) with shielding. In choosing the shielding material, one must pay attention to any undesirable contamination with radioactive substances. For example, a small amount of 210Pb(RaD), which is a member of the 226Ra series, is inevitably contained in lead. Its daughter nuclide 210Bi(RaE) emits energetic β-rays (Emax ¼ 1.17 MeV), resulting in Bremsstrahlung which may contribute to an increase in background rate. The concentration of 210Pb is very dependent on the mine where it was produced, and a careful check on the content of 210Pb is important for achieving ideal shielding characteristics. Since the half-life of 210Pb is 22 years, old lead has much better characteristics. Iron blocks are sometimes used instead of lead. However, the modern iron/steel is often contaminated with 60Co, and careful checking is necessary prior to the construction of the shielding assembly. This 60Co contamination originates from the 60Co source which was used to monitor the abrasion of the furnace wall. For shielding purposes, the use of old iron, e.g., from materials from a sunken ship, is recommended. The background may also originate from the trace amount of radioactivity in the detector itself and its assemblies, and cannot be eliminated by shielding alone and hence, the contamination of the detector materials with radioactive substances should be carefully examined. A typical example is 40K in the glass used for phototubes. This can be eliminated by the use of a quartz phototube. The 226Ra
173
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CHAPTER 4 Measurements of radioactivity
present in solder sometimes gives rise to a considerable rate of background. The molecular sieve used for keeping a vacuum in the cryostat of a Ge detector sometimes causes a background spectrum. In this sense, electrolytic copper, stainless steel and Perspex are recommended as materials to be used in the shielding box. Even when very thick materials are used for shielding, it is very difficult to stop the penetration of highly energetic cosmic rays, which have energies in excess of 103 MeV. In order to overcome this undesirable background, an electrical guard technique is employed. The cosmic rays which give a pulse to the main detector should produce another signal to one of the guard counters. The two signals from the main counter and from the guard counter are produced simultaneously. Therefore an anticoincidence technique is applied quite effectively to reduce the background rate, and an ultimate low background rate of the order of 0.1 cpm can be achieved by the combination of heavy shielding and the anticoincidence guard technique. The coincidence technique is often used for the low-level detection of electrons. As an example, we mention the low background β-ray spectrometer as developed by Tanaka (1961). A small Geiger-M€uller counter is mounted inside a large plastic scintillator, and scintillation pulses from only those events which are coincident in the two detectors are analysed by a multichannel analyser. In this case, signals caused by cosmic rays are also subject to the analysis, but the minimum width of the scintillator is designed so as to produce a very large pulse height, and this signal is rejected in the course of the pulse-height analyses. In these instruments, the β-ray spectra can be measured for even very weak samples. By the addition of a logarithmic amplifier, the analysis of β-ray spectra becomes even easier. Pulse-shape discrimination (PSD) is also used sometimes for background reduction. Electrical noise is very often caused by the discharge of high-potential circuits, the transition signal of thyristors and the induction signal from spark gaps, etc., which result in a background signal to the counter. To prevent this noise, electrical filtering in the main power supply, and electrical and/or magnetic shielding are sometimes required. Another essential requirement for reliable low-level counting is to increase the counting efficiency and/or to use a larger amount of source sample. In order to increase the counting efficiency, one should use a large volume (large size) detector, and try to improve the source-to-detector geometry. However, it should be noted that with increasing detector size or volume, the background may also increase. As for the improvement of the geometry, the use of a well-type detector is advantageous, but the source size is sometimes limited. On the other hand, to use a larger amount of source, it is convenient to use a Marinelli-beaker, especially for the measurement of samples such as milk or soil. In principle, one can perform the measurement of activity of any sample by using radiation detection systems and standard sources. Standard sources and solutions are available from many national and international operations. However, for sources one can perform radioactivity measurement without the help of standards. Absolute counting implies of method of radioactivity measurement performed without the
4.3 Radiometric methods
help of a standard/reference source. The methods for the absolute measurements can be classified into two categories. The first comprises the direct methods, which include the defined solid angle method, the 4π counting method, the internal gas counting method and the liquid scintillation counting (LSC) method. These methods use relatively simple procedures, and each provides unique applicability. However, some uncertain corrections, such as for self-absorption, are involved in these methods, and the accuracies obtainable are somewhat limited. The second category is based on the coincidence technique, by which we can determine the radioactivity more accurately without any uncertain estimation of correction terms. The most often used methods include: the 2π α-counting method, 4π β-counting with a 4π gas-flow counter, 4π β-counting with a liquid scintillation spectrometer and the 4π β–γ coincidence counting method.
4.3.1.1 2π α-Counting method
In cases of radioactivity measurements of an α-emitter such as 241Am or uranium source electrodeposited onto a metal disk, a simple 2π α-counting with a gas-flow proportional counter is practical and a reasonable accuracy can be expected. If the applied potential to the proportional counter is set at the α-plateau region, the counter responds only to α-particles. Since the backscatter of α-rays is small (a few per cent) compared with that of β-rays, and the value can be estimated as a function of the atomic number of the backing material, the radioactivity, n0, can be readily obtained from the observed counting rate, n, by the following simple relation: n0 ¼ 2n
1 1 ð1 aÞ ð1 + bÞ
(4.42)
where the factors a and b represent the corrections for the self-absorption and backscattering of the α-particles (Hino and Kawada, 1990). Self-absorption, a, of a uniformly distributed source with superficial density, d, can be estimated by the simple equation: a ¼ d=ð2RÞ
(4.43)
where R is the range (mg/cm ) of the α-particles in source material; this value can be referred from some data books, e.g., from Ziegler (1977). The effective thickness (superficial density), d, of the source can be determined from the line width of the pulse-height spectrum for α-rays obtained with a Si surface-barrier detector mounted in a vacuum chamber as shown in Fig. 4.24. The effective thickness, d (in mg/cm2 unit), of the source material can be obtained from 2
d¼
qffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi Γ 2obs Γ 2i =s
(4.44)
where ΓOBS and Γi are the observed and intrinsic line widths (FWHM), respectively, and s is the stopping power of the source material for the specified α-particle. In the cases of the measurements of electrodeposited sources of RaDEF (210Pb–210Bi–210Po), e.g., the stopping power of 294 keV cm2/mg for 5.3 MeV
175
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CHAPTER 4 Measurements of radioactivity
α-particle in PbO2 should be used. If the thickness of the sources ranges from 0.59 to 1.10 mg/cm2, the self-absorption values from 1.79% to 3.3% would result. Backscatter-values of α-rays are less than a few per cent, and can be estimated from the published data as a function of the atomic number of source backing. The correction, b, for the backscattering can therefore be easily obtained with an acceptable reliability. This method is very simple in procedures, and can be applied to the absolute measurements of very weak sources without any standard source. For these advantageous features, this method is useful for the measurements of samples for environmental radioactivity analysis and also for the preparation of laboratory-made standard αsources.
4.3.1.2 4π β-Counting with a 4π gas-flow counter
The classical method of 4π β-counting with a 4π gas-flow proportional counter is still useful for the absolute measurements of β-emitting nuclides provided that good sources with small self-absorption can be prepared. From the observed counting rate, after fundamental corrections for background and counting loss due to dead time, the radioactivity, n0, can be calculated as n0 ¼ n
1 1 1 af 1 as
(4.45)
where af and AS are the corrections for absorptions in the source-supporting films and in the source material itself. While the foil-absorption can be measured experimentally, e.g., by the sandwich method, it is difficult to make direct determination of the self-absorption values only by the use of 4π-counting. In practice, it will be convenient to estimate the self-absorption values from the published data taken as a function of β-end point energy for various source conditions. The self-absorption values are very dependent on both the β-ray energy and source conditions, and are in the range from 1% to several 10s%. In order to get reliable data with this technique, minimizing the foil-and self-absorptions is essential. The source preparation techniques are therefore of great importance. As materials for source support, a VYNS film (copolymer of vinyl chloride and vinyl acetate) whose thickness ranges from about 10 to 30 μg/cm2 can be easily prepared in the following manner. 1. Dissolve the VYNS powder (i.e. from the Union Carbide Co.) in cyclo-hexanone (CsH10O) 5 times in volume, and keep it for at least 1 week in sealed glassware to ensure homogeneous mixing. 2. Attach a small amount of VYNS solution onto the edge of a glass plate (e.g. a slide glass for a microscope observation) and contact the edge onto the surface of clean water. A thin VYNS layer will be spread over the water surface. 3. Scoop up the film with thin metal rings for film support, and dry naturally. In this procedure, many source supports can be prepared simultaneously if a number of rings are set on a holder.
4.3 Radiometric methods
Table 4.5 Source preparation techniques for 4π counting Name of technique
Procedures
Seeding
After adding a drop of freshly prepared 1:5000 dilution of Ludox-SM (Dupond) with a distilled water, dry in air with or without an infrared lamp irradiation After adding a drop of a 1:10,000 dilution of Catanac, dry in open air with or without an infrared irradiation For example: (1) Iodine isotopes: After adding a drop of AgNO3 solution, dry naturally in open air (2) Mercury isotopes: Dry in an atmosphere of H2S After adding a drop of pure water, dry in air with an infrared lamp irradiation. Add a drop of water again, and dry in an atmosphere of ammonium
Wetting Precipitation
Ammonium complex forming
4. In order to avoid the accumulation of electrical charges on the source mount in a 4π counter, the thin VYNS film thus prepared should be rendered electrically conducting by the vacuum deposition of gold or aluminium to both sides (15 μg/cm2 each side). The sources are prepared by a simple deposition of a known amount of radioactive solution of sample on the support. The solid contents of the sample solution must be kept to a minimum to minimize the self-absorption. Even when carrier concentration is small (0.01–0.05 mg/cm2), a considerable amount of self-absorption is unavoidable. This is mainly due to the local agglomeration or crystallization formed in the dry. Several procedures as illustrated in Table 4.5 are useful to extend the source material uniformly onto the source support and consequently to reduce the selfabsorption. For volatile materials such as iodine, mercury and selenium, special care and treatment are needed to form a more stable chemical form. In order to measure the radioactive concentration, i.e., radioactivity in a unit mass or unit volume of sample solution, quantitative deposition of sample solution is necessary. This is achieved by gravimetrical means with a microbalance or with volumetrical ones with a micropipette. For precise measurements, the former method is preferable, in which a differential weighing with a pycnometer is essential to avoid the appreciable amount of evaporation of solution in the course of the weighing (0.5%). Although the attainable accuracy is limited to around a few per cent in this simple 4π β-counting method, this method can be applied to the measurements of weak source, and hence can play a special role in the environmental radioactivity measurements after chemical separation.
4.3.1.3 4π β-Counting with a liquid scintillation spectrometer A liquid scintillation counter is considered as a version of a 4π counter, since the radioactive material is mixed in a scintillator, and the effective solid angle sustained from the source to the detector is nearly 4π. A large quantity of sample solution is
177
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CHAPTER 4 Measurements of radioactivity
usable in a LSC compared with the samples used for the original 4π. This is one of the advantageous possibilities of LSC. Although the pulse height may be decreased and distorted due to the quenching phenomena, and accordingly the counting efficiency changes because of the source conditions, the possible counting loss due to the quenching effect can be corrected by using the pulse-height information. The classical means to correct the counting loss are the external standard method and the channel ratio method. However, the efficiency tracing technique has been proved to be more advantageous and convenient to determine the absolute activity. In this improved technique, most of the pure β- and β–γ emitting nuclides, including 3H and 14 C, can be measured accurately and readily by the use of only one or two common standard samples without the need to prepare quenching correction curves. In the relationship between the efficiencies for the standard sample and the count rates of the sample to be measured, the counting efficiency of 100% with respect to the standard sample serves to derive the radioactivity of the sample to be measured. Signal pulses resulting from liquid scintillation are analysed in separate discriminating regions with a multichannel analyser; the relation between the efficiency, E, and the count rate, n, is empirically expressed by the quadratic regression equation: n ¼ aE2 + bE + c:
(4.46)
Extrapolated value to E ¼ 100% gives the true disintegration rate of the sample. This technique has universal applicabilities, and weak radioactive samples can be measured in terms of absolute activity. These features are considered to be advantageous in the environmental radioactivity measurements.
4.3.1.4 4π α-Counting with a liquid scintillation spectrometer
Since α-rays emitted from radionuclides are energetic (4 MeV), and almost monochromatic, the detection efficiency of a liquid scintillation counter for α-rays is expected to be near unity. However, if β-rays are emitted from the source in addition to α-rays, the spectra originated from α-rays and from β-rays are sometimes overlapped partially. In order to overcome this effect, PSD technique is useful to select α-signals from α–β mixing signals. The background due to β-rays is entirely diminished after pulse-height discrimination. It should be noted that a special scintillator cocktail such as NE-213 should be used in this α–β PSD method. A similar cocktail can be prepared by mixing 5 g of PPO, 0.3 g of DMPOPOP, 100 g of naphthalene and 1 L of xylene. This special α-counting with a liquid scintillation spectrometer is useful for the α-radioactivity measurements of environmental samples on an absolute basis.
4.3.1.5 4π β–γ Coincidence counting method
Among various techniques used for radioactivity standardization, the 4π β–γ coincidence counting method is undoubtedly the most important technique, and most radionuclides can be standardized by this technique or its extended techniques. Actually, most national standardizing laboratories in the world employ this method for the establishment of the national standards of various kinds of radionuclides.
4.3 Radiometric methods
The 4π β–γ coincidence counting is an improved version of the β–γ coincidence counting method. Here, a 4π β gas-flow-type proportional counter is usually used as the β-detector, and a NaI(Tl) scintillation counter located near the 4π β-counter wall is employed to detect the γ-rays. Source preparation techniques are the same as those explained in the 4π β-counting. Even in this 4π β–γ coincidence method, the source preparation to minimize the self-absorption is important to reduce the correction terms and also to shorten the extrapolation range in the 4π β–γ coincidence extrapolation technique. Hence the special treatments as explained in Table 4.4 are important in the course of source preparation. For the measurements of α–γ emitting nuclide such as 241Am, the same technique can be used, provided that the applied potential to the 4π proportional counter is set at the α-plateau region. This 4π β–γ coincidence counting technique can be readily applied to the absolute measurements of electron capture nuclides such as 51Cr, 54Mn, 88Y, 75Se, etc., since a 4π gas-flow counter filled with P-10 gas (gas mixture of Ar + 10% methane) has considerable detection efficiency for the electron capture events through the detection of the characteristic X-rays and Auger electrons. This should be called 4π X–γ coincidence counting, but the measuring procedures are essentially the same as those for 4π β–γ coincidence technique. Here the extrapolation technique is again useful for making various corrections. In cases of 4π X–γ coincidence counting, an alternation of the flowing gas, from P-10 gas to pure methane, is effective to change the counting efficiency of the β-counter to X-rays. This 4π β–γ coincidence counting method and its extended technique are very useful for the precise absolute measurements of radioactivity, and therefore these methods are employed for the establishment of national standards of radioactivities of various kinds of nuclides in most of the national standardizing laboratories. Laboratory-made standard source can be prepared by these techniques. However, these techniques are not adequate for the measurements of low-level radioactivities, and hence the direct application to the measurements of environmental samples is difficult. In this arrangement, apart from some corrections required, the disintegration rate (radioactivity), n0, and counting efficiencies, εβ and εγ, of the β- and γ-channels are given by the following simple equations for a β–γ emitter with a simple decay scheme: n0 ¼ nβ nγ =nc εβ ¼ nc =nγ and εγ ¼ nc =nβ ,
(4.47)
where nβ, nγ and nc represent counting rates, after correcting the background, accidental coincidences, etc., in the β-, γ- and coincidence channels, respectively. The above fundamental equations can be easily derived from the following three equations: nβ ¼ n0 εβ , and nc ¼ n0 εβ εγ nγ ¼ n0 εγ ,
(4.48)
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CHAPTER 4 Measurements of radioactivity
The explanation along this line is usually made in most textbooks. However, the ideal conditions are seldom achieved in any practical counting system, and some modifications of the fundamental equations are required in order to correct the possible effects which may disturb the ideal conditions. For example, the 4π βproportional counter has an appreciable sensitivity to γ-rays. Furthermore, the γtransition is detected by the β-detector through the internal conversion process, if any. Besides, because a coincidence mixer has a finite resolving time, false accidental coincidences are inevitably produced by chance. In addition to this problem, further consideration must be given when a nuclide with a complex decay scheme is measured. Taking account of all of these effects the coincidence equation becomes nβ nγ 1 τR n∗β + n∗γ εc εβ 1 n0 ¼ αεce + εβγ + C 1+ εβ 1 + α εγ nc 2τR n∗β n∗γ ð1 n∗c τÞ
(4.49)
and εβ ¼
nc 1 nγ 1 n∗ τ
(4.50)
β
where εβγ is the γ-sensitivity of the β-counter; εce is the counting efficiency of the β-detector to internal conversion electrons; α is the total internal conversion coefficient; εc is the probability of obtaining γ–γ coincidence signal when a β-particle is not detected; C is the correction for the complex decay scheme; τR is the resolving time of the coincidence mixer and τ is the dead time of the β-channel. The notations with an asterisk denote the counting rate including background rate. These coincidence equations should be valid for all types of coincidence measurements. The effects due to the γ-sensitivity of the β-counter, the response of the β-counter to internal conversion electrons and the effect originating from the complex decay scheme are all reduced by a factor of (1 εβ)/εβ. In the 4π β–γ coincidence counting, the β-efficiency, ε, is expected to be nearly unity, and the corrections for these effects remain small. The attainable accuracy is therefore improved very much in this technique. This is one of the reasons why the 4π β–γ coincidence counting method is superior to the ordinary β–γ coincidence counting. However, when a nuclide with a complex decay scheme is measured, considerable correction is sometimes required. As a practical approach to overcome these difficulties, a series of 4π β–γ coincidence countings should be made for various source conditions which provide different β-efficiencies, and an extrapolation to (1 – εβ)/εβ ! 0 gives the true disintegration rate. It should be noted that most correction terms such as for the γ-sensitivity of the β-counter and the response to the internal conversion electrons as well as for the complex decay scheme are all corrected automatically in this procedure. The shape of the efficiency function, graph of the nβnγ/nc against (1 εβ)/εβ, is dependent on the decay scheme and the γ-gate setting, etc. In the radioactivity measurements by means of coincidence technique, the counting statistics are given by the following formula:
4.3 Radiometric methods
rffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi sffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi 1 εβ 1 εγ 2εβ εγ εβ εγ + 1 1 ¼ + : σ¼ nc T n0 T n0 εβ εγ T
(4.51)
The second term of Eq. (4.51) is a function of (1 – εβ), which will be reduced considerably in the use of a high-efficiency detector, and excellent counting statistics are obtained in a short measuring time. This is another advantageous future use of the 4π β–γ coincidence method.
4.3.2 γ-Spectrometry High-purity detection systems having a very low background are suitable tools for the direct measurement of low-level radioactivity in environmental samples. The background features of the detection system are of considerable importance because they have to be known for one to obtain an estimate of the detection limit and of the MDA. The natural radioactivity background originates from the uranium and the thorium series from 40K and from cosmic rays. Natural radioactivity is found in most materials, and it is necessary to shield the detector using carefully selected materials of high density. The materials used in the detector assembly, and the shielding materials need to have the lowest possible inherent background radiation (see Gilmore and Hemingway, 1995). Containers such as Marinelli-beakers can be filled with aliquots of environmental samples and placed on top of the end cap of the detector in an accurately reproducible beaker–detector geometry (Park and Jeon, 1995; Dryak et al., 1989). The radioactivity of the samples can be measured if the detector has been calibrated with Marinelli-beaker standard sources (MBSS) of the same dimensions, density and chemical composition. The calibration procedure should follow as closely as possible that defined in the IEC Standard 697 (1981). This method enables the simultaneous detection of several γ-emitters present in the sample matrix without the need for separation of the radionuclides from the matrix. The method can be applied to a large variety of environmental and biological materials such as air, water, soil, sediments, vegetation (grass, hay, etc.) and, particularly, individual foods of vegetable and animal origin as well as total diet mixtures. This method is suitable for the surveillance and monitoring of radioactivity originating from the operation of nuclear plants, nuclear weapons tests and releases from nuclear accidents (Kanish et al., 1984). It is recommended that the γ-spectrometry system should be a fully integrated data acquisition and computation system comprising the following items (IAEA 295, 1989): •
A vertical, high-purity germanium (HPGe) detector is recommended. The detector should have an efficiency of 18%–20%. Generally, the efficiency of germanium detectors is specified as the photo peak efficiency relative to that of a standard 7.62 7.62 cm2 cylindrical NaI(Tl) scintillation crystal and is normally based on the measurement of the 1.33 MeV γ-ray photopeak of a 60 Co source with a source-detector spacing of 25 cm used in both measurement systems. The resolution of the detector which is normally specified for germanium detectors as the full-width (in keV) at half-maximum
181
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CHAPTER 4 Measurements of radioactivity
•
•
• •
•
•
• •
(FWHM) of the full-energy peak of the 1.33 MeV peak of 60Co should be between 1.8 and 2.2 keV. It is recommended that the peak-to-Compton ratio of the detector be >46:1. The peak-to-Compton ratio is defined as the ratio of the count in the highest photopeak channel to the count in a typical channel just below the associated Compton edge and is conventionally quoted for the 1.33 MeV γ-ray photopeak of 60Co. A preamplifier is necessary. This is normally an integral part of the detector unit and it is located very near the detector in order to take advantage of the cooling which is necessary for the operation of the detector and which helps the preamplifier to operate with low noise. A biased high-voltage power supply is required to supply high voltage to the detector through the preamplifier. A power supply of 1500–5000 V is adequate for the operation of germanium detectors. A linear amplifier is required to process the output signals from the preamplifier. A detector shield will be needed with a cavity which is able to accommodate large (up to 41) samples, constructed of either lead or steel with some type of graded line to degrade X-rays. Lead shields have a much lower backscatter effect than steel shields. Typically, lead shields have walls 5–10 cm thick, lined inside with graded absorbers made of cadmium (l–6 mm) and copper (0– 4 mm). Other materials, such as Plexiglas and aluminium, are also used as absorbers. A multichannel analyser (MCA) with a minimum of 4096 channels should be connected to a keyboard and display screen for input and output of data and interaction with a computer. Several kits are available for the conversion of personal computers (PCs) into MCAs. Basically there are three types of conversion kits. One makes use of board with an analogue-to-digital converter (ADC) that simply clips into the PC; a second type uses a clip-in board with an external ADC and the third type uses a multichannel buffer (MCB) connected to the PC. All of these PC-based MCA systems are relatively inexpensive and are very suitable for use in germanium and sodium iodide γ-ray spectrometry. A rapid data-storage and recovery system is needed. It can consist of magnetic tape, hard disk, floppy disk, or a combination of these media. This system can be used for programming, short-term storage of data, and archiving data. A high-speed printer is required for data output. Useful, but not absolutely necessary, is a plotter for archiving spectral drawings. Software for system operation and data reduction is usually supplied with the MCA system. Software packages with varying features and capabilities are available for MCAs based on PCs.
Several aspects of γ-ray spectrometry with such a system deserve some discussion. Interferences associated with γ determinations may be caused by improper spectral identities, changes in background, errors in calibration and/or geometry and lack of homogeneity in samples. Many of the problems in γ-ray spectrometry are due to malfunctions of
4.3 Radiometric methods
electronic components. Very important also is the calibration of the measuring systems; both energy and efficiency calibration should be performed with care. Energy calibration of a germanium detector system (i.e. establishing the channel number of the MCA in relation to a γ-ray energy) is achieved by measuring mixed standard sources of known radionuclides having well-defined energies within the energy range of interest, usually 60 to 2000 keV (IAEA-619, 1991). The use of the lower energy photons emitted by 241Am may indicate changes in the intercept. Mixed γ-ray standards are available in various forms and containers from reliable suppliers. A partial list of radionuclides usually available with γ-ray energies in the range of interest includes: 241Am, 109Cd and 57Co, 139Ce, 51Cr, 22Na, 54Mn and 60Co. The energy calibration source should contain a selection of radionuclides with at least four different γ-ray energies. It is recommended that one of the nuclides should be 137Cs. The gain of the system should be adjusted so as to position the 662 keV photopeak of 137Cs at about one-third full scale. It is also recommended that the gain of the system be adjusted to 0.5 keV/channel. Once these adjustments are made, the gain of the system should remain fixed. As an example we show energy calibration with a point source incorporating 60 Co, 133Ba and 137Cs. The peaks corresponding to the following energies are present in the measured spectra: 81.00, 301.85, 356.00, 661.65, 1173.24 and 1332.50 keV (see Fig. 4.39). The energy calibration, represented by a quadratic equation: E ¼ a + bX1 + cX2
(4.52)
is determined (from the six pair data of peak centre channel (X) and the energy (E)) by using least square method for determination of the coefficient (a, b and c).
6
1332.50
Counts
661.65
4
10
1173.24
5
302.85 356.00
10
81.00
10
103 10
2
101 0
10
0
1000
2000 Pulse height/Channel
FIG. 4.39 Gamma-ray spectrum of an energy calibration source.
3000
4000
183
184
CHAPTER 4 Measurements of radioactivity
An accurate calibration of the efficiency of the system is necessary to quantify the radionuclides present in a sample. It is essential that this calibration be performed with great care because the accuracy of all quantitative results will depend on it. It is also essential that all system settings and adjustments be made prior to determining the efficiencies and should be maintained until new calibration is undertaken. Small changes in the settings of the system components may have slight but direct effects on the counting efficiency. In general, it is not so easy to obtain the absolute peak efficiency ε(E, x) which is a function of both γ-ray energy (E) and geometry (x) between source and detector. The following method also includes an approximation that the energy dependency of the peak efficiency is scarcely affected by geometry, and the peak efficiency consists of two components being independent of each other. εðE, xÞ ¼ f ðEÞ εðxÞ
(4.53)
where, f(E) represents the energy dependency, which was normalized at the specific energy and is often called “relative peak efficiency”. ε(x) gives the absolute peak efficiency that was obtained for various geometries using the radioactivity standard sources emitting γ-rays of the specific energy. The experimental procedure for determination of peak efficiency vs energy curve involves the following steps: •
• •
•
Place a standard source (51Cr, 54Mn, 57Co, 60Co, 85Sr, 88Y, 109Cd, 137Cs, 139Ce) at the position of 2 cm apart from the top face of the Ge detector using a source holder. Fig. 4.40 shows the γ-ray spectrum of such a standard source. Accumulate the spectrum for a time period (t) long enough to reduce the counting error; approximately 1000 s is recommended. After the measurement has finished, the spectrum shown in Fig. 4.40 is accumulated (integral net count of peak region) for each peak corresponding to the energy: 88, 122, 136, 166, 320, 514, 662, 835, 898, 1173, 1332 and 1836 keV. Calculate the peak efficiencies for all peaks using the following equation Peak efficiency : ε ¼
peak area=t 100% aA
(4.54)
where a is a fraction of γ-ray emission per disintegration, and A is the present radioactivity (Bq). •
Plot data of ε(%) vs E(keV) on log–log graph paper, then draw a smooth curve through all points. The result of such work is shown in Fig. 4.41, as an example.
The equation for the efficiency can be written in the form: h i f ðEÞ ¼ exp a + b ln ðEÞ + c ln ðEÞ2
(4.55)
Quantitative γ-ray spectrum analysis of environmental samples such as soil, water or ash of food requires the peak efficiency for volume sample. A Marinelli-beaker is often used as a container for a large quantity of sample such as water or soil. As
f (E)
Log / Counts/Channel 5
4
0.0
0.1 0.05
3
0.5
0.1
0.2
FIG. 4.41
Relative peak efficiency as a function of energy. 898.02 keV, γ-BS
0.5 Energy (MeV)
1.0 Gamma ray energy (MeV)
1
1836.04 keV, γ-88
S.E.(1836 keV) 1332.47 keV, Co-60
1173.21 keV, Co-50
D.E(1836 keV)
661.65 keV, Gs-137
511.00 keV, γ-rays 513.99keV, Sr-85
391.69 keV, Sn-113
255.12 keV, Sn-113 279.19 keV, Hg-203
83.032 keV, Cd-109 122.063 keV, Co-57 136.47 keV, Co-57 165.855 keV, Ce-139
4.3 Radiometric methods
6
2
1
1.5 2.0
FIG. 4.40
γ-Ray spectrum of the standard for efficiency calibration.
9
5
2
1
0.5
0.2
2
3
185
186
CHAPTER 4 Measurements of radioactivity
the volume of sample to be measured is usually fixed, the absolute peak efficiency is dependent on energy only, but affected by self-absorption which depends on density of matrix and the elemental composition. The method to be used for determination of the peak efficiency is based on experimental procedures involving the following steps: First, the Q-peak efficiency for an aqueous sample is determined by: •
placing an aqueous standard sample (51Cr, 54Mn, 57Co, 60Co, 85Sr, 88Y, 109Cd, Cs, 139Ce) in the Marinelli-beaker with the Ge detector; accumulating the spectrum for a time period (t) long enough to reduce the counting error (approximately 1000 s of preset time is recommended); after the measurement has finished, one should analyse the spectrum. Read the peak area for each following peak from the printed data 88, 122, 136, 166, 320, 514, 662, 835, 898, 1173, 1332 and 1836 keV and calculate the observed peak efficiency
137
• •
•
Observed efficiency : ε ¼
peak area=t 100% aA
(4.56)
where a is a fraction of γ-ray emission per disintegration and A is the present radioactivity (Bq) of each nuclide. Let the peak efficiency for 662 keV be ε1. Second, the Q-peak efficiency for a heavy liquid sample is determined by: •
• • •
Replacing the above sample with a heavy liquid sample (ZnBr2 in water; 80% in weight; ρ ¼ 2.5 g/cm3; weight composition: Zn(0.232), Br(0.568), H(0.02), 0 (0.178)) in a Marinelli-beaker of the same size. Then, measure the spectrum for approximately 1000 s. After the measurement has finished, analyse the spectrum, read the peak area for 662 keV from the printed data, then calculate the peak efficiency. Peak efficiency should be corrected by self-absorption. Determine the self-absorption correction factor (K) by Eq. (4.57). K¼
1 ε1 ln μ2 ρ2 μ1 ρ1 ε2
(4.57)
where ρ is density (g/cm3) of matrix: ρ1 ¼ 1.0, ρ2 ¼ 2.5; μ is the mass attenuation coefficient; μ1 ¼ μ of water for 662 keV ¼ 0.0857 cm2/g and μ2 ¼ μ of heavy liquid for 662 keV ¼ 0.0742 cm2/g. • Calculate the peak efficiency ε* without self-absorption for each energy: ε∗ ¼ ε=eKμρ
where μ is mass attenuation coefficient (cm2/g) of water and ρ is density of water (¼1). • Plot the data of ε* vs E(keV) on log–log graph paper, then draw a smooth curve through all points.
4.3 Radiometric methods
• •
From the curve, read the efficiency for each energy shown below as exactly as possible. The coefficients (a, b and c) of absolute peak efficiency without selfabsorption can be calculated separately for two regions of energy: h i ε∗ ðEÞ ¼ exp a + b ln ðEÞ + c ln ðEÞ2
•
(4.58)
Thus, the peak efficiency for a sample with linear attenuation coefficient (μρ) is represented by the following equation: εðEÞ ¼ ε∗ ðEÞ eKμρ
(4.59)
Most laboratories involved in radiation measurements now use personal computers and commercially available software for the analysis of γ-ray spectra. Some of these programs allow the user to control the multichannel analyser (MCA), calibrate the detector for various geometries, and provide analysis results. The programs are easy to use and do not require the user to be an expert in γ-ray spectrometry. As an example we refer to work by Heimlich et al. (1989). In their work, the γ-ray spectrum analysis program, GAMANAL, has been modified to operate on a small computer. The program uses an algorithm involving a Gaussian and a tailing term for fitting and resolving peaks obtained from spectrometers using germanium detectors. γ-ray energies, intensities and absolute photon emission rates can be determined. A graphical output showing the original and fitted data can also be obtained. The results generated by the program are stored on disk for further analysis. This allows the use of other computer programs and languages in tasks such as decay curve analysis, radionuclide activity measurements and neutron activation analysis. Sanderson (1989) evaluated commercially available IBM PC-compatible software in 1987. At that time, it was reported that most of the programs satisfactorily detected peaks and resolved doublets of equal intensity. Problems arose when the doublets were of unequal intensity or the analysis of a complex spectrum was needed. The suppliers of the programs involved in that study have corrected some of these deficiencies. Since many of these programs have undergone numerous revisions, and a few programs have become available, a re-evaluation was performed (Decker and Sanderson, 1992). Six programs were evaluated in this study: GAMMA-W (Gesellschaft f€ur Kernspektrometrie, Germany), INTERGAMMA (Intertechnique Instrumentation Nucleaire, France), QSA/Plus (Aptec Nuclear, Inc., Canada), OMNIGAM (EG&G Ortec, USA), GDR (Quantum Technology, Inc., USA) and SAMPO 90 (Canberra Nuclear Products Group). Hardware requirements were similar for all the programs tested (IBM PC compatibility, 384–640 K of memory, a hard-disk drive). Two programs, QSA/Plus and SAMPO 90, also required a maths coprocessor. The QSA/Plus program had to be installed in Windows 3.0. All of the other programs operated directly from DOS. Except for GAMMA-W, all the programs control data acquisition. GAMMA-W, which is written by an independent company that does not manufacture nuclear
187
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CHAPTER 4 Measurements of radioactivity
instrumentation, does not allow the user to read 10 different formats of γ-ray spectra. The conclusions reached by the authors were as follows. All of the programs satisfactorily found small peaks using sensitivity values recommended in the program manuals. However, these values may not be optimal for every situation. When the sensitivity values were lowered, additional valid peaks were found. When the recommended sensitivity values were used, only two programs did not report any false peaks. All of the programs were able to resolve equal-intensity doublet peaks with only a 2 keV (four-channel) separation. The resolution of doublets of unequal intensity, especially where the smaller is on the high-energy side of the predominant one, has improved since the last evaluation. However, some programs still require improvement in this area. All the programs for analysis of γ-ray spectra measured with germanium detectors generally allow the following steps (among others) to be carried out: a. b. c. d. e.
transfer of spectra from the MCA to the computer; search for and identification of photopeaks; energy and efficiency calibrations; resolution of multiplets and calculation of activities and the dosimetric magnitudes deriving from them.
Most of these programs, however, are not specifically designed for the development of plans for environmental radiological monitoring, where one must measure periodically and systematically activity levels of a set of man-made γ-emitters in different types of environmental samples. Such samples often present very low counting statistics for these emitters. Moreover, since most of these programs employ elaborate algorithms for the search and fit to the shape of the γ-peaks, they are coded in FORTRAN (Koskelo et al., 1981) and are therefore implemented on mini computers or, in more reduced versions, on personal computers. Also commercially available programs are difficult for the user to modify, requiring usually a great deal of programming effort to adopt them to the peculiarities of each laboratory. Baeza et al. (1992) have developed ESPEC: a set of programmes specially designed to undertake γ-spectrometric analyses of environmental samples. The processes enumerated earlier (a–e) are carried out in a simple form. The package is designed so that it can be implemented not only on an IBM-type personal computer, but even on one with less memory and lower performance, such as a 64 KB memory microcomputer. The code is written in BASIC and it is easily modified to suit the specifications of any MCAs.
4.3.3 β-Particle spectrometry Table 4.6 illustrates β-emitting radionuclides; since β-emitters show a continuous emission spectrum, the average energy EB and the maximum energy Em B are given. Measurement uncertainties are shown in brackets after the respective value (in units of the last significant digit).
4.3 Radiometric methods
Table 4.6 Some common β emitters Nuclide
Half-life
Em B
EB (%)
3
12.35(1) a 5730(40) a 14.29(2) d 87.44(7) d 2.75(2)a 96(4) a 50.5(1) d 28.7(3) a 64.1(1) h 59.3(2) d 22.3(2) a
18.6 156.48 1710.4 167.47
5.68 49.47 695.0 48.80
65.87 1492 546.0 2284
17.13 583.1 195.8 934.8
16.5 63.0
4.15 (80) 16.13 (20)
H C 32 P 35 S 55 Fe 63 Ni 89 Sr 90 Sr 90 Y 125 I 210 Pb 14
Er
Kα 5.9, Kβ 6.5
Kα 27.4, Kβ 31.0
C Amplifer R
Scaler
Pulse shaper and discriminator Timer
High-voltage power supply Source
FIG. 4.42 Block diagram of the counting electronics associated with a G-M tube.
The classical way to measure low-level β-particle activity is with a Geiger-M€uller (GM) gas-flow counter. The block diagram of the GM counter is shown in Fig. 4.42. In anticoincidence with a guard detector the GM gives a fairly low background (0.003 s1) and a counting efficiency of 40%. The disadvantage, however, is its inability to give energy resolution. As the characteristics of the GM tube may gradually change over a long period of use, it is necessary to examine the characteristics of the tube and background count occasionally. The counting efficiency of a GM counter for detecting β-rays is defined as the ratio of the count rate to β-activity of the source. Thus, the counting efficiency is one of the most important factors to determine the characteristics of the detector. There is the following relation between activity A (Bq) and net count rate n (s1) in the case of activity measurement using the GM counter n ¼ A fg fi fsa fs fa
(4.60)
189
190
CHAPTER 4 Measurements of radioactivity
where fg is the geometry factor of counting setup; fi is the intrinsic efficiency of GM tube for β-rays; fsa is the source self-absorption factor; fs is the source-mount backscattering factor and fa is the correction due to absorption between source and GM tube. Eq. (4.60) is rewritten as Eq. (4.61), if fg, fi, fsa, fs is replaced by η¼
n Af a
(4.61)
where η is called counting efficiency; it represents the rate of count rate n corrected for the adsorption fa to activity A. The counting efficiency η is obtained by using the standard source of the known activity A, and the absorbers with the known thickness dm. If the count rates for β-rays are defined as n and n0, respectively, which correspond to those passing through with and without absorber, fa is equal to n/n0, where fa represents the transmission rate of β-rays for the absorber thickness dm. When the thickness of absorber dm is not so large compared with the maximum range of β-rays, n can be represented by the experimental formula n ¼ n0 eμm dm
(4.62)
Therefore, if one plots the n values as a function of the thickness dm on semilog graph paper, the n values decrease linearly with the dm values increase. The slope of this line is μm, which is called the mass absorption coefficient. If one puts dm ¼ 0, the count rate n0 (¼ n/fa) can be obtained. Thus the equation becomes η¼
n0 A
(4.63)
If η is already known, the unknown activity A0 is obtained by A0 ¼
n00 η
(4.64)
where n0 0 is the count rate extrapolated to dm ¼ 0. Furthermore, in the case that the actual count rate n is larger than 100 s1, it is necessary for the count rate to correct the resolving time τ[s] of the GM counter. It has been shown (Holm et al., 1990) that ion-implanted detectors can be used not only for α-particle spectrometry but also for β-particle spectrometry. Their drawback, however, is the high background around 100 keV and noise below 1 keV. Olsson et al. (1992) have described a detector system for low-level β-particle spectrometry where the good characteristics of gas-flow and silicon detectors are used. The gas-flow GM counter used in the detector system is a windowless single-channel version of the GM-25-5 multicounter developed in the Riso National Laboratory, Denmark. The gas-flow counter utilizes a GM counter with a diameter of 22 mm and a guard of 80 90 10 mm3, using argon (99%)/isobutane (1%) as counting gas. Background counts, produced in the sample counter by cosmic radiation, are reduced by means of the guard and attached anticoincidence circuits. As the energy-discriminating detector, a passivated implanted silicon detector, with an
4.3 Radiometric methods
active area of 450 mm2 and a depth of 300 μm, was used. The PIPS detector has a 0.5-μm-thick aluminium coating on the front surface. The detector is placed on top of the gas-flow counter, and integrated into the gas detector unit, allowing the aluminized front surface to act as one of the ground electrodes of the sample counter (Holm et al., 1990). The source is placed in a cavity inside the detector, between the sample and guard counter. The instrumentation used for PIPS detectors is identical to that used for α-particle spectrometry. The authors have demonstrated that it is possible, by the coincidence technique, to reduce the contribution of noise and background from a PIPS detector by a factor of 10 and to improve its energy resolution. Such a detector system could be a useful tool for quality control of low level, low energy, pure β-emitters such as 63Ni from environmental samples.
4.3.4 α-Particle spectrometry Common radioisotope sources of α-particles are listed in Table 4.7. All but the first one listed are members of radioactive decay chains. Decay chains are classified into four groups according to their mass numbers. They are Th-series whose mass number is 4N (N is integer), U-series of 4N + 2, Ac-series of 4N + 3 and Np-series of 4N + 1. An Np-series does not exist naturally because the half-life of its longest-lived member is three orders of magnitude shorter than the age of the Universe. In three natural decay chains activities of each parent and succeeding daughter nuclides are equal. This condition is called radioactive equilibrium, because half-lives of parent nuclides, 232Th, 238U and 235U are much longer than those of their daughter nuclides. Table 4.7 Common α emitters Nuclide
Half-life
α-Energy/MeV (% branching)
147
1.07 1111 years 60.55 m 138.38 d 0.296 μs 55.6 s 3.824 d 1600 years 1.405 1010 years 2.45 105 years 7.038 108 years 4.468 109 years 2.413 104 years 6570 years 432 years 18.1 years
2.232(100) 6.051(72), 6.090(28) 5.304(100) 8.784(100) 6.288(99.93) 5.490(99.9) 4.602(5.5), 4.784(49.5) 3.954(23), 4.013(77) 4.723(27.5), 4.775(72.5) 4.368(12.3), 4.400(57) 4.150(23), 4.197(77) 5.105(11.5), 5.144(15.1) 5.157(73.3) 5.124(26.4), 5.168(73.5) 5.443(13.1), 5.486(85.2) 5.763(23.6), 5.805(76.4)
Sm Bi 210 Po 212 Po 220 Rn 222 Rn 226 Ra 232 Th 234 U 235 U 238 U 239 Pu 240 Pu 241 Am 244 Cm 212
191
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CHAPTER 4 Measurements of radioactivity
α-particles emitted from nuclides which decay to a single level are observed as mono-energy particles. On transitions given the branching ratio in Table 4.6, multiple α-energies are observed. Such a fine structure in the α-spectrum comes about because an α-emitter may decay to any one of several discrete energy levels of its daughter. 241Am is commonly used as a standard source. The detection of α-particles is based on the physics of the processes which take place when the particles pass through matter. During this passage the α-particle loses its energy by excitation and ionization of the atoms. The energy loss per unit path length is called stopping power(dE/dx). Stopping powers of various elements are given (Ziegler, 1977; Northcliff and Shilling, 1970). Stopping power at energy above 1 MeV is inversely proportional to α-energy. At the energy range of <1 MeV stopping power is nearly proportional to the energy. Because the velocity of α-particles is slow compared with orbital electrons, α-particles capture electrons and their effective changes for ionization decrease. When α-particles pass through matter, they make small-angle scattering. Therefore an asymmetric line spectrum which has a tail extending to the lower energy side is observed after penetrating relatively thick foils, because of asymmetry in their trajectory path. Source thickness has an important effect on the observed α-particle spectrum. As α-particles have a relatively large stopping power, the observed spectra for thick sources show some degradations caused by self-absorption. For high-resolution spectrometry, thin α sources or samples are required. They are usually prepared by electrodeposition or vacuum evaporation. With a source thickness <10 μg/ cm2, no effects due to self-absorption are observed. For such thin sources, a Gaussian-shape line spectrum whose width is limited by the detector energy resolution, is observed for a mono-energetic α source. When there is an absorber between an α source and a detector, the observed energy is reduced by the energy loss in the absorber. Absorbers are the entrance window of a detector, the covering material which prevents contamination of a source, air, and so on. In the case of a thin source, prepared by chemical separation, the emitted α-particles are observed as line spectra by spectrometers having high-energy resolution. The intensity of each α-emitter is easily estimated from the area of the corresponding peak. Even if the peak is super-imposed on the tail of another peak, the peak area can be calculated from the shape of the line spectrum. An example of this situation is plutonium isotopes, 239Pu and 240Pu. They are used for the estimation of a burn up of nuclear fuel. As the energy difference of these α-emitters is only 10 keV, the α-particle spectrum is observed as an overlapped single peak. However, when a Si detector is used, which has an energy resolution of <10 keV (FWHM), the overlapped peaks can be analysed by the least squares fitting technique. α-particle spectrometry is usually performed using Si detectors, which are especially useful for thin and small area α sources. Two types of Si detectors are commonly used; the surface-barrier type and the ion-implanted PN-junction type. As an alternative to Si detectors, a Frich-grid ionization chamber is sometimes used as a spectrometer of α-particles. This ionization chamber has a grid between a
4.3 Radiometric methods
cathode and anodes and a sample is put on the cathode electrode. Electrons and ions, which are generated between the grid and the cathode by ionization, drift towards the anode or the cathode, and a signal pulse is obtained. The pulse height obtained from the cathode depends on the emission angle of α-particles. Only the drift of electrons is observed due to their drift velocity being 1000 times greater than that of ions. However, the height of the anode pulse is proportional to the α-energy. The advantage of this counter is that the areas of samples can be made larger than with Si detectors. A commercially available counter of this type has an area larger than 1000 cm2. The energy resolution of this counter is 40–50 keV (FWHM) for α-particles. Several other types of spectrometers can be used in some specific applications. For example, an organic liquid scintillation counter is useful for detection of extremely low-level α-activities. α-emitters, chemically separated from samples, are mixed into the scintillator. The geometrical efficiency is then 100% and a relatively large amount of source can be introduced into the scintillator. The main drawbacks of the counter are its poor energy resolution and its relatively low light output in excitation by α-particles compared to that by electron. Track detectors are also often used for α-detection. The length of tracks produced by α-particles is measured with a microscope after appropriate chemical etching. The energy spectrum of α-particles is obtained from the distribution of their trace lengths. Advantages of this detector are its high sensitivity, good discrimination against β- and γ-ray backgrounds, and the acquisition of α-emitter distribution in samples. Its main drawbacks are a poor energy resolution and the time-consuming processing of the microscope reading. It should be mentioned that the energy of α-radiation is one of the most important characteristics of radionuclide sources. Knowledge of the α-particle energy is necessary for the determination of other major characteristics of external radiation of α sources, such as the flux and energy flux density, and the absorbed dose rate. Information on the α-radiation energy is also used for calibrating semiconductor α spectrometers. The results of accurate measurements of α-particle energies play an important role in the evolution of the atomic mass scale and comprise nearly 60% of the input data for atoms with A > 200, and are deciding factors in the design of precision spectrometers for the high-precision energy measurements of α-particles from radionuclide sources with the minimum attainable uncertainties. A very precise measurement of α-particle energies can be achieved by the measurement of the α-particle time of flight (Frolov, 1992). A special problem is the assay of α-particle emitters in water samples. There are several procedures in the literature for assaying α-particle-emitting radionuclides in natural waters. However, few of them are well suited for low-level analysis since they require the handling of large water samples and the application of concentration techniques. Typical water samples do not contain sufficient amounts of some important nuclides for a precise or reasonably rapid measurement. The most common concentration technique is the coprecipitation of the heavy elements by the addition of a carrier. Some of the best-known methods are coprecipitation with iron as the hydroxide and with calcium and strontium as the oxalates (Livingston et al., 1975).
193
194
CHAPTER 4 Measurements of radioactivity
However, these methods present some disadvantages. These include the transportation of large volumes of water (several, to hundreds of litres) to the laboratory, the isotopic exchange of tracers, the use of large containers and laboratory ware and the difficulty of recovering the precipitate. As a consequence of these difficulties, interest in studying in situ concentration techniques with appropriate adsorbents has increased. These methods eliminate the need for preservation, storage, evaporation or coprecipitation of large water samples. One such approach has been described in detail by Crespo et al. (1992). According to the authors MnO2 and Al2O3 can be used as absorbents in the assay of the low levels of α-emitters in the waters. This technology requires more development work before it can become routine. Another problem of importance to safeguards and reactor fuel technology is the measurement of the relative abundance of plutonium isotopes. In addition to γ-and X-ray spectrometry (Gunnick, 1982), α-particle spectrometry has also been used (Bland et al., 1992; Amoudry and Burger, 1984a; Bortels and Collaers, 1987). This usually results in a complex α-spectrum, with difficulties in obtaining correct amplitudes for overlapping peaks which tail towards lower energies. The 239Pu and 240Pu peaks overlap when α-particle energies are measured using Si detectors. Some sort of deconvolution procedure is required in this case (see, for example, Bland et al., 1992). The radiation dose due to radon and its decay products in the environment amounts to more than 50% of the total radiation dose to man from natural radiation sources. Recently the risk of developing lung cancer from exposure to inhaled radon daughters has been attracting considerable attention. Thus it is very significant to measure α-activity levels in the air. Let us briefly describe a filter-sampling method used to assay the gross α-activity concentration of radon daughters in the air by means of the gross α-particle counting using ZnS(Ag) scintillation counters. An air sampler is used to collect natural radionuclides onto a filter paper by passing air through, because some nuclides of radon progeny (RaA ¼ 218Po, RaB ¼ 214Pb, RaC ¼ 214Bi, RaC0 ¼ 214Po, etc.) tend to attach to aerosol particulates floating in the air. Supposing only one radionuclide is attached to the surface of a filter paper, the rate of change with time in the total number of the radioactive atoms on the filter paper can be written as dN ¼ ξqn λN dt
(4.65)
where t is the sampling time, N is the total number of radioactive atoms on filter paper at time t, ξ is the flow rate of sample air (constant), n is the concentration of radioactive atoms in the air (assumed to be constant), λ is the decay constant of the radionuclide, ξqn is the rate of collection and λN is the rate of decay. The solution of Eq. (4.65) for the initial condition N ¼ 0 at t ¼ 0 is N¼
ξqn 1 eλt λ
(4.66)
4.3 Radiometric methods
The radioactivity A is given by λN, or
A ¼ ξqn 1 eλt
(4.67)
The counting rate R of α-particles emitted from the filter paper at time t is related to the radioactivity A: R ¼ εζA
4.68)
where ε is the counting efficiency of α-ray counter and ζ is the emerging efficiency of filter paper. From Eqs (4.67) and (4.68) the concentration n of the α-emitting nuclide in the air is obtained n¼
R εζξqð1 eλt Þ
(4.69)
The α-activity concentration Q in the air is given by λn or Q¼
λR R 0:693 1 ¼¼ εζξqð1 eλt Þ εζξq T 1 ð1=2Þt=T
(4.70)
where T (¼0.693/λ) represents the (α-decay) half-life of the radionuclide. For longlived radionuclides such as 238U, 239Pu and 226Ra, Eqs ((4.66), (4.67) and (4.70), respectively, become N ¼ ξqnt
(4.71)
A ¼ λξqnt
(4.72)
R : εζξqt
(4.73)
Q¼
4.3.5 Liquid scintillation measurement method LSC has been accepted as the generally preferred method of counting weak β-emitters and is useful to a lesser extent for α- and γ-emitters. The counting sample consists of three components, the radioactive material, an organic solvent or solvent mixture and one or more organic phosphors. A particle or radiation emitted by the sample material is absorbed in, and its energy transferred to the solvent and then to the phosphor which emits a scintillation of light photons. These photons are absorbed by the photocathode of a photomultiplier tube which converts them into an electronic pulse. The pulse, after suitable amplification, is registered as a count corresponding to the emission of the particle or radiation (Neame and Howewood, 1974). Configuration of a typical LSC is shown in Fig. 4.43. The technique has the following characteristics: 1. Self-absorption is usually negligible. 2. There is no absorption of radiation by air or a detector’s window between the radioactive sample and the sensitive region of the detector.
195
CHAPTER 4 Measurements of radioactivity
Photomultiplier tube
Photomultiplier tube
Sample
High voltage
Coincidence
Pneumatic
196
Pump
Summation Amplifier Gate Analyzer Data output
Source external standard
FIG. 4.43 Configuration of a typical liquid scintillation counter.
3. There is no radiation scattering prior to incidence on the detector. 4. 4π counting is performed, because the radioactive material is completely surrounded by the liquid scintillator. Based on the above merits, the liquid scintillation measurement is extremely sensitive to the low-level radioactivity existing in the environment and food. Since the radioactive sample material from most methods of sample preparation is in intimate contact or in actual solution with the phosphor, the detection of emitted particles or radiation is highly efficient and may even approach 100%. Problems of self-absorption of the emissions are thus absent, or considerably smaller than those associated with planchette counting of solid samples. This is of particular importance for the measurement of low-energy β-emitters such as tritium and carbon-14. On the other hand, the measurement method has intrinsic drawbacks such as quenching and chemiluminescence.
4.3 Radiometric methods
Currently, the liquid scintillation counter has been employed not only for the measurement of low-energy β-emitters, but also for pure β, β–γ and α-emitters and further Cherenkov radiation. The liquid scintillator consists mainly of organic solvent and fluorescent material (i.e. solute), and sometimes a surfactant or other material is added to the solution. The characteristics of the liquid scintillator depend mostly on the sort and amount of these chemicals. The liquid scintillator plays the role of an energy transducer, converting radiation energy into photons. The organic solvent which comprises most of the liquid scintillator should satisfy the following conditions (Neame and Howewood, 1974): 1. The energy should efficiently transfer in the process of luminescence. 2. An absorption spectrum of solvent should never overlap an emission spectrum of solute. 3. The radioactive sample and solute must be able to be incorporated with the solvent. 4. The solvent must be of high purity. In the past many kinds of chemicals were employed as solvents, but nowadays only a few typical solvents are being used including toluene, xylene, pseudocumene and droxane. The characteristics of typical solvents used for liquid scintillators are shown in Table 4.8. The solute is classified into a primary solute and a secondary one. The former is a main fluorescent material, and the latter serves as a wavelength shifter which gives rise to an emission spectrum having long wavelength. These characteristics and molecular structures are respectively represented in Table 4.9. The sample to be measured is easily prepared by incorporating the radioactive sample into the liquid scintillator such as xylene-base-, toluene-base- or emulsive scintillator. The xylene (or toluene)-base-scintillator is appropriate for hydrophobic samples, to form a homogeneous solution which provides efficient energy transfer and light-counting efficiency. Table 4.8 Typical solvents used for liquid scintillator
Solvent
Molecular weight
Solidifying point (°C)
Absorption spectrum (A) λmax a
Toluene Xylene Pseudocumene Dioxane
92.13 106.16 120 88.1
95 20 60.5 12
2.620 2.660 2.690 1.880
a
Wavelength giving maximum value. Wavelength giving mean value.
b
Emission spectrum (A) λmax a λmean b
Relative pulse height
2.870 2.890 2.930 2.470
1.00 1.09 — 0.65
2.840 2.880 2.900 —
197
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CHAPTER 4 Measurements of radioactivity
Table 4.9 Typical solutes used for liquid scintillator Absorption spectrum λmax
Emission spectrum
Solute
a
λmax a
λmean b
Optimum concentration (g/L)
Primary solute PPO Secondary solute DMPOPOP bis-MSB
3.030
3.640
3.703
4–7
1.6
3.630
4.290
4.273
0.2–0.5
1.5
3.470
4.120
4.219
0.5
1.3
Decay time of scintillation
a
Wavelength giving maximum value. Wavelength giving mean value.
b
The simplest mixture consists of a primary scintillator and a secondary scintillator dissolved in a primary solvent. The primary solvent converts the kinetic energy of radiations into excitation energy. The primary solvents are usually aromatic hydrocarbons, and therefore only nonpolar radioactive materials can be dissolved in them. The most widely used primary solvent is toluene. Others are p-, and m- and mixed xylenes. Pseudocumene (1,2,4-trimethylbenzene) is becoming a popular solvent for new, commercially produced scintillation cocktails. It offers the highest energy conversion efficiency of the solvents known and has fewer restrictions on shipping and storage as a combustible liquid because of its high flashpoint, low volatility and lower toxicity. Owing to the presence of organic solvent, the mechanism of luminescence for the liquid scintillator is much more complicated than that for a crystal scintillator. The energy successively transfers to generate finally the luminescence, as follows: 1. 2. 3. 4.
excitation of solvent molecule due to the absorption of radiation energy; solvent–solvent energy migration; solvent–solute energy transfer and luminescence from solute molecule.
Since the solvent molecules have a majority in the liquid scintillator, they are initially excited by the radiation energy. The π electron of the solvent molecule plays an important role in the process of energy transfer due to its active mobility. When the solute absorbs the excitation energy liberated from the solvent, the solute molecules in the ground state are excited up to the excited electron level or its vibrational one. Then, the following processes compete with one another as intramolecular behaviours of the solute: 1. 2. 3. 4.
internal conversion of molecule, fluorescence emission, intersystem crossing and phosphorescence emission.
4.3 Radiometric methods
The quenching is a phenomenon caused by the energy loss in the process of the energy transfer inside the liquid scintillator. It should be taken into account, whenever the liquid scintillation measurement is performed. Owing to the existence of a quencher, which means the material giving rise to the quenching, the counting efficiency finally decreases as follows: 1. Energy is partly lost during the energy transfer. 2. The number of photons emitted from the solute decreases, and/or some of the photons are wastefully absorbed in solution. 3. The height of electric pulse created by a photomultiplier tube becomes lower. 4. A pulse-height distribution shifts towards a lower pulse height. 5. The number of pulses which are present within a countable region decreases, leading to the lowering of the counting efficiency. Most materials are regarded as quenchers whose quenching strength depends on the material itself and its amount. The quenching and optical (or colour) quenching are practically significant. The impurity quenching is the quenching which arises in the process where the excitation energy of the solvent is provided to the solute. Its mechanism can be explained by the theory of an exciplex which means the excited complex formed by two different kinds of molecules. Final energy loss is represented by the following scheme: 1
M∗ + Q ! M + Q ∗ ! M + Q ðdislocationÞ
(4.74)
M + Q ! 1 M + Qðtransition to ground stateÞ ðthrough intersystem crossingÞ
(4.75)
exciplex
3
where 1M is the solvent molecule in singlet state, 3M is the solvent molecule in triplet state, Q is the quencher molecule and * is the excited state. When the material whose absorption spectrum overlaps, more or less, the emission spectrum of the solute is contained in a prepared sample, the fluorescence generating from the solute is partly absorbed to cause the optical quenching. The strength of the optical quenching depends on the sort and density of colours. In general, red and yellow cause strong quenching, but blue causes weak. Even though a quencher is colourless, the optical quenching occurs if the quencher has an absorption spectrum which partly overlaps an emission spectrum of solute. The liquid scintillation measurement method is a quantitative technique to determine radioactivity, but not a qualitative one to find radiation energy or the sort of nuclide. Based on the internal-sample counting, the measurement method has great merits for low-level radioactivity. The following three measurement methods are in general often used.
4.3.5.1 External standard method The index of quenching with respect to sample is found from Compton spectrum created by γ-ray irradiation to the sample. A set of quenched standards are employed for the construction of a quenching correction curve which represents the relationship
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CHAPTER 4 Measurements of radioactivity
between the index of quenching and the counting efficiency. The radioactivity of the sample to be measured can be determined by using the quenching correction curve.
4.3.5.2 Sample channel ratio method The index of quenching in this method is obtained from an areal ratio or a centre of gravity with respect to a β-ray spectrum. The radioactivity is determined through a similar quenching correction curve to that of the external standard method.
4.3.5.3 Automatic efficiency tracing method As a matter of fact, the foregoing two methods are confined to the measurements of only tritium and carbon-14. The automatic efficiency tracing method, however, makes it possible to determine the radioactivity of many sorts of radionuclides such as pure β-, β–γ and α-emitters. The radioactivity can be found from a quadratic regression equation which is constructed by the relationship between the counting efficiencies of a standard sample and the counting rates of a sample to be measured. Sample preparation techniques involve the chemical separation of the specific chemical forms of a radionuclide, and the preparation of the counting sample by mixing the separated nuclides with the proper liquid scintillator. A large number of recipes for scintillation mixtures have been published, but for many purposes a few simple ones will suffice. The properties of a suitable mixture are as follows: 1. It should generally be clear, colourless and homogeneous after addition of the radioactive sample, although in certain cases suspensions or gels may be satisfactory. 2. It should quench as little as possible. This is particularly important when measuring radioisotopes such as tritium whose β-particles are of low energy. 3. It should not be expensive. 4. Its constituents should be stable. Some scintillators are known to be unstable when exposed to light and other materials may deteriorate on storage, with the formation of impurities which cause quenching or chemiluminescence. Scintillation mixtures should be stored in dark bottles. The simplest mixture consists of a primary scintillator and a secondary scintillator dissolved in a primary solvent. The primary solvent converts the kinetic energy of radiations into excitation energy. Primary solvents are usually aromatic hydrocarbons, and therefore only nonpolar radioactive materials can be dissolved in them. The most widely used primary solvent is toluene. Other primary solvents are p-, m- and mixed xylene. Pseudocumene (1,2,4-trimethyl benzene) is becoming a popular solvent for new, commercially produced scintillation cocktails. It offers the highest energy conversion efficiency of the solvents known, and has fewer restrictions on shipping and storage of a combustible liquid because of its high flash point, low volatility and lower toxicity. The primary scintillator converts excitation energy into light. Two commonly used primary scintillators are PPO (2,5-diphenyl-oxazole) and butyl-PBD (2-(40 -
4.3 Radiometric methods
tert-butyl-phenyl)-5-(40 -biphenyl)-1,3,4-oxadiazole). The pulse voltage produced by butyl-PBD is about 20% greater than that produced by PPO, but its usefulness is reduced because of its limited solubility. Suitable concentrations of PPO and butyl-PBD are about 5 and 7 g/L, respectively. With alkaline samples butyl-PBD produces a brownish colour in the scintillation cocktail. In highly quenched samples, a higher concentration may be required for optimum efficiency. The secondary scintillator shifts light wavelength and may be needed in the mixture if the emission wavelength of the primary scintillator does not match the wavelength to which the photomultiplier is most sensitive. Commonly used secondary scintillators are POPOP (1,4-bis-(2-(5-phenyloxazolyl))-benzene) and dimethylPOPOP (1,4-bis-(2-(4-methyl-5-phenyloxazolyl))-benzene), used at concentrations varying from 0.05 to 0.5 g/L (usually 0.1 g/L). POPOP is the less soluble of the two but is slightly cheaper. Both should be given adequate time to dissolve completely, preferably overnight. The best general purpose secondary scintillator is bis-MSB (p-bis-(o-methylstyryl)-benzene). Bis-MSB is readily soluble in solvents and has a fast rate of dissolution. Bis-MSB is used at 0.5–1.5 g/L and does not react chemically with most liquid scintillation samples, while dimethyl-POPOP can react with acids to produce a yellow to greenish colour. A suitable mixture for toluene-soluble materials is toluene containing 5 g PPO/L or 7 g butyl-PBD/L 0.1 g POPOP or dimethyl-POPOP/L. This mixture cannot be used for measuring the radioactivity in aqueous samples since water is immiscible with toluene. A further solvent (secondary solvent) miscible with both water and primary solvent must be added to this mixture to enable water to be incorporated. This amount of aqueous solution which can be incorporated in a particular mixture without phase separation depends on the identity of the secondary solvent, the relative proportions of primary and secondary solvents and the temperature. The secondary solvents used for this purpose are a surface-active agent or emulsifiers. Triton X-100 and Triton N-101 are nonionic surfactants commonly used in laboratory-prepared scintillation cocktails. Anionic surfactants give better sampleholding capacities than the nonionic surfactants for certain types of salt solutions. As a general rule, the greater the ratio of secondary to primary solvent, the greater the amount of water which can be incorporated, but the extra secondary solvents and the extra water increase the quenching. The properties of a good emulsifier scintillator are as follows: 1. Mixture of scintillator and water should show high counting efficiency and a low background counting rate. 2. It should show high water capacity. 3. The region of phase separation between the homogeneous liquid region and the stable emulsion or gel region should be narrow. 4. It should produce minimum chemiluminescence. 5. Mixture of aqueous solution should be stable for a long period. Mixtures having water content between 10 and 20 vol% should be avoided because they are unstable and show two phase separation. Table 4.10 shows some useful mixtures used in liquid scintillation measurements.
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Table 4.10 Some useful mixtures used in liquid scintillation measurements Mixture 2 PPO DMPOP Toluene Triton X-100 Conc. HNO3
Mixture 3 6 g/L 04 g/L 667 mL 333 mL 10 μL
PPO bis-MSB Xylene Triton N101 Conc. HNO3
6 g/L 0.5 g/L 700 mL 300 mL 10 μL
The scintillation vial is the container for the analyte and the scintillator and permits light transfer from the scintillation cocktail to the photomultiplier tubes. Many materials have been used for scintillation vials. Low-potassium, borosilicate glass is the best material to meet the requirements for most LSC. It is nonpermeable and nonreactive with chemicals used in LSC. It has good optical clarity for light transmission. The glass should be selected for low radioactivity content, providing a low background counting rate. Polyethylene vials are popular because of their low cost. They show a low background counting rate and good counting efficiency. The opacity of polyethylene vials makes it difficult to detect phase separation and precipitation in the sample mixture. The major disadvantage of polyethylene vials is that they are permeable to aromatic hydrocarbons. As the primary solvent permeates into polyethylene, it carries the scintillators with it, changing the background counting rate and affecting the efficiency. The radioactivity may also be carried into the vial wall leading to geometry and self-absorption losses. Other materials for vials are quartz and Teflon. They are the best materials, but they are very expensive to make and used only in applications requiring very low backgrounds and long counting times. When one wants to measure low-level radionuclides with good precision and accuracy, one has to adjust the counter settings properly and to select the most suitable scintillators and vials to ensure the highest sensitivity and reproducibility. For these purposes, the “Figure of Merit” is adopted as a quantitative criterion: Figure of Merit ðFMÞ ¼ ðEMÞ2 =B,
(4.76)
where E is the counting efficiency (%), M is the mass (g) of a sample introduced in the vial and B is a background counting rate (cpm). The performances of several emulsifier scintillators prepared in a laboratory and available in the market show no significant difference in sensitivity. Some of the radionuclides commonly distributed in environmental samples include 3H, 14C, 6CO, 89Sr and 90Sr. Tritium and 14C emit low-energy βs which are efficiently counted by LSC. Here we present some details of sample preparation for LSC. Aqueous samples containing tritium are distilled to eliminate impurities and mixed with an emulsifier scintillator with the proportion of 40–50 vol% water. Tritium in biological samples exists as tissue free water tritium (TFWT) and organically bound tritium (OBT). Water containing TFWT is obtained by vacuum distillation (lyophilization) or azeotropic distillation using organic solvents such as toluene
4.3 Radiometric methods
and benzene. The OBT is converted into HTO for counting by combusting dried samples. The tritium concentration in the natural environment is less than 10 Bq L1 and sometimes less than 1 Bq L1. Only low-background-type LSCs detect these low tritium levels. To measure tritium below 10 Bq L1 with good accuracy, using the conventional type LSCs, it is necessary to enrich the tritium in a large water sample by electrolysis. This procedure can increase the tritium concentration 20–30-fold. Carbon-14 exists mostly as carbon dioxide (CO2) and as organic compounds in the environment. CO2 in the atmosphere or in the effluent gas from a combustion system for biological samples is first absorbed in alkaline solutions such as aqueous NaOH or NH4OH. If necessary, calcium chloride or barium chloride is added to the alkaline solution to precipitate CaCO3 or BaCO3, which are then purified and stored in a sealed bottle for future analysis. The alkaline solution containing CO2 or carbonate is acidified by titrating with a strong acid solution to generate CO2. The CO2 gas is then absorbed in a solution of organic amine absorber and liquid scintillator mixture (1:1). The procedure for 60Co measurement is as follows: after adding cobalt salt as carrier to seawater, 60Co is first precipitated as cobalt hydroxide in alkaline solution. The precipitate is separated from seawater and dissolved in HCl solution. 60Co is separated and recovered in the effluent from an anion-exchange column. The effluent containing 60Co is evaporated to dryness. The residue is redissolved in dil. HCl solution and transferred into a counting vial. The emulsifier scintillator is added into the vial and mixed well. The counting sample is counted by means of a low background LSC. This technique can be applied to 60Co in biological samples by using an ashing procedure beforehand. Strontium isotopes can be measured by utilizing the fact that high-energy β-rays in aqueous solution emit Cerenkov photons which can be counted by photomultipliers in a LSC without any phosphors. Strontium radioisotopes in a seawater sample are separated and purified by precipitation of strontium carbonate and subsequently strontium oxalate precipitate followed by the cation exchange. Yttrium-90, the daughter nuclide of 90Sr, is scavenged from the solution by Fe(OH)3 coprecipitation. The acidified aqueous solution containing strontium isotopes is transferred into a counting vial and counted after the total volume of the solution is adjusted by adding distilled water. By setting the lower level of the discriminator properly, the LSC can count only Cerenkov light signals produced by 1.49 MeV β-rays of 89Sr, without counting the 0.546 MeV β-rays of 90Sr. 90Sr is determined by counting Cerenkov photons generated by 2.28 MeV β-rays of 90Y which grows into the solution after separation.
4.3.6 Radiochemical analysis 4.3.6.1 Introduction There is a considerable amount of attention focused on long-lived artificial radionuclides such as 90Sr and 137Cs released into natural environments. Their inventories began to grow as a result of nuclear weapon testing programs since 1950 and the
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nuclear power reactor accident of Chernobyl in 1986. The new sources contaminating the environment with radionuclides are recently thought to be liquid effluents from nuclear power stations and nuclear fuel reprocessing facilities. The long-lived fission products and their characteristics are given in Table 4.11 (World Health Organisation, 1966; Sugihara, 1961; Norton, 1967; Chase and Rabinowitz, 1967; Harley, 1972; Greenberg et al., 1981). It is not practicable to present here all the methods used; we outline only the reliable methods used for 89Sr and 90Sr for tritium and for actinides (for details, see IAEA-295, 1989).
4.3.6.2 Analysis of strontium Strontium-90 is one of the most important fission products because of its relatively high yield (about 6%), long physical half-life (29 years) and its uptake and retention in biological systems. For assessing the integrated exposure to large populations, not only the direct measurements of biological materials must be done, but also the monitoring of the transportation of the nuclide in the environment, e.g., in oceans and streams. The assay of low-level strontium-90 in biological and radiotoxicological samples requires time-consuming and laborious techniques because both the nuclide 90Sr and its daughter 90Y are pure β-ray emitters. Therefore, the radiostrontium must be completely separated from other radionuclides prior to the β-ray counting. Table 4.11 Long-lived fission products Nuclides
Decay mode
Half-life (years)
79
β β (γ) β β β β β (γ) β β (γ) β (γ) β (γ) β (γ) β (γ) β (γ) β β (γ) β (γ) β (γ)
6.5 104 10.7 4.8 1010 28.8 1.5 106 2.14 105 1.00 6.5 106 14.6 55 2.7 1 105 1.57 107 2.06 3 106 30.17 2.62 4.9
Se Kr 97 Rb 90 Sr 93 Zr 99 Tc 106 RU 107 Pd 113m Cd 121m Sn 125 Sb 126 Sn 129 I 134 Cs 135 Cs 137 Cs 147 Pm 155 EU 95
4.3 Radiometric methods
The most commonly used method for separating strontium is by nitrate precipitation. With some modifications this method can be applied to all kinds of environmental samples and foods. The chemical yield varies according to the type of material. The use of 85Sr tracer to determine chemical yield is a general practice. When determining yield in this manner, it is important that the tracer is pure 85Sr, i.e., free from 89Sr and 90Sr. Although the method is time consuming, it is reliable and safe. Rapid methods for 90 Sr analysis exist, and it has been shown that they can be used after short-lived nuclides have decayed. In fresh fallout situations, the nitrate precipitation method has been shown to be more reliable, also, during periods of fresh fallout, the amount of 89Sr is of interest and the rapid methods can only analyse for 90Sr. In the case of higher contamination with 90Sr, the daughter 90Y can be separated without waiting for equilibrium. Within 10 h the activity concentration of 90Y will be approximately 10% of the equilibrium value and may be sufficient for a reliable 90Sr analysis (IAEA-295, 1989). A special application of liquid scintillation counters is in the measurement of Cerenkov radiation produced by β-emitters with β-energies >260 keV. This application can be used for screening samples for 90Sr (Carmon and Dyer, 1986; Eakins et al., 1986; IAEA-118, 1970; Johns et al., 1979; Kleinberg and Smith, 1982; Krieger and Whittaker, 1980; Regan and Tyler, 1976; Szabo, 1982). An outline of the method used for determination of radiostrontium in various materials by nitrate precipitation is as follows. The ashed material is dissolved in nitric acid in the presence of strontium and barium carriers. The nitric acid concentration is then increased to precipitate all the strontium and barium (and part of the calcium) as nitrates. After further nitric acid separations, barium chromate and iron hydroxide scavenges are carried out. The subsequent treatment depends somewhat on the circumstances but the following is normal practice. Yttrium carrier is added to the purified strontium solution and, after a delay of about 14 days for the growth of 90Y, the yttrium is separated, mounted and counted. The storage period for the growth of 90Y can be reduced if sufficient 90Sr is known to be present, and the appropriate growth factor applied. For samples of very low activity, as well as for measurement of 89Sr, strontium is precipitated from the solution remaining after the removal of yttrium and mounted for counting. In many cases the determination of the natural inactive strontium content of the material is required so that the strontium chemical yield can be corrected. In the case of milk, direct application of the nitric acid separation to a solution of the ash usually gives low strontium yields. The calcium, strontium and barium are therefore concentrated by an initial phosphate precipitation. The mixed phosphates are then dissolved in an acid and the general procedure continued from that point. In the case of cereals, and vegetation generally, the ash is very variable in composition and contains numerous elements other than calcium: a mixture of hydrofluoric and perchloric acids is necessary to decompose and dissolve the ash. After heating to remove the hydrofluoric and most of the perchloric acid, the residue is dissolved in dilute acid and the alkaline earth precipitated as phosphates. For the details of procedure, see report IAEA (1989) and references therein.
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Here is procedure designed to separate 90Y from a partially purified Sr fraction isolated from various marine samples. It has been used successfully in soil, sediment and ashed biological samples up to 300 g in weight; and, in fresh water and seawater samples, up to 600 L in volume. 90Sr levels as low as 1 dpm have been measured in these samples. Several important facts should be kept in mind when planning 90Sr analysis on marine samples. Average seawater contains 8 ppm of stable Sr, or about 8 mg/kg, while average carbonate soil is about 1% (10,000 mg/kg) Sr. Thus even modest size samples require some thought in sample handling and about chemical yield measurements, especially if stable Sr is to be used for a yield monitor. This procedure has been developed to accommodate a Sr carrier from 20 mg to 4 g. For most routine analysis, however, 85Sr, a γ-emitter readily measured by γ-spectrometry and free from 90Sr contamination, is the preferred yield monitor. If stable strontium is used for chemical recovery, the strontium content of the original sample has to be determined. In average seawater the 226Ra content is about 0.2 dpm/L and 238U (or 234U) is about 2 dpm/L. The uranium usually presents no problem, but a good separation of the final 90Y from Ra and its daughters is important, especially for deep ocean water. During the separation steps, this procedure effectively removes all radionuclides that are chemically similar to yttrium or rare earths from the Sr fraction; and, after the establishment of 90Sr/90Y radioactive equilibrium, the final purification steps further remove any interfering radionuclides in the ingrowth 90Y fraction before measurement by β-counting. A great number of analytical methods has been developed and applied for the determination of 90Sr and 89Sr in environmental and nuclear samples using various measuring techniques like β-counting, liquid scintillation spectrometry and mass spectrometry. Most techniques use radiochemical procedures for the separation of strontium and/or yttrium including the classical procedure based on a series of semi-selective precipitations, the ion exchange, solvent extraction and extraction chromatographic procedures. All of this has been discussed in details in the review article by Vajda and Kim (2010).
4.3.6.3 Analysis of tritium Tritium is measured by LSC of a portion of a distilled sample. Several reagents (such as sodium sulphite and silver iodide) can be added in the distillation to prevent interference by radioiodine. The allowed concentration of tritium in water for human consumption is relatively high; thus the method presented here is normally adequate for routine determinations. However, if required, lower concentrations of tritium in water can be determined by electrolytic enrichment. The principles of the tritium determination procedure are as follows. The water sample is distilled to remove nonvolatile quenching materials and nonvolatile radioactive materials. Prior to distillation, sodium carbonate (Na2CO3) and sodium thiosulphate (Na2S2O3) are added to the sample. The majority of the constituents that might interfere remain in the residue together with any radioactive iodide and bicarbonate that might be present. If the tritium content of nonaqueous biological
4.3 Radiometric methods
samples is required, the sample can be converted into water by oxidation. The distillation is carried out to dryness to ensure complete transfer of the tritium to the distillate. An aliquot of the distillate is mixed with a scintillation solution in a counting vial. The mixture is cooled and counted in a liquid scintillation spectrometer (coincidence type). In this sample (usually an emulsion) the kinetic energy of the tritium β-particles is partly converted into light photons. When certain boundary conditions are satisfied (e.g. simultaneous detection by two or more photomultiplier tubes connected in coincidence) these photons are counted as pulses. Standard tritium and background samples are prepared and counted identically to minimize errors produced by ageing of the scintillation medium or instrumental drift. The counting rate is a measure of the tritium activity concentration, the sensitivity (counting time 100 min) is generally of the order of 20–200 Bq L1. Details of the method are described in IAEA-295 (1989). For additional reading, see Krieger and Whittaker (1980), Budnitz (1974), Bush (1968), Fox (1976), Horrocks (1974), IAEA-246 (1981), Konno and Suguro (1986), Lieberman (1984), Lieberman and Moghissi (1970), Momoshima et al. (1986) and Volchok and de Planque (1983).
4.3.6.4 Caesium analysis For many samples, the concentration of 137Cs is determined directly by γ-spectrometry. Only water samples and a small percentage of the soil, sediment and biota samples require preconcentration of 137Cs and measurement by β-counting and/or in some cases in low background well-type Ge detectors. The radiochemical procedure for the determination of 137Cs in aqueous samples is based on the batch extraction of caesium onto a microcrystalline cation exchanger, ammonium molybdophosphate (AMP), and subsequent purification from potassium and rubidium activities by ion-exchange separation using a strongly acidic cation exchange resin (BIO-REX-40). Natural K and Rb have radioactive isotopes that interfere with the β-counting of 137Cs. The purification of caesium is also necessary to determine the chemical recovery. γ-Spectrometry is used when samples contain both 134Cs and 137Cs activities. In theory, using the proper absorbers, 134Cs can also be resolved from 137Cs activity because of the difference in β-energy. However, the absorbers greatly reduce the counting efficiency, thereby eliminating any gain in the β-measurements of samples containing both 134Cs and 137Cs. The following are steps to be followed for sample preparation and preconcentration in the case of aqueous samples. If other radionuclides (Pu, Am and Sr) are also analysed in the same sample, the MnO2 preconcentration steps are generally performed first to remove the transuranics, followed by Cs extraction with AMP; and then, the Sr fraction is removed last as the oxalate. After the completion of MnO2 separation, the water sample is transferred to the proper size polyethylene processing container. Adjust to pH 1–4 with nitric acid, add AMP as a slurry in water to extract the caesium (use 0.2 g AMP/L of sample), stir the sample thoroughly and let the AMP settle, filter or decant the supernatant (discard the
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CHAPTER 4 Measurements of radioactivity
supernatant or save for Sr analysis, if required), separate the AMP by centrifugation and purify the Cs for β- or γ-counting. The steps to be followed in the sample preparation procedures used for soil, sediment or ashed biota are: dissolve or leach the sample with conc. HNO3 and separate the residual materials. The acid sample is diluted with water and adjusted to pH 2–4 with NaOH. Add 1–2 of AMP to extract the Cs from solution. Use a minimum of 1 g AMP/L of sample. Larger amounts of AMP are required for soil, sediment or biota samples compared with seawater samples, because of the higher ionic strength in the acid-leached sample solution, which reduces the extraction efficiency of AMP for Cs. Separate the AMP as described earlier for the water sample and purify as described further. The steps required for 137Cs purification are described next. The amount of AMP collected from the preconcentration step, especially from large water samples, is invariably >1 g. The following steps reduce the amount of AMP to about 1 g in order to perform the ion-exchange procedure for the separation of Cs from K and Rb. 1.
2.
3.
4.
5.
6.
7.
Dissolve the AMP with a minimum amount of 10 M NaOH. Add an 0.5 mL excel of 10 M NaOH for each gram of AMP dissolved. Transfer the sample solution to a 400 mL beaker with a few millilitre of water, and heat the solution on a hot plate at a medium setting, without a cover glass, to decompose the AMP and to evaporate the ammonia. Periodically check the vapour phase (just above the hot beaker) with a wet pH paper or ammonium specific test paper to check for the presence of ammonia fumes. The sample solution must be kept strongly basic with NaOH for the AMP decomposition to be effective. When the vapour no longer shows the presence of ammonia, and the solution is strongly basic (pH >13+), stop the heating, cool the sample to room temperature, and add water to dissolve any salts that may be formed during cooling. Dilute the sample to about 200 mL with water, add 1–2 drops of methyl red indicator, adjust the pH to 1–4 with 8 N HNO3, add 1.0 g of AMP, let the AMP settle, decant the clear liquid, transfer the AMP slurry to a clean 50 mL C-tube with water, centrifuge and discard the supernatant. Dissolve the AMP with a minimum amount of 10 M NaOH and add 10 mL of 2% EDTA-07.5 NaOH solution. The sample solution should be clear. If any precipitate forms, centrifuge and decant the clear solution into another clean 50-mL C-tube. Discard the precipitate. Load a medium-size ion-exchange column with a goose-neck adapter (see Fig. 4.44) with 20 mL of BIO-REX-40 20–50 mesh cation exchange resin. Precondition the column with 100 mL of 3 N HCl followed by 150 mL of 5% NaCl solution and 50 mL of water. Using a Teflon-coated stirring rod, carefully pour the sample solution from Step 4 directly onto the top of the resin column. Try not to splatter any solution on the upper part of the reservoir. After the sample has drained to the top of the glass-wool plug, rinse the column walls and the reservoir three times with about 5–10 mL aliquots of water.
4.3 Radiometric methods
1.5 cm
15 cm
8 cm
5 cm cm
25 mL reservoir
250 mL reservoir
12 cm
Glass wool plugs top and bottom
1000 mL reservoir
20 cm
3 cm top resin 100–200 mesh
3 cm top resin 100–200 mesh
6 cm 50–100 mesh 1.0 cm
0.7 cm
21 cm 20–50 mesh Total resin bed = 24 cm
1.5 cm
21 cm 20–50 mesh Total resin bed = 24 cm
Goose-neck adapter for Cs analysis only
(A)
(B)
(C)
FIG. 4.44 Glass columns for anion-exchange separation.
8.
Rinse the column with another 40 mL of water (necessary to remove any dissolved AMP from the resin and preventing the possibility of AMP reforming in the column when the acid rinse is added in the next step). Total water rinses in Steps 7 and 8 are not critical but should not exceed about 60–70 mL. Discard the rinses. 9. Wash the column with 160 mL of 0.75 N HCl (to remove K and Rb). Discard the wash. 10. Elute the Cs with 125 mL of 3 N HCl and collect the sample in a 150 mL beaker. 11. Evaporate the caesium eluate to dryness and prepare the Cs for β-counting as described in the following section. Later Cs samples should be prepared for β-counting following the steps: 1. Dissolve the caesium salts from Step 11 with 1–2 drops of 8 N HNO3 and 2–3 mL of water. Transfer the solution to a 50 mL C-tube. Rinse the beaker twice with 2–3 mL of water. 2. Add 1 mL of 10 N NaOH, dilute the sample to about 10 mL with water, and add 2 mL of 0.12 chloroplatinic acid (H2PtCl6) to precipitate Cs2PtCl6. 3. Cool sample in a refrigerator or ice bath for 30–40 min. 4. Prepare a tarred glass-fibre filter paper. (a) Assemble a filtering apparatus with a 2.54 cm base. (b) Cut a 2.54 cm diameter glass-fibre filter paper disc. (c) With the vacuum off, centre a filter disc on the base of the filter holder, wet the filter with
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CHAPTER 4 Measurements of radioactivity
5. 6.
7. 8. 9.
water, apply vacuum, wash the filter with 2–3 mL of water, and 2–3 mL of acetone. (d) Dry the filter under a heat lamp. (e) Weigh the filter to 0.01 mg. Filter the sample through the tarred filter from Step 4. With the vacuum on, remove the filter chimney, wash the filter and the Cs2PtCl6 thoroughly with a few millilitre of cold acetone. Turn off vacuum, transfer the filter and precipitate to a petri dish. Dry the filter under a heat lamp and cool the filter to room temperature. Weigh the sample to constant weight ( 0.01 mg). Mount the filter containing the Cs2PtCl6 on the ring and disc holder as shown in Fig. 4.45. Count the sample using a low background β-detector or by γ-spectrometry.
4.3.6.5 Determination of actinides Actinides in the environment can be classified into two groups: (i) the uranium and thorium series of radionuclides in the natural environment and (ii) neptunium, plutonium, americium and curium which are formed in a nuclear reactor during the neutron bombardment of uranium through a series of neutron capture and radioactive decay reactions. Transuranics thus produced have been spread widely in the atmosphere, geosphere and aquatic environment on the earth, as a result of nuclear bomb Platinum electrode Wing nut
Anode Insulation tape
Stainless steel support ring Teflon top support
Glass chimney Constat current power supply Teflon washer Stainless steel polished disc, 2.45 cm Cathode
Stainless steel base and posts
FIG. 4.45 Electroplating cell.
4.3 Radiometric methods
Table 4.12 Nuclear data of actinide isotones Nuclide
α-Energy (MeV)
Yield (%)
Half-life (years)
244
5.81, 5.76 6.11, 6.07 5.28, 5.23 5.49, 5.44 4.90, 4.86 5.16, 5.12 5.16, 5.14, 5.10 5.50, 5.46 5.77, 5.72 4.79, 4.77 4.20, 4.15 4.40, 4.37 4.77, 4.72 5.32, 5.26 4.02, 3.96 4.69, 4.62 4.90, 4.85 5.42, 5.34
77, 74, 88, 86, 74, 76, 73, 71, 69, 51, 77, 57, 72, 69, 77, 76, 11, 73,
1.81 10 6.09 106 7.37 103 4.32 102 3.76 105 6.57 103 2.41 104 8.77 10 2.85 2.14 106 4.47 109 7.04 105 2.45 105 7.18 10 1.41 1010 8.03 104 7.34 103 1.91
Cm Cm 243 Am 241 Am 242 Pu 240 Pu 239 Pu 238 Pu 238 Pu 237 Np 238 U 235 U 234 U 232 U 232 Th 230 Th 229 Th 228 Th 242
23 26 11 13 26 24 15, 12 29 31 36 23 18 28 31 23 23 56 27
tests in the atmosphere, and accidental release from nuclear facilities (Sakanoue, 1987). Most of these radionuclide inventories have deposited in the northern hemisphere following the tests conducted by the United States and the Soviet Union. In actinide series, the elements of greatest interest as environmental contaminants are neptunium, plutonium, americium and curium, because their presence at relatively high concentrations in ecosystems would represent potential health problems (Katz et al., 1986). Nuclear data for actinide isotopes are presented in Table 4.12. As a result of nuclear bomb tests in the atmosphere, accidental release from nuclear facilities, and the accidental fall of artificial satellite SNAP 9A, plutonium isotopes have been spread widely in the earth’s atmosphere, geosphere and aquatic environment from the 1960s. They include 241Pu, 240Pu, 239Pu and 239Pu (Harley et al., 1973). These plutonium isotopes in the geosphere and the aquatic environment are incorporated metabolically into plants and ultimately find their way into man through the food chain; these radionuclides in the atmosphere are also incorporated into man directly by inhalation. From the standpoint of a dose assessment, these radionuclides are important because they are mostly α-emitting radionuclides and have half-lives of 13–2.4 104 years. From the standpoint of the safety assessment of the disposal of high-level radioactive waste, it is important to clarify the migration behaviour of plutonium in ground strata. Consequently, it is essential to know the plutonium concentrations in food and environmental samples for these studies.
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The commonly applied methods for determination of actinides in environmental samples may be classified as follows: 1. Preparation: (a) drying, (b) ashing and (c) scavenging. 2. Solubilization and equilibration: (a) fusion and (b) leaching. 3. Concentration and separation: (a) coprecipitation, (b) ion exchange and (c) solvent extraction. 4. Electrodeposition and α-spectrometry. For a description of a procedure for plutonium separation in large volumes of fresh and saline water by manganese dioxide coprecipitation, see Wong et al. (1975). Among ion-exchange separation methods for transuranics, strong base anion exchange in hydrochloric and nitric acids is important (Keder, 1962; Keder et al., 1960; Horwitz et al., 1990; Chu, 1971; Wong, 1971; Diamond et al., 1954; Korkisch, 1989). Among solvent-extraction reagents for transuranics, thenoyltrifluoroacetone (TTA) and trioctylamine (TOA) are important (Keder, 1962; Chieco et al., 1990). Each transuranic element has many valencies and their behaviour in aqueous solution is very complicated because of disproportionation reactions. As stated above, the ion-exchange and solvent-extraction behaviours of transuranics are dependent on their valency state. Therefore valency control is very important in their analysis (Katz et al., 1986). Additional references discussing this problem include Korkisch (1989), Diamond et al. (1954), Stevenson and Nervik (1961), Abuzwida et al. (1987), Bernabee et al. (1980), Budnitz (1973), Chu (1971), Fukai et al. (1976), Hampson and Tennant (1973), Hindman (1986), Holm et al. (1979), Irlweck and Veselsky (1975), Jiang et al. (1986), Johns (1975), Scott and Reynolds (1975) and Sekine et al. (1987). Procedures are described for the determination of plutonium and americium in environmental samples by anion exchange (HNO3). Procedures are also described for the determination of plutonium, americium and their sequential analysis by anion exchange (HNO3) and TOA extraction (Chieco et al., 1990). Livens and Singleton (1989) developed the method for the determination of Am in environmental samples (Fig. 4.46). Yamamoto et al. (1989) developed the method for the determination of Np in environmental samples (Fig. 4.47). There follows a description of plutonium separation by anion exchange. Plutonium as well as several other heavy elements (e.g. U, Th, Np, Am, Cm) may be separated and purified by anion-exchange chromatography. If only 239+240Pu activity is to be determined, a single anion-exchange column separation, done carefully, is
4.3 Radiometric methods
Sample leachate (HCl)
Fe solvent extraction (diisopropyl ether)
Anion exchange (HCl)
Pu fraction
Cation exchange (HCl)
BiPO4 or Fe(OH)3 precipitation Anion exchange (HNO3 – CH3OH)
Anion exchange (HNO3)
Am fraction
FIG. 4.46 Flowsheet for Pu/Am method. After Livens, F.R., Singleton, D.L., Evaluation of methods for the radiometric measurement of Americium-241 in environmental samples. Analyst 114 (1989) 1097.
usually sufficient. However, if the 238Pu activity is also required, then a second small column separation will be necessary to eliminate any trace of interfering activities from the naturally occurring elements, Ra and Th, present in all soil and sediment samples. The steps to be followed in plutonium purification are •
• •
•
Prepare anion-exchange columns as shown in Fig. 4.44. Use the medium-size column for sample volumes <500 mL or the large size column for samples with final volumes >500 mL. Precondition the columns by passing about 20 mL of 8 M HNO3 through each column. Check the sample solution for particulate material. If insoluble silicate (white, gelatinous materials) or other particles are noted, centrifuge or filter the solution through a glass-fibre filter before loading the sample onto the preconditioned column. This step is important in order to maintain a continuous flow of the sample solution through the column. Load the sample solution onto the column as fast as the column will allow. The plutonium is retained on the ion-exchange column. If insoluble material has been removed from the sample and there is no outgassing in the column, the average flow rate in the medium-size column is about 3–6 mL per minute, depending on the viscosity of the sample solution.
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CHAPTER 4 Measurements of radioactivity
Soil or sediment sample 239
Heat
Res.
Np tracer Aqua-regia
Sup. (Pa, U)
Sup.
Evaporate to dryness 10 M HCl Isopropyl ether extraction
ppt. (LaF3) 8M HNO3 saturated with Al(NO3)3 Anion exchange
Feed solution 8 M HNO3 10 M HCl 10 M HCl-0.1 MHI 4 M HCl
Dewox 1´8, 100–200 mesh (U) (Th) (Pu) (Np)
Evaporate to dryness Aq.
Org. (Fe) 0.05 M HI (Heat)
10% TOA-xylene extraction (x3)
Anion exchange
Feed solution 4M CH3COOH (Np)
Aq. (Th, Pu, Am)
Org.
Electroplate from 2 M NH4Cl colution
10 M HCI wash Plated Np
1 M HCl-0. 1 M HF strip Continued to Pu and Am separation procedure
NH2OH x HCl Conc. H2SO4 La carrier
r-Ray spectrometry α-Ray spectrometry
FIG. 4.47 Separation scheme of neptunium for environmental soil or sediment sample. After Yamamoto, M., Chatani, K., Komura, K., Ueno, K., Development of alpha-ray spectrometric techniques for the measurement of low-level 237Np in environmental soil and sediment. Radiochim. Acta 47 (1989) 63.
•
• •
• • •
After the sample has drained to the top glass-wool plug in the column, rinse the wall of the column reservoir thoroughly with 8 M HNO3. This can be done conveniently with a polyethylene wash bottle. Do this at least three times with about 5–10 mL of 8 M HNO3 and allow each rinse to drain to the top glasswool plug. After the last rinse, wash the column-volumes of 8 M HNO3 and discard the rinse (about 150 mL for the medium column or 350 mL for the large column). After the 8 M HNO3 rinse, wash the wall of the column reservoir with about 5–10 mL of conc. HCl. Repeat the conc. HCl wash 3 times, each time draining the liquid to the top of the glass-wool plug. Now wash the column-volumes of conc. HCl (about 150 mL for med. Column or 350 mL for the large column). Discard HCl rinse. Elute the Pu with five column-volumes of NH4I-HCl solution. Use 100 mL of NH4I-HCl for the medium-size column. Collect the Pu eluent in a 150 mL beaker. Add approximately 5 mL of conc. HNO3 to the Pu eluent, mix, and evaporate to dryness on a hot plate. A small residue may be visible on the bottom of the beaker after the evaporation, which appears to have no significant effect on the analysis.
4.3 Radiometric methods
This material is normally seen and probably results from the decomposition products of strong acids with the resin. The following is a description of plutonium purification by anion-exchange separation. 1.
Cool the sample to room temperature. Rinse the wall of the beaker with 8–10 mL of conc. HCl. Add 2–3 drops of 30% H2O2, 3–5 drops of 1 M NaNO2, and heat on a hot plate for about 5 min. Again, cool the sample to room temperature. 2. Prepare a small sample with Dowex 1 8.50–100 mesh and precondition it with 1 mL of conc. HCl. Collect the eluent in a 50 mL C-tube. 3. Load the sample from the beaker onto the column. 4. Rinse the beaker two times with 2–3 mL of conc. HCl and load the rinse onto the column. 5. When the liquid has drained to the top glass-wool plug of the column, rinse the wall of the reservoir with 2–3 mL of conc. HCl. 6. Rinse the column 2 more times with 2–3 mL of conc. HCl. 7. Wash the column with 10 mL of conc. HCl. 8. Discard the rinse from Step 2 through Step 7. 9. Elute the Pu with 20 mL of NH4I-HCl solution. Collect the Pu in a 50-mL glass beaker. 10. Add 2–3 mL of conc. HNO3 and evaporate the solution to dryness on a hot plate. The steps to be followed in the implementation of the electrodeposition of plutonium procedure are: 1. 2. 3. 4. 5.
Assemble the plating cell. Fill the cell with water to test for leakage. Add 1 mL of conc. Sulphuric acid to the sample. Heat the sample on a hot plate until copious white fumes evolve. Cool the sample to room temperature, carefully rinse the wall of the beaker with 2 mL of 1 N H2SO4. Add two drops of 0.1% methyl red indicator. 6. Transfer the sample to the electroplating cell. 7. Rinse the beaker with 2 mL of 1 N H2SO4. Add the rinse to the plating cell. 8. Repeat Step 7 three times. 9. Add conc. NH4OH dropwise until the colour of the sample changes from red (pH 4.4) to yellow (pH 6.2). Mix the solution by swirling the plating cell. 10. Add 1 N H2SO4 dropwise to the red end point (pH 4.4) and then add two drops excess. 11. Complete the assembly of the electroplating cell by attaching the platinum anode with plastic insulation tape. Position the platinum wire anode about 0.5 cm from the stainless steel disc (cathode). 12. Connect the anode and cathode of the electroplating cell to a constant current power supply.
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13. 14.
15.
Electroplate at 1.0 A for 60–70 min. Before turning off the power supply when electrodeposition is completed, add 1 mL of conc. NH4OH to the cell and continue plating for about 1 min then turn off the power supply and as quickly as possible: (a) disconnect the cell from the power supply, (b) discard the solution from the cell, (c) rinse the cell with diluted NH4OH from a wash bottle. (Make the diluted NH4OH solution by adding 0.5 mL conc. NH4OH to 500 mL of water), (d) disassemble the plating cell, (e) rinse the plated disc with diluted NH4OH and (f) rinse the plated disc with acetone and let the disc air dry on a clean paper tissue. Count the plated disc and determine the activity of plutonium isotopes by α-spectrometry.
4.3.7 Rapid methods The need for rapid methods is apparent in accident and similar situations. An impulse to the development of rapid methods is also provided through the so-called coordinated research program of IAEA, Vienna. Here we present some of the methods reported in a Research coordination meeting on Rapid Methods held in Vienna in 1991. As an example of rapid methods, we mention here the work by Brodzinski and Perkins (1992). They have described a completely portable instrument, operable by one man, which is capable of quantifying the radioactive content of drums. Eleven radioisotopes are measured simultaneously in just a few minutes. The assayer uses two measuring techniques: segmented γ-ray spectrometry and neutron counting. A drum (or other container) to be assayed is placed on a rotating turntable by a selfcontained electric hoist. A collimated high-resolution germanium γ-ray spectrometer vertically scans the rotating drum to measure the intensity of γ-rays as a function of the energy emanating from the drum. Most fission and activation products and some transuranic radionuclides emit measurable quantities of monochromatic photons that serve as “fingerprints” of those radioisotopes. Comparison with emission rate from known standards provides a quantitative measure of radioactivity from each γ-ray emitter in the drum. This germanium spectrometer is used to measure the bremsstrahlung radiation from 90Sr. By manipulating the software with the on-board computer, the intensity of the 90Sr bremsstrahlung in the assayed drum is also compared to that of standards and the 90Sr concentration is quantified. The reported sensitivity for transuranic radionuclides is approximately 1 nCi/g, while that for γ-emitters is of the order of 0.1–1 pCi/g. Also, based on the bremsstrahlung radiation measurement, 90 Sr can be determined at concentrations of 100 pCi/g.
4.3 Radiometric methods
4.3.7.1 Rapid determination of transuranic elements and plutonium The rapid methods are based on fast removal of the transuranic elements from interfering materials so that they can be electrodeposited as a group, and measured by α-energy analysis. The procedure involves the following basic steps (Thomas, 1991). 1. 2.
The sample is first brought into solution. Radioisotope tracers, including 242Pu, 243Am and 234Th (if appropriate), are added. 3. A small amount of Fe carrier (10 mg) plus sodium sulphite is added to this solution, which is subsequently made basic by addition of ammonium hydroxide to allow the formation of an iron hydroxide precipitate. This precipitate serves to carry the thorium and the transuranic elements. 4. The mixture is centrifuged and the solution discarded: the precipitate is dissolved in dilute hydrochloric acid then diluted with water. It is then made basic with ammonium hydroxide, which results in a second iron hydroxide precipitate forming. 5. Following centrifuging and discarding of the solution, the precipitate is dissolved in dilute hydrochloric acid, diluted with water and a small amount of sodium sulphite added to maintain the transuranic elements in their lower valence states. 6. Small amounts of calcium and oxalic acid are then added and the pH is adjusted to approximately three to allow formation of an oxalate precipitate. The iron forms a very soluble oxalate, thus remaining in solution. This and two subsequent oxalate precipitations serve to remove any remaining iron. 7. The final oxalate precipitate is then dissolved in a small amount of sulphuric acid (0.5 mL of concentrated H2SO4), and the pH adjusted using dilute ammonium hydroxide. 8. The solution is then placed in an electrodeposition cell, where the transuranic elements are electrodeposited on a 1 cm2 area of a 2.5 cm diameter stainless steel disc. 9. Electrodeposition is conducted for a 1-h period at a current of 1 A. 10. Immediately before turning off the current, 1 mL of concentrated NH4OH is added to the cell and the electrodeposition continued for an additional minute. The current is then turned off, the solution discarded, and the electrode washed with water, then ethanol and air dried. Following α-energy analysis, the radiochemical yield, as determined from the radioisotope tracer content and the concentrations of the radionuclides of interest, are calculated. Samples with a large amount of iron such as soil extracts or vegetation ash may require partial removal of iron prior to initiation of this procedure (Thomas, 1991).
4.3.7.2 Rapid determination of
90
Sr
Up to now, two different approaches have been used successfully in fast radiochemical separation procedures for the determination of strontium-90 in environmental sample:
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• •
investigations dealing with the extraction of yttrium-90 and investigations which tailor precipitation methods for the strontium-90 separation to the needs of special sample types.
We mention only some of the work in these fields. Vajda et al. (1991) have described a simple and rapid method for the separation and successive determination of total radiostrontium in soil by using a crown ether. The method consists of three basic steps: oxalate precipitation to remove bulk potassium, chromatographic separation of strontium from most inactive and radioactive interferences utilizing crown ether, oxalate precipitation of strontium to evaluate the chemical yield. Radiostrontium is then determined by LSC of the dissolved precipitate. When 10-g samples of soil are used, the sensitivity of the method is about 10 Bq/kg. The chemical yield is about 80%. The separation and determination of radiostrontium can be carried out in about 8 h. Another method for 90Sr determination in food and environmental samples has been described by Shuzhong et al. (1991). It is based on the use of a tributyl phosphate for extraction of 90Y, the daughter of 90Sr. The method is shown to be sensitive to 0.2 Bq/kg of dry grass and milk powder and 2 Bq/kg of soil. In the case of a nuclear accident, most of the radioisotopes in the environment and food can be reliably and quickly assayed by γ-ray spectroscopy. There is a problem with some important isotopes which are pure β- or α-emitters and which cannot be identified directly by γ-ray spectroscopy. The activity of the isotopes of the strontium group 89Sr, 90Sr and 91Y after a 3-year reactor fuel cycle can reach about 8% of the total in-core activity and one of them, 90Sr(Y), is important for the long-term health consequences. It has been shown (Vapirev and Hristova, 1991) that the Ba/Sr reactor core ratio can be used for estimation of the upper limit of strontium activity in the fallout immediately after an accident. The Cs/Sr ratio can be used for estimation of the strontium fallout in the late postaccident period.
4.4 Nonradiometric methods There is a variety of situations in which it is better to determine the concentration of a radionuclide by a mass measurement rather than by measuring the activity present. This approach is possible using a wide range of instrumental methods of nonradiometric elemental analysis; analytical measurements can be performed also by element-specific chemical methods, some of which are extremely sensitive. The most important criterion for selecting an analytical method is whether the technique is sufficiently sensitive to measure the amount of radionuclide present in the sample. This is a very different problem when considered from the viewpoint of analytical chemists who use radiometric methods and those who use nonradiometric methods. Limits of detection in radiometric methods can be as low as 104 Bq, although 1 mBq is a more generally attainable detection limit. For nonradiometric methods, the detection limit is expressed in terms of mass and the
4.4 Nonradiometric methods
Table 4.13 The mass of 1 mBq for a selection of radionuclides with a variety of half-lives Radionuclide
Half-life (years)
Mass of 1 mBq (g)
232
1.4 1010 4.5 109 1.7 107 2.1 105 2.4 104 5.8 103 30
2.5 107 8.1 108 1.6 1010 1.6 1012 4.3 1013 6.1 1015 3.1 1016
Th U 129 I 99 Tc 239 Pu 14 C 137 Cs 238
After McMahon, A.W., 1992. An intercomparison of non-radiometric methods for the measurements of low levels of radionuclides. Appl. Radiat. Isot. 43, 289–303.
relationship between radiometric and nonradiometric limits of detection will depend on the half-life of the radionuclide of interest. Table 4.13 (from McMahon, 1992) shows the mass corresponding to 1 mBq for a number of important radionuclides. Although 1 mBq of 232Th is readily measured by a number of nonradiometric methods, 1 mBq of 137Cs could only be detected by the most sensitive of methods and is probably best determined radiometrically. Nonradiometric methods offer a variety of features and their use may be favoured for reasons other than improved sensitivity or isotopic selectivity. They can, in some instances, be used to perform analyses with less sample preparation and greater speed or sample throughput, and allow remote analysis or provide elemental or isotopic maps or depth profiles (McMahon, 1992). The instrumental methods of elemental analysis can be conveniently grouped as follows: (i)
Methods based on X-ray fluorescence analysis X-ray fluorescence analysis (XRF) Total reflection X-ray fluorescence analysis (TXRF) Electron microprobe analysis (EMA) Particle-induced X-ray emission (PIXE) Synchrotron radiation-induced X-ray emission (SRIXE) (ii) Methods based on ultraviolet or visible spectroscopy Atomic absorption spectroscopy (AAS) Graphite furnace AAS (GFAAS) Atomic fluorescence spectroscopy (AFS) Inductively coupled-plasma optical-emission spectroscopy (ICPO-ES) Glow-discharge optical-emission spectroscopy (GC-OES) Laser-excited resonance-ionization spectroscopy (LERIS) Laser-excited atomic fluorescence spectroscopy (LEAFS) Laser-induced-breakdown spectroscopy (LIBS) Laser-induced photoacoustic spectroscopy (LIPAS)
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Table 4.14 Analytical techniques classified by amount of isotopic information and amount of sample required Bulk samples No isotopic information Minor isotope determination Trace isotope determination
ICP-OES ICP-MS, GDMS, SSMS
Small samples profiling
Imaging and depth
XRF, GFAAS, LEAFS, TXRF ETV-ICP-MS
PIXE, SRIXE SIMS, SNMS, LMS, SIRIS
TIMS, RIMS, AMS
After McMahon, A.W., 1992. An intercomparison of non-radiometric methods for the measurements of low levels of radionuclides. Appl. Radiat. Isot. 43, 289–303.
Resonance-ionization spectroscopy (RIS) (iii) Methods based on mass spectrometry Spark-source mass spectrometry (SSMS) Glow-discharge mass spectrometry (GDMS) Inductively coupled-plasma mass spectrometry (ICP-MS) Electrothermal vaporization-ICP-MS (ETV-ICP-MS) Thermal-ionization mass spectrometry (TIMS) Accelerator mass spectrometry (AMS) Secondary-ion mass spectrometry (SIMS) Secondary neutral mass spectrometry (SNMS) Laser mass spectrometry (LMS) Resonance-ionization mass spectrometry (RIMS) Sputter-initiated resonance-ionization spectroscopy (SIRIS) Laser-ablation resonance-ionization spectroscopy (LARIS) McMahon (1992) has reported on intercomparison of nonradiometric methods for the measurement of low levels of radionuclides. He has classified the earlier analytical techniques according to the amount of isotopic information and the amount of sample required. The conclusions are presented in Table 4.14.
4.4.1 Methods based on X-ray spectrometry The electronic transitions which give rise to X-ray emission spectra involve core electrons and are therefore relatively insensitive to the chemical and physical form of the determinant (Bertin, 1978). As a result, analyses can be performed with a minimum of sample preparation directly on materials in the condensed phase. This insensitivity of sample matrix applies to the wavelength of the emitted X-rays, not to their intensities and as quantitation is based on intensity measurement, closely matched standards are required. X-ray emission spectra can be excited by primary X-rays in a fluorescence experiment or by changed particles via collisional excitation.
4.4 Nonradiometric methods
The cross sections for excitation of X-ray emission are rather low and this is combined with the low efficiency of collection, collimation, diffraction and detection of the emitted X-rays. This low overall efficiency leads to a relatively low sensitivity in some cases and is compounded by high backgrounds either from scattered primary radiation in a fluorescence experiment or due to bremsstrahlung in the charged-particle-excitation methods. Methods based on X-ray spectrometry do not provide isotopic information about the sample. Nonetheless, there are a number of radio analytical problems which can be solved by methods based on X-ray spectrometry. The following instrumental methods of elemental analysis are based on X-ray spectrometry: X-ray fluorescence analysis (XRF) Total reflectance X-ray fluorescence analysis (TXRF) Electron microprobe analysis (EMA) Particle-induced X-ray emission (PIXE) Synchrotron radiation-induced X-ray emission (SRIXE) XRF is the simplest of these methods. It allows bulk analysis of solid or liquid samples with detection limits of approximately 0.1 μg. The method can thus only compete with radiometric methods for the longest lived of radionuclides. It has approximately the same sensitivity for 232Th as α-spectrometry but has the advantage that little sample preparation is required and that analysis is rapid and easily automated. XRF would be the method of choice for measurement of airborne thorium collected onto filter papers, for example. The more sophisticated methods address the problem of the low overall efficiency of generation and acquisition of the X-ray spectrum. The low-fluorescence cross section is addressed by using a highly intense X-ray source, a synchrotron in the SRIXE method. The high intensity of synchrotron X-rays allows the beam to be focused and collimated while retaining significant intensity. The method can therefore be used in a microprobe mode and by moving the sample in a raster pattern across the incident X-ray beam, elemental images can be generated with micron spatial resolution. The scattered primary radiation background can be reduced by using the total reflectance technique in TXRF (Knoth and Schwenke, 1978). The instrumental geometry limits scattering of primary X-rays in the direction of the detector, however this is at the expense of increased sample preparation. The gains in sensitivity achieved by each of these methods may be compounded in a method which uses a total reflectance sample geometry in combination with a synchrotron X-ray source. The charged-particle-beam methods EMA and PIXE also allow elemental imaging within the sample. These methods generally require that the sample be enclosed in a vacuum. The approximately 15 keV electrons used in an EMA instrument, penetrate only 1–2 μm into the sample. This rapid declaration of the charged particles generates bremsstrahlung X-rays which generate a strong background signal in the spectral region of interest. EMA thus has relatively poor detection limits. The method can be used for analysis of electrodeposits such as sources prepared for
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α-particle spectrometry where the element of interest is present at high concentration in a very thin surface layer. The approximately 2.5-MeV-proton beam used in PIXE analysis penetrated much deeper into the sample than the EMA electron beam. The resulting proton bremsstrahlung is less intense and backgrounds are therefore reduced. PIXE can thus achieve much lower detection limits. PIXE ( Johansson and Campbell, 1988) and SRIXE ( Jones and Gordon, 1989) have similar imaging capabilities and detection limits but both suffer from the drawback that they rely on major pieces of hardware, an accelerator in the PIXE experiment and a synchrotron X-ray source for SRIXE.
4.4.2 Methods based on ultraviolet-visible spectroscopy Atomic spectroscopy in the ultraviolet-visible region involves transitions of valence shell electrons and the spectra are thus sensitive to the chemical and physical form of the element of interest. For sensitive quantitative work the sample is normally converted to free atoms in the gas phase. This can be achieved by vaporization from a furnace, by aspiration of a solution into a flame or inductively coupled plasma, by sputtering in a glow discharge or by laser ablation. Producing free, gas-phase atoms is a particular problem for thorium, uranium and plutonium as these elements react with traces of oxygen, even in a high vacuum system, to give oxides and dioxides. The following methods are based on types of atomic spectroscopy in the ultravioletvisible region: Atomic absorption spectroscopy (AAS) Graphic furnace AAS (GFAAS) Atomic fluorescence spectroscopy (AFS) Inductively coupled-plasma optical-emission spectroscopy (ICPOES) Glow-discharge optical-emission spectroscopy (GDEOS) Laser-excited atomic fluorescence spectroscopy (LEAFS) Laser-induced-breakdown spectroscopy (LIBS) Resonance-ionization spectroscopy (RIS) The methods range from simple, inexpensive absorption spectroscopy to sophisticated tunable-laser-excited fluorescence and ionization spectroscopies. AAS has been used routinely for uranium and thorium determinations (see for example Pollard et al., 1986). The technique is based on the measurement of absorption of light by the sample. The incident light is normally the emission spectrum of the element of interest, generated in a hollow-cathode lamp. For isotopes with a shorter half-life than 238U and 232Th, this requires construction of a hollow-cathode lamp with significant quantities of radioactive material. Measurement of technetium has been demonstrated in this way by Pollard et al. (1986). Lawrenz and Niemax (1989) have demonstrated that tunable lasers can be used to replace hollow-cathode lamps. This avoids the safety problems involved in the construction and use of active hollow-cathode lamps. Tunable semiconductor lasers were used as these are low-cost
4.4 Nonradiometric methods
devices. They do not, however, provide complete coverage of the spectral range useful for AAS and the method has, so far, only been demonstrated for a few elements, none of which were radionuclides. Absorption spectroscopy measures the difference in intensity between an incident and transmitted signal. Lower detection limits can be potentially obtained by monitoring a single low-intensity signal, as in emission or fluorescence spectroscopy. LEAFS uses tunable lasers to efficiently excite fluorescence and, by passing the sample atoms repeatedly through excitation–fluorescence cycles, very high sensitivities can be obtained. Again, LEAFS has been demonstrated for only a limited number of elements, none of which were radionuclides. A particularly sensitive approach is to excite fluorescence by a two photon process. In this way the wavelength of the fluorescent light is much shorter than that used to excite fluorescence and scattered primary radiation can easily be discriminated from the fluorescent signal. As an alternative to observing the fluorescent signal in a LEAFS experiment, the state which has been resonantly excited by a tunable laser can be further excited by further laser photons to produce an ion. Ions can be collected and detected with electron multipliers with high-efficiency leading to the extremely high sensitivity of ionization spectroscopies. LEAFS and RIS combine the high selectivity of laser spectroscopy with high sensitivities. Both these components are required to give low detection limits. Isotope effects are observable in high-resolution ultraviolet-visible spectroscopy. At very low atomic numbers and at very high atomic numbers (especially for the actinides) isotope shifts can be observed and coupling of the electron spin with the nuclear spin in odd-mass-number elements. Ultraviolet-visible spectroscopy can thus potentially provide isotopic information in these regions of the periodic table although routine methods are not yet available. For those elements where isotopes cannot be distinguished from their simple atomic spectra an isotope-specific resonance-ionization method has been suggested by Lethokov (1987). An instrument is being developed, based on this suggestion, for the determination of 90Sr in which strontium ions are accelerated to an energy of about 50 keV and neutralized collisionally. At this kinetic energy, the different strontium isotopes are travelling at sufficiently different velocities that collinear resonance-ionization spectroscopy can differentiate between isotopes on the basis of their different Doppler shifts (Monz et al., 1993). Thus some methods of atomic spectroscopy can provide isotopic information (Hurst and Payne, 1988).
4.4.2.1 Inductively coupled-plasma-optical emission spectrometry (ICP-OES) Atomic spectroscopy is widely used in inorganic chemistry to determine total element concentrations in many sample types, and generally allows rapid sample throughput. The optical techniques allow determination of atomic concentrations down to subnanogram/millilitre levels (108 M and below) in samples of a few
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CHAPTER 4 Measurements of radioactivity
millilitres or less. The recent introduction of a new mass spectrometric technique allows isotope-specific measurements to be made with the ease of use and sample throughput of the atomic spectroscopic techniques. The inductively coupled plasma (ICP) is stable argon plasma heated by inductive coupling of argon cations and free electrons, but is perhaps best thought of as simply a hot flame. Temperature measurement indicates that the plasma has a temperature approaching 6–7000 K. Samples in solution are nebulized (at about 0.4 mL min1 solution consumption) to produce an aerosol of fine droplets. A spray chamber is used to select only the smallest droplets for analysis, in practice those below approximately 5 μm. The selected droplets are swept into the centre of the plasma by an argon stream. In the plasma, droplets undergo rapid heating causing firstly desolvation of droplets, and then breakage of molecular bonds. The resulting free atoms are electronically excited; many are ionized. As atoms leave the plasma and cool, they relax leading to emission of light. Detection of this light is the basis of ICP-optical emission spectrometry (OES). The wavelengths emitted are characteristic of the elements present and the intensity proportional to their concentrations. ICP-OES limits of detection for many elements lie in the range 1–100 ng/mL (ppb) in solution. A few elements, notably Li, Be, Mg, Ca, Sc, Ti, Mn, Cu, Sr, Y and Ba, have limits <1 ng/mL. ICP-MS on the other hand generally is more sensitive (by 1–3 orders of magnitude) and gives isotope-specific information; ICP-OES to a first approximation only gives total element concentrations. The principal advantages of the technique are that it is multielement and that data acquisition takes approximately 1 min with a changeover time between samples of a similar order. The technique has drawbacks: spectral interference is possible, depending on other elements present; therefore in ICP-OES a high-resolution optical spectrometer may be required. The technique is best suited to solution, although direct solid sampling techniques are being developed.
4.4.2.2 Laser-excited resonance-ionization spectroscopy (LERIS) To achieve both high isotopic selectivity and high sensitivity at the same time, collinear laser spectroscopy is combined with resonance ionization. The principle of resonance-ionization spectroscopy is the following: the atoms are excited by one or several resonant optical excitation steps into an energetically high-lying state. Subsequently the atoms are ionized either by absorption of another photon, by collisions, or by field ionization. The photo-ions produced in this process can then be detected with high efficiency. This technique has proved to be extremely useful and sensitive in numerous applications. For the combination of collinear fast-beam laser spectroscopy with resonance-ionization detection, the excitation into highlying Rydberg states with subsequent field ionization is best suited because of the effective suppression of background. This technique has already been successfully applied for trace analysis of 3He in environmental samples as well as for the
4.4 Nonradiometric methods
Laser system
Ion source
Deflector Ion filter Separator magnet
Ion filter Neutralization Optical excitation
Energy selection
Ion detector
Field ionization
FIG. 4.48 Simplified scheme of the experimental set-up for detector of 89,90Sr by collinear fast-beam laser excitation and resonance-ionization detection.
sensitive study of radioactive Yb-isotopes at the online mass separator facility ISOLDE at CERN. Monz et al. (1993) have described the use of laser-excited resonance-ionization spectroscopy (LERIS) for low-level detection of 90Sr and 89Sr in environmental samples. The experimental set-up is shown schematically in Fig. 4.48. After chemical separation from the environmental sample, the Sr is inserted into the ion source. The ions are accelerated to an energy of 30 keV and pass through the mass separator, where the stable isotopes are strongly suppressed. The 89Sr or 90Sr ions enter the apparatus for resonance ionization in collinear geometry and are deflected by 10 degree to enable collinear superimposition of the laser beam. Neutralization takes place inside a charge exchange cell filled with caesium vapour. Remaining ions are removed afterwards from the resulting fast atomic beam by different electrostatic deflectors. Subsequently, the selective excitation into high-lying Rydberg states is induced by the laser light. The Rydberg atoms are field-ionized in a longitudinal electric field and the resulting ions are deflected out of the atomic beam for counting with a particle detector. A total isotope selectivity of 88Sr/90Sr > 1010 and an overall efficiency of 5 106 have been achieved. With these values, a detection limit of 1 108 atoms of 90Sr in the presence of >1017 atoms of stable isotopes has been demonstrated. The trace determination of such a contamination can be carried out with an accuracy of 30% within one working day including all chemical extraction steps. The chemical procedure of separation of strontium from air filters is carried out without the usual addition of strontium carrier to keep the content of stable strontium low. Such a chemical procedure has been worked out. Water and soil samples may have higher contents of stable strontium and thus require still higher values for the selectivity of the method.
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CHAPTER 4 Measurements of radioactivity
The performance of the technique might be increased by an optimization of the ion source efficiency and higher optical excitation probability affecting both the overall efficiency and the selectivity. These improvements should enable us to lower the detection limit for 90Sr and extend the measurements to 89Sr (see also Bolshov et al., 1988; Lethokov, 1987).
4.4.3 Methods based on mass spectrometry The electric and magnetic fields, used for the analysis of ions, provide only information about the two quantities E/q and M/q where E, M, and q are the energy, mass, and charge of the ion respectively. There are four ways in which the quantities E/q and M/ q may be determined: 1. 2. 3. 4.
magnetic selection, (Bρ)2 ¼ 2(M/q)(E/q) electrostatic selection, Eρ ¼ 2(E/q) cyclotron selection, 1/f ¼ (π/B)(M/q) velocity selection, v2 ¼ 2(E/q)(M/q)
where B is the magnetic field, ρ the radius of the ion path, E the electric field, f the cyclotron frequency and v the ion velocity. Low-resolution measurements that separate neighbouring isotopes as their final output have the quantities M/q and E/q. Since M can be regarded as an integer, ambiguities can arise if M and q have common factors. It is for this reason that some flexibility in the choice of q is desirable. If it is possible to measure the energy of the ion also, then it is possible to determine q and so determine the mass from the ratio M/q. The use of energy, mass and charge signatures, at energies such that charge state 3 + or higher is dominant, is the basis for the accelerator mass spectrometry (AMS) of almost all stable isotopes. The methods listed in the following are based on mass spectrometry, differing mainly in the design of the ion source used: Spark-source mass spectrometry (SSMS) Glow-discharge mass spectrometry (GDMS) Inductively coupled-plasma mass spectrometry (ICP-MS) Electrothermal vaporization-ICP-MS (ETV-ICP-MS) Thermal-ionization mass spectrometry (TIMS) Accelerator mass spectrometry (AMS) Secondary-ion mass spectrometry (SIMS) Secondary neutral mass spectrometry (SNMS) Laser mass spectrometry (LMS) Resonance-ionization mass spectrometry (RIMS) Sputter-initiated resonance-ionization spectroscopy (SIRIS) Laser-ablation resonance-ionization spectroscopy (LARIS) Let us briefly discuss again the limitations of radioactive decay measurement. The observation of the radioactive decay of a single atom is possible, consequently, with efficient apparatus for the detection of the decay particles and a radioactive species
4.4 Nonradiometric methods
with a half-life of seconds and minutes, it is possible to detect all or nearly all of a small number of radioactive atoms in the presence of a large number of nonradioactive atoms with radiation detection techniques. However, as the half-life increases, the time taken to carry out an experiment with a small number of radioactive atoms naturally increase, for half-lives, of say, 106 years efficient detection of the radioactive decay products becomes impossible unless the measurement can be continued for 106 years. Therefore, studies of long-lived radioactive isotopes invariably use very large numbers of atoms and the apparatus detects the decay of only a small fraction of the total during the experiment. In this situation the mass spectrometric detection sensitivity surpasses by far the sensitivity of radioactive counting methods. An important example is the study of 14C (half-life ¼ 5730 years), generated in the atmosphere by cosmic rays, in connection with radiocarbon dating. The observed β-ray counting rate from 1 g of contemporary carbon of biological origin is about 15/ min/g. However, this low counting rate is supported by the presence of 6.5 1010 atoms of 14C in the 1-g sample. If 14C atoms could be counted efficiently by accelerator mass spectrometry, it would be possible to determine the 14C content of very small quantities of carbon. This has now been accomplished for milligram carbon samples even though the ratio 14C/12C is near or below 1012. A comparison of measurements of long-lived radioisotopes at natural levels with β-ray counting and AMS (after Elmore and Phillips, 1987) is shown in Table 4.15. Atomic mass spectrometry is inherently sensitive and by its nature provides isotopic information. The goal of methods of elemental analysis based on mass spectrometry is to produce a spectrum of singly charged atomic species. Again this can be a problem for elements, such as uranium, which readily form oxides. If molecular or multiply charged species “contaminate” the atomic mass spectrum, they can give rise to background signals at the mass of interest, or if these molecular ions contain the element of interest, then the signal due to that element is distributed between the atomic singly charged ion, the multiply charged ions and any molecular species formed. Atomic mass spectra are simple to interpret, however great care must be taken to avoid molecular interferences especially at very low concentrations. SIMS analysis of electrodeposited 232Th α-particle sources gives rise to higher signals for ThO+ and ThO+2 than for Th+. This leads to difficulties in quantitation as the oxide to atomic ion ratios will be sensitive to local oxygen concentrations. Isobaric atomic interferences also present a problem: 99Ru and 99Tc, for example, have the same nominal mass and cannot be discriminated between on the basis of mass except by high-resolution mass spectrometers. Even with high mass resolution, if the interfering isobar is present in excess then discrimination at high mass resolution will be difficult and in any spectrometric method there is a trade-off between resolution and sensitivity. If there is a vast excess of an isotope of adjacent mass, even this may interfere with the signal of interest. The ability for a mass spectrometer to discriminate against such an interference is termed the “abundance sensitivity.” Methods such as ICP-MS and TIMS must discriminate against isobaric interferences by chemical separation methods prior to instrumental analysis.
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Table 4.15 A comparison of measurements of long-lived radioisotopes at natural levels with β-ray counting and AMS Radioisotope 10
Half-life (years) Stable isotopes Stable isobar Chemical form Sample size (mg) Atom per sample AMS run time (min) Decaycounting interval (years)
14
B
C
26
5730
7.05 10
9
13
12
10
14
1.6 10
6
Be
C N
36
Al 5
129
Cl
4.05 10
5
I
1.57 107
C
27
26
36
Ar AgCl
S AgI
Al
35
Cl
36
B BeO
C
Mg Al2O3
0.2
0.25
3
2
2
2 105
2 105
4 105
5 105
2 106
10
7
40
30
20
110
3
250
86
1130
After Elmore and Phillips (1987).
AMS, which is most commonly used for radiocarbon dating, discriminates against interferences in a number of ways. The 14N interference in 14C measurement is removed by generating a beam of anions and relying on the instability of the nitrogen anion. Molecular interferences are removed by high-energy (several MeV) collisions in a gas cell or thin foil. Further discrimination can be achieved by charge stripping in the same collisional processes to produce highly charged ions or even ions in their maximum charge state. The combination of discrimination processes used depends on the isotope of interest and the potential interferences. RIMS approaches the same problem by selectively ionizing only the element of interest prior to mass spectrometric separation. A selectivity of approximately 1 in 105 can be achieved per resonant excitation step in the ionization process and two or three such steps are frequently used. In combination with the selectivity of the mass spectrometer the method should potentially offer elemental selectivities in excess of 1015.
4.4.3.1 Inductively coupled-plasma mass spectrometry ICP-MS uses inductively coupled plasma as an ion source for a mass spectrometer. The basic units of an ICP-MS system, in the order used, are the sample introduction device, the plasma, the plasma/mass spectrometer interface, the ion focusing/ion filtering system, the detector and the data acquisition/data handling system (Fig. 4.49). Beauchemin (1992) gives a helpful comparison between ICP-MS and ICP-OES. The
4.4 Nonradiometric methods
Computer
Ion lenses Quadrupole
MCS
To vacuum
Plasma
Ion filtration
Interface
Detector
Data acquisition and handling
Sample introduction
FIG. 4.49 Component of a typical ICP-MS system.
sample introduction device introduces liquid samples as either a dry vapour or a fine mist into the plasma, with several options available. These include pneumatic nebulization (the most common), ultrasonic nebulization, electrothermal vaporization (ETV, which uses a graphite furnace), flow injection (Denoyer and Stroh, 1992) and direct injection (Wiederin et al., 1991). The transport efficiencies of sample into the plasma for pneumatic nebulization, ultrasonic nebulization and ETV are about 1%, and 100%, respectively. Laser ablation is a common method for introducing solid samples into the plasma. This and other sampling methods for solids are reviewed by Baumann (1992). Inductively coupled argon plasma is used most frequently in ICP-MS, with argon as the cooling, carrier and auxiliary gas. The high-temperature plasma (5000–8000 K) is sustained by radio frequency fields at the tip of a quartz torch. The plasma desolvates (if necessary), atomizes, and ionizes the sample. Horlick (1992) notes the use of helium or nitrogen-based microwave-induced plasmas to eliminate interference from argon-based background species. Smith et al. (1991) and Lam and McLaren (1990) have used mixed carrier gases to reduce argon-based background ions. The plasma/mass spectrometer interface allows the import of a stream of ions from the plasma, at atmospheric pressure, into the mass spectrometer, which is under vacuum. The interface has a sample cone and a skimmer cone, usually of nickel and typically with 1.0 mm and 0.8 mm apertures, respectively. The turbomolecular pump is presently the most widely used type in the vacuum system. The extraction lens, held at a negative voltage, attracts the positive ions as they emerge from the skimmer cone. The negative ions are repelled and the neutral species diffuse away. The accelerated positive ions are focused by the ion-lens stack and then enter the quadrupole region. Positioned between the extraction lens and the first ion lens, the photon stop prevents stray photons from reaching the detector. Voltages
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applied to the quadrupole, four metal rods mounted in a square array, produce an electric field that affects the trajectories. For any specific applied voltage, only ions of a very narrow range of mass/charge (M/Z) ratios have stable trajectories and reach the detector. A single M/Z ratio may be continuously focused onto the detector, or a selection of M/Z ratios can be sequentially focused onto the detector. A channel electron multiplier, the commonly used detector, responds to each incoming positive ion and produces a measurable pulse. The number of pulses measured is proportional to the number of ions of the selected M/Z ratio reaching the detector. In the pulse counting mode, maximum gain is obtained by applying a high voltage to the multiplier so that individual ion arrivals at the detector are recorded and ultimate detection limits are obtained. The analogue mode uses lower voltages, which produce lower gains. The use of lower gains extends the useful analytical range but results in higher detection limits. A Faraday cup replaces the analogue mode in some instruments. The data acquisition/data handling system consists of a multichannel scalar (MCS) and a computer system. Signal pulses from the detector accumulate into memory channels of the MCS according to their M/Z ratios. Signal pulses from replicate scans are sorted into the appropriate channels and accumulated until the analysis of that sample is complete. A computer program retrieves the totals from the MCS memory and stores the data for later manipulation or display. As mass spectrometry has continued to gain sensitivity and reliability, inductively coupled-plasma/mass spectrometry (ICP-MS) has become increasingly useful in the measurement of radionuclides. The optimization of ICP-MS is improving our ability to use the atomic detection of radionuclides in that it allows the near-complete isotopic analysis of any form of sample. Aqueous samples are generally introduced into the plasma source, and solids or individual particles, and organic solutions, may be atomized and continuously introduced into the plasma source. ICP-MS sensitivity, which is currently 8 109 atoms, can be improved by • • •
the use of more efficient sample introduction techniques, understanding of the basic principles of ion and gas dynamics in the ICP-MS interface and the use of high-resolution mass spectrometers with high-ion transmission.
The ultimate sensitivity could approach 107 atoms, which would result in a superior detection capability for all radionuclides with half-lives >1 year. For radionuclides with half-lives of thousands of years and longer, ICP-MS has two principal advantages over radiation counting, which are speed of measurement and sensitivity. Most radiation counting times range from 50 to 2500 min for most samples and most backgrounds. In contrast, an ICP-MS analysis requires only a few minutes per sample or blank, whether it is introduced via nebulizer, ETV unit or other device. The analysis time is independent of the half-life or decay scheme of the radionuclide. The analysis time is also not greatly lengthened by a lower required MDA. Indeed, this advantage of ICP-MS over radiation counting becomes greater with increasing half-life and decreasing MDA. Analysis by ICP-MS may be the preferred
4.4 Nonradiometric methods
method even when its sensitivity is not as great as the one obtained by radiation counting, because of its speed. Using ICP-MS quantitatively becomes feasible for radionuclides with half-lives greater than about 1 103 years. However, the sensitivity that is routinely achievable with ICP-MS is not as high as that achieved with radiation counting for radionuclides with half-lives less than about 1 104 years, unless the decay scheme is unfavourable for radiation counting. For present-day instruments, at least 107–108 atoms are necessary to qualify a nuclide by an ICP-MS with an ETV unit (Smith et al., 1992). To use an ultrasonic nebulizer requires at least 108–109 atoms; while a pneumatic nebulizer requires at least 109–1010 atoms. The corresponding masses vary according to the atomic weight of the nuclide. (For 239Pu, 2 107 atoms equals 8 fg). Under favourable operating conditions, an instrument with an ETV unit should just meet an 8 fg detection limit for 239Pu (t1/2 ¼ 2.41 104 years). This equals an MDA of 0.001 dpm (1.7 105 Bq), which is 5 times lower than the contractual MDA given earlier. This corresponding MDA for 2 107 atoms of 240Pu (t1/2 ¼ 6.57 103 years) is 0.004 dpm (6.7 105 Bq). The sensitivity of ICP-MS for heavier elements is better than for lighter ones, due to lower background in the higher mass region and more stable trajectories for more massive ions ( Jarvis et al., 1992). This is an advantage when one is interested in analysing for long-lived radionuclides of the rare earth and heavier elements. Essentially the entire inert sample matrix needs to be removed when performing radiation counting of all α-particle emitters and low-energy β-particle emitters because of sample self-absorption. Complete matrix removal may not be necessary for analyses by ICP-MS, depending on the elemental composition of the sample, the analytes, the sampling device and the required sensitivity. Partial or less-complex matrix decompositions and separations of analytes may suffice. For example, Hursthouse et al. (1992) compare the extent of chemical purification necessary to obtain good results for 237Np via ICP-MS, α-particle spectrometry and neutron activation analysis. Depending on the dissolved solids content, natural waters may need filtration only and/or treatment with acid. A chemical separation of a group of elements may be satisfactory. Preparation of purified samples for α-particle spectrometry is usually by either electrodeposition or micro-coprecipitation. Either technique takes at least an hour. Many β-particle emitters are precipitated with several milligrams of carrier and weighted for determining the chemical yield prior to counting, which also takes time. In contrast, a few millilitres of solution are satisfactory for ICP-MS. If the analyte concentration should exceed the linear part of the calibration curve, a simple dilution overcomes the problem. Mixtures of β-particle-emitting nuclides of more than one element usually have overlapping spectra. This is also often true for α-particle-emitting nuclides. Some of these mixtures can be analysed by ICP-MS without internal interferences. For example, the α-particle spectra of 237Np and 242Pu partially overlap even under the best of conditions, whereas ICP-MS is appropriate for analysing the long-lived radionuclides
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in a NpdPu mixture. Only X-and γ-ray counting are comparable to ICP-MS in the number of radionuclides that can be measured simultaneously. As with all analytical techniques, ICP-MS has its problem areas (Olesik, 1991). Although the mass spectrum of a sample is usually much simpler than an atomic emission spectrum for the same sample, spectral interferences from isobaric interferences and peak overlap can still be a problem. Isobaric interferences result from two situations. The first occurs when two elements in a sample have nearly identical M/Z values. An example of this is the presence of 113In+ interfering with the analysis of 113 Cd+. The second situation occurs due to the formation of “background species.” These are ions, usually polyatomic, formed from the plasma gas alone or in combination with elements from the solvent used in the sample preparation. Examples, together with the ions for which they cause the greatest interference, are Ar+2 (80Se+), ArO+ (56Fe+), Ar+ (40Ca+ and 40K+) and O+2 (32S+). Tables of common background species for an argon plasma are readily found in the literature. Most of these interfering species have m/Z < 81. Fortunately, most long-lived radionuclides have masses >81 amu. Peak overlap occurs when major constituents in the sample have such massive peaks at particular M/Z value channels. Examples of this type of interference would be a massive peak for uranium at mass 238 that tails into the 237 and 239 mass channels, thus complicating the analysis of 237Np+ or 239Pu+. Other problems may be caused by the matrix of the sample itself. If chloride is present, a series of polyatomic chloride-containing species may cause major interferences. As an example, 40Ar35Cl+ is an intense peak that interferes with 75As+ Arsenic is monoisotopic; therefore appreciable levels of chloride in the sample will seriously compromise the precise determination of arsenic. Components of the sample matrix may also contribute to oxide formation. Oxides of the form MO+ give rise to peaks at the (M/Z) + 16 position. One or more of these may interfere with nuclides of interest. An example of this is 48Ti16O+ interfering with the analysis of 64Zn+. The four other naturally occurring titanium isotopes would, of course, also give interference at their respective (M/Z) + 16 values to any analytes with these masses. Formation of the oxide of the analyte also reduces the signal measured at M/Z. The sample matrix may also induce changes in the analyte signal intensity. High concentrations of contaminant elements generally cause suppression of an analyte signal, although under certain conditions signal enhancement has been observed. In general, the lower mass elements are more subject to suppression than higher mass elements, and higher mass elements are more likely to cause signal suppression of lower mass elements than the reverse ( Jarvis et al., 1992). An ICP-MS instrument will not tolerate dissolved solids at concentrations that can be run with an ICP-atomic emission spectrometer. In addition to increasing the probability of interelement (isobaric) interferences and signal suppression, high levels of dissolved solids condense on the sample cone orifice. This deposition degrades the sensitivity and stability of the analytical signal. Typically, a maximum of 0.1% dissolved solids is recommended for continuous nebulization with a pneumatic nebulizer. Dissolved solids should be kept below about 0.01% with an
4.4 Nonradiometric methods
ultrasonic nebulizer, due to its desolvation effect. Liquid samples containing up to about 1% dissolved solids can be run with ETV and flow injection. The sensitivity achievable for an element is inversely related to its ionization potential (I.P.). Thus, for example, the ICP-MS sensitivity for iodine (I.P. ¼ 10.34 eV) is not as good as it is for nearby caesium (I.P. ¼ 3.89 eV). Finally, the sample aliquant is consumed during an ICP-MS measurement, whereas with radiation counting the sample aliquant usually can be retained and can be remeasured. Some of the reported ISP-MS applications to radionuclide determination are presented in the following. Kim et al. (1989a) measured the 240Pu/239Pu ratio in two soils and an estuary silt after performing radiochemistry. They also measured this ratio in these soils by the fission track-etch technique and found that ICP-MS gave better precision. They measured the 239Pu concentration on separate aliquants by α-spectrometry, with 3.76 105-year 242Pu tracer as the chemical yield monitor. However, at the 239Pu concentrations in their samples, ICP-MS could have measured these directly. Determination of low levels of 99Tc in environmental samples by ICP-MS was reported by Nicholson et al. (1991). In earlier work Nicholson et al. (1989) used ICP-MS with the β-particle-emitting nuclide 95Tc, in salt marsh soil, seaweed and seawater. They employed 95mTc as a yield tracer, and radiochemically isolated Tc from the matrix. They confirmed the chemical removal of any interfering isobaric 99 Ru by monitoring for other stable Ru isotopes. The chemical yield was measured by γ-ray spectrometry. Beals (1992) used 2.6-million year 97Tc as the yield monitor in ICP-MS measurements of 99Tc in river water, thereby eliminating the need for a separate yielding measurement. In purified water, they have reliably detected 0.05 ng/mL of very low specific-activity 113Cd by ICP-MS with pneumatic nebulization. For environmental waters, better sensitivities could be achieved by sample concentration and a clean-up that includes the removal of interfering 113In. It is entirely impractical to detect 0.05 ng of 113Cd by β-particle counting. James et al. (1989) demonstrated, with diluted aqueous standards, an MDA of approximately 8 fg of 239Pu (1.7 105 Bq) with ETV-ICP-MS in the peak dwelling mode. Comparison measurements by ICP-MS and α-spectrometry on radiochemically processed urine with moderately higher activity gave good agreement, considering the very low amount present. Plutonium-244 (t1/2 ¼ 8.3 107 years) is also amenable to analysis by ICP-MS, whereas 238Pu is not. ICP-MS was found to be compatible with LC for the trace metal speciation. The role of ICP-MS in trace element speciation studies at the FSL was described (Crews et al., 1987). The characteristics of LC-ICP-MS for the study of metalloprotein species were assessed and the chromatographic efficiency of ICP-MS was found to be similar to that obtained with a UV detector (Dean et al., 1987a, 1987b). Information about the chemical nature of trace elements from food can be obtained by first treating the foods in vitro with enzymes to broadly simulate the action of enzymes in the gastrointestinal tract (Crews et al., 1988). The soluble components can be separated by size-exclusion chromatography (SEC) and an estimate of their molecular size obtained. By coupling SEC directly to ICP-MS, the trace element content of the
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chromatographic fractions can also be measured. This approach has been used at the FSL to investigate the speciation of cadmium in raw and cooked pig kidney (Crews et al., 1989). The sensitivity of ICP-MS enabled the researchers to study retail samples in which the levels of multielement data obtained indicated that, while the feeds were contaminated with a number of elements, only lead presented a serious problem in parts of the rest of the food chain. For example, while samples from affected cattle were not allowed to enter the food chain, experiments with meat on contaminated bones showed that lead did not migrate significantly from the bone under a variety of cooking conditions (Baxter et al., 1992). Kim et al. (1991) have reported the measurement of some long-lived radionuclides, such as 99Tc, 226Ra, 232Th, 237Np, 238U, 239Pu and 240Pu using high-resolution inductively coupled-plasma mass spectrometry (HR-ICP-MS). By using HR-ICPMS with an ultrasonic nebulizer, the detection limits of these nuclides were 0.002–0.02 pg mL1 and the sensitivities were 10 times better than those obtained using HR-ICP-MS without the ultrasonic nebulizer. More accurate isotopic data were also obtained using HR-ICP-MS than with quadrupole ICP-MS at lower concentrations of the analyte because of improvement in counting statistics that can be obtained with HR-ICP-MS due to the greater efficiency of ion transmission. Morita et al. (1993) have applied ICP-MS to the determination of technetium-99 in environmental samples. The determination of eliminating the interfering element (Ru) before the ICP-MS measurements is made. Technetium-95m is used as the chemical recovery tracer. Compared with conventional methods, the method sensitivity is 10–100 times higher and the counting time is 300–10,000 times shorter. Tye and Mennie (1994) have reported development of the performance of a new interface for the Plasma Quad ICP-MS, which enhances the signal-to-noise performance of the standard instrument by a factor of 10. In order to maintain the flexibility of the instrument, the new interface is designed such that the enhanced performance can be easily switched on and off, offering the benefit of routine performance plus the high sensitivity mode when required. These improvements in signal to noise make it possible for the routine monitoring of many actinide elements directly by ICP-MS, potentially shortening the analytical cycle from days to hours. A further improvement on these impressive limits of detection is possible if a high-efficiency nebulizer is used to introduce samples into the new instrument, giving the capability of single figure ppq detection limits in an analysis which takes minutes, not hours.
4.4.3.2 Accelerator mass spectrometry For isotopes with long lifetimes (>1 year), it may often be more advantageous to use atom-counting techniques rather than traditional decay-counting methods. This is especially true for measurements where efficiency is a criterion, as for small samples, or if high precision is required. While atom counting has a counting rate that is essentially independent of decay lifetime and sample size, the decaycounting rates are comparable only if the isotopic half-life is less than 1 year for a sample size of the order of 1 mg. Of course, if sufficient material is available,
4.4 Nonradiometric methods
Table 4.16 Long-lived cosmogenic isotopes detected with accelerator mass spectrometry Interfering stable Isotope 10
Be C 26 A1 36 Cl 41 Ca 129 I
14
a
Half-life (years) 1.5 106 5.7 103 7.2 105 3.1 105 1.3 105 15.9 106
Isotopes
Isobars
9
Be
10
12, 13
14
C
27
Al
35,37 40,42 127
I
Cl Ca
B N 26 Mg 36 S 41 K 129 Xe
AMS detection limita
Range of terrestrial concentration a
7 1015 0.3 1015 10 1015 36 Ar 500 1015 100 1015
108–1014 1012–1016 1014 0.2 1015 1015–1016 1016
Compared to the stable isotope of the same element.
the decay-counting rate can always be improved by using more material (Litherland, 1987; Kilins et al., 1992). AMS extends the capabilities of atom counting using conventional mass spectrometry, by removing whole-mass molecular interferences without the need for a mass resolution very much better than the mass difference between the atom and its molecular isobar. This technique has been used with great success for the routine measurement of 14C, 10Be, 26Al, 36Cl and, recently, 129I (see Table 4.16). Analysis of 14C by AMS can, for example, generate dates with a precision that is at least equal to the best conventional β-particle-counting facility. In many cases, where small sample analysis is required, the AMS method has proved superior (Beukens, 1990). A complete description of AMS can be found in review articles (Litherland et al., 1987; Elmore and Philips, 1987) or recent conference publications. The application of AMS to 129I measurement has been discussed in detail in Kilins et al. (1992). AMS is an analytical technique that uses an ion accelerator and its beam transport system as an ultrasensitive mass spectrometer. AMS was introduced in Muller (1977), who suggested that a cyclotron could be used for detecting 14C, 10Be and other long-lived radioisotopes, and independently by the Rochester group, who demonstrated that 14C could be separated from the isobar 14N by relying on the instability of negative ion 14N. Presently AMS measurements are being made at about 30 accelerator laboratories around the world, and half of these are dedicated to AMS measurements of long-lived radioisotopes. Six long-lived radionuclides beyond uranium exist which have half-lives >100 ka (236Np, 237Np, 242Pu, 244Pu and 248Cm). The first two are natural byproducts of the nuclear industry. Nuclear weapons tests will generate the plutonium and curium isotopes although attempts have been made to detect presolar system 244 Pu in ores (Hoffman et al., 1971) or 244Pu from more recent supernova debris. The detection of these isotopes is still in the development stage. Unlike the natural elements, isobaric interferences are not a major problem as all isotopes will be
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CHAPTER 4 Measurements of radioactivity
Stripper AT +HV terminal
Negative ion source
Velocity selector
ES analyzer
Magnetic analyzer
Preaccelaration Injector
Tandem accelerator
Magnetic analyzers
Gas-filled ionization chamber Time-of-flight Gas-filled magnet
ES analyzer Detector system
Positive ion Analysis
FIG. 4.50 Components of a typical AMS system.
equally rare or nonexistent because of their very short decay half-lives compared to the lifetime of the solar system. The components of accelerator mass spectrometry (AMS) system are shown in Fig. 4.50 and they include ion source, injector, tandem accelerator, positive ion analysis and detection system. A caesium sputter ion source is used for most AMS work. This is essentially a secondary-ion mass spectrometry (SIMS) instrument that has been refined to produce high current of negative ions. Generally, solid samples are used; gas samples can give intense beams, but the problem of contamination from the previous sample (memory) is difficult to overcome. For radioisotope studies, sample sizes are 1–10 mg of processed material and beam currents of 1–50 μm are typical, depending on the element and ion source model. Some sort of multiple sample changing system is used at most AMS installations. For example, the main features of the 846B model high-intensity sputter source (High-Voltage Engineering Europa) include: a hemispherical ionizer giving a focused Cs beam spot of <0.5 mm, an x–y scanning stage to limit cratering effects and a 60-sample carousel with automated remote loading for throughput work. Currents of up to μA 12C have been quoted for this source from graphite targets. Mass analysis of the negative ion beam with a resolution sufficient to separate isotopes of heavy elements is needed prior to acceleration. For example, an electrostatic analyser is used at the University of Toronto to sharpen the energy distribution of ions produced from a caesium sputter ion source. A preacceleration of the negative
4.4 Nonradiometric methods
ion beam to 100–400 keV is used with large tandem accelerators to ensure that the injected ion beam is focused at the central terminal where the stripper canal is located. The name “tandem” refers to a dual acceleration design. The negative ions are accelerated to the terminal of the accelerator, which is held at a constant positive voltage, typically in the range 2–10 MV. The electron stripper at the terminal removes several electrons while energetic negative ions pass through; positive ions are then accelerated from the terminal to the end of the accelerator (ground potential). Tandetrons operate reliably below 3 MV using a solid-state power supply, and tandem Van de Graaff accelerator use a rotating belt or chain to charge the terminal up to 25 MV in some models. Tandem accelerators have the following characteristics: (i) the ion source and detector are located at ground potential for tandetrons; (ii) they do not require pulsed beam; (iii) the electron-stripping step need to eliminate molecules is an integral step in the operation of the accelerator and (iv) transmission through the accelerator and subsequent analysers can be made independent of small changes in the terminal voltage. Analysers positioned after the accelerator remove scattered particles accepted by the injector analyser, molecular fragments and unwanted charge states. Magnetic analysers alone are not sufficient. An electrostatic analyser or velocity selector is necessary to remove particles that have different mass but would otherwise have the correct mass-energy product to pass through the magnetic analysers. At 1 MeV/amu energies, the dE/dx and total energy measurements are made with either gas ionization detectors or silicon surface-barrier detectors or a combination of these. The time-of-flight detector serves as an additional positive ion mass analysis stage. It is most useful for the heaviest (slowest) ion such as 129I and consists of two time-pickoff detectors with time resolution of a few hundred picoseconds. Isotope ratios are obtained by alternately selecting each stable isotope and measuring its beam current in a removable or offset Faraday cup and then by measuring the radioisotope (rare nuclide) counting rate in the detector. Standards (samples with a known isotope ratio) are periodically measured for normalization, and blanks (samples containing no detectable nuclides to be measured) are used to measure background. Ratios are corrected for time-varying linear mass fractionation when more than one stable isotope is measured and for nonlinear fractionation, which arises from the stripping process and from stray magnetic fields in the accelerator, by comparison to the standard. The precision of ratios ranges from 1% to 10%, in the AMS measurement, depending on the value of ratios and counting time (if background is low enough). The long-lived radioisotopes 10Be, 14C, 26A1, 36C1, 41Ca and 129I can be measured now in small (mg) natural samples having isotopic abundances in the range 1012–1015 and as few as 105 atoms. At elevated energies (>1 MeV/amu), ions can pass through thin or equivalent gas with virtually no attenuation of the particle beam and little energy loss. As a result of electron capture and loss interaction, an ion passing through matter isPcharacterized by the fraction of the total ions (Fq) in a given charge state (q) where Fq ¼ 1. The resulting charge-state distribution is determined by the electron capture and loss
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CHAPTER 4 Measurements of radioactivity
cross section of an ion in a gas or solid. An equilibrium distribution will be established, the character of which depends only on the ion velocity and the target material. This equilibrium distribution is independent of the initial ionic charge or the target thickness, and approximation is valid as long as the energy loss remains insignificant. The passage of an ion through matter with the subsequent removal of electrons from molecules will decrease the bond strength among the constituents. Generally, after a reduction of two electrons in ionically bound molecules, no bond is possible and the molecule is broken up by Coulomb force. A sufficient number may remain to leave the charged molecule in a stable or metastable configuration. To avoid the possibility of long-lived (>1 ms) metastable molecules, at least three electrons must be removed. At present no 3 + molecules are known to exist. For light ions (Z < 20) an energy of at least 3 MeV is needed to maximize the production of charge state +3 ions in a gas cell. As the mean ionic charge rises approximately linearly with energy, the higher charge states will dominate at the 8 MV accelerating potentials. Usually one of these charge states was chosen to provide the highest conversion efficiency and no molecular interference. Separation of isobars can be accomplished by using different approaches: (i)
By chemical separation
No two isobars will have the same atomic number and rarely will they belong to the same chemical group of elements. Consequently, an initial reduction of the isobaric interference is always possible by some form of chemical processing. The degree to which this is effective depends on the specific chemical differences among the isobars and the level of isobaric contamination. Because the mass ambiguity is always formed from stable elements which have had the benefit of the last 4–5 109 years to achieve some level of contamination in all materials, chemistry cannot usually eliminate this type of ambiguity significantly below 1 ppb. One part in 106 (1 ppm) is the typical level of purity for most reagents. At this level of refinement, the interference is still at least many orders of magnitude greater in concentration than the rare long-lived radioisotopes. For some very rare isobars, specifically 36S (0.017%) in the case of 36Cl, careful chemical processing is the primary means of isobaric reduction. (ii)
Using negative ions
Some isobars can be eliminated by exploiting the instability of negative ions. For example, noble gas negative ions are known to be metastable or unstable, thereby removing 36Ar and 129Xe as a source of interference in the measurement of 36Cl and 129I. Others such as 14N and 26Mg do not form readily or have metastable states which decay in a time scale ≪ 1 ms. Since the lifetimes of these metastable states are small in comparison to the transit time (>1 ms) through the ion analysis system, attenuation factors in excess of 106 were possible with negative ions for the detection of 14C and 26Al. In contrast to the high probability for scattering and multiple charge changes for positive ions used at low energies (keV), the scattered negative ions from the more intense isobaric beams are greatly reduced in intensity after an interaction with the
4.4 Nonradiometric methods
residual gas in the ion source. This can be attributed to the low binding energy of the negative ions. (iii)
By fully stripping the electrons
At sufficiently high energies and for cases where the radioisotope has a higher Z than the stable isobar, separation by fully stripping with subsequent magnetic analysis can be accomplished. This method has been investigated for the systems 3He2+–3H1+, 26 Al13+–26Mg12+, 36Cl17+–36S16+, 41Ca20+–41K19+, 53Mn25+–53Cr24+ and 59Ni25 + 59 – Co24+. Isobar separation by fully stripping actually allows the simultaneous acceleration of the radioactive and stable isobar, with the latter of sufficient intensity to provide beam feedback signals. (iv)
By energy loss measurement
When ions with energies about 1 MeV/amu are passing through matter, the energy loss per unit path length (dE/dx) of an ion (Z) traversing a solid or gas with velocity (V) is governed by the Bethe–Block equation: dE/dx ¼ k (Z/V)2 where k is a constant. The effectiveness of energy loss measurements decreases as the atomic number increases. The percentage difference in energy loss between 14C and 14N is 30% at 40 MeV in isobutane gas. This rapidly decreases to 6.8% between 36Cl and 36S. Beyond calcium, very high energies (>100 MeV) are required. (v)
By gas-filled magnet
The gas-filled magnet is a powerful isobar separation instrument developed only recently for AMS. An ion passing through a gas changes its charge frequently by electron capture and loss. If this charge-changing occurs frequently enough in a magnetic field region, the trajectory is determined by the average charge state of the ion, which depends on Z. Development of the gas-filled magnet should make possible AMS of 36Cl at lower energies and aid in AMS detection of other heavier isotopes. (vi)
By lasers
Lasers are very powerful instruments to separate elements. Since the separation of isobars from different elements is the most difficult task in AMS, the use of lasers in connection with AMS could provide a very effective clean-up of background. The basic idea in a recent proof-of-principle experiment at the Rehovot AMS facility was to clean a negative ion beam from unwanted isobaric background ions by selective electrons detachment. 32S ions which have an electron affinity of 2.08 eV were effectively neutralized by interaction with 2.33 eV photons from a pulsed Nd: YAg laser. The same photons did not affect 37Cl ions whose electron affinity is 3.62 eV. This clearly demonstrated that a laser depletion of 36S background in 36 Cl measurements is feasible, opening up the possibility for sensitive 36Cl measurement at small AMS facilities where the ion energy is too low to perform isobar separation. However, for actual applications in AMS measurements, a substantial improvement in overall efficiency of the laser depletion process is necessary. The developments in instrumentation include production of dedicated machines. For example, with the introduction of the “Attomole 2000,” HVEE (High Voltage
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Engineering Europa B.V., P.O. Box 99, 3800 AB Amersfoort, The Netherlands) has made available a compact 14C isotope ratio mass spectrometer (14C IRMS) for biomedical applications. The system provides 14C/12C ratios down to 1013 from submilligram samples in typically a few minutes. Both solid samples (carbon) as well as CO2 can be analysed. The Attomole 2000 combines a compact instruments package with the extreme sensitivity of large tandem accelerators that are normally found in big research centres. The main application of the instrument will be for tracer kinetic and pharmacological measurements in biomedical studies. Essentially, 14C IRMS is an alternative technique to 13C Isotope Ratio Mass Spectrometry (13C IRMS), which is widely accepted within the biomedical community. 14C IRMS is at least 100–1000 times more sensitive than 13C IRMS. This allows, e.g., that the kinetics of 14C-labelled enzymes, aminoacids or carcinogens can be studied in the human body at levels that are comparable to actual environmental exposure. However, a widespread acceptance of 14C IRMS in the biomedical community is presently handicapped by the size of the existing 14C IRMS systems and their need for expensive support personnel. These shortcomings of the present instrumentation are overcome with the introduction of the Attomole 2000. The system is a compact, turnkey instrument that is user friendly. Its characteristics are reflected in the following specifications: System footprint Sample medium Required sample size Output Accuracy Counting rate Background Detection efficiency Detection limit Throughput
2.22 1.25 m2 Solid graphite or CO2 Solid graphite minimum 100 μg; CO2 sample 0.2 μmol CO2 (minimum 10 modern) 14 C/12C ratios Better than 3% > 250 cnts/s. for 10 modern samples < 1013 for 14C/12C ratios (0.1 modern) 1%–2% Approx. 4 attomole (1018 mol) 14C for 3% accuracy 400 sample/day (based on samples with an average of 4 modern 14 C content)
Furthermore the system is fully automated, self-tuning and needs little or no maintenance. The operator will consider the instrument as an analytic tool; the fact that an accelerator is involved is incidental. Up to 50 solid graphite samples can be loaded in a carousel prior to analysis. CO2 samples can be admitted on line to the ion source. The ion source uses a primary caesium beam to sputter the sample under investigation to form a negative carbon ion beam. The ion beam is accelerated through the system to reach the detector with an energy of 2.5 MeV. A detailed description of the system can be found in Mous et al. (1997).
4.4 Nonradiometric methods
4.4.4 Laser-induced photoacoustic spectroscopy Growing interest has been recently directed to the application of photoacoustic sensing techniques to the spectroscopic analysis of various optical absorbers in very dilute concentrations. For this purpose a laser is commonly used as a light source. Since the discovery of the photoacoustic effect by Bell, 1880, its application has a long history of development. Renewed interest in photoacoustics has emerged, starting with the work of Kreuzer (1971) who analysed trace amounts of gas molecules by laser-induced photoacoustic spectroscopy (LPAS) generation. The theory, instrumentation and application of laser-induced photoacoustic generation developed in recent years have been thoroughly reviewed by Patel and Tam (1981) and more recently by Tam (1983, 1986). Other reviews are also available in the literature from different authors: Pao (1977), Somoano (1978), Rosencwaig (1978), Colles et al. (1979), Kirkbright and Castleden (1980), Lyamshev and Sedov (1981), Kinney and Staley (1982), West et al. (1983) and Zharov (1985). Because of difficulties involved in handling radioactive preparations, the photoacoustic sensing technique had not been applied until some years ago to the spectroscopy of aqueous actinide ions. A relatively simple detection apparatus of photoacoustic spectroscopy for the spectral work of actinide ions using pulsed laser as a light source has been developed. This detection apparatus can be used for radioactive α-emitting aqueous samples without restriction to corrosive solutions and
Modulated light source, e.g., Laser pulse
Ion-specific absorption
Generation of heat by nonradiative relaxation
Modulated volume expansion
Generation and propagation of acoustic wave
Detection of compression wave by piezoelectric detector
FIG. 4.51 Generation and detection of photoacoustic signals.
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CHAPTER 4 Measurements of radioactivity
Radial intensity distribution
Sample cell
Temporal intensity distribution
Laser beam Halfwidth t P
Halfwidth R
Detector
FIG. 4.52 Geometry of laser beam and photoacoustic signal generation.
facilitates the spectroscopic investigation of actinide solutions, particularly transuranic ions, in very dilute concentrations. The creation of the photoacoustic signal in the solution is based on the conversion of absorbed optical radiation into heat by nonradiative relaxation processes. This principle is illustrated in Fig. 4.51. The geometry of laser beam and photoacoustic signal generation is schematically presented in Fig. 4.52. The spectroscopic system has been introduced to different nuclear chemical laboratories and further developed for a variety of purposes. Most of these developments are confined primarily to the spectroscopic investigation (i.e. speciation) of actinides in very dilute solutions or natural aquatic systems in which the solubility of actinides is, in general, very low (<106 mol L1). Optical spectroscopy of high sensitivity is an indispensable tool for the study of the chemical behaviour of actinides in natural aquatic systems, which has a newly developing research field in connection with nuclear waste disposal in the geosphere. For this reason, not only is photoacoustic spectroscopy attracting great attention but also thermal lensing spectroscopy and fluorescence spectroscopy, all using laser light sources, are in growing use for the same purpose. Actinides have particular spectroscopic properties which are characterized primarily by the f ! f transitions within the partially filled 5f shell and thus by a number of relatively weak, sharp absorption bands. The optical spectra of actinides are characteristic for their oxidation states, and to a lesser degree dependent on the chemical environment of the ion. Thus spectroscopic investigation provides information on the oxidation state of an actinide element and also serves to characterize the chemical states, such as hydrolysis products, various complexes and colloids. Hence, laserinduced photoacoustic spectroscopy with its high sensitivity can be conveniently used for the speciation of aqueous actinides in very dilute concentrations. For a summary of the present knowledge of laser-induced photoacoustic spectroscopy, as regards theoretical backgrounds, instrumentation and radiochemical applications to particular problems in aquatic actinide chemistry, see Kim et al. (1990). Since there is no other radiochemical application known in the literature,
4.5 QA/QC procedures
except the measurement of tritium decay by an acoustic sensing technique, the present discussion is limited to application to actinide chemistry, particularly in aquatic systems. The most interesting field of application is and will be the geochemical study of long-lived radionuclides, namely man-made elements (transuraniums). The main importance for such a study is not only the detection of a migrational quantity of radioactivity but also the characterization of their chemical states and hence their chemical behaviour in given aquifer systems. Knowledge of this kind will facilitate a better prediction of the environmental impact of transuranic elements which are being produced in ever-growing quantities and will be disposed of in the geosphere. Since LPAS application to actinide chemistry is in its infancy, only a limited number of works are available in the published literature. Experiments hitherto performed are confined to either hydrolysis, complexation reactions with carbonate, EDTA and humate ligands and a variety of speciation works for Am(III) and too much lesser extent for U(IV), U(VI); Np(IV), Np(V), Np(VI); Pu(IV) and Pu(VI). Of considerable interest is the LPAS application to the direct speciation of actinides in natural aquifer systems, where the solubility of actinides is in general very low and multicomponent constituent elements as well as compounds are in much higher concentrations than actinide solubilities. The study of the chemical behaviour of actinides in such natural systems requires a selective spectroscopic method of high sensitivity. LPAS is an invaluable method for this purpose but its application to the problem is only just beginning.
4.5 QA/QC procedures Quality assurance to determine radionuclides in food and environmental samples ensures that the quality of data obtained is maintained at an adequate confidence level, and is objectively evaluated. Quality assurance includes quality control, which involves all those actions by which the adequacy of equipment, instruments and procedures are assessed against established requirements. For the purpose of quality assurance, the following items must be ensured: (1) equipment and instruments function correctly, (2) procedures are correctly established and implemented, (3) analysis are correctly performed, (4) errors are limited, (5) records are correctly and promptly maintained, (6) the required accuracy of measurements is maintained and (7) systematic errors do not arise. In general, the design of a quality assurance program should take the following factors into account: a. quality of equipment and instruments, b. training and experience of personnel, c. verification of procedures by the routine analysis of control samples and the use of standard methods for analysis, d. frequency of calibration and maintenance of equipment and instruments (variability in the measuring system is an important aspect of this),
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e. the need for traceability of the results of determinations to a national standard and f. the degree of documentation needed to demonstrate that the required quality has been achieved and is maintained. It is important to have each item of the quality assurance program established. Intercomparison is also necessary to generally evaluate the quality assurance of the determinations. By this process, it is possible for data to be compared between laboratories or within a laboratory at different times. The concept of “quality control” should be discussed in general comparison with that of the concept of “quality assurance.” The basic concept of quality assurance is that quality should be assured comprehensively and wholly from the beginning to the end of a fixed volume of successive procedures. It assures that whole data acquired by using the fixed volume come to have a signification result to meet intended objectives. On the other hand, quality control is related only to definite and practical control of respective procedures, and is limited to only some portions of those procedures. Its main objective is being maintenance of quality of results within a specific limit. The scope of quality assurance is extended from one laboratory, to a group of laboratories in a region, then on to those in a country, and then to international groups of laboratories. The wider the scope of subjects to be assured the more effective the quality assurance. A smaller scope can be utilized as part of a larger one. This means that more effects are found in scopes of quality assurance in ascending order from a single laboratory to a region, a country, a continent, and to the whole world. The causes of errors which are treated as problems in quality assurance for radioactivity analysis and measurements are (1) collection of samples; (2) sampling; (3) transportation of samples; (4) labelling of samples; (5) storage of samples; (6) pretreatment of samples; (7) procedures for measurements; (8) measuring instruments; (9) human errors; (10) erroneous conversion; (11) reporting and notification; (12) environmental changes and (13) misinterpretation of data. The causes from items (1) collection of samples to (5) storage of samples are related to procedures to handle them. Quality assurance is established by comprehensively quantifying the partial uncertainty of errors which are to be generated from these causes. There are a few items for implementation: a. Organization must be implemented to aid in establishing quality assurance. First, a laboratory or research institute to play the central role has to be set up or appointed in a region or a country. It should act as a centre, play a leading role and handle the clerical work in that country. A committee may also be set up in the central laboratory for specialist members to offer guidance and advice. Ideally, a network should be formed to cover all the laboratories concerned, with the central laboratory acting as leader. In some smaller operations, a network will be composed only of laboratories who agree to join. The items to be executed for QA involve: • To form a network of quality assurance covering all laboratories concerned, with a central laboratory acting as leader. • For the central laboratory to check, compare and analyse the work of all the laboratories including itself.
4.5 QA/QC procedures
•
b.
c.
d.
e.
For the central laboratory and/or all the laboratories to conduct periodic calibration and stability checks of instruments. • For the central laboratory and other related laboratories to make comparative measurement and analyses either continually or periodically. • To carry out exercises related to the network. It is easier to maintain technological levels if instruments are subjected to periodic calibration and stability checks. These can be performed by the respective laboratories. Needless to say, checks should be done whenever operators are changed, instruments are installed, replaced, or moved, or environmental conditions are changed. Intercomparison and comparative measurements have to be conducted continually and periodically for assuring quality. Irregular and/or short-term checks never represent real assurance and have the least effect as a quality assurance system. Quality cannot be assured completely from the very beginning. Quality assurance will take time before it is refined. To improve the level of quality assurance, the following steps must be carried out: • unify the subjects to be sampled and the sampling methods; • standardize measuring procedures; • standardize specifications of measuring instruments; • have specialists to operate the measuring instruments; • standardize forms to make results accessible to all concerned and • understand regional characteristics. The procedures to keep up quality assurance levels must be incorporated into the quality assurance system by all laboratories. The list includes: • secure (a) operators with full expertise, (b) appropriate methods and (c) appropriate, well organized locations and space; • supply standard samples; • examine materials prior to application; • calibrate and adjust instruments; • make use of reference and standard samples which have appropriate records; • check quality assurance procedures; • effect continuous review of related data; • check whether objectives are met or not; • use divided samples; • compare data with those of other laboratories; • correctly handle requests; • review results in an organized manner and • correct errors, if any, by means of continuous measurements.
Quality control measures are necessary to provide documentation to show that the analytical results are reliable. This is very important since analytical results can form a basis on which economic, administrative, medical and/or legal decisions are made.
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It is essential to develop a quality assurance (QA) programme that covers sample collection, sample handling and methods for on-site and laboratory analysis, data handling and record keeping. The QA programme should address the variety of different scenarios likely to be encountered. Appropriate calibration and analytical standards and a variety of reference materials will be needed. To keep costs down, one should carefully design a QA programme that recognizes that for some signatures high-precision data are not required. If, for example, one analyses for a typical short-lived radionuclide which does not exist in nature, background measurements are unnecessary, however low the reported concentration. In other cases where one looks for faint anomalies in certain isotope ratios, the QA programme should demand knowledge of background values and their variability; this would be much more expensive. The protocols should include “blank” samples as well as “background” samples. In the case that an attempt is made to find an undeclared facility adjacent to a declared one, the analyst should try to take “background” samples from a plant somewhere else, which is similar to that part of the installation which is being examined. When attempting to find an undeclared nuclear facility at a declared site, the optimum background samples would be from similar facilities which are a part of the declared installation. Reliability of results is a function of precision (reproducibility) and accuracy (true value). The precision of results can easily be determined by internal measurement. The determination of accuracy in most cases, however, requires more detailed procedures such as the following: • • •
•
analysis by as many different methods, analysts and techniques as possible. In cases where agreement is good, the results are assumed to be accurate. control by as many different methods, analysts and techniques as possible. In cases where agreement is good, the results are assumed to be accurate. control analysis with reference materials that are as similar as possible to the materials to be analysed. Agreement between certified and observed values is then a direct measure of accuracy for that particular determination. Participation in an interlaboratory comparison. Samples used in such an intercomparison should be, as far as possible, similar in composition and concentration to the samples to be analysed on a routine basis. The agreement between the results received from a particular laboratory and the most probable mean value obtained from statistical evaluations of all the results will be a measure of the accuracy for that particular determination.
4.5.1 Intercomparison For practical reasons, most analytical laboratories are not in a position to check accuracy internally, without an external source of reference materials. To overcome some of the difficulties in checking the accuracy of analytical results, the IAEA provides the Analytical Quality Control Services (AQCS) programme to assist laboratories in
4.5 QA/QC procedures
Table 4.17 Analytical quality control services (AQCS) Year
Intercomparison
Reference available
Materials distributed
1986 1987 1988 1989 1990
24 24 33 27 19
39 38 46 50 58
1450 1680 2700 1800 1850
assessing the quality of their work. AQCS coordinates intercomparison studies and supplies reference materials. Participation is on a voluntary basis and at minimum cost. Information supplied by laboratories taking part in the intercomparisons is treated as confidential. The IAEA has traditionally played an important role in the development and testing of analytical methodology for the determination of radionuclides and through the AQCS programme provides a service by offering laboratories the option of determining their accuracy by distributing reference and intercomparison materials containing radionuclides in different types of materials. The analytes of interest in these samples include naturally occurring radionuclides and radionuclides of fission and activation products. The activities of the IAEA AQCS programme are shown in Table 4.17. Currently the orders for reference and intercomparison materials are running at the level of about 3000 units per year for the whole AQCS programme. The distribution of reference and intercomparison materials is coordinated by the Chemistry Unit of the Agency’s Laboratories at Seibersdorf, but it also receives input from other Sections of the IAEA, including the Hydrology Section, the Nutrition and Health Related Environmental Studies Section, the Safeguards Analytical Laboratory, Monaco. Intercomparison studies organized over the last 20 years are generally based on recommendations of consultants’ group meetings, and in response to the demands of many of the IAEA Member States for assistance in developing methodologies for the measurement of radioactivity. The Chemistry Unit distributes every 4 years a questionnaire concerning the need for organizing intercomparison tests and the preparation of reference materials. Using this data the AQCS programme collects different kinds of environmental and foodstuff bulk samples, some of which are affected by fallout radioactivity following the Chernobyl nuclear reactor accident. The general policy is to organize intercomparisons with those materials which are in most demand and have various levels of activity. Collection of a sufficient quantity of the raw materials (typically of the order of 2–400 kg) is first organized. The samples obtained by a sampling operation are generally dried, ground and homogenized. Aliquots are then taken at this stage and analysed to check the homogeneity of the bulk materials. Other preparation steps include aliquoting into bottles in amounts of about 25–100 g per bottle. To ensure long-term stability of the material, the sealed bottles are sterilized by γ-ray irradiation (Co-60 at a dose of 2.5 Mrad). A further control of
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homogeneity takes place after the materials have been distributed into bottles. Within-bottle and between bottle homogeneity is determined separately, usually by determining 40K, 137Cs, 90Sr and U. When this has been done, the material is announced in the AQCS Catalogue as an intercomparison material. Participants in such intercomparisons are provided with information about the material and special forms on which they are requested to report, for each element, up to six individual net results on a dry-weight basis, the sample weights used, information about the analytical method, and various other items. To preserve anonymity, each participant is assigned a code number, known only to him and the AQCS programme, by which he is identified in the report that is subsequently prepared on the results of the intercomparisons. The number of participants in each intercomparison varies but at present it is around 50. A chronological list of materials for intercomparisons which have been organized by the AQCS programme during a 9-year period is given in Table 4.18. Table 4.18 IAEA intercomparison exercises involving radionuclides during the 9 years: 1983–92 Matrix
Level
Alga, marine Milk powder Sediment, marine
Environmental Environmental Environmental
Soil Sediment marine Fish flesh
Environmental Environmental Environmental
Sediment, deep sea Sediment, like Air-filter, simulated
Environmental Environmental Artificial
Milk powder
Elevateda
Whey powder
Elevateda
Soil
Environmental
Sediment, stream
Environmental
Sediment, stream
Environmental
Milk powder
Environmental
IAEA code AG-B-1 A-14 SD-N-1/ 2 Soil-6 SD-N-2 MA-B-3/ RM SD-A-1 SL-2 IAEA083 IAEA152 IAEA154 IAEA312 IAEA313 IAEA314 IAEA321
Year
Certified as RM
1983 1983 1983
+ + +
1983 1983 1986
+ + +
1986 1986 1986
+ + +
1987
+
1987
+
1988
+
1988
+
1988
+
1988
+
4.5 QA/QC procedures
Table 4.18 IAEA intercomparison exercises involving radionuclides during the 9 years: 1983–92 Continued Matrix
Level
Clover
Elevateda
Seaweeds, Mediterranean Sediment, Baltic Sea
Elevateda
Sea-plant, Posidonia oceanica Uranium ore, phosphate
Elevateda Elevateda Environmental
Tuna homogenate, Mediterranean Sediment, Pacific Ocean
Elevateda
Soil
Elevateda
Grass
Elevateda
Cockle flesh
Environmental
Sediment, marine
Environmental
a
Natural
IAEA code IAEA156 IAEA308 IAEA306 IAEA307 IAEA364 IAEA352 IAEA368 IAEA375 IAEA373 IAEA134 IAEA135
Year
Certified as RM
1988
+
1988
+
1988
+
1988
+
1989
+
1989
+
1990
+
1991– 92 1991– 92 1992
1992
Contaminated with radioactive fallout from Chernobyl.
The results submitted by the participants are in all cases evaluated by the AQCS programme. A specific feature of any intercomparison is that gross errors occur quite frequently and results differing by as much as two or three orders of magnitude may be reported by participating laboratories. Various approaches and criteria for the detection and rejection of the highest and the lowest values or outliers have been discussed in the literature. The analytical data received in intercomparison exercises by the AQCS programme are treated using two different methods in order to derive a consensus value, which is considered to be a reliable estimate of the true value. Applying the first method, four different criteria, namely Dixon’s test, Grubbs’ test, the coefficient of Dewness test and the coefficient of kurtosis test are used at a significance level of α ¼ 0.05. If a laboratory mean for each element as single unweighted value was declared to be an outlier by any criterion, it is rejected and the whole procedure repeated until no more outliers could be identified. The remaining laboratory means are then combined in the usual way to provide estimates of the overall mean (consensus value) and its associated standard deviation, standard error and 95% confidence interval.
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The consensus values cannot automatically be accepted as recommended to certified values because their analytical validity usually requires a re-assessment in the light of additional analytical information such as concentration level, number of different analytical methods used, per cent of outliers and other criteria. In practice, certified or recommended values are always based on the following requirements: data should be available from a certain number of participants and two or more different analytical methods; there should be no significant differences between the groups of accepted results; outliers should not exceed 20%–30% of the submitted results. Depending on the extent to which the data satisfy such acceptance criteria, the consensus values are then assigned to one of the following conclusions: certified or recommended concentration, information value or not recommended. The Agency’s AQCS programme provides mainly four types of materials as follows: •
•
• •
Materials which can be used in analytical laboratories working in the fields of nuclear technology and isotope hydrology. These include uranium ore reference materials and other substances of interest for nuclear fuel technology as well as stable isotope reference materials for mass spectrometric determination of isotope ratios in natural waters. Materials with a known content of uranium, thorium and/or transuranium elements or fission products for the determination of environmental radioactivity or control of nuclear safety. Materials for use in the determination of stable trace elements in environment, biomedical and marine research. Materials which can be used in analytical laboratories working in the fields of monitoring organic microcontaminants in the marine environment.
Many countries practise national intercomparison programmes. For example, the Japanese nationwide intercomparison programme is based on the following: a. Comparison method Two methods of comparison, the “sample dividing method” and the “reference sample method,” were adopted for comparing the results of radionuclide analysis. b. Item for analysis and measurement method γ-Ray spectrometry is used. Participating laboratories are requested to determine artificial radionuclides as 40K, 54Mn, 59Fe, 60Co, 131I, 137Cs and 144 Ce, for the “reference sample method,” but as 40K and 137Cs for the “sample dividing method.” c. Samples and materials for intercomparison The environmental samples are soil, milk and crops. The reference samples are agar gel, alumina powder and liquid milk, which are all spiked with known radioisotopes.
4.5 QA/QC procedures
Table 4.19 IAEA reference materials for measurements of natural and fallout radioactivity in environmental and food samples Matrix
Analytes
Sediment, lake Sediment, stream
K-40, CS-137 Ra-226, Th, U
Soil Soil Bone, animal Clover Milk powder Milk powder Milk powder Whey powder Fish, flesh
Ra-226, Th, U Sr-90, Cs-137, Ra-226, Pu-239 Sr-90, Ra-226 K-40, Sr-90, CS-134, Cs-137 K-40, Sr-90, Cs-137 K-40, Sr-90, Cs-134, Cs.137 K-40, Sr-90, Cs-134, Cs-137 K-40, Sr-90, CS-134, Cs-137 K-40, Cs-137
Seaweeds, Mediterranean Sea-plant, posidonia oceanica Sediment, marine Sediment, Pacific Ocean Sediment, Pacific Ocean Tuna homogenate, Mediterranean Water, Pacific Ocean
K-40, Ru-106, Ag-110m, Cs-134, Cs-137, Pb210, Th-228, Pu-238, Pu-239 + 240, Am-241 K-40, Ru-106, Ag-110m, Cs-134, Cs-137, Ra226, Pu-238, Pu-230 + 240, Am-241
IAEA code SL-2 IAEA-313, IAEA, 314 IAEA-312 Soil 6 A-12 IAEA-156 A-14 IAEA-152 IAEA-154 IAEA-154 MA-B-3/ RM IAEA-308 IAEA-307
K-40, Cs-137, Th-232, Pu-239 + 240 Co-60, Sr-90, Cs-137, Pu-239 + 240
SD-N-2 IAEA-367
Co-60, Eu-155, Pb-210, Ra-226, Pu-238, U-238, Pu-239-240 K-40, Cs-137, Pb-210, Po-210
IAEA-368
Sr-90, Cs-137, Pb-210, Po-210
IAEA-352
IAEA-352
4.5.2 Reference materials All of the IAEA reference materials which are currently available have been certified on the basis of previously conducted intercomparison exercises. Natural matrix reference materials with certified values for the activities of various radionuclides are listed in Table 4.18. Some of the materials listed in Table 4.19 are the first “post-Chernobyl” natural matrix radionuclides reference materials that are internationally available. Including those reference materials available before Chernobyl, activities range for 137Cs from 0.8 (marine sediment, IAEA SD-N-2) to 3.7 kBq/kg (whey powder, IAEA-154).
251
252
CHAPTER 4 Measurements of radioactivity
Table 4.20 Reference materials of a similar matrix with different levels of analytes Analytes
Activity (Bq/kg)
Reference date
Matrix
Code
Ra-226
342
30.01.88
IAEA-313
732
30.01.88
1.5 3.3 7.7 1.79 72.6 2159
31.08.87 01.01.90 31.08.87 31.08.87 01.01.90 31.08.87
Stream sediment Stream sediment Milk powder Milk powder Milk powder Milk powder Milk powder Milk powder
Sr-90
Cs-137
IAEA-314 A-14 IAEA-321 IAEA-152 A-14 IAEA-321 IAEA-152
Ideally, there is a need for several reference materials which have a similar matrix type to the samples being analysed and which contain a concentration of the analyte representative of the whole working range that is of interest. Table 4.20 lists stream sediments and milk powder reference materials which reflect the fact that such materials have a different level of activity with practically the same matrix type. Reference materials for radioactivity measurements can also be obtained from the following specialized international or national organizations. 1. Central Bureau for Nuclear Measurements, Commission of the European Communities, Joint Research Centre, Geel, Belgium. 2. Office des Rayonnements Ionisants Commissariat a` l’Energie Atomique BP 21, 91910, Gif-Sur-Yvette, France. 3. Commission d’Etablissement des Methodes d’Analyse Commissariat a` l’Energie Atomique BP 6, 92265, Fontenay aux Roses, France. 4. AEA Fuel Services, Chemistry Division, Harwell Laboratory, Oxfordshire OX11 0EA, UK. 5. New Brunswick Laboratory, US Department of Energy 9800 South Cass Avenue, Argonne, IL 60439-4899, USA. 6. All Union Foreign Economic Association “Techsnabexport,” Staromonetniy Per. 26, 109180, Moscow, Russia. The IAEA AQCS Programme provides the following three main types of material: •
Materials that can be used in analytical laboratories working in the fields of nuclear technology and isotope hydrology. These include uranium ore
4.5 QA/QC procedures
•
•
reference materials and other substances relevant to nuclear fuel technology as well as stable isotope reference materials for mass spectrometric determination of isotope ratios in natural waters. Materials with known contents of uranium, thorium and/or transuranic elements or fission products for the determination of environmental radioactivity or control of nuclear safety. Materials for use in the determination of stable trace elements in environmental or biomedical research. Radiochemical methods such as neutron activation or isotope dilution analysis are often used in the determination of such trace elements and constitute an important contribution of nuclear techniques to applied science (Strachnov et al., 1993).
Table 4.20 lists the radionuclides referenced by IAEA, their activity, matrix and sample code. Table 4.21 includes also materials of marine origin (Ballestra et al., 1992). The intercomparison samples cover a range of materials and contain radionuclides with very different levels. IAEA intercalibration exercises are conducted with the involvement of many laboratories. As an example, Fig. 4.53 shows the results of an intercomparison run for 137 Cs determination in milk powder. Some laboratories had difficulties in determining the activity level. This situation is rapidly improving with time. Table 4.22 shows the improvements in the quality of the work at the participating laboratories. This is also seen in Table 4.23 where the mean values and relative standard deviations of three intercomparison runs for the 90Sr determination in simulated air filters are presented. The intercomparison exercises show a need for greater standardization of the analytical techniques used for radionuclide determination. This is indicated in McGee et al. (1992), where the bias and measurement errors in radioactivity data from four European radiation research laboratories were reported. Within the framework of the International Chernobyl Project, the IAEA’s Seibersdorf Laboratories organized an intercalibration exercise (Cooper et al., 1992) among some of the laboratories which were involved in assessing the environmental contamination in the former USSR caused by the accident. The objective was to assess the reliability of the radioanalytical data for food and environmental samples, which were used to assess the doses. The initial study reference materials from the stocks of the IAEA’s Analytical Quality Control Service (AQCS) were relabelled and submitted to 71 laboratories as blind samples in June and July of 1990. These natural matrix materials included samples of milk (containing two different levels of radioactivity), soil, air filters and clover. The concentrations of radionuclides (137Cs, 134Cs, 40K, 90Sr, 239Pu, 226Ra, 60Co, 133Ba and 210Pb) in these samples were known from previous intercalibration exercises. The overall range in performance was broad, which is as observed in previous international intercomparisons. This is illustrated in Table 4.24 where the results
253
254
Table 4.21 Radionuclides referenced by IAEA (Bq/kg)a
3
H C
14
60 40
Co K
Activity or conc.
Confidence (Bq/ kg)
37.2 15.03 pMC 0.12 pMC 2.50 391
0.5
0.17 pMC 0.21 pMC 2.40–2.60 379–405
527 150 220 240 272 539 552 657 1381 1575 103 11.4 607 6.8 1188 1080 72.1 660 373 481 234
510–543 141–161 189–226 211–269 252–299 510–574 563–569 637–676 1320–1456 1511–1644
0.4 604–612 6.5–7.1
30
56
3.2 626–671 360–380 470–486
12
Matrix
Reference date
Sample code
Irish Seawater Oxalic acid Wood Fangataufa Sediment Tuna homogenate, Mediterranean Milk powder Sea-plant, Posidonia oceanica Sediment, marine Sediment, lake Fish flesh Milk powder Milk powder Clover Seaweeds, Mediterranean Hay powder Irish Seawater Irish Sea Sediment Fangataufa Sediment Spinach Grass Brown rice Baltic Sea Seaweed Mussel from Mediterranean Sea Fish Soil from oil field
1 1 1 1 1
IAEA-443 IAEA-C8 IAEA-C9 IAEA-384 IAEA-352
Jan 2007 Jan 1998 Jan 2004 Aug 1996 Jan 1989
31 Aug 1987 1 Jan 1988 1 Jan 1985 31 Jan 1986 1 Jan 1986 31 Aug 1987 1 Jan 1990 1 Aug 1986 1 Jan 1988 31 Aug 1987 1 Jan 2007 1 Jan 1996 1 Aug 1996 15 Oct 2007 1 Jun 2006 1 Jan 2015 1 Aug 2008 1 Nov 2003 1 Jan 1997 1 Jan 2009
A-14 IAEA-307 SD-N-2 SL2 MA-B-3/RN IAEA-152 IAEA-321 IAEA-156 IAEA-308 IAEA-154 IAEA-443 IAEA-385 IAEA-384 IAEA-330 IAEA-372 IAEA-464 IAEA-446 IAEA-437 IAEA-414 IAEA-448
CHAPTER 4 Measurements of radioactivity
Ref. analyte
89 90
Sr Sr
106
Ru
110m
134
Ag
77 1.33–1.57 3.16–3.44 6.0–8.0 7.0–8.3 13.4–16.3 24.2–31.67 46.3–59.2
5
2.1
0.005 22–25 30.0–36.5 1–2.27 4.8–5.5 1.5–1.8 1.5–1.9 14.8–16.2 126–138 722–802 1295–14 17
0.4
7
Milk powder Milk powder Milk powder Hay powder Milk powder Clover Soil Bone, animal Milk powder Spinach Irish Seawater Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Milk powder Clover Milk powder Whey powder Brown Rice Milk powder
1 Nov 2014 31 Aug 1987 1 Jan 1990 31 Aug 1987 31 Aug 1987 1 Aug 1986 30 Jan 1983 15 Dec 1981 1 Nov 2014 15 Oct 2007 1 Jan 2007 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1990 1 Aug 1986 31 Aug 1987 31 Aug 1987 1 Jan 2015 1 Nov 2014
IAEA-473 A-14 IAEA-321 IAEA-154 IAEA-152 IAEA-156 SOIL-6 A-12 IAEA-473 IAEA-330 IAEA-443 IAEA-308 IAEA-307 IAEA-308 IAEA-307 IAEA-308 IAEA-307 IAEA-321 IAEA-156 IAEA-152 IAEA-154 IAEA-464 IAEA-473 Continued
4.5 QA/QC procedures
Cs
2405 1.5 3.3 6.9 7.7 14.8 30.34 54.8 209 20.1 0.110 23 33.5 20 5.1 1.6 1.6 15.5 132 764 1355 12.0 357
255
256
Ref. analyte 137
155
Activity or conc.
Confidence (Bq/ kg)
Cs
2.7
2.5–2.8
Eu
0.8 1.79 2.4 4.9 5.6 14.2 53.65 72.6 264 2159 3749 11,320 38.6 0.36 33.0 1235 18.8 5.18 425 224 7.0
0.5–1.0 1.62–1.97 22–2.6 4.5–5.2 5.3–6.0 13.7–15.3 51.43–57.91 71.1–74.2 254–274 2503–22 09 3613–38 87
360
0.8
0.01 32.7–33.6
35 18.2–19.2 5.12–5.22
10
5 6.7–7.3
Matrix
Reference date
Sample code
Tuna homogenate, Mediterranean Sediment, marine Milk powder Sediment lake Sea-plant, Posidonia oceanica Seaweeds, Mediterranean Fish flesh Soil Milk powder Clover Milk powder Hay powder Grass Brown Rice Irish Seawater Irish Sea Sediment Spinach Baltic Sea Seaweed Fish Moss-soil Milk powder Fangataufa Sediment
1 Jan 1989
IAEA-352
1 Jan 1985 31 Aug 1987 31 Aug 1986 1 Jan 1988 1 Jan 1988 1 Jan 1986 30 Jan 1983 1 Jan 1990 1 Aug 1986 31 Aug 1987 31 Aug 1987 1 Jun 2006 1 Jan 2015 1 Jan 2007 1 Jan 1996 15 Oct 2007 1 Aug 2006 1 Jan 1997 15 Nov 2009 1 Nov 2014 1 Aug 1996
SD-N-2 A-14 SL-2 IAEA-307 IAEA-308 MA-B-3/RN SOIL-6 IAEA-321 IAEA-156 IAEA-152 IAEA-154 IAEA-372 IAEA-464 IAEA-443 IAEA-385 IAEA-330 IAEA-446 IAEA-414 IAEA-447 IAEA-473 IAEA-384
CHAPTER 4 Measurements of radioactivity
Table 4.21 Radionuclides referenced by IAEA (Bq/kg)a Continued
210
0.6
0.36–1.0
210
Po
73 680 0.42 103 2.2
66–75
58
0.02 103 1.7–27
226
Ra
228
Ra
228
Th Th
423 3.1 5.2 79.92 269 342 732 21.9 19.05 103 25.1 780 32.0 25 2.50 211 30.6
10 21–4.4 4.4–6.7 69.56–93-43 250–287 307–379 678–787 21.6–22.4
0.26 103
2.0
62 31.3–32.5 2.2–3.6 2.38–2.61
9 30.0–33.6
230
Tuna homogenate, Mediterranean Seaweeds, Mediterranean Phosphogypsum Moss-soil Tuna homogenate, Mediterranean Moss-soil Sea-plant, Posidonia oceanica Bone, animal Soil Soil Sediment, stream Sediment, stream Irish Sea Sediment Soil from oil field Moss-soil Phosphogypsum Irish Sea Sediment Seaweeds, Mediterranean Fangataufa Sediment Phosphogypsum Irish Sea Sediment
1 Jan 1989
IAEA-352
1 Jan 1988 1 Jan 2008 15 Nov 2009 1 Jan 1989
IAEA-308 IAEA-434 IAEA-447 IAEA-352
15 Nov 2009 1 Jan 1988 15 Dec 1981 30 Jan 1983 30 Jan 1988 30 Jan 1988 30 Jan 1988 1 Jan 1996 1 Jan 2009 15 Nov 2009 1 Jan 2008 1 Jan 1996 1 Jan 1988 1 Aug 1996 1 Jan 2008 1 Jan 1996
IAEA-447 IAEA-307 A-12 SOIL-6 IAEA-312 IAEA-313 IAEA-314 IAEA-385 IAEA-448 IAEA-447 IAEA-434 IAEA-385 IAEA-308 IAEA-384 IAEA-434 IAEA-385 Continued
4.5 QA/QC procedures
Pb
257
258
Ref. analyte 232
Th
234
U
235
U
238
U
Activity or conc.
Confidence (Bq/ kg)
Matrix
Reference date
Sample code
4.9 0.028 33.7 1.22 2.3 1.02 10.5 120 27 21.8 0.044 0.050 0.00185 1.11 35.5 1.87 0.95 120 29 22.2 0.039
4.5–5.4 0.025–0.031 32.8–33.9 1.15–1.26 2.2–2.4
0.07 10.0–11.0
9 26–28
0.8
0.002 0.045–0.055
0.00010 1.07–1.15 33.4–36.8 1.80–1.92
0.05
11 28–30
0.8
0.002
Sediment, marine Fish Irish Sea Sediment Fish Mussel from Mediterranean Sea Spinach Baltic Sea Seaweed Phosphogypsum Irish Sea Sediment Moss-soil Irish Seawater Fish Irish Seawater Fish Fangataufa Sediment Mussel from Mediterranean Sea Spinach Phosphogypsum Irish Sea Sediment Moss-soil Irish Seawater
1 Jan1985 1 Jan 1997 1 Jan 1996 1 Jan 1997 1 Nov 2003 15 Oct 2007 1 Aug 2006 1 Jan 2008 1 Jan 1996 15 Nov 2009 1 Jan 2007 1 Jan 1997 1 Jan 2007 1 Jan 1997 1 Aug 1996 1 Nov 2003 15 Oct 2007 1 Jan 2008 1 Jan 1996 15 Nov 2009 1 Jan 2007
SD-N-2 IAEA-414 IAEA-385 IAEA-414 IAEA-437 IAEA-330 IAEA-446 IAEA-434 IAEA-385 IAEA-447 IAEA-443 IAEA-414 IAEA-443 IAEA-414 IAEA-384 IAEA-437 IAEA-330 IAEA-434 IAEA-385 IAEA-447 IAEA-443
CHAPTER 4 Measurements of radioactivity
Table 4.21 Radionuclides referenced by IAEA (Bq/kg)a Continued
238
Pu
239
Pu Pu
239
241
a
0.017 0.025 0.0230 39.0 0.44 1.04 8.8 0.50 0.72 0.036 0.17 0.0197 3.84 7.1 0.197
Am
The
232
Th is in equilibrium with 228Ra and
0.016–0.023 0.022–0.028 0.0221–0.0250 38.6–39.6 0.42–0.48 0.962–1.11 6.51–4.0 0.46–0.52 0.66–0.79 0.030–0.050 0.16–0.25
0.0002 3.78–4.01 6.7–7.4 0.193–0.204
Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Fish Fangataufa Sediment Irish Sea Sediment Soil Sediment, marine Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Sea-plant, Posidonia oceanica Seaweeds, Mediterranean Irish Seawater Irish Sea Sediment Fangataufa Sediment Fish
1 Jan 1988 1 Jan 1988 1 Jan 1997 1 Aug 1996 1 Jan 1996 30 Jan 1983 1 Jan 1985 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 2007 1 Jan 1996 1 Aug 1996 1 Jan 1997
IAEA-308 IAEA-307 IAEA-414 IAEA-384 IAEA-385 SOIL-6 SD-N-2 IAEA-307 IAEA-307 IAEA-307 IAEA-308 IAEA-443 IAEA-385 IAEA-384 IAEA-414
228
Th.
4.5 QA/QC procedures 259
CHAPTER 4 Measurements of radioactivity
2800
2400 2000 Bq (kg)
260
1600
1200
800
400
36
29
5
37
16
4
13
34 398
6
10
28
12
35
14
6
30
33
27
32
26
20
17
21
31
2 25
9 18A 22 11 1 39A 23 15
19 3
Lab. Code No.
FIG. 4.53 Results of IAEA-152 intercomparison 137Cs determinations. Recommended value—2065 Bq/kg; confidence interval—1991–2143 Bq/kg.
of the original IAEA intercomparison run (worldwide) and former Soviet Union laboratories, for high-level (H) milk are presented. The Central Service for Protection against Ionizing Radiation (SCPRI), a service of the French Ministry of Public Health, National Institute of Health and Medical Research, was nominated at the end of 1969 as the International Reference Centre (IRC) of the World Health Organisation for Radioactivity measurements. Four laboratories in the world have been officially designated as WHO collaborating laboratories. These laboratories are: • • • •
Radiation Protection Bureau in Ottawa, Canada. National Institute of Radiation Protection in Stockholm, Sweden, Environmental Monitoring and Support Laboratory (EPA) in Las Vegas, USA and National Radiation Laboratory in Christchurch, New Zealand.
At the present time, 28 laboratories from 17 countries are interested in the WHO-IRC Intercomparisons. Its programme of intercomparison shows the following characteristics: 1. the radioactivity of the samples is the present environmental monitoring level; 2. generally, the samples present real radioactivity due to the fallout or releases of nuclear facilities; 3. the amount of the product provided allows several tests to be carried out; 4. standard materials can be provided and 5. a preliminary study of the results of each intercomparison is given to the participants as soon as possible.
4.5 QA/QC procedures
Table 4.22 Determination of 137Cs in the same milk powder during intercomparisons in 1983 and 1989
1983 1989
Mean value (Bq/kg)
Rel. SD (%)
% outliers/lab.
2.08 1.70
45 19
10 0
Table 4.23 Determination of Sr-90 in simulated air filters during intercomparison runs in 1973, 1976, 1988
1973 1976 1988
Mean value
Rel. SD (%)
178 Bq/filter 179.5 Bq/filter 231 Bq/filter
27 17 3
Table 4.24 Comparison of performance of the two groups of laboratories: worldwide vs Soviet Union Range of reported values for milk(H) (Bq/kg) Radionuclide
Worldwide
USSR
137
469.3–2491.3 58.0–652.5 103.6–3650.0 5.53–8.54
175–3070 184.7–542.5 429–4959 1.43–68.8
Cs Cs 40 K 90 Sr 134
Table 4.24 shows concrete contents of intercomparisons which WHO has carried out in the period 1970–91. The first column of Table 4.25 indicates periods when samples were sent to participating laboratories, the second their nature, the third their numbers, the fourth nuclides and stable elements to be measured and determined for the intercomparison purposes, and the fifth and last column, the scope of radioactivity levels in the samples. As demonstrated by this table, a wide variety of samples has been adopted since 1970, among which are liquid milk, animal bones, human bones, foods, low-level radioactive liquid waste, ground water, mineral water, river sediment, seaweed, pond water, fresh water fish, cereals, seawater, rain water, drinking water, soil and vegetation. Concerning the general conditions of the intercomparison programmes in progress and the results obtained, it can be noted that: •
The IRC has diversified its programme by introducing new categories of samples (waters from various origins, sediments, fish, seaweed, liquid waste, cereals, soil, etc.) in which laboratories involved in environmental monitoring of nuclear power plants are interested.
261
262
Period of dispatch June 1970 Feb 1971 June 1971 Feb 1972 Nov 1972
Nature of the sample
No. sample
Proposed determinations
Radioactivity level
Sr,
Cs, Ca, K
90
Sr, 137Cs, Ca, K Sr, Sr, Ca
90
Liquid milk
A010
90
Liquid milk Animal bones
A338 A504 A505 A806 B078
90
Human bones Dried total diet
90
90 90
137
Sr, Sr, Ca Sr, 137Cs, Sr, Ca, K
B845 C140 C141
90
C617 C618 D256
3
River water
D402
3
Sept 1976 Dec 1976
Liquid milk
D499
90
River sediment
D601
March 1977 July 1977
Animal bones
D736 D737 D925
Activation and fission products (54Mn, 58Co, 60 Co, 90Sr, 134Cs, 137Cs, ...) 90Sr, Sr, Ca
Feb1975 March 1976 June 1976
Sea fish
3
Sr, H
Sr 20 pCi/L, 137Cs 20 pCi/L Sr 3 pCi/g-ash 90 Sr 3 pCi/g-ash 90 Sr < 0.5 pCi/g-ash 90 Sr 40 pCi/kg-dry 137 Cs 90 pCi/kg-dry 90 Sr 20 pCi/L, 137Cs 30 pCi/L 3 H 10,000 pCi/L 3 H 150,000 pCi/L 90
Liquid milk Ground water low-level liquid waste Rain water spiked ground water Mineral water
July 1973 April 1974
Sr 35 pCi/L, 137Cs 120 pCi/L
137
Cs, Ca, K
H 1000 pCi/L H 5000 pCi/L Nat. U 20 μg/L,
3
H
3
Gross β, K nat. U, H,
90
90
Sr,
Sr,
Sr,
106
226
Ra
Ru + 106Rh, 125Sb, 137Cs...
137
Cs, Ca, K
134
Cs, 137Cs, Sr, Ca, K
226
Ra 15 pCi/L
H < 5000 pCi/L 106Ru + 106Rh y 1000 pCi/L 90Sr, 125Sb, 137Cs < 50 pCi/ L 90 Sr 10 pCi/L, 137Cs 15 pCi/L 3
5000–20,000 pCi per kg-dry for each radionuclide 90 Sr 7 pCi/g-ash 90 Sr 3 pCi/g-ash 90 Sr 150 pCi/kg-dry 134 Cs 400 pCi/kg-dry 137 Cs 4000 pCi/kg-dry
CHAPTER 4 Measurements of radioactivity
Table 4.25 WHO-IRC intercomparisons
Dec 1977
Marine sediment
E114
90
Sr, 95Zr + 95Nb, 103Ru, 106Ru + 106Ru, Ag, 125Sb, 137Cs, ...
110m
March 1978 June 1978 Oct 1978
Liquid milk
E414
90
Seaweed
E468
Fission products
Pond waters
3
H
March 1979
Low-level liquid waste
E888 E889 F100
3
H,
June 1979 Nov 1979
Liquid milk
F140
90
Fresh water fish
F290
U,
Sr,
37
Cs, Ca, K
54
Sr,
Mn, 58Co,
60
Co
137
Cs
226
Ra,
90
Sr,
137
Cs
106 Ru + 106Ru, 144Ce + 144Pr < 10,000 pCi/kg-dry other nuclides <500 pCi/kg-dry 90 Sr 15 pCi/L, 137Cs 30 pCi/L
20–10,000 pCi/kg-dry according to the radionuclides 3 H 5000 pCi/L 3 H 20,000 pCi/L 3 H 2.5 μCi/L, 54Mn 800 pCi/L 58 Co 50,000 pCi/L 58 Co 2500 pCi/L 90 Sr 40 pCi/L, 137Cs 250 pCi/L U 1500 μg/kg-dry Ra 3500 pCi/kg-dry 90 Sr 6000 pCi/kg-dry 137 Cs 1500 pCi/kg-dry 90 Sr, 137Cs 20 pCi/kg-dry 226
F553
90
Seawater
F712
Fission products
Total activity 5000 pCi/L
Rainwater Liquid milk Drinking water
F856 G041 G336
3
3
90
90
Oct 1981
Soil
G477
90
Sr,
137
Feb1982
Total diet
G660
90
Sr,
137
Sr,
137
Cs
H Sr, 137Cs, Ca, K 3 H, 90Sr, 106Ru
Cs natural radionuclides Cs, Ca, K, nat. U
H 2500 pCi/L Sr 10 pCi/L, 137Cs 300 pCi/L 3 H < 10,000 pCi/L 106 Ru 75 pCi/L 90 Sr 3 pCi/L 90 Sr 200 pCi/kg-dry 137 Cs 500 pCi/kg-dry 90 Sr 0.4 Bq/kg-dry (10 pCi/kg-dry) 137 Cs 0.4 Bq/kg-dry (10 pCi/kg-dry)
263
Continued
4.5 QA/QC procedures
Cereals (wheat)
March 1980 June 1980 Oct 1980 Feb 1981 June 1981
264
Period of dispatch
Nature of the sample
No. sample
Proposed determinations
Radioactivity level Be 200 Bq/kg-dry (5000 pCi/kgdry) 90 Sr 20 Bq/kg-dry (500 pCi/kg-dry) 137 Cs 5 Bq/kg-dry (100 pCi/kg-dry) U 50–100 μg/L 226 Ra 0.4 Bq L1 (10 pCi/L) 90 Sr 0.3 Bq L1 (pCi/L) 137 Sr 0.3 Bq L1 (pCi/L) 90 Sr 2 Bq/kg-dry (50 pCi/kg-dry) 137 Cs 20 Bq/kg-dry (500 pCi/kg-dry) 5–500 Bq/kg dry, according to the radionuclides 200 Bq L1 10,000 Bg L1
June 1982
Vegetation
G730
7
Dec 1982
Mineral water
nat U,
Feb 1983
Liquid milk
G972 G973 H071
90
Sr,
137
June 1983 Feb 1984
Sea fish
H264
90
Sr,
137
River sediment
H519
June 1984
Rain water Ground water
Feb 1985
Liquid milk
40 PM 300 41 P 300 42 L 300
Activation and fission products (54Mn, 58Co, 60Co, 90Sr, 134Cs, 137Cs, 106Ru) 3 H 3 H
3
Be,
90
Sr,
137
9
Cs, Ca, K
226
Ra
Cs, Ca, K Cs, Ca, K, Sr
100 Bq L1
H
90
Sr Cs 3 H γ-emitters (54Mn, 58Co, 137
May 1985
Low-level radioactive liquid effluent
43 E 300
Feb 1986
Urine
44 UR 300
3
H
60
Co,
134
Cs,
137
Cs)
0.3 Bg L1 0.6 Bq l1 3 H 5 105 Bq L1 54 Mn 1 102 Bq L1 58 Co 2 103 Bq L1 60 Co 2 103 Bq L1 134 Cs 1 102 Bq L1 137 Cs 1 102 Bq L1 500,000 Bq L1
CHAPTER 4 Measurements of radioactivity
Table 4.25 WHO-IRC intercomparisons Continued
May 1986
Aquatic plants
45 V 300
April 1987
Liquid milk
46 L 3000
Activation and fission products (54Mn, 58Co, 60 Co, 90Sr, 137Cs, 106Ru) natural radionuclides 90 Sr Cs Ca 134 Cs 137 Cs 134 Cs 137 Cs 226 Ra
April 1988 Oct 1988
Dry may blossom
April 1989 Nov 1989
Radioactive effluent Herba Drosera longifolia
Aug 1990
River sediments
Dec 1990
Seaweeds
May 1991
River sediments
Mineral water
47 V 300 48 V 300 49 SH 300 50 E 300 51 V 300 52 V 300 53 SD 300 54 AL 300 55 SR 300
3
H Mn, 58Co, 60Co, 124Sb, 228 Th, 228Ra, 232Th 54
134
Cs, 137Cs
<2 Bq L1 5–100 μg L1 103–106 Bq L1 0–104 Bq L1 for each nuclide 5 102 Bq/kg dry for each nuclide
Activation and fission products
300 Bq/kg dry
Transuranium elements
5 Bq L1
Natural radionuclides
4.5 QA/QC procedures
Shelled hazelnuts
0–1 Bq L1 0–50 Bq L1 0–100 Bq L1 0–200 Bq/kg, according to the radionuclides 0–5000 Bq L1
134
137
Oct 1987
10–2000 Bq/kg dry, according to the radionuclides
265
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• •
With regard to the quality of the analyses, the situation presented in the preceding annual report has not evolved much. The regular participation of different laboratories in the intercomparisons provides a comprehensive view of their technical capabilities and of the quality of their analytical work.
According to the Articles 35/36 of the Euroatom Treaty (2012) and the Commission Recommendation (EC, 2000) derived from the Treaty, the Member states of European Union have the legal obligation to inform the European Commission on a regular basis on the radioactivity levels in their environment (drinking water, soil air and food staff). In order to obtain information on the quality of reported values an International Comparison Scheme for Radioactivity Environmental Monitoring (ICS-REM) has been established by European Commission, EC (W€atjen, 2008). Several EC Interlaboratory Comparisons have been organized the Institute for Reference Materials and Measurements in Geel, Belgium, most notably for the determination of 137Cs in air filters (W€atjen et al., 2007; Ma´te et al., 2016). In the last case spiked air filters were prepared for each 76 participating laboratory from Europe and 93.4% of the Laboratories Reported within 33% Interval of the Reference Value. The ALMERA network (Analytical Laboratories for the Measurement of Environmental Radioactivity) was established by the IAEA in 1995, and is a cooperative effort of analytical laboratories worldwide. Members of the network are nominated by their respective IAEA. Member States as those laboratories which would be expected to provide reliable and timely analysis of environmental samples in the event of an accidental or intentional release of radioactivity. ALMERA currently (December 2017) consists of 166 laboratories representing 87 countries. The Agency’s Environment Laboratories in Seibersdorf and Monaco are additional members of the network. The International Atomic Energy Agency’s Environment Laboratories are the central coordinator of the ALMERA network’s activities. IAEA proficiency tests and interlaboratory comparison exercises are organized on a regular basis, specifically for the members of the ALMERA network. These exercises are designed to monitor and demonstrate the performance and analytical capabilities of the network members, and to identify gaps and problem areas where further development is needed. At least one exercise is organized per year by the IAEA for the ALMERA network (see IAEA, 2017; Osvath et al., 2016).
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Karol, I.L., 1970. Calculation of planetary distributions of radon and its long-lived daughter concentrations in the troposphere and lower stratosphere. Tellus 22, 219–227. Kathren, R.L., 1984. Radioactivity in the Environment—Sources, Distribution, and Surveillance. Harvard Academic Publishers. Kawada, Y., 1972. Nucl. Instrum. Methods 98, 21. Kim, C.-K., Oura, Y., Takaku, Y., Nitta, H., Igarashi, Y., Ikeda, N., 1989b. Measurement of 240 Pu/239Pu ratio by fission track method and inductively coupled plasma mass spectrometry. J. Radioanal. Nucl. Chem. Lett. 136 (5), 353. Koppenaal, D.W., 1992. Atomic mass spectrometry. Anal. Chem. 64, 320R. Kotrappa, P., Dempsey, J.C., Hickley, J.R., Steiff, L.R., 1988. Health Phys. 54, 47–56. Kownacka, L., 1980. Vertical distribution of fission products, 226Ra and Pb in the atmosphere. Institute of Nuclear Research, Warsaw. Ph.D. thesis. Report INR-1895/IA/D/a. Kownacka, L., Jaworowski, Z., Baranski, A., Suplinska, M., Lewitowicz, J., Wesolek, R., Romicki, Z., 1985. Measurements of concentration of 90Sr, 106Ru, 125Sb, 137Cs, 144Ce, 210Pb, 226Ra, U and stable Pb in the tropospheric and lower stratospheric air from 1973 to 1984. Central Laboratory for Radiological Protection, Warsaw. Report No. CLOR-117/D. Kownacka, L., Jaworowski, Z., Suplinska, M., 1990. Vertical distribution and flows of lead and natural radionuclides in the atmosphere. Sci. Total Environ. 91, 199–221. Larsen, R.J., Submicrometer Aerosol Collection Characteristics of Dynaweb DW7301L and Freudenberg FA2311 Filter Media, Abstract Book, American Association for Aerosol Research, 1990 Annual Meeting, pp. 266, June 18–22 (1990). Larsen, R.J., 1993. Letter to editor: global decrease in Beryllium-7 in surface air. J. Environ. Radioact. 18, 85–87. Larsen, R., Juzdan, Z., 1986. Radioactivity at barrow and Mauna Loa following the Chernobyl accident. In: Geophysical Monitoring for Climatic Change. pp. 130–133. No. 14, Summary Report 1985. December. Larsen, R.J., Sanderson, C.G., 1991. EML surface air sampling program, 1989 data. USDOE Report-541, August, NTIS, Springfield, VA. Larsen, R.J., Sanderson, C.G., Rivera, W., Zamichieli, M., 1986. The characterisation of radionuclides in North America and Hawaiian surface air and deposition following the Chernobyl accident. USDOE Report EML-460. pp. 1–104. Larsen, R.J., Haagenson, P., Reiss, N., 1989. Transport process associated with the initial elevated concentrations of Chernobyl radioactivity in surface air in the United States. J. Environ. Radioact. 10, 1–18. Larsen, R.J., Sanderson, C.G., Lee, H.N., Decker, K.M., Beck, H.L., 1994. Letter to editor: fission products detected in Alaska following the Tomsk-7 accident. J. Environ. Radioact. 23, 205–209. Larsen, R.J., Sanderson, C.G., Kada, J., 1995. EML surface air sampling program, 1990–1993 data. USDOE Environmental Measurements Lab, New York, NY, pp. 4–247. Report EML-572. Lederer, C.M., Shirley, V.S., 1978. Table of Isotopes, seventh ed. Wiley, New York. Lee, H.N., Feichter, J., 1995. An intercomparison of wet precipitation scavenging schemes and the emission rates of 222Rn for the simulation of global transport and deposition of 210Pb. J. Geophys. Res. 100, 253. Lee, H.N., Feichter, J., Rehfeld, S., 1993a. Simulation of global transport and deposition of 222Rn and 210Pb. In: Proceedings of Topical Meeting on Environmental Transport and Dosimetry of the American Nuclear Society, Charleston, SC, September 1–3. Lee, H.N., Larsen, R.J., Sanderson, C.G., 1993b. Tomsk-7 debris at BRW: detection and transport. In: Climate Monitoring and Diagnostic Laboratory. pp. 104–105. No. 21, Summary
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Report 1992, U.S. Department of Commerce, National Oceanic and Atmospheric Administration, Environmental Research Laboratories. Lepp€aneu, A., Feb. 1994. Ground-level air sampling in Finnish centre for radiation and nuclear safety. In: Proc. Second International Meeting on Low-Level air Radioactivity Monitoring, Madralin, near Warsaw, Poland, 14–18. Lucas, H.F., 1957. Improved low-level alpha scintillation counter for radon. Rev. Sci. Instrum. 28, 258–259. Machta, L., 1965. Some aspects of the USA fallout programme. Meteorological aspects of atmospheric radioactivity. World Meteorological Organisation, Geneva, pp. 155–177 Tech. Note No. 68. Martell, E.A., Moore, H.E., 1974. Tropospheric residence times: a critical review. J. Rech. Atmosph. 8, 903–910. McKlveen, J.W., McDowell, W.J., 1984. Nucl. Instrum. Methods Phys. Res. 223, 372. McMahon, W., 1989. A critical intercomparison of techniques for the determination of low levels of long lived radionuclides. AERE R 13617, Harwell Laboratory, Oxfordshire. Millies-Labroix, J.-C., Bourlat, Y., Mosnier, R., 1994. Airborne radioactivity measurements in Montlhery (France). In: Proc. Second International Meeting on Low-Level Air Radioactivity Monitoring, Madralin, Near Warsaw, Poland, 14–18. Monitoring and Assessment Research Center, 1987. United Kingdom United Nations Environmental Programme, Environmental Data Report. Basil Blackwell Ltd.. Moore, H.E., Poet, S.E., Martell, E.A., 1980. Size distribution and origin of lead-210, bismuth210 and polonium-210 on airborne particles in the troposphere. In: Gesell, T.F., Lowder, W.M. (Eds.), Natural Radiation Environment III. Technical Information Center, U.S. Dept. of Energy, Springfield, VA, pp. 415–429. Morita, S., Kim, C.-K., Takaku, Y., Seki, R., Ikeda, N., 1991. Determination of Technetium-99 in environmental samples by inductively coupled plasma mass spectrometry. Appl. Radiat. Isot. 42 (6), 531. Mouchel, D., Wordel, R., 1992. Measurement of low-level radioactivity in environmental samples by gamma-ray spectrometry. Appl. Radiat. Isot. 43, 49–59. Munro, S., Ebdon, L., McWeeny, D.J., 1986. Applications of ICP-MS for trace determinations in foods. J. Anal. At. Spectrom. 1, 211–219. National Council of Radiation Protection, 1985. A Handbook of Radioactivity Measurements Procedures, NCRP-Report No. 58, second ed. Owen, L.M.W., Crews, H.M., Hutton, R.S., Walsh, A., 1992a. Preliminary study of metals in proteins by HPLC-ICP-MS using multi-element time-resolved analysis. Analyst 117, 649–655. Owen, L.M.W., Crews, H.M., Massey, R.C., 1992b. Aluminium in tea: SEC-ICP-MS speciation studies of infusions and simulated gastro-intestinal digests. Chem. Spec. Bioavailab. 4 (3), 89–96. Perkins, R.W., Thomas, C.W., 1980. Worldwide fallout. In: Hanson, W.C. (Ed.), Transuranic Elements in the Environment. U.S. Department of Energy, NTIS, Springfield, VA. DOE/ TIC-22800. Pierson, D.H., Cambray, R.S., Spicer, G.S., 1966. Lead-210 and polonium-210 in the atmosphere. Tellus 18, 427–433. Pin, C., Lacombe, S., Telouk, P., Imbert, J.-L., 1992. Isotope dilution inductively coupled plasma mass spectrometry: a straightforward method for rapid and accurate determination of uranium and thorium in silicate rocks. Anal. Chim. Acta 256, 153.
Further reading
Porstendorfer, J., Mercer, T.T., 1979. Influence of electric charge and humidity upon the diffusion coefficient of radon decay products. Health Phys. 15, 191–199. Price, W.J., 1964. Nuclear Radiation and Detection. McGraw-Hill Book, Co, New York. Pritchard, H.M., Marien, K.A., 1985. A passive diffusion Rn-222 sampler based on activated carbon absorption. Health Phys. 48, 797–803. Prospero, J.M., Savoie, D.L., Saltzman, E.S., Larsen, R.J., 1991. Impact of oceanic sources of biogenic sulphur on sulphate aerosol concentrations at Mawson, Antarctica. Nature 350, 221–223. Purser, K.H., et al., 1977. An attempt to detect stable N ions from a sputter ion source and some implication of the results for the design of tandems for sensitive carbon analysis. Phys. Rev. App. 12, 1487. Rama, Honda, M., 1961. Natural radioactivity in the atmosphere. J. Geophys. Res. 66, 3227–3231. Rangarajan, C., Gopalakrishnan, S., Chandrasekaran, V.R., Eapen, C.D., 1975. The relative concentrations of radon daughter products in surface air and the significance of their ratios. J. Geophys. Res. 80, 845–848. Rangarajan, C., Gopalakrishnan, S.S., Eapen, C.D., 1976. Global Variation of Lead-210 in Surface Air and Precipitation. Health and Safety Laboratory, New York, pp. 163–182. Report No. HASL-298. Rehfeld, S., Heimann, M., 1994. Three-dimensional atmospheric transport simulation of radioactive tracers 210Pb, 7Be, 10Be and 90Sr. Max-Planck-Institut f€ ur Meteorologie. Report No. 144. Riglet, C., Provitina, O., Dautheribes, J.-L., Revy, D., 1992. Determination of traces of Neptunium-237 in enriched uranium solutions using inductively coupled plasma mass spectrometry. J. Anal. At. Spectrom. 7, 923. Robertson, D.E., Smith, M.R., Doppenaal, D.W., Kiddy, A.R., Strebin, R.S., 1991. Concentrations and behaviour of 129I, 99Tc, and 14C in low-level radioactive wastes from commercial nuclear power stations. vol. 2. Waste Management, p. 287. Rosencwaig, A., 1980. Photoacoustics and Photoacoustic Spectroscopy. In: Chemical Analysis. 57,Wiley, New York. 309 pp. Ross, R.R., Nayce, J.R., Lardy, M.M., 1993. Inductively coupled plasma-mass spectrometry: an emerging method for analysis of long-lived radionuclides. Radioact. Radiochem. 4, 24–37. Samuellson, C., 1987. Hopke, P.K. (Ed.), ACS Symp. Series 331. American Chemical Society, Washington, DC. Chapter 16. Sanderson, C.G., Latner, N., Larsen, R.J., 1994. Environmental gamma-ray spectroscopy at remote sites with satellite data transmission. Nucl. Instrum. Methods Phys. Res., Sect. A 271–277. Savoie, D.L., Prospero, J.M., Saltzman, E.S., Larsen, R.J., 1992. Nitrogen and sulphur species in aerosols at Mawson, Antarctica and their relationship to natural radionuclides. J. Atmos. Chem. 14, 181–204. Savoie, D.L., Prospero, J.M., Larsen, R.J., Huang, F., Izaguirre, M.A., Huang, T., Snowden, T.H., Custals, L., Sanderson, C.G., 1993. Nitrogen and sulphur species in Antarctic aerosols at Mawson, Palmer Station, March (King George Island). J. Atmos. Chem. 17, 95–122. Schmit, J.-P., Youla, M., Gelinas, Y., 1991. Multi-element analysis of biological tissues by inductively coupled plasma mass spectrometry. Anal. Chim. Acta 249, 495.
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Shaw, T.J., Francois, R., 1991. A fast and sensitive ICP-MS assay for the determination of 230 Th in marine sediments. Geochim. Cosmochim. Acta 55, 2075. Shiraishi, K., Takaku, Y., Yoshimizu, K., Igarashi, Y., Masuda, K., McInroy, J.F., Tanaka, G., 1991. Determination of thorium and uranium in total diet samples by inductively coupled plasma mass spectrometry. J. Anal. At. Spectrom. 6, 335. Sill, C.W., Olson, D.G., 1970. Sources and prevention of recoil contamination of solid-state alpha detectors. Anal. Chem. 42, 1596–1607. Spitz, H. and Wrenn, M.E., Design and Application of a Continuous, Digital-Output, Environmental Radon Measuring Instrument. Health and Safety Laboratory, New York; HASL325 (1977) 48–64. Steinkopff, T., Fay, B., Blaab, H., Jacobsen, I., Kiesewetter, W., 1994. The surveillance of radioactivity in the atmosphere by the Deutscher Wetterdienst. In: Proceedings of the Second International Meeting on Low-Level Air Radioactivity Monitoring. Madralin, Near Warsaw, Poland, pp. 14–18. Sun, X.F., Ting, B.T.G., Zeisel, S.H., Janghorbani, M., 1987. Accurate measurement of stable isotopes of lithium by ICP-MS. Analyst 112, 1223–1228. Thompson, J.J., Houk, R.S., 1986. ICP-MS detection for multi-element flow injection analysis and elemental speciation by reversed-phase liquid chromatography. Anal. Chem. 58, 2541–2548. Ting, B.T.G., Janghorbani, M., 1985. ICP-MS applied to isotope analysis of iron in human faecal matter. Anal. Chem. 58, 1334–1340. Ting, B.T.G., Mooers, C.S., Janghorbani, M., 1989. Isotopic determination of selenium in biological materials with ICP-MS. Analyst 114, 667–674. UNSCEAR, 1993. Ionising Radiation: Sources and Biological Effects. United Nations Scientific Committee on the Effects of Atomic Radiation. Wellisch, E.M., 1913. Philos. Mag. 6 (xxvi), 623–635. Wershofen, H., Arnold, D., 1992. Die Aktivit€atskonzentrationen γ-strahlender Radionuklide im Staub der bodennahen Luft von Braunschweig und Berlin. Physikalisch-Technische Bundesanstalt. Report PRB-6.301-93-1. Bundesallee 100, D-38116 Braunschweig. Wershofen, H., Arnold, D., 1995. Radionuclides in ground-level air in Braunschweig—A report of the PTB trace survey station for 1993 and 1994. PTB-6.33-95-3, Braunschweig, Germany. Wessman, R.A., Leventhal, L., 1978. Historical Review of Transuranic Analytical Methodology. NVO. W€olfli, W., Polach, H.A., Andersen, H.H., 1984. Proceedings of the III international symposium on accelerator mass spectrometry, Z€ urich, Switzerland (1984). Nucl. Instrum. Methods Phys. Res. B 5, 91. Wrenn, M.E., Spitz, H., Cohen, N., 1975. IEEE Trans. Nucl. Sci. 22, 645–648. Yamato, A., 1982. J. Radioanal. Chem. 75, 265.
Measurements of radioactivity
4
Chapter outline 4.1 Radiation interaction with matter .....................................................................124 4.1.1 Heavy-charged particles .................................................................125 4.1.1.1 Energy loss due to ionization ................................................ 126 4.1.1.2 Ranges of charged particles ................................................. 129 4.1.1.3 Rutherford scattering .......................................................... 132 4.1.2 Electrons ......................................................................................133 4.1.3 γ- and X-rays .................................................................................134 4.1.3.1 Absorption coefficients ........................................................ 137 4.1.4 Neutrons .......................................................................................138 4.1.5 Penetrating powers of ionizing radiations .........................................141 4.2 Radiation detectors .........................................................................................143 4.2.1 Charged particle detection ..............................................................144 4.2.2 γ- and X-ray detection ....................................................................145 4.2.2.1 Gas-filled detectors ............................................................. 147 4.2.2.2 Scintillation detector ........................................................... 152 4.2.2.3 Semiconductor detector ...................................................... 155 4.2.2.4 Thermoluminescent detectors .............................................. 167 4.2.2.5 Nuclear track detectors ....................................................... 168 4.2.2.6 Photographic film as a radiation detector ............................... 168 4.2.2.7 Neutron detection ............................................................... 168 4.2.3 New developments .........................................................................169 4.2.4 Personal dosimetry detectors ..........................................................171 4.3 Radiometric methods .......................................................................................172 4.3.1 Counting .......................................................................................172 4.3.1.1 2π α-counting method ........................................................ 175 4.3.1.2 4π β-counting with a 4π gas-flow counter .............................. 176 4.3.1.3 4π β-counting with a liquid scintillation spectrometer .............. 177 4.3.1.4 4π α-counting with a liquid scintillation spectrometer .............. 178 4.3.1.5 4π β–γ coincidence counting method .................................... 178 4.3.2 γ-Spectrometry ..............................................................................181 4.3.3 β-Particle spectrometry ..................................................................188 4.3.4 α-Particle spectrometry ..................................................................191 4.3.5 Liquid scintillation measurement method ........................................195 4.3.5.1 External standard method .................................................... 199 Radioactivity in the Environment. https://doi.org/10.1016/B978-0-444-64146-5.00004-5 # 2019 Elsevier B.V. All rights reserved.
123
124
CHAPTER 4 Measurements of radioactivity
4.3.5.2 Sample channel ratio method .............................................. 200 4.3.5.3 Automatic efficiency tracing method ..................................... 200 4.3.6 Radiochemical analysis ..................................................................203 4.3.6.1 Introduction ....................................................................... 203 4.3.6.2 Analysis of strontium ........................................................... 204 4.3.6.3 Analysis of tritium ............................................................... 206 4.3.6.4 Caesium analysis ................................................................ 207 4.3.6.5 Determination of actinides ................................................... 210 4.3.7 Rapid methods ..............................................................................216 4.3.7.1 Rapid determination of transuranic elements and plutonium .... 217 4.3.7.2 Rapid determination of 90Sr ................................................. 217 4.4 Nonradiometric methods ..................................................................................218 4.4.1 Methods based on X-ray spectrometry ..............................................220 4.4.2 Methods based on ultraviolet-visible spectroscopy ............................222 4.4.2.1 Inductively coupled-plasma-optical emission spectrometry (ICP-OES) .......................................................................... 223 4.4.2.2 Laser-excited resonance-ionization spectroscopy (LERIS) ........ 224 4.4.3 Methods based on mass spectrometry ..............................................226 4.4.3.1 Inductively coupled-plasma mass spectrometry ...................... 228 4.4.3.2 Accelerator mass spectrometry ............................................ 234 4.4.4 Laser-induced photoacoustic spectroscopy .......................................241 4.5 QA/QC procedures ...........................................................................................243 4.5.1 Intercomparison ............................................................................246 4.5.2 Reference materials .......................................................................251 References ............................................................................................................266 Further reading ......................................................................................................274
4.1 Radiation interaction with matter The detection of radiation is based on the interactions of the various types of ionizing radiations with matter. The differences between the interactions and the penetrating abilities of the various radiations are very relevant to radiation detection and measurement—e.g., they partly explain the variety of detector types and designs. Radiations can be grouped into directly ionizing radiations and indirectly ionizing radiations. Directly ionizing radiations include all charged particles such as α-particles and heavier ions and β-particles. All charged particle radiations lose energy interaction with the orbital electrons or nuclei of atoms in the materials they traverse. There are two main processes involving the orbital electrons: 1. Atomic or molecular excitation, with the emission of light resulting from subsequent deexcitation. 2. Ionization, which involves the ejection of an orbital electron, resulting in the creation of an ion pair. Indirectly ionizing radiations include some types of electromagnetic radiations and neutrons. These radiations interact with matter by giving rise to secondary radiation
4.1 Radiation interaction with matter
which is ionizing. Indirectly ionizing radiations lose energy by collisions with electrons, or atomic nuclei, and the charged particles thus set in motion interact in turn with the orbital electrons or nuclei.
4.1.1 Heavy-charged particles Heavy-charged particles lose energy by Coulomb interaction with the electrons and the nuclei of the absorbing materials. The collision of heavy-charged particles with free and bound electrons results in the ionization or excitation of the absorbing atom, whereas the interaction with nuclei leads only to a Rutherford scattering between two types of nuclei. Thus the energy spent by the particle in electric collisions results in the creation of electron–hole pairs, whereas the energy spent in nuclear collisions is lost to the detection process. The interactions of charged particles with matter are generally expressed in terms of the energy loss per unit path length (also known as stopping power) and the total range of the particle. The range is most important in detector selection, although some applications require the stopping power for particle identification. For heavy-charged particles in which the path is a straight line, the theoretical energy loss while undergoing collisions with electrons is given by dE 14πZ1 Z2 e2 N 2mv2 ln ¼ I dz mv2
(4.1)
where m, e is the mass, charge of electron; Z1 is the atomic number of moving particle; Z2 is the atomic number of stopping material; v is the ion velocity; N is the number of atoms per unit volume and I is the average ionization potential of stopping atom. The logarithmic term varies slowly with energy so that it is approximately correct to write the energy loss as dE Constant ¼ dx E
(4.2a)
dE Z1 Constant ¼ dx E
(4.2b)
or as
to identify the charged particle atomic number. The range is obtained by integrating the stopping power dE/dx over the energy: range ¼
ð0
dE ð dE=dx Þ Emax
(4.3)
Nuclear collisions can become an important part of the energy loss process, especially in the case of heavy ions and fission fragments. The theory describing this process is too complicated for a brief summary. Finally, it should be mentioned that channelling effects (the steering of charged particles in open regions in the lattice) could reduce the specific ionization loss.
125
126
CHAPTER 4 Measurements of radioactivity
4.1.1.1 Energy loss due to ionization A charged particle moving through matter loses energy as a consequence of collisions with atomic electrons. Let us first consider the effect of a fast-charged particle on an unbound electron. As the particle passes by, the electron “sees” the rapidly changing electric field as an impulse
Δp ¼ FΔt
Zi e2 b2
2b v
(4.4)
where Zi is the atomic number of the incident-charged particle, v is the centre-ofmass velocity and b is the distance of closest approach, called the impact parameter. The energy transferred to the electron is then ðΔpÞ2 2Zi2 e4 1 EðbÞ ¼ ¼ 2me me v2 b2
(4.5)
In order to find the total loss of the kinetic energy by the charged particle to the atomic electrons in the medium, the number of electrons, dn, affected by the charged particle must be known. It follows
dE ¼ EðbÞdn ¼
2Zi2 e2 m e v2
1 ðNZi 2πdbdxÞ b2
(4.6)
where N is the number density of the target atoms and Zt is the atomic number of the target material. The above expression includes the estimate of the number of electrons within a cylinder having its axis along the path of the charged particle. The energy loss per distance traversed is then dE 4πZi2 Zt e4 N ¼ dx m e v2
ð bmax bmin
db 4πZi2 Zt e4 N bmax log ¼ bmin b me v2
(4.7)
We shall now discuss the values of bmax and bmin. For bmax we shall consider electrons bound in atomic orbits. In order to perturb their orbit the duration of perturbation, b/v, must be smaller than the periods, τ, of the system under consideration: b/v < τ ¼ 1/ν. This determines bmax as bmax ¼
v hνi
(4.8)
where hνi is the appropriate average of the electron frequencies. As a result, the electron binding energy becomes important, it impedes energy absorption by the bound electron. Taking the relativistic corrections into account, the duration of the perturbation is shortened by a factor γ ¼ (1 β2)1/2; therefore, bmax ¼ γv=hνi
(4.9)
The lower limit on b, bmin, is determined by quantum mechanics since the electron can be localized with respect to the heavy ion only to the accuracy of its de Broglie wavelength: h ħ bmin > ¼ p γme v
(4.10)
4.1 Radiation interaction with matter
By taking these values for bmin and bmax, one obtains dE 4πZi2 Zt e4 N m e v2 γ 2 log ¼ ħhνi dx m e v2
(4.11)
The quantity 2πħhi is a special average of the excitation and ionization potentials of the atoms in the stopping material. 2πħhνi ¼ I
(4.12)
A more complete theory along these lines was developed by Bethe (1930) using Born’s approximation. He obtained the following expression: dE 4πZi2 Zt e4 ¼ m e v2 dx
"
! # wme ðγvÞ2 2 β Ck ln I
(4.13)
where Ck is a correction term for the “nonparticipating” K-shell electrons that are strongly bound and are partially screened by outer electrons. The mean ionization-excitation potential, I, can be experimentally determined. Its relation to Z is given by the semiempirical formula: I ¼ 9:1Z 1 + 1:9Z2=3 eV
(4.14)
The general form for the energy loss formula allows us to draw some conclusions. Energy loss of a charged particle is proportional to its mass, square of its charge and inversely proportional to its energy, dE Zi2 Zi2 M ∝ ¼ dx v2 2E
(4.15)
Therefore, simultaneous measurement of kinetic energy, E, and energy loss, dE/dx, allows the determination of the mass of the charged particle, since dE E∝ Zi2 M dx
(4.16)
Fig. 4.1 shows the rate of energy loss for protons in some materials (C, Al, Fe, Pb) commonly used as targets or stopping foils; the unit of dE/dx is keV/(mg/cm2). Fig. 4.2 shows energy loss curves for hydrogen and helium isotopes in silicon. These additional curves are given for convenience because silicon is a common detector material, but, in general, these curves can be obtained from the proton curves by using the following relations: dE dE ðdeuteronÞ ¼ ðprotonÞ1 dx dx E E 2
(4.17a)
dE dE ðtritonÞ ¼ ðprotonÞ1 dx dx E E 3
(4.17b)
127
CHAPTER 4 Measurements of radioactivity
1000 500
Al Fe
200
dE/dx (keV/mg/cm2)
100
C
50 20
Pb
10 5
2 0.1
0.2
0.5
1.0
2
5
10
20
50
Proton energy (MeV) FIG. 4.1 The rate of energy loss for protons in C, Al, Fe and Pb.
103
Energy loss (keV μm–1)
128
10
2
4
He
101
3
He
3
H 2
H H
1
100
1
10 Energy (MeV)
100
FIG. 4.2 Energy vs energy loss curves for hydrogen and helium isotopes in silicon.
100 200
Stopping power (MeV / g cm–2)
4.1 Radiation interaction with matter
1000
2000
500
1000
200
500
100
200
50
100
20
α d
50
10
20
5
10
p
5 2 2
1
1 Si Ge
0.05 0.1 0.2 0.05 0.1 0.2
0.5 0.5
1 1
5
2 2
5
10 20 10 20
50 100 200 50 100 200
500 500
Energy (MeV)
FIG. 4.3 Stopping power vs energy for protons, deuterons and α-particles in Si and Ge.
dE 3 dE He ¼ 4 ðprotonÞ1 dx dx E E 3
(4.17c)
dE dE ðaÞ ¼ 4 ðprotonÞ1 dx dx E E 4
(4.17d)
The factor 4 appearing in Eqs (4.17c) and (4.17d) is valid only when the equilibrium charge of the helium ion is essentially 2. The specific ionization loss measures the amount of energy lost by the particle per unit-length of its track; the range indicates how deeply the particle penetrates the absorbing material. Silicon and germanium are the two most common materials in semiconductor industry and especially in radiation detector manufacturing. Therefore, we present here the stopping power of Ge and Si for p, d and α-particles as a function of energy, as shown in Fig. 4.3.
4.1.1.2 Ranges of charged particles The range of a charged particle of incident energy Ei in a material in which its rate of energy loss is dE/dx is given by RðEi Þ ¼
ð Ei 0
dE dE=dx
(4.18)
If dE/dx is known for 0 E Ei, then the range can easily be calculated. For example, Fig. 4.4 shows the range (μm) of protons in Fe for proton energies up to 3 MeV,
129
CHAPTER 4 Measurements of radioactivity
40
30
Range (μm)
130
20
10
0 0
1.0
2.0
3.0
Proton energy (MeV) FIG. 4.4 Range of protons in Fe.
while the range as a function of the energy in silicon and germanium for α-particles, protons and deuterons is shown in Fig. 4.5. Unfortunately, stopping cross sections have not been measured for very low energies nor can they be calculated with reliability. Therefore, computed range–energy relations are subject to considerable uncertainty at low energies. On the other hand, range differences from, say, 1 MeV to Ei can be calculated with confidence. The following curves give such range differences, i.e., Rdif ðEi Þ ¼
ð Ei
dE
1 MeV dE=dx
(4.19)
The total range is given by Rdif(Ei) + R(1 MeV). Table 4.1 lists some estimates of R (1 MeV) for the materials considered here. With better estimates (or actual measurements) this table can be corrected. Fig. 4.6 shows range differences of p, d, t, 3He and 4 He for carbon. All of the range differences in Table 4.1 (except for carbon) are given in mm. The density of carbon, however, depends on its manufacture and it is recommended that the density be measured for the sample used before converting the range differences in mg/cm2 to range differences in mm. The following equations are frequently useful for obtaining ranges for particles in terms of proton ranges: Re ðEÞ ¼ 2Rp
1 E 2
(4.20a)
4.1 Radiation interaction with matter
100 50
p (Ge)
20 10
p (Si)
5
d (Ge)
2
Range (g cm–2)
1
d (Si)
0.5 0.2 0.1
α (Ge)
0.05 0.02
α (Si)
0.01 0.005 0.002 0.001 0.0005
0.0002 0.05 0.1 0.2 0.5 1 2 5 10 20 50 100 200 500 Energy (MeV)
FIG. 4.5 Proton, deuteron and α-particle ranges in Si and Ge.
Table 4.1 Approximate ranges for 1 MeV Particle Material
Unit
p
d
t
He3
He4
C Al Si Fe Ge Pb NaI
mg/cm2 mm mm mm mm mm mm
2.7 0.0146 0.0170 0.0075 0.0130 0.0116 0.0218
1.9 0.011 0.013 0.0061 0.0108 0.0095 0.014
1.7 0.010 0.012 0.0057 0.0100 0.0087 0.011
0.55 0.0035 0.0041 0.0019 0.0033 0.0027 0.0029
0.59 0.0037 0.0043 0.0020 0.0034 0.0028 0.0025
1 E 3
(4.20b)
1 E + 0:25mg=cm2 4
(4.20c)
Rt ðEÞ ¼ 3Rp Rα ðEÞ Rp
131
CHAPTER 4 Measurements of radioactivity
500
d
t
200
α
100
Energy (MeV)
132
3
He
50
p 20 10 5
2 1 1
2
5
10
5
2
2
10
2
5
10
3
2
5
10
4
2
Range difference (1 MeV mg/cm2)
FIG. 4.6 Range differences for carbon of 1 MeV particles.
RHe
3
3 4 ðEÞ ¼ Rα E 4 3
(4.20d)
where the subscripts p, d, t, α and He3 refer to protons, deuterons, tritons, α-particles and He3 ions, respectively. Of course, when range differences are considered, the approximate additive factor for helium ion ranges does not contribute. When considering the range of charged particles it is useful to look at the property of the beam of charged particles. Fig. 4.7 shows the range curve for a beam of particles penetrating to a given depth.
4.1.1.3 Rutherford scattering In addition to interacting with atomic electrons, charged particles interact with nuclei when passing through matter. When approaching the nucleus the charged particle feels a potential, U ðr Þ ¼
Zi Zt e2 ‘ð‘ + 1Þħ2 + r 2mr 2
(4.21)
The first term is due to the Coulomb repulsion, while the second term is the result of angular momentum of relative motion. The approaching charged particle will undergo scattering on the above potential. As early as 1906, Rutherford detected anomalously large α-particle scattering by thin sheets of mica, gold and some other materials. He has explained this scattering in one of the most cited papers ever (Rutherford, 1911). Rutherford’s most significant assumption was that the scattering
4.1 Radiation interaction with matter
N
R0
s
R1
x
FIG. 4.7 Range curve showing the number of particles in a beam penetrating to a given depth.
centre of the target atom was the atomic charge concentrated into a nucleus of 1012 cm. Assuming a Coulomb potential (U(r) ¼ ZiZte2/r) between the α-particle and the target nuclei, the differential scattering cross section is classically derived as dσ Zi Zt e2 ¼ 2μv dΩ
1 θ sin 4 2
(4.22)
where Zie is the electronic charge of the impinging particle (in this case an α-particle), Zte is that of the target nucleus, μ is the reduced mass of the two particles, v is the velocity of the centre of mass and θ is the centre-of-mass scattering angle.
4.1.2 Electrons The interaction of electrons with matter is similar to the interaction of heavy particles, with the following differences: 1. Nuclear collisions are not part of the interaction because of the very light electron mass. 2. At energies higher than a few MeV, radioactive processes (bremsstrahlung) must be considered in addition to the inelastic electron collision. 3. Again because of their light mass, electrons are so intensely scattered that their trajectory in the material is a jagged line; therefore, the concept of range as previously used cannot be applied. Rather, the concept of zero-transmission range is introduced. This is done by means of absorption experiments, which permit definition of the absorber thickness resulting in zero-electron transmission at a given energy. Fig. 4.8 shows the zero-transmission range as a function of energy in silicon and germanium.
133
CHAPTER 4 Measurements of radioactivity
10
Energy (MeV)
134
1
100 Ge Si 10–1
10
–2 0
10
1
10
2
10 Range (μm)
3
10
4
10
FIG. 4.8 Zero-transmission range vs energy for electrons in Si and Ge.
4.1.3 γ- and X-rays The interaction of ionizing electromagnetic radiation with matter is different from the processes previously mentioned, and the concept of ranges and specific ionization loss cannot be applied. Only the three most important absorption processes are considered: the photoelectric effect, the Compton effect and the pair-production effect. The corpuscular description of electromagnetic radiation is the most appropriate for these effects, as one photon in a well-collimated beam of N0 photons disappears at each interaction. The attenuation of the photon beam can be described by a simple exponential law N ¼ N0 exp ðμxÞ
(4.23)
where N is the remaining photons in the beam after traversing distance x, and the absorption coefficient μ is the sum of three terms due to the three above-mentioned processes. These processes are strongly dependent on the energy of effect, such as Rayleigh scattering. Thomson scattering and others are much less important and can be ignored in detection processes (see Fig. 4.9). In the photoelectric interaction, the photon ejects a bound electron from an atom. All of the photon energy, hv, is given to the atom, which ejects the electron with an energy hv – E1, where E1 is the binding energy of the electron. The excited atom then releases energy E1 by decaying to its ground state. In this process, the atom releases one or more photons (and possibly an electron, called an auger electron). The cross section of the photoelectric effect increases rapidly with the atomic number Z and decreases with increasing energy. The Compton effect is essentially an elastic collision between a photon and an electron; during this interaction, the photon gives a fraction of its energy to the electrons, and its frequency ν is therefore decreased. The cross section for this effect decreases with increasing energy, but the decrease is less rapid than for the photoelectric effect.
4.1 Radiation interaction with matter
1.0 0.8 Compton
Fraction of total attenuation
0.6
Compton
0.4 Photoelectric
0.2
Pair
(A) 1.0 Photoelectric
Pair
0.8 0.6 0.4 Compton 0.2
0.01
0.1
1
(B)
10
100
hϖ (MeV)
FIG. 4.9 Relative contributions of various photon interactions to the total attenuation coefficient for (A) carbon and (B) lead.
The energy of electron and the scattered photon are given by
1 Eγ ¼ E0 1 + E0 =mc2 ð1 cos θÞ E0 ¼ E0 Eγ ¼ E0
ðE0 =mc2 Þð1 cos θÞ 1 + ðE0 =mc2 Þð1 cos θÞ
(4.24) (4.25)
where E0 is the incident photon energy; Eγ is the scattered photon energy; Ee is the electron energy; m is the electron mass; c is the velocity of light and θ is the angle between incident and scattered γ-ray directions. The maximum energy loss by the photon is for a head-on collision (θ ¼ 180 degree) and is equal to: E0 Eγ ¼
E0 1 + mc2 =2E0
(4.26)
The fractional energy loss of low-energy photons is small since the scattering is nearly elastic, but becomes appreciable at higher energies. The probability of scattering to a particular angle is a complicated function of energy and angle, but can be generally described as becoming increasingly peaked at small angles as the photon energy increases. The total cross section depends on the
135
CHAPTER 4 Measurements of radioactivity
number of electrons available, or as the atomic number Z of the material. The energy distribution of Compton electrons for several γ-ray energies is shown in Fig. 4.10. In the pair-production effect, a high-energy photon near a nucleus gives up its energy to produce an electron–positron pair. The photon energy goes into the rest-mass energy and the kinetic energy of the electron–positron pair. The minimum energy necessary for this effect is set by elementary relativistic considerations at the value of 1.022 MeV, an amount equivalent to two electron rest masses. The cross section P for pair production increases with energy. Up to energies of 10 MeV, the P/Z ratio remains constant with energy. At higher energies the cross section starts to decrease. Fig. 4.11 summarizes values of the linear absorption coefficients of the above-mentioned effects as a function of γ-ray energy for silicon and germanium. In the pulse-height distributions of Compton interactions of γ-rays in scintillation detectors there are two prominent features usually present: (1) the Compton edge, which corresponds to the maximum energy that can be transferred to an electron by the γ-ray and (2) the backscatter peak, which corresponds to the absorption of a photon which has been scattered through 180 degree in the material surrounding the detector. The energy of the Compton edge is given by Ec ¼
Eγ
(4.27)
ð1 + m0 c2 Þ=2 Eγ
where Eγ is the energy of the incident γ-ray. The energy of the backscatter peak is given by Eb ¼ Eγ Ec ¼ m0 c2 = 2 + m0 c2 =Eγ :
(4.28)
Eγ = 0.5 MeV
Differenntial cross section
136
Eγ = 1.0 MeV Eγ = 2.0 MeV
0.5
1.0
Electron energy (MeV) FIG. 4.10 Compton scattered electron energy distribution.
1.5
2.0
4.1 Radiation interaction with matter
1
10
Linear absorption coefficients (cm-1)
Ge(PE) 0
10
Si(PE) Ge(C)
–1
10
Ge(PP)
Si(C)
Si(PP) 10–2
10–3 10–2
10–1
100 Energy (mev)
101
102
FIG. 4.11 Linear absorption coefficients vs γ-ray energy for Si and Ge (PE ¼ photoelectric, C ¼ Compton, PP ¼ pair production).
The quantities Ec and Eb are shown in Figs. 4.12 and 4.13 as functions of Eγ for both low- (<0.7 MeV) and high-energy (0.5–4 MeV) regions.
4.1.3.1 Absorption coefficients The probability of interacting with matter in one of these three processes can be expressed as a cross section or as an absorption coefficient. The absorption coefficient contains the cross section and is therefore more practical in calculating absorption fractions (Hubbel, 1982). The attenuation coefficient for a beam of γ-rays is related to the number of γ-rays removed from the beam, either by absorption or scattering. For the Compton effect, the absorption cross section is determined by the energy absorbed by the electron, which is the total collision energy minus the average scattered photon energy. For all three processes, the total attenuation coefficient μ is the sum of the three partial attenuation coefficients: μtotal ¼ μphotoelectric + μCompton + μpair production
(4.29)
The attenuation coefficients themselves are defined in terms of thickness of material or surface weight of material. This is just using a thickness x (cm) or a surface weight ρx (g/cm2) where ρ is the density (in g/cm3). The number of primary photons n removed from a beam of n photons is
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CHAPTER 4 Measurements of radioactivity
0.5
Energy of compton edge or scattered γ-ray (mev)
138
0.3 0.2 Back-scater peak 0.1 0.07 0.05 0.03
Compton edge
0.02
0.01 0
0.1
0.2 0.3 0.4 0.5 Incident γ-ray energy (mev)
0.6
0.7
FIG. 4.12 Energy of Compton edge and backscattered peak for γ-rays of energy <0.7 MeV.
dn ¼ μdx n
(4.30)
n ¼ n0 eμdx
(4.31)
which integrates to for an initial beam intensity n0. For surface density ρx the equation is n ¼ n0 e
μ ðρxÞ ρ
(4.32)
where μ is known as the linear attenuation coefficient and μ/ρ as the mass attenuation coefficient. Graphs of the values for NaI are shown in Fig. 4.14, showing the total μ/ρ value and Compton, photoelectric and pair-production components. Fig. 4.15 shows total μ/ρ values for lead and copper. The K and L edges in the photoelectric absorption are at the energy where the photon can eject a K or L shell electron, thus providing an additional absorption mechanism.
4.1.4 Neutrons The interaction of neutrons with matter is quite different from that of either charged particles or γ-rays. Depending on their energy, neutrons interact with matter by various processes.
4.1 Radiation interaction with matter
Energy of comption edge or scattered γ-ray (mev)
5 3 compton edge
2
1.0 0.7 0.5 0.3
Back-scatter peak
0.2
0.1 0.5
1.0
3 2 Incident γ-ray energy (mev)
4
FIG. 4.13 Energy of Compton edge and backscattered peak for high-energy γ-rays (0.5–4 MeV).
1. Elastic scattering: The neutron shares its initial kinetic energy with the nucleus, which suffers recoil only and is not left in an excited state. The smaller the mass of the nucleus, the greater the fraction of the kinetic energy taken by it. The average fraction of the neutron energy transferred per collision to a medium of atomic weight A is given by 2A/(1 + A)2. A 2-MeV neutron gets thermalized in about 18 collisions in water and in about 420 collisions in lead. 2. Inelastic scattering: Inelastic scattering is possible only with fast neutrons: the scattered neutron carries less energy than the incident neutron and the nucleus goes into an excited state. The excited nucleus either emits a γ-ray or remains in a metastable state. 3. Capture: The incident neutron is captured by the target nucleus forming a compound nucleus which may be excited and emitted γ-radiation. This reaction is probably the most common, since thermal neutrons can induce this reaction in nearly all nuclides. The excitation energy of the target nucleus may be emitted in a single photon or in several. Every such capture results in energy emission amounting to about 6–10 MeV. Hence, materials in which neutron capture is allowed to take place for purposes of attenuation are so chosen that, as a result of the capture, charged particles or photons are emitted that can be easily absorbed. Cadmium and boron are commonly used for capturing thermal neutrons. 4. Particle emission: In this type of reaction, the interaction of the incident neutron with the target nucleus may lead to the emission of particles such as protons,
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CHAPTER 4 Measurements of radioactivity
102
Iodine K edge
101
NaI
100
(m/r) (cm2/g)
140
–1
10
–2
10
Compton
Total
Photoelectric Pair production
–3
10
–4
10
0.1
1.0
10.0
Gamma ray energy (mev)
FIG. 4.14 Absorption coefficients for NaI.
neutrons and αs. Since the charged particles will have to overcome the Coulomb barrier before escaping the nucleus, this type of reaction is most probable for light nuclides and fast neutrons. 5. Fission: In this process the compound nucleus splits into two fission fragments with the emission of one or more neutrons. Fission reactions take place with thermal neutrons in 235U, 239Pu and 233U and with fast neutrons in many heavy nuclides. Essentially, the absorption of neutrons occurs in two distinct stages. Fast neutrons are slowed down by elastic and inelastic scattering processes with nuclei, particularly light nuclei like carbon and hydrogen. The slowed-down neutrons are then captured, as the capture cross section for low-energy neutrons is high for most elements. It should be noted that neutron capture and certain nuclear reactions are the only interactions that can make the receiving medium radioactive, α-, β-, γ- and X-rays cannot make a medium radioactive. They can ionize a medium, but that does not make the medium radioactive. An ionized stable atom is not radioactive, because ionization alters only the electron structure of an atom, not the nuclear structure, which determines whether an atom is radioactive or stable.
4.1 Radiation interaction with matter
L edges 2
10
1
(m/r) (cm2/g)
10
K edge
Lead
100
Copper –1
10
10–2
102
102
102
Gamma ray energy (mev) FIG. 4.15 Absorption coefficients for Pb and Cu.
4.1.5 Penetrating powers of ionizing radiations Ionizing radiation gives up some or all its energy for each interaction it undergoes. This means that radiation gradually loses its energy as it passes through a medium, and that at best, only some of the incident amount of radiation can pass entirely through any medium. The radiation that does not emerge from the medium is not trapped inside it—the radiation energy has been transferred to the medium, resulting in ionization and excitation. Some or all of the radiation in a beam entering a medium may cease to exist, some may exit from the medium with diminished energy and some may emerge without having undergone any interactions, depending on the penetrating power of the specific radiation. In general, the penetrating power of ionizing radiation is determined by • • •
the type of radiation, the energy of the radiation, the medium the radiation passes through.
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CHAPTER 4 Measurements of radioactivity
The penetrating power of: • • • •
α-particles are the heaviest of the four most important ionizing radiations and they move the slowest; an α has two positive charges; an α therefore has a high chance for interactions, rapidly loses its energy and so only travels for short distances, especially in dense media and it is possible to stop α-particles completely—after having lost all its energy, an α captures two free electrons and becomes an ordinary stable 4He atom and is then no longer an ionizing particle.
The penetrating power of: • • • •
β-particles are much lighter than α and travel much faster; a β has only one negative charge; a β therefore has a much smaller chance for interactions and loses its energy slower than an α, and so travels longer distances in media and it is possible to stop β-particles completely—after having lost all its energy, a β may be captured by a positive ion, becoming an ordinary orbital electron, and so ceases to be an ionizing particle.
The penetrating power of X- and γ-radiation of: • • •
photons travel at the speed of light and are not electrically charged; a photon therefore has a much smaller chance for interactions than either α or β and γ-rays give up their energy only a little at a time and it is almost impossible to stop all photons in a beam completely; the beam is only weakened or attenuated by a medium.
The penetrating power of: • •
• •
Neutrons have a very small chance for interactions, because they are not electrically charged. Neutrons travel long distances through dense media and shorter distances in less dense media. Media containing lots of hydrogen atoms slow down neutrons the quickest. When a neutron collides with an atom that has about the same mass as a neutron (like a hydrogen atom), it loses the most energy (think of a marble colliding with another marble, as opposed to a marble colliding with a soccer ball). Slower neutrons travel shorter distances in materials that have a greater chance of capturing (absorbing) them. Eventually a neutron is either absorbed by a nucleus or it decays into a proton and an electron; these are eventually neutralized much like α and β. It is very difficult to completely stop neutrons.
4.2 Radiation detectors
Table 4.2 Ranges of ionizing radiations with (maximum energy of 1 MeV) Radiation
Electrical charge
Range in air
Range in tissue
α β γ n
+2 –1 0 0
6 mm 3m >500 m >500 m
0.008 mm 4 mm >65 cm >65 cm
Data taken from US Bureau of Radiological Health. 1970. Radiological Health Handbook, Edition 1970. U.S. Dept. of Health, Education, and Welfare, Public Health Service, Food and Drug Administration, Bureau of Radiological Health, Rockville, MD and from Cember, H. 1983. Introduction to Health Physics. Pergamon Press, Oxford.
Alpha particles
Beta particles
Gamma rays
Paper
Aluminum
Lead
FIG. 4.16 Penetrating powers of ionizing radiations: α-particles can be stopped by paper, β-particles can be stopped by aluminium, γ-radiation is weakened by lead, but is never greatly totally blocked; neutrons will pass through lead, but will be stopped by thick wax or concrete.
The penetrating abilities of the various radiations are compared in Table 4.2 and Figs. 4.16 and 4.17. The ranges given in Table 4.2 are approximate and are only valid for the specified energy, since radiation ranges are energy dependent.
4.2 Radiation detectors Ionizing radiation cannot be detected by any of the human senses. Nor can we perceive the cause of radiation, namely radioactivity (“glowing in the dark” is, unfortunately, a myth). Man has to use special instruments to detect radiation. These instruments are based on the interactions of ionizing radiation with matter, in particular • •
ionization in gases and ionization and excitation in certain solids.
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144
CHAPTER 4 Measurements of radioactivity
Concrete
Alpha particles Beta particles Gamma rays
FIG. 4.17 Penetrating powers of ionizing radiation in tissue: γ and n radiation penetrates easily; βparticles penetrate a few centimetres at max. α-Particles do not penetrate the dead layer of the skin.
A variety of instruments based on these principles exists for the assessment of radiological hazards in the environment. The operating principles and practical applications of these instruments are discussed further. Measurement techniques often also require calculations using the instrument reading, in order to obtain the desired quantity.
4.2.1 Charged particle detection Charged particle detection is a process in which the particles interact directly with the material in the region of ionization, and the path of the particle is clearly defined. The efficiency for detecting a particle once it has entered the detector is nearly unity, so that efficiencies for standard geometries are generally calculated from the solid angle alone. For a small detector at sufficiently large distance from a point source (sourceto-detector distance greater than about three times detector diameter), the solid angle is A/x2 steradians, where A is the detector area and x is the source-to-detector distance. At closer distances the following formula should be used where θ1 is the maximum angle subtended by a circular detector solid angle ¼ 2π
ð θ1
sinθdθ ¼ 2π ð1 cosθ1 Þ
(4.33)
0
For extended sources at close distances experimental calibration should be made. Many types of detectors, such as Geiger-M€uller counters, proportional counters and scintillation detectors, are used for charged particle detection. The selection is made on the basis of resolution and range of particle in the gas or scintillator. In some cases, the particles are not completely stopped within the detector for an energy measurement, but deposit only a portion of their energy. This is related to the relative ionization of the particle and can be used to identify different kinds of particles.
4.2 Radiation detectors
Semiconductor detectors are also used for charged particle detection, but differ from those used for γ-rays in being very thin and operating at room temperature. They have very good resolution and are used for total energy, or partial energy loss, known as stopping power or dE/dx measurements. Silicon-charged particle detectors have a p-i-n structure in which a depletion region is formed by applying reverse bias, with the resultant electric field collecting the electron–hole pairs produced by an incident-charged particle. The resistivity of the silicon must be high enough to allow a large enough depletion region at moderate bias voltages. A traditional example of this type of detector is the gold–silicon detector. In this detector, the n-type silicon has a Schottky barrier contact as the positive contact, and deposited aluminium is used at the back of the detector as the ohmic contact. A modern version of the charged particle detector is called PIPS, an acronym for passivated implanted planar silicon. This detector employs implanted rather than surface-barrier contacts and is therefore more rugged and reliable than the silicon surface-barrier (SSB) detector it replaces. At the junction there is a repulsion of majority carrier (electrons in the n-type and holes in p-type) so that a depleted region exists. An applied reverse bias widens this depleted region which is the sensitive detector volume, and can be extended to the limit of breakdown voltage. Detectors are generally available with depletion depths of 100 and 1000 μm, with the cost approximately proportional to the depletion depth. Detectors are specified in terms of surface area and α- or β-particle resolution as well as depletion depth. The resolution depends largely on detector size, being best for small area detectors. Α resolution of 12–35 keV and β-resolution of 6–30 keV are typical. Areas of 25–3000 mm2 are available as standard, with larger detectors available in various geometries for custom applications. Additionally, PIPS detectors are available with fully depleted silicon wafers so that a dE/dx energy loss measurement can be made, with the particle exiting into another detector to measure the remaining energy.
4.2.2 γ- and X-ray detection The kinds of detectors commonly used for γ- and X-ray detection can be categorized as follows: a. gas-filled detectors, b. scintillation detectors and c. semiconductor detectors. Gas-filled detectors are used for X-rays or low-energy γ-rays. These include ionization chambers, proportional counters and Geiger-M€uller counters. Scintillation detectors are used in conjunction with a photomultiplier tube to convert the scintillation light pulse into an electric pulse. Solid crystal scintillators such as CsI or NaI are commonly used, as well as plastics and various liquids.
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CHAPTER 4 Measurements of radioactivity
Scintillator Gas proportional Counts
146
Si(Li)
0
100
200 300 Channel number
FIG. 4.18 The pulse resolving ability of three types of X-ray detection: scintillator, gas proportional and Si(Li).
Semiconductor detectors, made from single crystals of very pure germanium or silicon, are the highest performance detector type. The superior resolution of these detectors has revolutionized data-gathering for X-ray and γ-ray measurements. The comparison of the pulse resolving ability of the three types of X-ray detectors: scintillator, gas proportional and Si(Li) is shown in Fig. 4.18. The choice of a particular detector type for an application depends on the γ-energy range of interest and the application’s resolution and efficiency requirements. The detector must have sufficient material to absorb a large fraction of the γ-ray energy. Thus, a gas-filled proportional counter is suitable for 14.4 keV γ-rays or for X-rays, but would not “see” 1 MeV γ-rays because the probability of absorbing the γ-ray energy is too low. Further, the higher γ-ray energies are more effectively absorbed by higher Z materials. Other considerations are count rate capability, resolution and if timing applications are involved, pulse rise time. The efficiency of a detector is a measure of how many pulses occur for a given number of γ-rays. Various kinds of efficiency definitions are in common use for γ-ray detectors: 1. Absolute efficiency: The ratio of the number of counts produced by the detector to the number of γ-rays emitted by the source (in all directions). 2. Intrinsic efficiency: The ratio of the number of pulses produced by the detector to the number of γ-rays striking the detector. 3. Relative efficiency: Efficiency of one detector relative to another; commonly that of a germanium detector relative to a 3 in diameter by 3 in long NaI crystal, each at 25 cm from a point source, and specified at 1.33 MeV only.
4.2 Radiation detectors
10–3 8 6
57
Co
109
Cd 139
Ce
4
203
Efficiency
2
Hg
113
Sn
85
Sr 137
–4
10 8 6
Cs
88
Y
60
Co
4
88
Y
2 –5
10
4
6 8 100
2
4 6
1000
2
Energy (kev)
FIG. 4.19 Efficiency calibration for Ge detector.
4. Full-energy peak (or photopeak) efficiency: The efficiency for producing fullenergy peak pulses only, rather than a pulse of any size for the γ-ray. An example of a full-energy peak efficiency curve for a germanium detector is shown in Fig. 4.19.
4.2.2.1 Gas-filled detectors A gas-filled detector is basically a metal chamber filled with gas and containing a positively biased anode wire. A photon passing through the gas produces free electrons and positive ions by the interactions previously described. The electrons are attracted to the anode wire and collected to produce an electric pulse (see Fig. 4.20). At low anode voltages, the electrons may recombine with the ions. Recombination may also occur for a high density of ions. At a sufficiently high voltage nearly all electrons are collected, and the detector is known as an ionization chamber. At higher voltages the electrons are accelerated towards the anode at energies high enough to Positive electrode
Meter Negative electrode Fill gas
FIG. 4.20 Basic elements of gas-filled detector.
– + Battery
147
CHAPTER 4 Measurements of radioactivity
10
10
Continuous discharge
12
GeigerMueller counter
10
8
Ions collected
148
10
106 Ionization chamber 104
10
2
Proportional counter
500 Anode voltage (V)
1000
FIG. 4.21 Gas detector output vs anode voltage.
ionize other atoms, thus creating a larger number of electrons. The detector is known as a proportional counter. At higher voltages the electron multiplication is even greater, and the number of electrons collected is independent of the initial ionization. € This detector is the Geiger-Muller counter, in which the large output pulse is the same for all photons. At still higher voltages continuous discharge occurs. The different voltage regions are indicated schematically in Fig. 4.21. The actual voltages can vary widely from one detector to the next, depending on the detector geometry and the gas type and pressure. The ionization chamber is the simplest form of gas-filled detector. It consists of a chamber provided with two electrodes coupled to an electric potential. Gas ions created by the radiation are attracted to the respective electrodes of opposite charge, causing an electric current (called the ion current) to flow between the electrodes. The ion current is electrically amplified and is measured with a micro-ampmeter calibrated to read in dose rate units. The electric current is related to the dose rate as follows: the electric current is proportional to the ion current, which is a measure of the rate of ionization in the gasfilled chamber, which in turn is a measure of the rate at which the gas in the chamber absorbs energy from the radiation passing through it, i.e., of the dose rate. Ions created inside the chamber walls may enter the chamber causing an incorrect reading. This can be avoided by manufacturing the walls from a material with similar ionization properties to the gas. For an air-filled chamber, such chamber walls are
4.2 Radiation detectors
Ionizing particle
-
+
+
+
Amplifier -
+
Ion pair
+ –
Gas-equivalent chamber wall
mA meter –
+
FIG. 4.22 Working principle of an ionization chamber.
called air-equivalent walls or simply air walls. If the detector is required to respond to β-radiation, the chamber must have thin walls or a thin window to allow the βs to enter (normally structural materials will stop the βs before they can enter the detector chamber). The working principle of an ionization chamber is illustrated in Fig. 4.22.
4.2.2.1.1 Proportional counter If the voltage applied to a gas-filled detector is increased beyond a certain point, it ceases to function like an ionization chamber, and an effect known as gas amplification occurs. Gas amplification occurs when electrons produced by the radiation through ionization of the gas in the chamber are accelerated by the applied voltage to such an extent that they gain enough kinetic energy to cause further ionization, resulting in a cascade of ionization that causes an electric pulse. The original amount of ionization caused by the radiation is thus amplified to create a much larger electric signal. The principle of gas amplification is illustrated in Fig. 4.23. High voltage
- - -
+ - - - - -
-
-
-
-
Initial ion pair +
– Gas amplification.
-
-
-
FIG. 4.23
-
149
150
CHAPTER 4 Measurements of radioactivity
If the applied voltage is not too high, the size of the output pulse is proportional to the amount of energy deposited in the detector by the incoming radiation particle/ photon. (This is why the detector is called a proportional counter.) It is called a counter because the number of output pulses is counted by a counting system. The readout can be either a total number of counts or a count rate (in cpm or cps). The count rate is a measure of the rate at which individual radiation particles/photons cause pulses in the detector. For this reason, proportional counters can be used to determine the amount of radioactivity in a sample. The proportionality of the detector can be used to distinguish between different radiation energies of types of radiation, based on pulse size. A specific type of proportional counter that is used for accurate counting of α- and β-activity on smear samples, is the gas-flow proportional counter. In some types of gas-flow proportional counters, the sample is put inside the detector for greater sensitivity—there is no structural material to absorb the radiation before it can be detected. Fig. 4.24 shows a diagram of the detector of such a gas-flow proportional counter. Other gas-flow proportional counters have very thin Mylar windows sealing off the detector’s gas chamber. During counting, samples are positioned very close to the window.
€ ller counter 4.2.2.1.2 Geiger-Mu If a very high electric potential is applied across the electrodes of a gas-filled detector, the gas amplification achieves a maximum value; this means that all pulses are amplified to the same maximum pulse size. All output pulses then have the same size and are no longer proportional to the energy deposited by radiation; an output pulse is only a sign of some radiation particle/photon having been detected—it cannot be said what type or energy. This type of detector is called a Geiger-M€uller counter (also generally known as a Geiger counter or GM tube). A GM tube is usually of the general construction shown in Fig. 4.20. If the detector is to be sensitive to β-radiation, it is fitted with a thin end-window (see Fig. 4.25) that allows the β-particles to penetrate into the detector’s sensitive volume. GM tubes Positive electrode
Gas inlet
Gas outlet
FIG. 4.24 Diagram of windowless gas-flow proportional counter used for α and β counting. The sample is inserted into the detector on a sliding tray.
4.2 Radiation detectors
Positive electrode
Thin end-window
Gas tight insulator
Negative electrode
FIG. 4.25 A typical thin end-window GM tube for β-detection.
are widely used because they are quite sensitive to radiation, but simple and rugged in construction. Gas-filled detectors do not respond instantaneously; their metres take some time to reach the eventual reading. This is caused by the time it takes to collect all the electric charges and the characteristics of the electric circuit. The time constant is used to quantify the response time. The time constant is the time an instrument takes to indicate 63% of its eventual reading. An instrument with a large time constant responds slowly, and one with a small time constant responds quickly. The significance of the response time is that in practice one must wait for the instrument reading to reach its full response and stabilize before taking a reading; this is especially important for portable instruments. A rule of thumb is that one must wait for at least three times the time constant before taking a precise reading. A time constant for an ionization chamber is in the order of 10 s; for Geiger counters it can vary from a few seconds to more than 20 s. The resolving or dead time of a detector is the minimum period of time that must elapse after the detector has detected a particle/photon, before it is able to detect the next particle/photon. Dead time is caused by the time the detector takes to collect all the charges created by an ionization avalanche, and recover for the start of the next impulse. While collecting the charge from one pulse, the detector is “dead,” that is it cannot register another pulse. A particle/photon that arrives in the detection volume while the detector is still “dead” will not be detected. The higher the activity/dose rate, the more particles/photons will be “missed.” In other words, detector dead time causes the instrument to increasingly under-respond at high dose rates/source activities. Dead time is particularly significant for Geiger counters; they have the largest dead times (up to 200 μs) because they have the most charge to collect (maximum gas multiplication, maximum size ionization avalanche), which takes time. At very high dose rates, a Geiger counter with large dead time may become completely “paralyzed”; its reading drops right down and stays there, even if the dose rate or activity is further increased. Geiger counters can be designed to electronically compensate for counts lost due to dead time. However, many are not. It is important to know whether a particular
151
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CHAPTER 4 Measurements of radioactivity
(A)
(B)
FIG. 4.26 Two generations of Geiger counters: (a) An old Geiger counter shown together with iPhone having an attached Geiger probe; (b) iPhone with its attachment and old Geiger counter probe.
Geiger counter is dead time compensated, so that if not, its drawback can be kept in mind. There has been a significant advance in miniaturization of Geiger counters. Today, one can acquire a “smart Geiger nuclear radiation detector” for iOS iPhone or Android phones, some of them of questionable quality. One such product is shown in Fig. 4.26 in comparison with classical instrument. There are number of sites on web offering the sale of such devices.
4.2.2.2 Scintillation detector Scintillation means the production of small flashes of light. Some crystals, e.g., sodium iodide (NaI) convert the ionization and excitation produced by radiation into a light pulse or scintillation. The amount of light that is produced is proportional to the energy deposited by the radiation particle/photon. The small light pulse is converted into an electric pulse by an electric component called a photomultiplier tube. The size of the amplified electric pulse is proportional to the energy deposited by the radiation photon/particle (for details, see Birks, 1964; Knoll, 1979). The basic components and operating principle of a scintillation detector are illustrated in Fig. 4.27. The proportionality of the output pulses of a scintillation counter to the deposited radiation energy enables the output to be used
4.2 Radiation detectors
Gamma ray in
NaI crystal
Optical pipe
Housing
Photomultiplier
Magnetic shield
Tube base
Preamplifier
HV in
Signal out
Preamplifier power supply
FIG. 4.27 Basic components and operating principle of a scintillator detector.
• •
to discriminate between different types and energies of radiation, e.g., with portable instruments and as input pulses for a γ-spectrometer.
There are also liquid scintillation counters that use a scintillation liquid instead of a crystal. The sample to be counted is dissolved in the scintillating liquid. This method is suitable for α- and β-counting.
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CHAPTER 4 Measurements of radioactivity
The properties of a scintillation material required for good detectors are transparency, availability in large size and large light output proportional to γ-ray energy. Relatively few materials have good properties for detectors. Thallium-activated NaI and CsI crystals are commonly used, as well as a wide variety of plastics. Both NaI and CsI require an activator such as Thallium for proper operation. NaI is the dominant material for γ-detection because it provides good γ-ray resolution and is economical. However, plastics have much faster pulse light decay and find use in timing applications, even though they often offer little or no energy resolution. The actual process by which light is produced is very complex. The high Z of iodine in NaI gives good efficiency for γ-ray detection. Energy dependence of absorption coefficient for NaI is shown in Fig. 4.14. A small amount of Tl is added in order to activate the crystal, so that the designation is usually NaI(Tl) for the crystal. The best resolution achievable is about 8.0% for the 662 keV γ-ray from 137Cs for 2 in diameter by 2 in long crystal, and is slightly better for larger sizes. Typical spectra are shown in Fig. 4.28 for 137Cs and 22Na γ-sources. NaI is slightly nonlinear (about 5%) at low energies (below 200 keV) because of light output variations with γ-energy. The light decay time constant in NaI is about 0.25 μs. Typical charge-sensitive preamplifiers translate this into an output pulse rise time of about 0.5 μs. Fast coincidence measurements cannot achieve the very short resolving times that are possible with plastic, especially at low γ-ray energies. 1200
800 0.662 MeV
137
Cs
400
Counts
154
0 1200
20
800
40
60
80
100
120
140
0.511 MeV 22
400
Na
1.274 MeV
0 20
40
60
80 100 Channel number
FIG. 4.28 NaI(Tl) spectra for
137
Cs and
22
Na γ sources.
120
140
160
4.2 Radiation detectors
Many configurations of NaI detectors are commercially available, ranging from very thin crystals for X-ray measurements to large crystals with multiple phototubes. Crystals built with a well to allow nearly spherical (4π) geometry counting of weak samples are also a standard configuration. Many types of plastic scintillators are commercially available and find applications in fast timing, charged particle or neutron detection, as well as in cases where the rugged nature of the plastic (compared to NaI), or very large detector sizes, are appropriate. Subnanosecond rise times are achieved with plastic detectors coupled to fast photomultiplier tubes, and these assemblies are ideal for fast timing work. Photomultiplier tubes are extremely sensitive light detectors, which multiply the current produced by incident photons by up to 108 times. Since their invention in the 1930s they have seen huge developments that have increased their performance significantly (see Wright, 2017).
4.2.2.3 Semiconductor detector Semiconductors are materials that do not normally conduct electricity because their crystals do not contain enough free-charged particles to carry the current, but that do become conducting when atoms in the crystal become ionized (Knoll, 1979; Bertolini and Coche, 1968). When a relatively small voltage (25–300 V) is applied across the crystal, and it is exposed to ionizing radiation, the electric field sweeps the free-charged particles formed by the radiation out of the crystal. This creates an electric pulse in the external circuit. The size of the pulse is proportional to the radiation energy deposited in the semiconductor. The number of pulses per size range can be counted, and the count rate can be used to determine the activity of the radiation source. Like scintillation detectors, semiconductor detectors are usually used in γ-spectrometer set-ups to identify radionuclides and determine their activities in a sample. A semiconductor detector is much more expensive and somewhat more troublesome to operate than a scintillation detector, but it can distinguish much better between different radiation energies and is better for nuclide identification. The group IV elements silicon and germanium are by far the most widely used semiconductors, although some compound semiconductor materials are finding use in special applications as development work on them continues. Semiconductor detectors have a P-I-N diode structure in which the intrinsic (I) region is created by depletion of charge carriers when a reverse bias is applied across the diode. When photons interact within the depletion region, charge carriers (holes and electrons) are freed and are swept to their respective collecting electrode by the electric field. The resultant charge is integrated by a charge-sensitive preamplifier and converted to a voltage pulse with amplitude proportional to the original photon energy. Since the depletion depth is inversely proportional to net electrical impurity concentration, and counting efficiency is also dependent on the purity of the material, large volumes of very pure material are needed to ensure high counting efficiency for high-energy photons.
155
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CHAPTER 4 Measurements of radioactivity
Table 4.3 Some of the key characteristics of various semiconductors used as detector materials Material
Z
Band gap (eV)
Energy/e–h pair (eV)
Si Ge CdTe HgI2 GaAs
14 32 48–52 80–53 31–33
1.12 0.74 1.47 2.13 1.43
3.61 2.98 4.43 6.5 5.2
Prior to the mid-1970s, the required purity levels of Si and Ge could be achieved only by counter-doping P-type crystals with the N-type impurity, lithium, in a process known as lithium-ion drifting. Although this process is still widely used in the production of Si(Li) X-ray detectors, it is no longer required for germanium detectors since sufficiently pure crystals have been available since 1976. The band gap figures in Table 4.3 signify the temperature sensitivity of the materials and the practical ways in which these materials can be used as detectors. Just as Ge transistors have much lower maximum operating temperatures than Si devices, so do Ge detectors. As a practical matter both Ge and Si photon detectors must be cooled in order to reduce the thermal charge carrier generation (noise) to an acceptable level. This requirement is quite aside from the lithium precipitation problem which made the old Ge(Li), and to some degree Si(Li) detectors, perishable at room temperature. The most common medium for detector cooling is liquid nitrogen; however, recent advances in electrical cooling systems have made electrically refrigerated cryostats a viable alternative for many detector applications. In liquid nitrogen (LN2)-cooled detectors, the detector element (and in some cases preamplifier components) is housed in a clean vacuum chamber which is attached to or inserted in a LN2 Dewar (see Fig. 4.29 as an illustration of the often used configuration). The detector is in thermal contact with the liquid nitrogen which cools it to around 77°C. At these temperatures, reverse leakage currents are in the range of 109 to 1012 A. In electrically refrigerated detectors, both closed-cycle Freon and helium refrigeration systems have been developed to eliminate the need for liquid nitrogen. Besides the obvious advantage of being able to operate where liquid nitrogen is unavailable or supply is uncertain, refrigerated detectors are ideal for applications requiring long-term unattended operation, or applications such as undersea operation, where it is impractical to vent LN2 gas from a conventional cryostat to its surroundings. There are various types and configurations of semiconductor detectors. One type, the surface-barrier detector, is especially useful for α-spectrometry. This type of
4.2 Radiation detectors
Detector holder
Electrical feedthroughs
End cap
Preamp housing
Tailstock
Fill/vent tubes
LN2 transfer collar
Necktube
Molecular sieves
Dewar
Insulation
Coldfinger
FIG. 4.29 Vertical dipstick cryostat.
detector uses a semiconductor wafer, and the sensitive volume of the detector is the surface layer of the wafer. It detects a radiation very effectively, because the α-particles are not absorbed before they reach the sensitive volume (the usual problem with other detectors). The energy lost by ionizing radiation in semiconductor detectors ultimately results in the creation of electron–hole pairs. Details of the processes through which incoming radiation creates electron–hole pairs are not well known, but the average energy ε necessary to create an electron–hole pair in a given semiconductor at a given temperature is independent of the type and the energy of the ionizing radiation. The values of ε are 3.62 eV in silicon at room temperature, 3.72 eV in silicon at 80 K and 2.95 eV in germanium at 80 K. Since the forbidden bandgap value is 1.115 eV for silicon at room temperature and is 0.73 eV for germanium at 80 K, it is clear that not all the energy of the ionizing radiation is spent in breaking covalent bonds. Some of it is ultimately released to the lattice in the form of phonons. The importance of these interactions for detectors is in how they relate to incident γ-ray energy deposited in the crystal. An ideal detector converts all of the energy of
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CHAPTER 4 Measurements of radioactivity
Double escape peak Counts
158
0.5
Full energy peak Single escape peak
1.0 1.5 Detected energy (mev)
2.0
FIG. 4.30 Idealized spectrum.
the γ-ray into an electric pulse that is directly proportional to the γ-ray energy, that is, linear. For γ-rays the Compton scattering often results in only a fraction of the energy being deposited because the γ-ray can scatter and then escape from the crystal without further interaction. The full-energy peak can be produced by a photoelectric absorption, or one or more Compton scatterings followed by photoelectric absorption. If pair production occurs, the positron slows down in the material and then annihilates, producing two 511 keV γ-rays. Each of these may escape from the detector totally, or leave part of their energy by Compton scattering. If one or both totally escapes, the deposited energy is the full energy 511, or 1022 keV, leading to designation of these peaks as “single escape” and “double escape” peaks. An idealized spectrum from a source of γ-rays at 2 MeV is shown in Fig. 4.30. The constant value of ε for different types of radiation and for different energies contributes to the versatility and flexibility of semiconductor detectors for use in nuclear spectroscopy. The low value of ε compared with the average energy necessary to create an electron–ion pair in a gas (typically 15–30 eV) results in the superior spectroscopic performance of semiconductor detectors. If all of the energy lost by ionizing radiation in a semiconductor was spent breaking covalent bonds in the detector’s sensitive volume, no fluctuations would occur in the number of electron–hole pairs produced by ionizing radiation of a given energy. At the other extreme, if that energy entering the semiconductor detector that is partitioned between breaking covalent bonds and lattice vibrations or phonon production were completely uncorrelated, Poisson statistics would apply. The variance in the number of electron–hole pairs n would then be hni2 ¼ n. In fact, neither of these suppositions simulates reality. As the incoming ionizing radiation gives up energy, a large shower of hot electrons is created. After many generations, the energy of these hot electrons gets close to the ionization energy necessary to create an electron–hole pair in the semiconductor detector, so that there are several
4.2 Radiation detectors
possible competing mechanisms for energy loss. Thus the Fano factor F is introduced to modify the more familiar Poisson relation for this case. The equation for the variance can be written as hn0 i2 ¼ Fhni2 ¼ Fn:
In the case where there are no fluctuations in the number of electron–hole pairs, F would be zero; in the case where Poisson statistics apply, F would be equal to 1. Since the energy necessary to create electron–hole pairs in semiconductor detectors is much smaller than that of the incoming ionizing radiation, it can be concluded that F is closer to zero than to 1. The true value of F for silicon and germanium is still unknown; the conflicting theories on the subject do not lead to experimentally distinguishable results. However, by assuming a value of 0.1 for F in both silicon and germanium, satisfactory agreement with measured results is found in most cases. By assuming a value of 0.1 for the Fano factor, the following formula gives the germanium detector resolution at LN2 temperature: pffiffiffi ΔE ¼ 1:27 E
(4.34)
with E measured in eV. ΔE must be summed in quadrature with the FWHM keV noise ΔN in order to obtain the measured energy resolution ΔEs: ΔEs ¼
qffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi ðΔEÞ2 + ðΔN Þ2
(4.35)
The value of K for silicon at room temperature is of little interest because, in such conditions, other factors such as fundamental statistics dominate energy resolution values. These simple formulas show that, as expected from the better statistics due to the lower value of ε, when the energy resolution is dominated by the detector contribution, germanium detectors have an advantage over silicon detectors. The equivalent circuit of a semiconductor detector operated as a spectrometer is shown in Fig. 4.31. In most cases, effects of high resistance of the reverse-biased junction are negligible. If a zero-electric-field radiation-insensitive region is present
Z
I(t)
CD
RD
FIG. 4.31 Equivalent circuit of semiconductor detector: I(t) ¼ current generator; CD ¼ capacitance of the depletion region; Z ¼ series impedance; RD is the resistance of the depletion region.
159
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CHAPTER 4 Measurements of radioactivity
in the detector, its impedance (a parallel RC combination) appears in series with the circuit and is indicated in Fig. 4.31 by the impedance Z. The impedance also accounts for any resistance (or resistance-capacitance combination) appearing in series with the contacts. When semiconductor detectors are used as spectrometers, they are invariably connected to a charge-sensitive (integrating) preamplifier with a high-dynamic input capacitance. The charge-sensitive preamplifier integrates on its feedback capacitance the current signal delivered by the detector and feeds the resulting voltage signal to the filter amplifier (main amplifier). The time behaviour of the current signal at the input of the charge-sensitive preamplifier is determined by the current signal’s shape and by the effect of the equivalent circuit shown in Fig. 4.31. The effect of the equivalent circuit is usually either negligible or easily calculated, whereas detailed considerations on the charge collection process in the detector are needed to calculate the induced current signal I(t). The current delivered by the signal generator I(t) is induced on the contacts of the detector by the motion of the charge carriers created by the ionizing radiation. Therefore, the first problem in determining I(t) is calculating the motion of the charge carriers in the detector’s electric field. When this problem is solved, the induced charge can be calculated by electrostatic considerations. The charge carriers created by the ionizing radiation drift to the contacts of opposite polarity, following the lines of force of the electric field established by the applied voltage. In the case of heavily ionizing particles such as fission fragments, the drift process does not begin immediately due to the creation of the charge cloud. The electric field E(r) in the detector can be calculated from known quantities: applied bias voltage, detector geometry and resistivity of the bulk material. Once the electric field is known, the motion of a charge carrier created at a given point r0 or the detector volume can be calculated by using the values for the drift velocity Vd as a function of the electric field E. Thus the differential equation dr ¼ Vd ½Eðr Þ dt
(4.36)
can be written for every charge carrier and can be solved if the initial positions r0 are known only when the charge carriers are created along a well-defined track (heavycharged particles). In the case of β, x and γ-radiation, the only information on r0 values is of a statistical nature. The integration of Eq. (4.36) leads to r(t) for every charge carrier created. The charge induced by every carrier can then be calculated by electrostatic considerations. For instance, in the case of a detector with parallel contacts and a field E(x) across a distance W, the charge induced by a carrier q moving along a length Δx in the direction of the field is given by Δq ¼ q
Δx W
(4.37)
independently of the shape of E(x). Eqs (4.36) and (4.37) (or the appropriate induction equation) yield the contribution to I(t) of every single charge carrier and, by integration over all the created charge carriers, the total I(t) function.
4.2 Radiation detectors
The rise time Tt of the pulse generated by a semiconductor detector can be measured at the output of a charge-sensitive preamplifier. If the preamplifier is sufficiency fast, Tt is determined by the following factors: 1. the charge collection time TR, 2. the rise time of the detector equivalent circuit, in most cases a negligible quantity and 3. the plasma time. In most cases TR is the dominant factor. Although a precise calculation of TR can be quite complex, the order of magnitude of TR can be easily obtained by the following formulas: TR ffi W 107 s
(4.38a)
for silicon detectors at room temperature and TR ffi W 108 s
(4.38b)
for germanium detectors at LN2 temperature. In these formulas, W is the thickness of the depletion region measured in mm. For silicon detectors and for planar HPGe detectors, the value of W is provided with each detector. For coaxial Ge detectors, W is the radius of the cylinder. The formulas given earlier are indicative only of orders of magnitude and do not give exact values. The previous discussion did not consider trapping effects, which result in a loss of charge to the collection process and consequent distortion of the shape of the peak as observed with a multichannel analyser. Trapping of a charge carrier in a semiconductor occurs when the carrier is captured by an impurity or imperfection centre and is temporarily lost to any charge transport process. In semiconductor detectors, it is useful to introduce the quantity τ+ (mean free drift time): τ + ¼ ðNt , σVth Þ1
(4.39)
where Nt is the density of trapping centres, σ is the trapping cross section and Vth is the thermal velocity. Note that τ+ does not ordinarily coincide with the classical lifetime in photoconductivity theories. This is because in photoconductivity the traps are generally filled, while in a depleted detector, the traps are generally empty. The trapped charge carrier can be reemitted in the relevant band and take part again in the charge transport process. The average time spent by a carrier in a trap is called the mean detrapping time τD and is strongly temperature dependent Eτ τD ¼ Cexp kT
(4.40)
where C is the constant, Eτ is the activation energy of the trap, k is the Boltzmann’s constant and T is the absolute temperature. If the mean detrapping time is of the same order of magnitude as, or larger than, the electronic clipping-time constants, the charge carrier is lost to the charge
161
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CHAPTER 4 Measurements of radioactivity
collection process or is collected with significantly reduced efficiency. The result is poor energy resolution and peak tailing. On the other hand, if the mean detrapping time is orders of magnitude shorter than the charge collection time due to drift of the carriers, then the trap has no effect on the charge collection process. For this reason, normally used dopants such as Li, P, B and Ga, which are shallow donors or acceptors, do not act as traps. It can be shown that, to the first-order approximation, the efficiency of collection of a charge carrier subjected to trapping with a mean free drift time τ+ is given by
TR η¼1 2τ +
(4.41)
where η is the collected fraction of the created charge. In a modern germanium γ-ray spectrometer the charge collection efficiency is of the order of 0.999 and as TR is of the order of 107 s, then τ+ according to Eq. (4.41), is of the order of 104 s. As typical values of Vth and σ are 107 cm s1 and 1013 cm2 respectively, the maximum concentration of trapping centres permissible in the detector is of the order of 1010 cm3 corresponding to approximately 1 for every 1012 atoms of germanium.
4.2.2.3.1 Germanium detector Germanium detectors are semiconductor diodes having a P-I-N structure in which the intrinsic (I) region is sensitive to ionizing radiation, particularly X-rays and γ-rays. Under reverse bias, an electric field extends across the intrinsic or depleted region. When photons interact with the material within the depleted volume of a detector, charge carriers (holes and electrons) are produced and are swept by the electric field to the P and N electrodes. This charge, which is in proportion to the energy deposited in the detector by the incoming photon, is converted into a voltage pulse by an integral charge-sensitive preamplifier. Because germanium has relatively low band gap, these detectors must be cooled in order to reduce the thermal generation of charge carriers (thus reverse leakage current) to an acceptable level. Otherwise, leakage current-induced noise destroys the energy resolution of the detector. Liquid nitrogen, which has a temperature of 77°K is the common cooling medium for such detectors. The detector is mounted in a vacuum chamber which is attached to or inserted into an LN2 Dewar. The sensitive detector surfaces are thus protected from moisture and condensable contaminants. A modern Ge detector is a suitably shaped cylinder of highly purified germanium; it is rather hard to imagine that on 1010 atoms of germanium we have only one atom of an impurity. The modern metallurgical methods, zone refining, allow us to obtain such extreme purity of material. There are very few laboratories where germanium can be refined to this purity, and a single crystal can be grown; there are only three that produce such crystal for commercial uses. And even these three have difficulties in producing really big crystals. The technology of making a γ-ray detector from a piece of pure germanium is in principle simple but in practice very demanding. The Ge crystal must be
4.2 Radiation detectors
• • •
•
properly shaped, equipped with suitable electrical contacts, mounted in a cryostat (two important criteria: it must have excellent contact with the cold finger, and it should be mounted in such a way as to have minimum microphonics) and connected to the first stage of the preamplifier (which is usually placed inside the cryostat).
After a good vacuum has been created in the cryostat, and the detector is cooled by placing the assembly in liquid nitrogen, the characteristics of the detectors are measured. If the detector shows the resolution of 1.68 keV, it will obtain a high price tag. With the resolution of 2.2 keV, it will cost only half as much. And if it has resolution of 3 keV, it will be thrown away. The production of the detector is only to a smaller extent science. Mainly, it is good workmanship, applying modern technologies—and hoping for good results. There are different types of germanium detectors. Their geometry and construction features are illustrated in Fig. 4.32. The “classical” coaxial Ge detector is made of p-type germanium, and is used for spectroscopy of γ-rays. It covers the energy range from 100 keV to several MeV. On the low-energy side, its efficiency is limited by the fact that low-energy γ-rays cannot penetrate the wall of the cryostat. High-energy γ-rays might not be detected because they just pass the volume of the detector without creating a signal that would reflect all the energy of the ray.
(C) (A)
(B)
(D)
FIG. 4.32 (A) p-Type germanium detector; (B) n-type germanium detector; (C) planar germanium detector; and (D) low-energy germanium detector.
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CHAPTER 4 Measurements of radioactivity
If the detector crystal is made of n-type germanium, the outer electrode can be made rather thin. Include a thin aluminium or a beryllium window at the face of the detector, so that low-energy electromagnetic radiation will be able to enter the volume of the detector, and you get a γ-ray detector that is good also for soft γ-rays, and X-rays. Instead of a cylindrically shaped detector, one can take only a slice of the monocrystal. This will make a planar Ge detector. With a beryllium entrance window, it can be used for X-rays. The last addition to the world of Ge detectors is the lowenergy germanium detector. If made by Canberra, it will be called LEGE. It is excellent for low-energy γ-rays; its energy range extends from 10 to 300 keV. This is indicated in Fig. 4.33. The detector geometries also result in different energy resolutions especially for lower γ-ray energies. This is shown in Fig. 4.34. Typical absolute efficiency curves for various Ge detectors in the Marinelli-beaker configuration are shown in Fig. 4.35, while Fig. 4.36 shows typical absolute efficiency curves for various Ge detectors with 2.5 cm source to the end-cap spacing. Detector type Ultra LEGe LEGe Coaxial Ge REGe and XtRa well 0
10 100 Energy (keV)
1
1000
FIG. 4.33 γ-Ray detection energy interval for different types of detection.
3 keV
Resolution
164
Well 1 keV Coax.
ReGe XtRa Large LeGe Small LeGe 100 eV
5.9 10
100
Energy
FIG. 4.34 Typical resolution vs energy.
122
1332
10,000
4.2 Radiation detectors
Efficiency
0.1
0.01
0.001 10
100
1000
10,000
Energy (keV) FIG. 4.35 Detection efficiency of the HPGe detector in the Marinelli-beaker configuration.
10.0 1
2
4
Efficiency (%)
3 1.0
0.1
0.01
5
10
20
50 100 200 Energy (keV)
500
1000
2000
FIG. 4.36 Typical absolute efficiency curves for various Ge detectors with 2.5 cm source to end-cap spacing: (1) REGe, 15% relative efficiency; (2) LEGe, 100 cm2 15 mm thick; (3) LEGe, 200 mm2 10 mm thick; and (4) coaxial Ge, 10% relative efficiency.
4.2.2.3.2 Silicon detector Occasionally, high-resolution X-ray detectors made of very pure silicon can be found on the market. It seems, however, that the purification process of silicon does not yield the same excellent results as in the case of germanium. It is more of a lucky
165
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CHAPTER 4 Measurements of radioactivity
chance that sometimes the metallurgists succeed in producing a batch of silicon with an extremely low amount of impurities. Then, intrinsic silicon detectors are produced. Most of the time, the silicon crystal with little leakage current is obtained by the lithium drifting process. It should be noted that this process in the case of silicon results in a very stable product. At room temperature (without high voltage connected), a Si(Li) detector can be stored for years, without losing its properties. When it is cooled again, it will display the same characteristics as when it was new. Therefore, the search for ultrapure silicon material does not have the same significance as in the case of germanium. The most frequently used silicon detector is a small pellet of lithium-drifted silicon, some 6 mm in diameter and 3–4 mm thick. It is good only for X-rays, in the energy range between 4 and 50 keV. In fact, there are some standard sizes of Si (Li) detectors, areas of 30 and 100 mm2 are considered normal, anything else is a detector made by specifications of the user. The energy deposited in a silicon detector by an X-ray is small, accordingly the electrical signal produced by collecting the charges from the detector is also small. Obviously, these signals will be correctly detected and analysed only if they are clearly above the electronic noise. Therefore, the problem of signal-to-noise ratio is particularly critical with silicon detectors. As a rule, the electronics contributes more than half to the FWHM of an X-ray peak. This was probably one of the reasons that a novel type of electronic gadget was developed: the optically reset preamplifier. The planar germanium and the silicon detector are used for spectroscopy of X-rays. In one aspect, the Si(Li) detector is better: its escape peaks are very small, and do not interfere with the spectrum proper. In the planar germanium detector, as much as 15% of the X-rays coming into the detector will be registered as escape peaks, making the spectrum which is complex anyway, even more difficult to analyse. The planar germanium detector can be recommended if high-energy X-rays; or very low-energy γ-rays are to be analysed. The efficiency of detection for X-rays at very low energies, say at 4 eV, strongly depends on the thickness of the entrance windows at the top of the detector. This is usually a beryllium foil, strong enough to hold vacuum, but thin enough to let X-rays enter the detector. Its thickness is specified in thousands of an inch; 1-mL-thick Be windows are frequently used, thinner are available. Thicker windows are recommended for countries with high humidity. For X-rays detected by Si(Li) crystal, the efficiency also depends on energy (see Fig. 4.37). At the low-energy end, the efficiency decreases towards zero: no wonder, very soft X-rays cannot penetrate into the detector and if they do not come in they cannot be registered. Some manufacturers are replacing Be windows with a window made of lower Z material which improves low-energy response of the detector (see Fig. 4.38). At the high-energy end, the X-ray might travel through the detector without an interaction; the efficiency will decrease with the energy. Note that X-rays interact with the silicon mainly by the photoeffect; this is good—there is very little of the Compton tail, the spectrum looks cleaner.
4.2 Radiation detectors
Full energy detection efficienty (%)
100 80 0.0075 mm (0.3 mil)
60
0.0075 mm (0.3 mil) 3 mm thick
40
0.0075 mm (0.3 mil)
5 mm thick
20 0 C
N
O
0.1
F Ne Na MgAl Si
1
10
100
Energy (keV)
FIG. 4.37 Efficiency of Si(Li) detector as a function of X-ray energy for different thicknesses of Be window and detector sizes.
Intensity
O
C Quantum window spectrum Be window spectrum Ca
0
1
2
3 Energy (keV)
4
5
FIG. 4.38 X-ray spectrum of a detector with low-Z material window.
4.2.2.4 Thermoluminescent detectors When radiation energy is absorbed by crystals of certain materials (e.g. lithium fluoride, lithium borate and calcium fluoride), the absorbed energy is trapped (stored) as displaced electrons within the crystal structure. If the material is heated later after the exposure, the trapped electrons are released and the stored absorbed energy is released in the form of visible light. This process is called thermoluminescence. Materials having this characteristic are called thermoluminescent.
167
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CHAPTER 4 Measurements of radioactivity
After its exposure to ionizing radiation, a thermoluminescent detector (TLD) is read-out in special apparatus (the TLD reader) as follows: The TLD material is heated to a suitable temperature (about 200°C) with a heating element or lamp; the trapped electrons in the TLD return to their normal energy state, releasing their extra energy as a light pulse. The light output is converted to an electric pulse by a photomultiplier tube (the working principle of which is illustrated in Fig. 4.27 for the scintillation counter). The size of the electric pulse is measured, and is proportional to the light output from the TLD material, which in turn is proportional to the total radiation energy absorbed, i.e., to the total radiation dose accumulated over the time that the TLD material was exposed. The TLD material can be used again after read-out. Thermoluminescent materials are commonly used in personal dose metres, in so-called TLD badges.
4.2.2.5 Nuclear track detectors When charged particles, e.g., α-particles, impinge on certain types of plastic materials like polycarbonate or cellulose nitrate, they cause radiation damage tracks in the material. The tracks can be made visually detectable through chemical or electrochemical etching procedures. The visible tracks can be counted using a microscope, microfilm reader or automatic image analysers. The number of tracks is used to calculate the total amount of radiation to which the detector material was exposed. Nuclear track detectors can be used to determine concentrations of α-emitting radon and thoron; the number of tracks is related to the radon or thoron concentration. Nuclear track detectors can also be used to indirectly detect fast neutrons. Fast neutrons interact with the base material of a special film and cause recoil protons to be released. These protons then cause damage tracks in the film which can be made visible and counted as described earlier. The number of tracks can be used to determine the neutron dose.
4.2.2.6 Photographic film as a radiation detector Ionizing radiation affects photographic film in the same way as light does. Films used for radiation measurement are sealed in light-tight packets. When the film is exposed, the radiation causes a latent image which becomes visible as blackening when the film is developed. The optical density (blackness) of the film is measured by passing a beam of light through it. The radiation dose that the film received can then be inferred from the optical density.
4.2.2.7 Neutron detection Neutrons have mass but no electrical charge. Because of this they cannot directly produce ionization in a detector, and therefore cannot be directly detected. This means that neutron detectors must rely on a conversion process where an incident neutron interacts with a nucleus to produce a secondary-charged particle. These
4.2 Radiation detectors
charged particles are then directly detected and from them the presence of neutrons is deduced. The most common reaction used in neutron detection today is n + 3 He ! p + 3 H + 765 keV
where both the proton and the 3H are detected by gas-filled proportional counters using 3He fill gas. In addition, the following reactions are used to detect neutrons: •
•
•
•
B(n, α) reaction: A neutron collides with a 10B atom and an α-particle is released in the process. The ionization caused by the α-particle can then be measured. This reaction is used in the BF3 gas-filled proportional tube. Boron-lined proportional counters and boron-loaded scintillators are other examples of neutron detectors using this reaction. Nuclear fission: A gas-filled detector, typically an ionization chamber, is coated inside with a thin film of fissile material like uranium. Absorption of neutrons by this material causes nuclear fission that produces highly ionizing fission fragments, which can be readily detected by the ionization chamber. Proton recoil: Fast neutrons can knock hydrogen nuclei right out of their atoms. The resulting so-called recoil protons can then cause ionization that can be detected and measured. This reaction can be utilized in gas-filled detectors, nuclear track films and scintillators. Neutron activation: This is the term used for neutron reactions that result in the formation of radioactive nuclides. This type of reaction is used in solid-state use-once devices containing one or more materials that are activated by neutron radiation. The induced activity of each material can be measured and the neutron exposure can then be calculated from these activities. 10
4.2.3 New developments In this section we mention some of the new developments in this very active field. Most of the effort is directed towards searching for new materials which would improve the detector resolution and make detectors more available to the researcher. The efforts are usually problem oriented trying to provide better detection capabilities for the solution of particular problems. For example, Sabet et al. (2016) have demonstrated the feasibility of a novel technique for fabrication of high-resolution CsI(Tl) scintillation detectors dedicated to single photon emission computed tomography (SPECT). The scintillators are fabricated using laser-induced optical barriers technique to create optical microstructures; or optical barriers, inside the CsI(Tl) crystal bulk. The laser-processed CsI(Tl) crystals were 3, 5 and 10 mm thickness. In their work, the authors focus on the simplest pattern of optical barriers in such a way that the barriers were created inside the crystal to form pixel-like patterns. The monolithic CsI(Tl) scintillator samples were fabricated with optical barrier patterns with 1.0 1.0 mm2 and 0.625 0.625 mm2
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pixels. A 4 4 array was exposed to 122 keV γ-rays and peak-to-valley (P/V) ratios were calculated. The P/V values greater than 2.3 were calculated suggesting that pixels can be resolved. The results obtained show that optical barriers can be considered as a robust alternative to mechanically pixelated arrays and can provide high spatial resolution while maintaining the sensitivity and providing cost-effective solution (see also Nagarkar et al., 2004; Sabet et al., 2012). Increasing interest in energetic particle (cosmic rays) effects on weather and climate has motivated the development of miniature scintillator-based detectors intended for deployment on meteorological radiosondes or unmanned airborne vehicles. Aplin et al. (2017) have reported the measurement of ionizing radiation in the atmosphere with a new balloon-borne detector. Their detector is made by a 1.0 1.0 0.8 cm3 caesium iodide thellium activated, CsI(Tl), scintillator with an integrated 1 cm2 silicon PIN photodiode, operated under reverse bias of 12 V. The CsI(Tl) scintillators are more mechanically robust with a greater intrinsic efficiency per unit volume than similar scintillators such as NaI(Tl) and in addition the wavelength of their optical output is optimal for photodiode detection (see Fioretto et al., 2000; Saint-Gobain, 2016). The principle of operation is that ionizing radiation creates a flash of light in the scintillator, which is detected by the photodiode and converted into a charge pulse, proportional to the energy of the incoming particle. The associated current pulses are converted to voltage pulses by fast rise time current to voltage convertor. Coupled with additional electronics units this detector can be run from low bias voltage (12 V) and low current (20 mA) and also has the important advantage of being able to resolve particle energy. Difficult challenges in the areas of nuclear proliferation, homeland security and defence have focused attention on nuclear detection having a goal of enabling the rapid and effective detection, location and characterization of threat materials. The science of radiation detection materials intended for nuclear detection has seen greatly increased attention in recent years. A number of researchers have pursued formalized discovery of novel radiation detection materials for nuclear detection applications. It is assumed that formalized discovery can play a role in finding radiation detection materials for nuclear detection similar to the role it has played in other areas of science. Within the domain of electro-optical materials there are several examples of materials that have been discovered as a result of a detailed understanding of material performance like negative index refraction materials (Shelby et al., 2001), photonic materials (Cregan et al., 1999) and some lasing materials (Faist et al., 1994). Within the domain of radiation detection there has been significant number of research activities on the formalized discovery of materials for applications in highenergy research and medical diagnostics (Peurrung, 2008). Formalized discovery involves the use of one or more methods intended to accelerate the pace and effectiveness of discovery. This could be first principles performance prediction, highthroughput screening, combinational methods, informatic methods, material design, etc. Application of this methodology to search for new detector materials, especially scintillators is described in the work of Deng et al. (2007), van Loef et al. (2007) and Awater (2017).
4.2 Radiation detectors
The review paper by Milbrath et al. (2008) describes the current state of radiation detection material science with particular emphasis on national security needs. In this chapter, the radiation detector materials physics is reviewed with the aim of setting the stage for performance metrics that determine the relative merit of existing and new materials. Semiconductors and scintillator are the two primary classes that are evaluated and the state of art and limitations for each are presented.
4.2.4 Personal dosimetry detectors Personal radiation detectors (PRDs) detect and localize radiation sources generated by man-made devices, they are worn on the person and provide radiation detection in the immediate area around the wearer. The instrument can quickly pinpoint the location of radioactive source, allowing the wearer to respond to the exact location of the threat. The wearer should be able to use PRD effectively while performing other tasks. The PRD should not require the attention of the wearer until it alarms (Stephens, 2018). PRDs offer the functionality of traditional Geiger counter as well as an improved range of detection of multiple types of ionizing radiation. Variations include units that can monitor dose rate, stay time and peak values; options are available for user that may operate in explosive environments. They can be conveniently grouped into following groups: • • • •
basic survey metres, γ-surveys, surface contamination and postaccident response instruments and portable survey metres with full range of modular probes (Stephens, 2018).
Researchers at the national Space Biomedical Research Institute and the U.S. Naval Academy have developed a new radiation detector to be used during space missions where energy levels are dangerous and approximate doses are estimated. The device, called a microdosimeter, is small and low powered and it can measure atmospheric radiation levels in real time (Sauser, 2007). NASA is using dosimeters on every mission, measure the total accumulated amount of radiation exposure and can take form of a badge, pen size tube or a digital readout. However, the drawback of the procedure used is that the doses of radiation astronauts received while in space are not known until they return to Earth. The microdosimeter described here contains an array of cells made from silicon, each the size of a red blood cell and arranged on an electronic board like the squares of chess board, total size of a package of cigarettes. Each cell continuously measures the deposited energy spectrum. This system takes real-time measurements of radiation levels and can alert astronauts immediately if they are at risk. While extremely important to manned space missions, the microdosimeter will have Earth-based application also, especially in areas of nuclear power, different medical and industrial applications. At this point we refer reader to a key dosimeter reference: Radiation Dosimeters for Response and Recovery, Market Survey Report (DHS/OSTP/NUSTL) June 2016
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Table 4.4 U.S. Radiation Dosimeter Standards Standard no. Pub. year ANSI N42.49A 2011 ANSI NJ42.49B 2013 ANSI N13.11 2009 ANSI N42.20 2003 ANSI N322 2003
Standard title American National Standard for Performance Criteria for Alarming Electronic Personal Emergency Radiation Detectors (PERDs) for Exposure Control American National Standard for Performance Criteria for Nonalarming Personal Emergency Radiation Detectors (PERDs) for Exposure Control Personnel Dosimetry Performance—Criteria for Testing (2009) Performance Criteria for Active Personnel Radiation Monitors Inspection, Construction, and Performance Requirements for Direct Reading Electrostatic/Electroscope Type Dosimeters
(see NUSTL, 2016). One should keep in mind that the choice of radiation dosimeters should be governed by the standards they have to satisfy. The appropriate American National Standards Institute (ANSI) standards that cover radiation dosimetry in various applications standards are listed in Table 4.4. Dosimeters differ from other radiation detection devices that are designed for the purpose of preventing a radiological release by alerting a responder to the presence of radiation. Commercially available dosimeters range from low cost, passive devices that store personnel dose information for later readout, to more expensive, battery operated devices that display immediate dose and dose rate information. Readout method, dose measurement range and precision, size, weight, ruggedness and price are important selection factors. In the report NUSTL (2016), instruments have been organized into three main categories: electronic personal dosimeters, EPD, selfreading dosimeters and processed dosimeters. The market survey (NUSTL, 2016) identified 53 radiation dosimeters including 42 EPDs, 2 self-reading dosimeters, 5 processed dosimetry systems designed for user readout (4 of which are portable) and 4 processed dosimeters offered by dosimetry provider services.
4.3 Radiometric methods 4.3.1 Counting The fundamental statistical treatment by Currie (1968) shows that the (low) limit of detection is proportional to the square root of the number of background continuum counts under the peak region of interest, where the proportionality factor varies with the confidence level chosen. Since the detection limit is expressed in counts, a more
4.3 Radiometric methods
interesting parameter is the minimum detectable activity (MDA), expressed in becquerels and defined as the smallest quantity of radioactive nuclide which can be determined reliably, given the prevailing conditions of the specific spectral measurement. The MDA is inversely proportional to the absolute detection efficiency at the full-energy photopeak. Smaller MDAs can be obtained by lowering the background and increasing the detection efficiency (see also Watt and Ramsden, 1964; Friedlander et al., 1964). Therefore, for the measurement of the concentrations of radionuclides in environmental samples and especially for low concentration levels, it is essential to reduce the background rate as well as improving the counting efficiency. When considering the background rate reduction, it is important to pay attention to the origins of the background. These are 1. 2. 3. 4.
natural and artificial radioactivity in the environmental circumstances, radioactivity in the detector and/or shielding material, cosmic radiation and instrumental noise.
In the surrounding environmental materials, e.g., a concrete wall of a building, the soil, air or water, there are various kinds of natural and artificial radionuclides which give rise to a background. The most important are 40K, 226Ra and its decay products and 232Th and its decay products. As a result, many lines originating from these nuclides are observed even in a heavy shielding box, during the spectrometry with a Ge detector. The background due to these radiations can be reduced considerably by shielding with heavy materials such as lead and iron. Typically, such background can be reduced to a hundredth by a 10 cm thickness of lead or 30 cm of iron. In making a shielding box, it is important that the detector is surrounded entirely (4π direction) with shielding. In choosing the shielding material, one must pay attention to any undesirable contamination with radioactive substances. For example, a small amount of 210Pb(RaD), which is a member of the 226Ra series, is inevitably contained in lead. Its daughter nuclide 210Bi(RaE) emits energetic β-rays (Emax ¼ 1.17 MeV), resulting in Bremsstrahlung which may contribute to an increase in background rate. The concentration of 210Pb is very dependent on the mine where it was produced, and a careful check on the content of 210Pb is important for achieving ideal shielding characteristics. Since the half-life of 210Pb is 22 years, old lead has much better characteristics. Iron blocks are sometimes used instead of lead. However, the modern iron/steel is often contaminated with 60Co, and careful checking is necessary prior to the construction of the shielding assembly. This 60Co contamination originates from the 60Co source which was used to monitor the abrasion of the furnace wall. For shielding purposes, the use of old iron, e.g., from materials from a sunken ship, is recommended. The background may also originate from the trace amount of radioactivity in the detector itself and its assemblies, and cannot be eliminated by shielding alone and hence, the contamination of the detector materials with radioactive substances should be carefully examined. A typical example is 40K in the glass used for phototubes. This can be eliminated by the use of a quartz phototube. The 226Ra
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present in solder sometimes gives rise to a considerable rate of background. The molecular sieve used for keeping a vacuum in the cryostat of a Ge detector sometimes causes a background spectrum. In this sense, electrolytic copper, stainless steel and Perspex are recommended as materials to be used in the shielding box. Even when very thick materials are used for shielding, it is very difficult to stop the penetration of highly energetic cosmic rays, which have energies in excess of 103 MeV. In order to overcome this undesirable background, an electrical guard technique is employed. The cosmic rays which give a pulse to the main detector should produce another signal to one of the guard counters. The two signals from the main counter and from the guard counter are produced simultaneously. Therefore an anticoincidence technique is applied quite effectively to reduce the background rate, and an ultimate low background rate of the order of 0.1 cpm can be achieved by the combination of heavy shielding and the anticoincidence guard technique. The coincidence technique is often used for the low-level detection of electrons. As an example, we mention the low background β-ray spectrometer as developed by Tanaka (1961). A small Geiger-M€uller counter is mounted inside a large plastic scintillator, and scintillation pulses from only those events which are coincident in the two detectors are analysed by a multichannel analyser. In this case, signals caused by cosmic rays are also subject to the analysis, but the minimum width of the scintillator is designed so as to produce a very large pulse height, and this signal is rejected in the course of the pulse-height analyses. In these instruments, the β-ray spectra can be measured for even very weak samples. By the addition of a logarithmic amplifier, the analysis of β-ray spectra becomes even easier. Pulse-shape discrimination (PSD) is also used sometimes for background reduction. Electrical noise is very often caused by the discharge of high-potential circuits, the transition signal of thyristors and the induction signal from spark gaps, etc., which result in a background signal to the counter. To prevent this noise, electrical filtering in the main power supply, and electrical and/or magnetic shielding are sometimes required. Another essential requirement for reliable low-level counting is to increase the counting efficiency and/or to use a larger amount of source sample. In order to increase the counting efficiency, one should use a large volume (large size) detector, and try to improve the source-to-detector geometry. However, it should be noted that with increasing detector size or volume, the background may also increase. As for the improvement of the geometry, the use of a well-type detector is advantageous, but the source size is sometimes limited. On the other hand, to use a larger amount of source, it is convenient to use a Marinelli-beaker, especially for the measurement of samples such as milk or soil. In principle, one can perform the measurement of activity of any sample by using radiation detection systems and standard sources. Standard sources and solutions are available from many national and international operations. However, for sources one can perform radioactivity measurement without the help of standards. Absolute counting implies of method of radioactivity measurement performed without the
4.3 Radiometric methods
help of a standard/reference source. The methods for the absolute measurements can be classified into two categories. The first comprises the direct methods, which include the defined solid angle method, the 4π counting method, the internal gas counting method and the liquid scintillation counting (LSC) method. These methods use relatively simple procedures, and each provides unique applicability. However, some uncertain corrections, such as for self-absorption, are involved in these methods, and the accuracies obtainable are somewhat limited. The second category is based on the coincidence technique, by which we can determine the radioactivity more accurately without any uncertain estimation of correction terms. The most often used methods include: the 2π α-counting method, 4π β-counting with a 4π gas-flow counter, 4π β-counting with a liquid scintillation spectrometer and the 4π β–γ coincidence counting method.
4.3.1.1 2π α-Counting method
In cases of radioactivity measurements of an α-emitter such as 241Am or uranium source electrodeposited onto a metal disk, a simple 2π α-counting with a gas-flow proportional counter is practical and a reasonable accuracy can be expected. If the applied potential to the proportional counter is set at the α-plateau region, the counter responds only to α-particles. Since the backscatter of α-rays is small (a few per cent) compared with that of β-rays, and the value can be estimated as a function of the atomic number of the backing material, the radioactivity, n0, can be readily obtained from the observed counting rate, n, by the following simple relation: n0 ¼ 2n
1 1 ð1 aÞ ð1 + bÞ
(4.42)
where the factors a and b represent the corrections for the self-absorption and backscattering of the α-particles (Hino and Kawada, 1990). Self-absorption, a, of a uniformly distributed source with superficial density, d, can be estimated by the simple equation: a ¼ d=ð2RÞ
(4.43)
where R is the range (mg/cm ) of the α-particles in source material; this value can be referred from some data books, e.g., from Ziegler (1977). The effective thickness (superficial density), d, of the source can be determined from the line width of the pulse-height spectrum for α-rays obtained with a Si surface-barrier detector mounted in a vacuum chamber as shown in Fig. 4.24. The effective thickness, d (in mg/cm2 unit), of the source material can be obtained from 2
d¼
qffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi Γ 2obs Γ 2i =s
(4.44)
where ΓOBS and Γi are the observed and intrinsic line widths (FWHM), respectively, and s is the stopping power of the source material for the specified α-particle. In the cases of the measurements of electrodeposited sources of RaDEF (210Pb–210Bi–210Po), e.g., the stopping power of 294 keV cm2/mg for 5.3 MeV
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α-particle in PbO2 should be used. If the thickness of the sources ranges from 0.59 to 1.10 mg/cm2, the self-absorption values from 1.79% to 3.3% would result. Backscatter-values of α-rays are less than a few per cent, and can be estimated from the published data as a function of the atomic number of source backing. The correction, b, for the backscattering can therefore be easily obtained with an acceptable reliability. This method is very simple in procedures, and can be applied to the absolute measurements of very weak sources without any standard source. For these advantageous features, this method is useful for the measurements of samples for environmental radioactivity analysis and also for the preparation of laboratory-made standard αsources.
4.3.1.2 4π β-Counting with a 4π gas-flow counter
The classical method of 4π β-counting with a 4π gas-flow proportional counter is still useful for the absolute measurements of β-emitting nuclides provided that good sources with small self-absorption can be prepared. From the observed counting rate, after fundamental corrections for background and counting loss due to dead time, the radioactivity, n0, can be calculated as n0 ¼ n
1 1 1 af 1 as
(4.45)
where af and AS are the corrections for absorptions in the source-supporting films and in the source material itself. While the foil-absorption can be measured experimentally, e.g., by the sandwich method, it is difficult to make direct determination of the self-absorption values only by the use of 4π-counting. In practice, it will be convenient to estimate the self-absorption values from the published data taken as a function of β-end point energy for various source conditions. The self-absorption values are very dependent on both the β-ray energy and source conditions, and are in the range from 1% to several 10s%. In order to get reliable data with this technique, minimizing the foil-and self-absorptions is essential. The source preparation techniques are therefore of great importance. As materials for source support, a VYNS film (copolymer of vinyl chloride and vinyl acetate) whose thickness ranges from about 10 to 30 μg/cm2 can be easily prepared in the following manner. 1. Dissolve the VYNS powder (i.e. from the Union Carbide Co.) in cyclo-hexanone (CsH10O) 5 times in volume, and keep it for at least 1 week in sealed glassware to ensure homogeneous mixing. 2. Attach a small amount of VYNS solution onto the edge of a glass plate (e.g. a slide glass for a microscope observation) and contact the edge onto the surface of clean water. A thin VYNS layer will be spread over the water surface. 3. Scoop up the film with thin metal rings for film support, and dry naturally. In this procedure, many source supports can be prepared simultaneously if a number of rings are set on a holder.
4.3 Radiometric methods
Table 4.5 Source preparation techniques for 4π counting Name of technique
Procedures
Seeding
After adding a drop of freshly prepared 1:5000 dilution of Ludox-SM (Dupond) with a distilled water, dry in air with or without an infrared lamp irradiation After adding a drop of a 1:10,000 dilution of Catanac, dry in open air with or without an infrared irradiation For example: (1) Iodine isotopes: After adding a drop of AgNO3 solution, dry naturally in open air (2) Mercury isotopes: Dry in an atmosphere of H2S After adding a drop of pure water, dry in air with an infrared lamp irradiation. Add a drop of water again, and dry in an atmosphere of ammonium
Wetting Precipitation
Ammonium complex forming
4. In order to avoid the accumulation of electrical charges on the source mount in a 4π counter, the thin VYNS film thus prepared should be rendered electrically conducting by the vacuum deposition of gold or aluminium to both sides (15 μg/cm2 each side). The sources are prepared by a simple deposition of a known amount of radioactive solution of sample on the support. The solid contents of the sample solution must be kept to a minimum to minimize the self-absorption. Even when carrier concentration is small (0.01–0.05 mg/cm2), a considerable amount of self-absorption is unavoidable. This is mainly due to the local agglomeration or crystallization formed in the dry. Several procedures as illustrated in Table 4.5 are useful to extend the source material uniformly onto the source support and consequently to reduce the selfabsorption. For volatile materials such as iodine, mercury and selenium, special care and treatment are needed to form a more stable chemical form. In order to measure the radioactive concentration, i.e., radioactivity in a unit mass or unit volume of sample solution, quantitative deposition of sample solution is necessary. This is achieved by gravimetrical means with a microbalance or with volumetrical ones with a micropipette. For precise measurements, the former method is preferable, in which a differential weighing with a pycnometer is essential to avoid the appreciable amount of evaporation of solution in the course of the weighing (0.5%). Although the attainable accuracy is limited to around a few per cent in this simple 4π β-counting method, this method can be applied to the measurements of weak source, and hence can play a special role in the environmental radioactivity measurements after chemical separation.
4.3.1.3 4π β-Counting with a liquid scintillation spectrometer A liquid scintillation counter is considered as a version of a 4π counter, since the radioactive material is mixed in a scintillator, and the effective solid angle sustained from the source to the detector is nearly 4π. A large quantity of sample solution is
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usable in a LSC compared with the samples used for the original 4π. This is one of the advantageous possibilities of LSC. Although the pulse height may be decreased and distorted due to the quenching phenomena, and accordingly the counting efficiency changes because of the source conditions, the possible counting loss due to the quenching effect can be corrected by using the pulse-height information. The classical means to correct the counting loss are the external standard method and the channel ratio method. However, the efficiency tracing technique has been proved to be more advantageous and convenient to determine the absolute activity. In this improved technique, most of the pure β- and β–γ emitting nuclides, including 3H and 14 C, can be measured accurately and readily by the use of only one or two common standard samples without the need to prepare quenching correction curves. In the relationship between the efficiencies for the standard sample and the count rates of the sample to be measured, the counting efficiency of 100% with respect to the standard sample serves to derive the radioactivity of the sample to be measured. Signal pulses resulting from liquid scintillation are analysed in separate discriminating regions with a multichannel analyser; the relation between the efficiency, E, and the count rate, n, is empirically expressed by the quadratic regression equation: n ¼ aE2 + bE + c:
(4.46)
Extrapolated value to E ¼ 100% gives the true disintegration rate of the sample. This technique has universal applicabilities, and weak radioactive samples can be measured in terms of absolute activity. These features are considered to be advantageous in the environmental radioactivity measurements.
4.3.1.4 4π α-Counting with a liquid scintillation spectrometer
Since α-rays emitted from radionuclides are energetic (4 MeV), and almost monochromatic, the detection efficiency of a liquid scintillation counter for α-rays is expected to be near unity. However, if β-rays are emitted from the source in addition to α-rays, the spectra originated from α-rays and from β-rays are sometimes overlapped partially. In order to overcome this effect, PSD technique is useful to select α-signals from α–β mixing signals. The background due to β-rays is entirely diminished after pulse-height discrimination. It should be noted that a special scintillator cocktail such as NE-213 should be used in this α–β PSD method. A similar cocktail can be prepared by mixing 5 g of PPO, 0.3 g of DMPOPOP, 100 g of naphthalene and 1 L of xylene. This special α-counting with a liquid scintillation spectrometer is useful for the α-radioactivity measurements of environmental samples on an absolute basis.
4.3.1.5 4π β–γ Coincidence counting method
Among various techniques used for radioactivity standardization, the 4π β–γ coincidence counting method is undoubtedly the most important technique, and most radionuclides can be standardized by this technique or its extended techniques. Actually, most national standardizing laboratories in the world employ this method for the establishment of the national standards of various kinds of radionuclides.
4.3 Radiometric methods
The 4π β–γ coincidence counting is an improved version of the β–γ coincidence counting method. Here, a 4π β gas-flow-type proportional counter is usually used as the β-detector, and a NaI(Tl) scintillation counter located near the 4π β-counter wall is employed to detect the γ-rays. Source preparation techniques are the same as those explained in the 4π β-counting. Even in this 4π β–γ coincidence method, the source preparation to minimize the self-absorption is important to reduce the correction terms and also to shorten the extrapolation range in the 4π β–γ coincidence extrapolation technique. Hence the special treatments as explained in Table 4.4 are important in the course of source preparation. For the measurements of α–γ emitting nuclide such as 241Am, the same technique can be used, provided that the applied potential to the 4π proportional counter is set at the α-plateau region. This 4π β–γ coincidence counting technique can be readily applied to the absolute measurements of electron capture nuclides such as 51Cr, 54Mn, 88Y, 75Se, etc., since a 4π gas-flow counter filled with P-10 gas (gas mixture of Ar + 10% methane) has considerable detection efficiency for the electron capture events through the detection of the characteristic X-rays and Auger electrons. This should be called 4π X–γ coincidence counting, but the measuring procedures are essentially the same as those for 4π β–γ coincidence technique. Here the extrapolation technique is again useful for making various corrections. In cases of 4π X–γ coincidence counting, an alternation of the flowing gas, from P-10 gas to pure methane, is effective to change the counting efficiency of the β-counter to X-rays. This 4π β–γ coincidence counting method and its extended technique are very useful for the precise absolute measurements of radioactivity, and therefore these methods are employed for the establishment of national standards of radioactivities of various kinds of nuclides in most of the national standardizing laboratories. Laboratory-made standard source can be prepared by these techniques. However, these techniques are not adequate for the measurements of low-level radioactivities, and hence the direct application to the measurements of environmental samples is difficult. In this arrangement, apart from some corrections required, the disintegration rate (radioactivity), n0, and counting efficiencies, εβ and εγ, of the β- and γ-channels are given by the following simple equations for a β–γ emitter with a simple decay scheme: n0 ¼ nβ nγ =nc εβ ¼ nc =nγ and εγ ¼ nc =nβ ,
(4.47)
where nβ, nγ and nc represent counting rates, after correcting the background, accidental coincidences, etc., in the β-, γ- and coincidence channels, respectively. The above fundamental equations can be easily derived from the following three equations: nβ ¼ n0 εβ , and nc ¼ n0 εβ εγ nγ ¼ n0 εγ ,
(4.48)
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The explanation along this line is usually made in most textbooks. However, the ideal conditions are seldom achieved in any practical counting system, and some modifications of the fundamental equations are required in order to correct the possible effects which may disturb the ideal conditions. For example, the 4π βproportional counter has an appreciable sensitivity to γ-rays. Furthermore, the γtransition is detected by the β-detector through the internal conversion process, if any. Besides, because a coincidence mixer has a finite resolving time, false accidental coincidences are inevitably produced by chance. In addition to this problem, further consideration must be given when a nuclide with a complex decay scheme is measured. Taking account of all of these effects the coincidence equation becomes nβ nγ 1 τR n∗β + n∗γ εc εβ 1 n0 ¼ αεce + εβγ + C 1+ εβ 1 + α εγ nc 2τR n∗β n∗γ ð1 n∗c τÞ
(4.49)
and εβ ¼
nc 1 nγ 1 n∗ τ
(4.50)
β
where εβγ is the γ-sensitivity of the β-counter; εce is the counting efficiency of the β-detector to internal conversion electrons; α is the total internal conversion coefficient; εc is the probability of obtaining γ–γ coincidence signal when a β-particle is not detected; C is the correction for the complex decay scheme; τR is the resolving time of the coincidence mixer and τ is the dead time of the β-channel. The notations with an asterisk denote the counting rate including background rate. These coincidence equations should be valid for all types of coincidence measurements. The effects due to the γ-sensitivity of the β-counter, the response of the β-counter to internal conversion electrons and the effect originating from the complex decay scheme are all reduced by a factor of (1 εβ)/εβ. In the 4π β–γ coincidence counting, the β-efficiency, ε, is expected to be nearly unity, and the corrections for these effects remain small. The attainable accuracy is therefore improved very much in this technique. This is one of the reasons why the 4π β–γ coincidence counting method is superior to the ordinary β–γ coincidence counting. However, when a nuclide with a complex decay scheme is measured, considerable correction is sometimes required. As a practical approach to overcome these difficulties, a series of 4π β–γ coincidence countings should be made for various source conditions which provide different β-efficiencies, and an extrapolation to (1 – εβ)/εβ ! 0 gives the true disintegration rate. It should be noted that most correction terms such as for the γ-sensitivity of the β-counter and the response to the internal conversion electrons as well as for the complex decay scheme are all corrected automatically in this procedure. The shape of the efficiency function, graph of the nβnγ/nc against (1 εβ)/εβ, is dependent on the decay scheme and the γ-gate setting, etc. In the radioactivity measurements by means of coincidence technique, the counting statistics are given by the following formula:
4.3 Radiometric methods
rffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi sffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi 1 εβ 1 εγ 2εβ εγ εβ εγ + 1 1 ¼ + : σ¼ nc T n0 T n0 εβ εγ T
(4.51)
The second term of Eq. (4.51) is a function of (1 – εβ), which will be reduced considerably in the use of a high-efficiency detector, and excellent counting statistics are obtained in a short measuring time. This is another advantageous future use of the 4π β–γ coincidence method.
4.3.2 γ-Spectrometry High-purity detection systems having a very low background are suitable tools for the direct measurement of low-level radioactivity in environmental samples. The background features of the detection system are of considerable importance because they have to be known for one to obtain an estimate of the detection limit and of the MDA. The natural radioactivity background originates from the uranium and the thorium series from 40K and from cosmic rays. Natural radioactivity is found in most materials, and it is necessary to shield the detector using carefully selected materials of high density. The materials used in the detector assembly, and the shielding materials need to have the lowest possible inherent background radiation (see Gilmore and Hemingway, 1995). Containers such as Marinelli-beakers can be filled with aliquots of environmental samples and placed on top of the end cap of the detector in an accurately reproducible beaker–detector geometry (Park and Jeon, 1995; Dryak et al., 1989). The radioactivity of the samples can be measured if the detector has been calibrated with Marinelli-beaker standard sources (MBSS) of the same dimensions, density and chemical composition. The calibration procedure should follow as closely as possible that defined in the IEC Standard 697 (1981). This method enables the simultaneous detection of several γ-emitters present in the sample matrix without the need for separation of the radionuclides from the matrix. The method can be applied to a large variety of environmental and biological materials such as air, water, soil, sediments, vegetation (grass, hay, etc.) and, particularly, individual foods of vegetable and animal origin as well as total diet mixtures. This method is suitable for the surveillance and monitoring of radioactivity originating from the operation of nuclear plants, nuclear weapons tests and releases from nuclear accidents (Kanish et al., 1984). It is recommended that the γ-spectrometry system should be a fully integrated data acquisition and computation system comprising the following items (IAEA 295, 1989): •
A vertical, high-purity germanium (HPGe) detector is recommended. The detector should have an efficiency of 18%–20%. Generally, the efficiency of germanium detectors is specified as the photo peak efficiency relative to that of a standard 7.62 7.62 cm2 cylindrical NaI(Tl) scintillation crystal and is normally based on the measurement of the 1.33 MeV γ-ray photopeak of a 60 Co source with a source-detector spacing of 25 cm used in both measurement systems. The resolution of the detector which is normally specified for germanium detectors as the full-width (in keV) at half-maximum
181
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CHAPTER 4 Measurements of radioactivity
•
•
• •
•
•
• •
(FWHM) of the full-energy peak of the 1.33 MeV peak of 60Co should be between 1.8 and 2.2 keV. It is recommended that the peak-to-Compton ratio of the detector be >46:1. The peak-to-Compton ratio is defined as the ratio of the count in the highest photopeak channel to the count in a typical channel just below the associated Compton edge and is conventionally quoted for the 1.33 MeV γ-ray photopeak of 60Co. A preamplifier is necessary. This is normally an integral part of the detector unit and it is located very near the detector in order to take advantage of the cooling which is necessary for the operation of the detector and which helps the preamplifier to operate with low noise. A biased high-voltage power supply is required to supply high voltage to the detector through the preamplifier. A power supply of 1500–5000 V is adequate for the operation of germanium detectors. A linear amplifier is required to process the output signals from the preamplifier. A detector shield will be needed with a cavity which is able to accommodate large (up to 41) samples, constructed of either lead or steel with some type of graded line to degrade X-rays. Lead shields have a much lower backscatter effect than steel shields. Typically, lead shields have walls 5–10 cm thick, lined inside with graded absorbers made of cadmium (l–6 mm) and copper (0– 4 mm). Other materials, such as Plexiglas and aluminium, are also used as absorbers. A multichannel analyser (MCA) with a minimum of 4096 channels should be connected to a keyboard and display screen for input and output of data and interaction with a computer. Several kits are available for the conversion of personal computers (PCs) into MCAs. Basically there are three types of conversion kits. One makes use of board with an analogue-to-digital converter (ADC) that simply clips into the PC; a second type uses a clip-in board with an external ADC and the third type uses a multichannel buffer (MCB) connected to the PC. All of these PC-based MCA systems are relatively inexpensive and are very suitable for use in germanium and sodium iodide γ-ray spectrometry. A rapid data-storage and recovery system is needed. It can consist of magnetic tape, hard disk, floppy disk, or a combination of these media. This system can be used for programming, short-term storage of data, and archiving data. A high-speed printer is required for data output. Useful, but not absolutely necessary, is a plotter for archiving spectral drawings. Software for system operation and data reduction is usually supplied with the MCA system. Software packages with varying features and capabilities are available for MCAs based on PCs.
Several aspects of γ-ray spectrometry with such a system deserve some discussion. Interferences associated with γ determinations may be caused by improper spectral identities, changes in background, errors in calibration and/or geometry and lack of homogeneity in samples. Many of the problems in γ-ray spectrometry are due to malfunctions of
4.3 Radiometric methods
electronic components. Very important also is the calibration of the measuring systems; both energy and efficiency calibration should be performed with care. Energy calibration of a germanium detector system (i.e. establishing the channel number of the MCA in relation to a γ-ray energy) is achieved by measuring mixed standard sources of known radionuclides having well-defined energies within the energy range of interest, usually 60 to 2000 keV (IAEA-619, 1991). The use of the lower energy photons emitted by 241Am may indicate changes in the intercept. Mixed γ-ray standards are available in various forms and containers from reliable suppliers. A partial list of radionuclides usually available with γ-ray energies in the range of interest includes: 241Am, 109Cd and 57Co, 139Ce, 51Cr, 22Na, 54Mn and 60Co. The energy calibration source should contain a selection of radionuclides with at least four different γ-ray energies. It is recommended that one of the nuclides should be 137Cs. The gain of the system should be adjusted so as to position the 662 keV photopeak of 137Cs at about one-third full scale. It is also recommended that the gain of the system be adjusted to 0.5 keV/channel. Once these adjustments are made, the gain of the system should remain fixed. As an example we show energy calibration with a point source incorporating 60 Co, 133Ba and 137Cs. The peaks corresponding to the following energies are present in the measured spectra: 81.00, 301.85, 356.00, 661.65, 1173.24 and 1332.50 keV (see Fig. 4.39). The energy calibration, represented by a quadratic equation: E ¼ a + bX1 + cX2
(4.52)
is determined (from the six pair data of peak centre channel (X) and the energy (E)) by using least square method for determination of the coefficient (a, b and c).
6
1332.50
Counts
661.65
4
10
1173.24
5
302.85 356.00
10
81.00
10
103 10
2
101 0
10
0
1000
2000 Pulse height/Channel
FIG. 4.39 Gamma-ray spectrum of an energy calibration source.
3000
4000
183
184
CHAPTER 4 Measurements of radioactivity
An accurate calibration of the efficiency of the system is necessary to quantify the radionuclides present in a sample. It is essential that this calibration be performed with great care because the accuracy of all quantitative results will depend on it. It is also essential that all system settings and adjustments be made prior to determining the efficiencies and should be maintained until new calibration is undertaken. Small changes in the settings of the system components may have slight but direct effects on the counting efficiency. In general, it is not so easy to obtain the absolute peak efficiency ε(E, x) which is a function of both γ-ray energy (E) and geometry (x) between source and detector. The following method also includes an approximation that the energy dependency of the peak efficiency is scarcely affected by geometry, and the peak efficiency consists of two components being independent of each other. εðE, xÞ ¼ f ðEÞ εðxÞ
(4.53)
where, f(E) represents the energy dependency, which was normalized at the specific energy and is often called “relative peak efficiency”. ε(x) gives the absolute peak efficiency that was obtained for various geometries using the radioactivity standard sources emitting γ-rays of the specific energy. The experimental procedure for determination of peak efficiency vs energy curve involves the following steps: •
• •
•
Place a standard source (51Cr, 54Mn, 57Co, 60Co, 85Sr, 88Y, 109Cd, 137Cs, 139Ce) at the position of 2 cm apart from the top face of the Ge detector using a source holder. Fig. 4.40 shows the γ-ray spectrum of such a standard source. Accumulate the spectrum for a time period (t) long enough to reduce the counting error; approximately 1000 s is recommended. After the measurement has finished, the spectrum shown in Fig. 4.40 is accumulated (integral net count of peak region) for each peak corresponding to the energy: 88, 122, 136, 166, 320, 514, 662, 835, 898, 1173, 1332 and 1836 keV. Calculate the peak efficiencies for all peaks using the following equation Peak efficiency : ε ¼
peak area=t 100% aA
(4.54)
where a is a fraction of γ-ray emission per disintegration, and A is the present radioactivity (Bq). •
Plot data of ε(%) vs E(keV) on log–log graph paper, then draw a smooth curve through all points. The result of such work is shown in Fig. 4.41, as an example.
The equation for the efficiency can be written in the form: h i f ðEÞ ¼ exp a + b ln ðEÞ + c ln ðEÞ2
(4.55)
Quantitative γ-ray spectrum analysis of environmental samples such as soil, water or ash of food requires the peak efficiency for volume sample. A Marinelli-beaker is often used as a container for a large quantity of sample such as water or soil. As
f (E)
Log / Counts/Channel 5
4
0.0
0.1 0.05
3
0.5
0.1
0.2
FIG. 4.41
Relative peak efficiency as a function of energy. 898.02 keV, γ-BS
0.5 Energy (MeV)
1.0 Gamma ray energy (MeV)
1
1836.04 keV, γ-88
S.E.(1836 keV) 1332.47 keV, Co-60
1173.21 keV, Co-50
D.E(1836 keV)
661.65 keV, Gs-137
511.00 keV, γ-rays 513.99keV, Sr-85
391.69 keV, Sn-113
255.12 keV, Sn-113 279.19 keV, Hg-203
83.032 keV, Cd-109 122.063 keV, Co-57 136.47 keV, Co-57 165.855 keV, Ce-139
4.3 Radiometric methods
6
2
1
1.5 2.0
FIG. 4.40
γ-Ray spectrum of the standard for efficiency calibration.
9
5
2
1
0.5
0.2
2
3
185
186
CHAPTER 4 Measurements of radioactivity
the volume of sample to be measured is usually fixed, the absolute peak efficiency is dependent on energy only, but affected by self-absorption which depends on density of matrix and the elemental composition. The method to be used for determination of the peak efficiency is based on experimental procedures involving the following steps: First, the Q-peak efficiency for an aqueous sample is determined by: •
placing an aqueous standard sample (51Cr, 54Mn, 57Co, 60Co, 85Sr, 88Y, 109Cd, Cs, 139Ce) in the Marinelli-beaker with the Ge detector; accumulating the spectrum for a time period (t) long enough to reduce the counting error (approximately 1000 s of preset time is recommended); after the measurement has finished, one should analyse the spectrum. Read the peak area for each following peak from the printed data 88, 122, 136, 166, 320, 514, 662, 835, 898, 1173, 1332 and 1836 keV and calculate the observed peak efficiency
137
• •
•
Observed efficiency : ε ¼
peak area=t 100% aA
(4.56)
where a is a fraction of γ-ray emission per disintegration and A is the present radioactivity (Bq) of each nuclide. Let the peak efficiency for 662 keV be ε1. Second, the Q-peak efficiency for a heavy liquid sample is determined by: •
• • •
Replacing the above sample with a heavy liquid sample (ZnBr2 in water; 80% in weight; ρ ¼ 2.5 g/cm3; weight composition: Zn(0.232), Br(0.568), H(0.02), 0 (0.178)) in a Marinelli-beaker of the same size. Then, measure the spectrum for approximately 1000 s. After the measurement has finished, analyse the spectrum, read the peak area for 662 keV from the printed data, then calculate the peak efficiency. Peak efficiency should be corrected by self-absorption. Determine the self-absorption correction factor (K) by Eq. (4.57). K¼
1 ε1 ln μ2 ρ2 μ1 ρ1 ε2
(4.57)
where ρ is density (g/cm3) of matrix: ρ1 ¼ 1.0, ρ2 ¼ 2.5; μ is the mass attenuation coefficient; μ1 ¼ μ of water for 662 keV ¼ 0.0857 cm2/g and μ2 ¼ μ of heavy liquid for 662 keV ¼ 0.0742 cm2/g. • Calculate the peak efficiency ε* without self-absorption for each energy: ε∗ ¼ ε=eKμρ
where μ is mass attenuation coefficient (cm2/g) of water and ρ is density of water (¼1). • Plot the data of ε* vs E(keV) on log–log graph paper, then draw a smooth curve through all points.
4.3 Radiometric methods
• •
From the curve, read the efficiency for each energy shown below as exactly as possible. The coefficients (a, b and c) of absolute peak efficiency without selfabsorption can be calculated separately for two regions of energy: h i ε∗ ðEÞ ¼ exp a + b ln ðEÞ + c ln ðEÞ2
•
(4.58)
Thus, the peak efficiency for a sample with linear attenuation coefficient (μρ) is represented by the following equation: εðEÞ ¼ ε∗ ðEÞ eKμρ
(4.59)
Most laboratories involved in radiation measurements now use personal computers and commercially available software for the analysis of γ-ray spectra. Some of these programs allow the user to control the multichannel analyser (MCA), calibrate the detector for various geometries, and provide analysis results. The programs are easy to use and do not require the user to be an expert in γ-ray spectrometry. As an example we refer to work by Heimlich et al. (1989). In their work, the γ-ray spectrum analysis program, GAMANAL, has been modified to operate on a small computer. The program uses an algorithm involving a Gaussian and a tailing term for fitting and resolving peaks obtained from spectrometers using germanium detectors. γ-ray energies, intensities and absolute photon emission rates can be determined. A graphical output showing the original and fitted data can also be obtained. The results generated by the program are stored on disk for further analysis. This allows the use of other computer programs and languages in tasks such as decay curve analysis, radionuclide activity measurements and neutron activation analysis. Sanderson (1989) evaluated commercially available IBM PC-compatible software in 1987. At that time, it was reported that most of the programs satisfactorily detected peaks and resolved doublets of equal intensity. Problems arose when the doublets were of unequal intensity or the analysis of a complex spectrum was needed. The suppliers of the programs involved in that study have corrected some of these deficiencies. Since many of these programs have undergone numerous revisions, and a few programs have become available, a re-evaluation was performed (Decker and Sanderson, 1992). Six programs were evaluated in this study: GAMMA-W (Gesellschaft f€ur Kernspektrometrie, Germany), INTERGAMMA (Intertechnique Instrumentation Nucleaire, France), QSA/Plus (Aptec Nuclear, Inc., Canada), OMNIGAM (EG&G Ortec, USA), GDR (Quantum Technology, Inc., USA) and SAMPO 90 (Canberra Nuclear Products Group). Hardware requirements were similar for all the programs tested (IBM PC compatibility, 384–640 K of memory, a hard-disk drive). Two programs, QSA/Plus and SAMPO 90, also required a maths coprocessor. The QSA/Plus program had to be installed in Windows 3.0. All of the other programs operated directly from DOS. Except for GAMMA-W, all the programs control data acquisition. GAMMA-W, which is written by an independent company that does not manufacture nuclear
187
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CHAPTER 4 Measurements of radioactivity
instrumentation, does not allow the user to read 10 different formats of γ-ray spectra. The conclusions reached by the authors were as follows. All of the programs satisfactorily found small peaks using sensitivity values recommended in the program manuals. However, these values may not be optimal for every situation. When the sensitivity values were lowered, additional valid peaks were found. When the recommended sensitivity values were used, only two programs did not report any false peaks. All of the programs were able to resolve equal-intensity doublet peaks with only a 2 keV (four-channel) separation. The resolution of doublets of unequal intensity, especially where the smaller is on the high-energy side of the predominant one, has improved since the last evaluation. However, some programs still require improvement in this area. All the programs for analysis of γ-ray spectra measured with germanium detectors generally allow the following steps (among others) to be carried out: a. b. c. d. e.
transfer of spectra from the MCA to the computer; search for and identification of photopeaks; energy and efficiency calibrations; resolution of multiplets and calculation of activities and the dosimetric magnitudes deriving from them.
Most of these programs, however, are not specifically designed for the development of plans for environmental radiological monitoring, where one must measure periodically and systematically activity levels of a set of man-made γ-emitters in different types of environmental samples. Such samples often present very low counting statistics for these emitters. Moreover, since most of these programs employ elaborate algorithms for the search and fit to the shape of the γ-peaks, they are coded in FORTRAN (Koskelo et al., 1981) and are therefore implemented on mini computers or, in more reduced versions, on personal computers. Also commercially available programs are difficult for the user to modify, requiring usually a great deal of programming effort to adopt them to the peculiarities of each laboratory. Baeza et al. (1992) have developed ESPEC: a set of programmes specially designed to undertake γ-spectrometric analyses of environmental samples. The processes enumerated earlier (a–e) are carried out in a simple form. The package is designed so that it can be implemented not only on an IBM-type personal computer, but even on one with less memory and lower performance, such as a 64 KB memory microcomputer. The code is written in BASIC and it is easily modified to suit the specifications of any MCAs.
4.3.3 β-Particle spectrometry Table 4.6 illustrates β-emitting radionuclides; since β-emitters show a continuous emission spectrum, the average energy EB and the maximum energy Em B are given. Measurement uncertainties are shown in brackets after the respective value (in units of the last significant digit).
4.3 Radiometric methods
Table 4.6 Some common β emitters Nuclide
Half-life
Em B
EB (%)
3
12.35(1) a 5730(40) a 14.29(2) d 87.44(7) d 2.75(2)a 96(4) a 50.5(1) d 28.7(3) a 64.1(1) h 59.3(2) d 22.3(2) a
18.6 156.48 1710.4 167.47
5.68 49.47 695.0 48.80
65.87 1492 546.0 2284
17.13 583.1 195.8 934.8
16.5 63.0
4.15 (80) 16.13 (20)
H C 32 P 35 S 55 Fe 63 Ni 89 Sr 90 Sr 90 Y 125 I 210 Pb 14
Er
Kα 5.9, Kβ 6.5
Kα 27.4, Kβ 31.0
C Amplifer R
Scaler
Pulse shaper and discriminator Timer
High-voltage power supply Source
FIG. 4.42 Block diagram of the counting electronics associated with a G-M tube.
The classical way to measure low-level β-particle activity is with a Geiger-M€uller (GM) gas-flow counter. The block diagram of the GM counter is shown in Fig. 4.42. In anticoincidence with a guard detector the GM gives a fairly low background (0.003 s1) and a counting efficiency of 40%. The disadvantage, however, is its inability to give energy resolution. As the characteristics of the GM tube may gradually change over a long period of use, it is necessary to examine the characteristics of the tube and background count occasionally. The counting efficiency of a GM counter for detecting β-rays is defined as the ratio of the count rate to β-activity of the source. Thus, the counting efficiency is one of the most important factors to determine the characteristics of the detector. There is the following relation between activity A (Bq) and net count rate n (s1) in the case of activity measurement using the GM counter n ¼ A fg fi fsa fs fa
(4.60)
189
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CHAPTER 4 Measurements of radioactivity
where fg is the geometry factor of counting setup; fi is the intrinsic efficiency of GM tube for β-rays; fsa is the source self-absorption factor; fs is the source-mount backscattering factor and fa is the correction due to absorption between source and GM tube. Eq. (4.60) is rewritten as Eq. (4.61), if fg, fi, fsa, fs is replaced by η¼
n Af a
(4.61)
where η is called counting efficiency; it represents the rate of count rate n corrected for the adsorption fa to activity A. The counting efficiency η is obtained by using the standard source of the known activity A, and the absorbers with the known thickness dm. If the count rates for β-rays are defined as n and n0, respectively, which correspond to those passing through with and without absorber, fa is equal to n/n0, where fa represents the transmission rate of β-rays for the absorber thickness dm. When the thickness of absorber dm is not so large compared with the maximum range of β-rays, n can be represented by the experimental formula n ¼ n0 eμm dm
(4.62)
Therefore, if one plots the n values as a function of the thickness dm on semilog graph paper, the n values decrease linearly with the dm values increase. The slope of this line is μm, which is called the mass absorption coefficient. If one puts dm ¼ 0, the count rate n0 (¼ n/fa) can be obtained. Thus the equation becomes η¼
n0 A
(4.63)
If η is already known, the unknown activity A0 is obtained by A0 ¼
n00 η
(4.64)
where n0 0 is the count rate extrapolated to dm ¼ 0. Furthermore, in the case that the actual count rate n is larger than 100 s1, it is necessary for the count rate to correct the resolving time τ[s] of the GM counter. It has been shown (Holm et al., 1990) that ion-implanted detectors can be used not only for α-particle spectrometry but also for β-particle spectrometry. Their drawback, however, is the high background around 100 keV and noise below 1 keV. Olsson et al. (1992) have described a detector system for low-level β-particle spectrometry where the good characteristics of gas-flow and silicon detectors are used. The gas-flow GM counter used in the detector system is a windowless single-channel version of the GM-25-5 multicounter developed in the Riso National Laboratory, Denmark. The gas-flow counter utilizes a GM counter with a diameter of 22 mm and a guard of 80 90 10 mm3, using argon (99%)/isobutane (1%) as counting gas. Background counts, produced in the sample counter by cosmic radiation, are reduced by means of the guard and attached anticoincidence circuits. As the energy-discriminating detector, a passivated implanted silicon detector, with an
4.3 Radiometric methods
active area of 450 mm2 and a depth of 300 μm, was used. The PIPS detector has a 0.5-μm-thick aluminium coating on the front surface. The detector is placed on top of the gas-flow counter, and integrated into the gas detector unit, allowing the aluminized front surface to act as one of the ground electrodes of the sample counter (Holm et al., 1990). The source is placed in a cavity inside the detector, between the sample and guard counter. The instrumentation used for PIPS detectors is identical to that used for α-particle spectrometry. The authors have demonstrated that it is possible, by the coincidence technique, to reduce the contribution of noise and background from a PIPS detector by a factor of 10 and to improve its energy resolution. Such a detector system could be a useful tool for quality control of low level, low energy, pure β-emitters such as 63Ni from environmental samples.
4.3.4 α-Particle spectrometry Common radioisotope sources of α-particles are listed in Table 4.7. All but the first one listed are members of radioactive decay chains. Decay chains are classified into four groups according to their mass numbers. They are Th-series whose mass number is 4N (N is integer), U-series of 4N + 2, Ac-series of 4N + 3 and Np-series of 4N + 1. An Np-series does not exist naturally because the half-life of its longest-lived member is three orders of magnitude shorter than the age of the Universe. In three natural decay chains activities of each parent and succeeding daughter nuclides are equal. This condition is called radioactive equilibrium, because half-lives of parent nuclides, 232Th, 238U and 235U are much longer than those of their daughter nuclides. Table 4.7 Common α emitters Nuclide
Half-life
α-Energy/MeV (% branching)
147
1.07 1111 years 60.55 m 138.38 d 0.296 μs 55.6 s 3.824 d 1600 years 1.405 1010 years 2.45 105 years 7.038 108 years 4.468 109 years 2.413 104 years 6570 years 432 years 18.1 years
2.232(100) 6.051(72), 6.090(28) 5.304(100) 8.784(100) 6.288(99.93) 5.490(99.9) 4.602(5.5), 4.784(49.5) 3.954(23), 4.013(77) 4.723(27.5), 4.775(72.5) 4.368(12.3), 4.400(57) 4.150(23), 4.197(77) 5.105(11.5), 5.144(15.1) 5.157(73.3) 5.124(26.4), 5.168(73.5) 5.443(13.1), 5.486(85.2) 5.763(23.6), 5.805(76.4)
Sm Bi 210 Po 212 Po 220 Rn 222 Rn 226 Ra 232 Th 234 U 235 U 238 U 239 Pu 240 Pu 241 Am 244 Cm 212
191
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CHAPTER 4 Measurements of radioactivity
α-particles emitted from nuclides which decay to a single level are observed as mono-energy particles. On transitions given the branching ratio in Table 4.6, multiple α-energies are observed. Such a fine structure in the α-spectrum comes about because an α-emitter may decay to any one of several discrete energy levels of its daughter. 241Am is commonly used as a standard source. The detection of α-particles is based on the physics of the processes which take place when the particles pass through matter. During this passage the α-particle loses its energy by excitation and ionization of the atoms. The energy loss per unit path length is called stopping power(dE/dx). Stopping powers of various elements are given (Ziegler, 1977; Northcliff and Shilling, 1970). Stopping power at energy above 1 MeV is inversely proportional to α-energy. At the energy range of <1 MeV stopping power is nearly proportional to the energy. Because the velocity of α-particles is slow compared with orbital electrons, α-particles capture electrons and their effective changes for ionization decrease. When α-particles pass through matter, they make small-angle scattering. Therefore an asymmetric line spectrum which has a tail extending to the lower energy side is observed after penetrating relatively thick foils, because of asymmetry in their trajectory path. Source thickness has an important effect on the observed α-particle spectrum. As α-particles have a relatively large stopping power, the observed spectra for thick sources show some degradations caused by self-absorption. For high-resolution spectrometry, thin α sources or samples are required. They are usually prepared by electrodeposition or vacuum evaporation. With a source thickness <10 μg/ cm2, no effects due to self-absorption are observed. For such thin sources, a Gaussian-shape line spectrum whose width is limited by the detector energy resolution, is observed for a mono-energetic α source. When there is an absorber between an α source and a detector, the observed energy is reduced by the energy loss in the absorber. Absorbers are the entrance window of a detector, the covering material which prevents contamination of a source, air, and so on. In the case of a thin source, prepared by chemical separation, the emitted α-particles are observed as line spectra by spectrometers having high-energy resolution. The intensity of each α-emitter is easily estimated from the area of the corresponding peak. Even if the peak is super-imposed on the tail of another peak, the peak area can be calculated from the shape of the line spectrum. An example of this situation is plutonium isotopes, 239Pu and 240Pu. They are used for the estimation of a burn up of nuclear fuel. As the energy difference of these α-emitters is only 10 keV, the α-particle spectrum is observed as an overlapped single peak. However, when a Si detector is used, which has an energy resolution of <10 keV (FWHM), the overlapped peaks can be analysed by the least squares fitting technique. α-particle spectrometry is usually performed using Si detectors, which are especially useful for thin and small area α sources. Two types of Si detectors are commonly used; the surface-barrier type and the ion-implanted PN-junction type. As an alternative to Si detectors, a Frich-grid ionization chamber is sometimes used as a spectrometer of α-particles. This ionization chamber has a grid between a
4.3 Radiometric methods
cathode and anodes and a sample is put on the cathode electrode. Electrons and ions, which are generated between the grid and the cathode by ionization, drift towards the anode or the cathode, and a signal pulse is obtained. The pulse height obtained from the cathode depends on the emission angle of α-particles. Only the drift of electrons is observed due to their drift velocity being 1000 times greater than that of ions. However, the height of the anode pulse is proportional to the α-energy. The advantage of this counter is that the areas of samples can be made larger than with Si detectors. A commercially available counter of this type has an area larger than 1000 cm2. The energy resolution of this counter is 40–50 keV (FWHM) for α-particles. Several other types of spectrometers can be used in some specific applications. For example, an organic liquid scintillation counter is useful for detection of extremely low-level α-activities. α-emitters, chemically separated from samples, are mixed into the scintillator. The geometrical efficiency is then 100% and a relatively large amount of source can be introduced into the scintillator. The main drawbacks of the counter are its poor energy resolution and its relatively low light output in excitation by α-particles compared to that by electron. Track detectors are also often used for α-detection. The length of tracks produced by α-particles is measured with a microscope after appropriate chemical etching. The energy spectrum of α-particles is obtained from the distribution of their trace lengths. Advantages of this detector are its high sensitivity, good discrimination against β- and γ-ray backgrounds, and the acquisition of α-emitter distribution in samples. Its main drawbacks are a poor energy resolution and the time-consuming processing of the microscope reading. It should be mentioned that the energy of α-radiation is one of the most important characteristics of radionuclide sources. Knowledge of the α-particle energy is necessary for the determination of other major characteristics of external radiation of α sources, such as the flux and energy flux density, and the absorbed dose rate. Information on the α-radiation energy is also used for calibrating semiconductor α spectrometers. The results of accurate measurements of α-particle energies play an important role in the evolution of the atomic mass scale and comprise nearly 60% of the input data for atoms with A > 200, and are deciding factors in the design of precision spectrometers for the high-precision energy measurements of α-particles from radionuclide sources with the minimum attainable uncertainties. A very precise measurement of α-particle energies can be achieved by the measurement of the α-particle time of flight (Frolov, 1992). A special problem is the assay of α-particle emitters in water samples. There are several procedures in the literature for assaying α-particle-emitting radionuclides in natural waters. However, few of them are well suited for low-level analysis since they require the handling of large water samples and the application of concentration techniques. Typical water samples do not contain sufficient amounts of some important nuclides for a precise or reasonably rapid measurement. The most common concentration technique is the coprecipitation of the heavy elements by the addition of a carrier. Some of the best-known methods are coprecipitation with iron as the hydroxide and with calcium and strontium as the oxalates (Livingston et al., 1975).
193
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CHAPTER 4 Measurements of radioactivity
However, these methods present some disadvantages. These include the transportation of large volumes of water (several, to hundreds of litres) to the laboratory, the isotopic exchange of tracers, the use of large containers and laboratory ware and the difficulty of recovering the precipitate. As a consequence of these difficulties, interest in studying in situ concentration techniques with appropriate adsorbents has increased. These methods eliminate the need for preservation, storage, evaporation or coprecipitation of large water samples. One such approach has been described in detail by Crespo et al. (1992). According to the authors MnO2 and Al2O3 can be used as absorbents in the assay of the low levels of α-emitters in the waters. This technology requires more development work before it can become routine. Another problem of importance to safeguards and reactor fuel technology is the measurement of the relative abundance of plutonium isotopes. In addition to γ-and X-ray spectrometry (Gunnick, 1982), α-particle spectrometry has also been used (Bland et al., 1992; Amoudry and Burger, 1984a; Bortels and Collaers, 1987). This usually results in a complex α-spectrum, with difficulties in obtaining correct amplitudes for overlapping peaks which tail towards lower energies. The 239Pu and 240Pu peaks overlap when α-particle energies are measured using Si detectors. Some sort of deconvolution procedure is required in this case (see, for example, Bland et al., 1992). The radiation dose due to radon and its decay products in the environment amounts to more than 50% of the total radiation dose to man from natural radiation sources. Recently the risk of developing lung cancer from exposure to inhaled radon daughters has been attracting considerable attention. Thus it is very significant to measure α-activity levels in the air. Let us briefly describe a filter-sampling method used to assay the gross α-activity concentration of radon daughters in the air by means of the gross α-particle counting using ZnS(Ag) scintillation counters. An air sampler is used to collect natural radionuclides onto a filter paper by passing air through, because some nuclides of radon progeny (RaA ¼ 218Po, RaB ¼ 214Pb, RaC ¼ 214Bi, RaC0 ¼ 214Po, etc.) tend to attach to aerosol particulates floating in the air. Supposing only one radionuclide is attached to the surface of a filter paper, the rate of change with time in the total number of the radioactive atoms on the filter paper can be written as dN ¼ ξqn λN dt
(4.65)
where t is the sampling time, N is the total number of radioactive atoms on filter paper at time t, ξ is the flow rate of sample air (constant), n is the concentration of radioactive atoms in the air (assumed to be constant), λ is the decay constant of the radionuclide, ξqn is the rate of collection and λN is the rate of decay. The solution of Eq. (4.65) for the initial condition N ¼ 0 at t ¼ 0 is N¼
ξqn 1 eλt λ
(4.66)
4.3 Radiometric methods
The radioactivity A is given by λN, or
A ¼ ξqn 1 eλt
(4.67)
The counting rate R of α-particles emitted from the filter paper at time t is related to the radioactivity A: R ¼ εζA
4.68)
where ε is the counting efficiency of α-ray counter and ζ is the emerging efficiency of filter paper. From Eqs (4.67) and (4.68) the concentration n of the α-emitting nuclide in the air is obtained n¼
R εζξqð1 eλt Þ
(4.69)
The α-activity concentration Q in the air is given by λn or Q¼
λR R 0:693 1 ¼¼ εζξqð1 eλt Þ εζξq T 1 ð1=2Þt=T
(4.70)
where T (¼0.693/λ) represents the (α-decay) half-life of the radionuclide. For longlived radionuclides such as 238U, 239Pu and 226Ra, Eqs ((4.66), (4.67) and (4.70), respectively, become N ¼ ξqnt
(4.71)
A ¼ λξqnt
(4.72)
R : εζξqt
(4.73)
Q¼
4.3.5 Liquid scintillation measurement method LSC has been accepted as the generally preferred method of counting weak β-emitters and is useful to a lesser extent for α- and γ-emitters. The counting sample consists of three components, the radioactive material, an organic solvent or solvent mixture and one or more organic phosphors. A particle or radiation emitted by the sample material is absorbed in, and its energy transferred to the solvent and then to the phosphor which emits a scintillation of light photons. These photons are absorbed by the photocathode of a photomultiplier tube which converts them into an electronic pulse. The pulse, after suitable amplification, is registered as a count corresponding to the emission of the particle or radiation (Neame and Howewood, 1974). Configuration of a typical LSC is shown in Fig. 4.43. The technique has the following characteristics: 1. Self-absorption is usually negligible. 2. There is no absorption of radiation by air or a detector’s window between the radioactive sample and the sensitive region of the detector.
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Photomultiplier tube
Photomultiplier tube
Sample
High voltage
Coincidence
Pneumatic
196
Pump
Summation Amplifier Gate Analyzer Data output
Source external standard
FIG. 4.43 Configuration of a typical liquid scintillation counter.
3. There is no radiation scattering prior to incidence on the detector. 4. 4π counting is performed, because the radioactive material is completely surrounded by the liquid scintillator. Based on the above merits, the liquid scintillation measurement is extremely sensitive to the low-level radioactivity existing in the environment and food. Since the radioactive sample material from most methods of sample preparation is in intimate contact or in actual solution with the phosphor, the detection of emitted particles or radiation is highly efficient and may even approach 100%. Problems of self-absorption of the emissions are thus absent, or considerably smaller than those associated with planchette counting of solid samples. This is of particular importance for the measurement of low-energy β-emitters such as tritium and carbon-14. On the other hand, the measurement method has intrinsic drawbacks such as quenching and chemiluminescence.
4.3 Radiometric methods
Currently, the liquid scintillation counter has been employed not only for the measurement of low-energy β-emitters, but also for pure β, β–γ and α-emitters and further Cherenkov radiation. The liquid scintillator consists mainly of organic solvent and fluorescent material (i.e. solute), and sometimes a surfactant or other material is added to the solution. The characteristics of the liquid scintillator depend mostly on the sort and amount of these chemicals. The liquid scintillator plays the role of an energy transducer, converting radiation energy into photons. The organic solvent which comprises most of the liquid scintillator should satisfy the following conditions (Neame and Howewood, 1974): 1. The energy should efficiently transfer in the process of luminescence. 2. An absorption spectrum of solvent should never overlap an emission spectrum of solute. 3. The radioactive sample and solute must be able to be incorporated with the solvent. 4. The solvent must be of high purity. In the past many kinds of chemicals were employed as solvents, but nowadays only a few typical solvents are being used including toluene, xylene, pseudocumene and droxane. The characteristics of typical solvents used for liquid scintillators are shown in Table 4.8. The solute is classified into a primary solute and a secondary one. The former is a main fluorescent material, and the latter serves as a wavelength shifter which gives rise to an emission spectrum having long wavelength. These characteristics and molecular structures are respectively represented in Table 4.9. The sample to be measured is easily prepared by incorporating the radioactive sample into the liquid scintillator such as xylene-base-, toluene-base- or emulsive scintillator. The xylene (or toluene)-base-scintillator is appropriate for hydrophobic samples, to form a homogeneous solution which provides efficient energy transfer and light-counting efficiency. Table 4.8 Typical solvents used for liquid scintillator
Solvent
Molecular weight
Solidifying point (°C)
Absorption spectrum (A) λmax a
Toluene Xylene Pseudocumene Dioxane
92.13 106.16 120 88.1
95 20 60.5 12
2.620 2.660 2.690 1.880
a
Wavelength giving maximum value. Wavelength giving mean value.
b
Emission spectrum (A) λmax a λmean b
Relative pulse height
2.870 2.890 2.930 2.470
1.00 1.09 — 0.65
2.840 2.880 2.900 —
197
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Table 4.9 Typical solutes used for liquid scintillator Absorption spectrum λmax
Emission spectrum
Solute
a
λmax a
λmean b
Optimum concentration (g/L)
Primary solute PPO Secondary solute DMPOPOP bis-MSB
3.030
3.640
3.703
4–7
1.6
3.630
4.290
4.273
0.2–0.5
1.5
3.470
4.120
4.219
0.5
1.3
Decay time of scintillation
a
Wavelength giving maximum value. Wavelength giving mean value.
b
The simplest mixture consists of a primary scintillator and a secondary scintillator dissolved in a primary solvent. The primary solvent converts the kinetic energy of radiations into excitation energy. The primary solvents are usually aromatic hydrocarbons, and therefore only nonpolar radioactive materials can be dissolved in them. The most widely used primary solvent is toluene. Others are p-, and m- and mixed xylenes. Pseudocumene (1,2,4-trimethylbenzene) is becoming a popular solvent for new, commercially produced scintillation cocktails. It offers the highest energy conversion efficiency of the solvents known and has fewer restrictions on shipping and storage as a combustible liquid because of its high flashpoint, low volatility and lower toxicity. Owing to the presence of organic solvent, the mechanism of luminescence for the liquid scintillator is much more complicated than that for a crystal scintillator. The energy successively transfers to generate finally the luminescence, as follows: 1. 2. 3. 4.
excitation of solvent molecule due to the absorption of radiation energy; solvent–solvent energy migration; solvent–solute energy transfer and luminescence from solute molecule.
Since the solvent molecules have a majority in the liquid scintillator, they are initially excited by the radiation energy. The π electron of the solvent molecule plays an important role in the process of energy transfer due to its active mobility. When the solute absorbs the excitation energy liberated from the solvent, the solute molecules in the ground state are excited up to the excited electron level or its vibrational one. Then, the following processes compete with one another as intramolecular behaviours of the solute: 1. 2. 3. 4.
internal conversion of molecule, fluorescence emission, intersystem crossing and phosphorescence emission.
4.3 Radiometric methods
The quenching is a phenomenon caused by the energy loss in the process of the energy transfer inside the liquid scintillator. It should be taken into account, whenever the liquid scintillation measurement is performed. Owing to the existence of a quencher, which means the material giving rise to the quenching, the counting efficiency finally decreases as follows: 1. Energy is partly lost during the energy transfer. 2. The number of photons emitted from the solute decreases, and/or some of the photons are wastefully absorbed in solution. 3. The height of electric pulse created by a photomultiplier tube becomes lower. 4. A pulse-height distribution shifts towards a lower pulse height. 5. The number of pulses which are present within a countable region decreases, leading to the lowering of the counting efficiency. Most materials are regarded as quenchers whose quenching strength depends on the material itself and its amount. The quenching and optical (or colour) quenching are practically significant. The impurity quenching is the quenching which arises in the process where the excitation energy of the solvent is provided to the solute. Its mechanism can be explained by the theory of an exciplex which means the excited complex formed by two different kinds of molecules. Final energy loss is represented by the following scheme: 1
M∗ + Q ! M + Q ∗ ! M + Q ðdislocationÞ
(4.74)
M + Q ! 1 M + Qðtransition to ground stateÞ ðthrough intersystem crossingÞ
(4.75)
exciplex
3
where 1M is the solvent molecule in singlet state, 3M is the solvent molecule in triplet state, Q is the quencher molecule and * is the excited state. When the material whose absorption spectrum overlaps, more or less, the emission spectrum of the solute is contained in a prepared sample, the fluorescence generating from the solute is partly absorbed to cause the optical quenching. The strength of the optical quenching depends on the sort and density of colours. In general, red and yellow cause strong quenching, but blue causes weak. Even though a quencher is colourless, the optical quenching occurs if the quencher has an absorption spectrum which partly overlaps an emission spectrum of solute. The liquid scintillation measurement method is a quantitative technique to determine radioactivity, but not a qualitative one to find radiation energy or the sort of nuclide. Based on the internal-sample counting, the measurement method has great merits for low-level radioactivity. The following three measurement methods are in general often used.
4.3.5.1 External standard method The index of quenching with respect to sample is found from Compton spectrum created by γ-ray irradiation to the sample. A set of quenched standards are employed for the construction of a quenching correction curve which represents the relationship
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between the index of quenching and the counting efficiency. The radioactivity of the sample to be measured can be determined by using the quenching correction curve.
4.3.5.2 Sample channel ratio method The index of quenching in this method is obtained from an areal ratio or a centre of gravity with respect to a β-ray spectrum. The radioactivity is determined through a similar quenching correction curve to that of the external standard method.
4.3.5.3 Automatic efficiency tracing method As a matter of fact, the foregoing two methods are confined to the measurements of only tritium and carbon-14. The automatic efficiency tracing method, however, makes it possible to determine the radioactivity of many sorts of radionuclides such as pure β-, β–γ and α-emitters. The radioactivity can be found from a quadratic regression equation which is constructed by the relationship between the counting efficiencies of a standard sample and the counting rates of a sample to be measured. Sample preparation techniques involve the chemical separation of the specific chemical forms of a radionuclide, and the preparation of the counting sample by mixing the separated nuclides with the proper liquid scintillator. A large number of recipes for scintillation mixtures have been published, but for many purposes a few simple ones will suffice. The properties of a suitable mixture are as follows: 1. It should generally be clear, colourless and homogeneous after addition of the radioactive sample, although in certain cases suspensions or gels may be satisfactory. 2. It should quench as little as possible. This is particularly important when measuring radioisotopes such as tritium whose β-particles are of low energy. 3. It should not be expensive. 4. Its constituents should be stable. Some scintillators are known to be unstable when exposed to light and other materials may deteriorate on storage, with the formation of impurities which cause quenching or chemiluminescence. Scintillation mixtures should be stored in dark bottles. The simplest mixture consists of a primary scintillator and a secondary scintillator dissolved in a primary solvent. The primary solvent converts the kinetic energy of radiations into excitation energy. Primary solvents are usually aromatic hydrocarbons, and therefore only nonpolar radioactive materials can be dissolved in them. The most widely used primary solvent is toluene. Other primary solvents are p-, m- and mixed xylene. Pseudocumene (1,2,4-trimethyl benzene) is becoming a popular solvent for new, commercially produced scintillation cocktails. It offers the highest energy conversion efficiency of the solvents known, and has fewer restrictions on shipping and storage of a combustible liquid because of its high flash point, low volatility and lower toxicity. The primary scintillator converts excitation energy into light. Two commonly used primary scintillators are PPO (2,5-diphenyl-oxazole) and butyl-PBD (2-(40 -
4.3 Radiometric methods
tert-butyl-phenyl)-5-(40 -biphenyl)-1,3,4-oxadiazole). The pulse voltage produced by butyl-PBD is about 20% greater than that produced by PPO, but its usefulness is reduced because of its limited solubility. Suitable concentrations of PPO and butyl-PBD are about 5 and 7 g/L, respectively. With alkaline samples butyl-PBD produces a brownish colour in the scintillation cocktail. In highly quenched samples, a higher concentration may be required for optimum efficiency. The secondary scintillator shifts light wavelength and may be needed in the mixture if the emission wavelength of the primary scintillator does not match the wavelength to which the photomultiplier is most sensitive. Commonly used secondary scintillators are POPOP (1,4-bis-(2-(5-phenyloxazolyl))-benzene) and dimethylPOPOP (1,4-bis-(2-(4-methyl-5-phenyloxazolyl))-benzene), used at concentrations varying from 0.05 to 0.5 g/L (usually 0.1 g/L). POPOP is the less soluble of the two but is slightly cheaper. Both should be given adequate time to dissolve completely, preferably overnight. The best general purpose secondary scintillator is bis-MSB (p-bis-(o-methylstyryl)-benzene). Bis-MSB is readily soluble in solvents and has a fast rate of dissolution. Bis-MSB is used at 0.5–1.5 g/L and does not react chemically with most liquid scintillation samples, while dimethyl-POPOP can react with acids to produce a yellow to greenish colour. A suitable mixture for toluene-soluble materials is toluene containing 5 g PPO/L or 7 g butyl-PBD/L 0.1 g POPOP or dimethyl-POPOP/L. This mixture cannot be used for measuring the radioactivity in aqueous samples since water is immiscible with toluene. A further solvent (secondary solvent) miscible with both water and primary solvent must be added to this mixture to enable water to be incorporated. This amount of aqueous solution which can be incorporated in a particular mixture without phase separation depends on the identity of the secondary solvent, the relative proportions of primary and secondary solvents and the temperature. The secondary solvents used for this purpose are a surface-active agent or emulsifiers. Triton X-100 and Triton N-101 are nonionic surfactants commonly used in laboratory-prepared scintillation cocktails. Anionic surfactants give better sampleholding capacities than the nonionic surfactants for certain types of salt solutions. As a general rule, the greater the ratio of secondary to primary solvent, the greater the amount of water which can be incorporated, but the extra secondary solvents and the extra water increase the quenching. The properties of a good emulsifier scintillator are as follows: 1. Mixture of scintillator and water should show high counting efficiency and a low background counting rate. 2. It should show high water capacity. 3. The region of phase separation between the homogeneous liquid region and the stable emulsion or gel region should be narrow. 4. It should produce minimum chemiluminescence. 5. Mixture of aqueous solution should be stable for a long period. Mixtures having water content between 10 and 20 vol% should be avoided because they are unstable and show two phase separation. Table 4.10 shows some useful mixtures used in liquid scintillation measurements.
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Table 4.10 Some useful mixtures used in liquid scintillation measurements Mixture 2 PPO DMPOP Toluene Triton X-100 Conc. HNO3
Mixture 3 6 g/L 04 g/L 667 mL 333 mL 10 μL
PPO bis-MSB Xylene Triton N101 Conc. HNO3
6 g/L 0.5 g/L 700 mL 300 mL 10 μL
The scintillation vial is the container for the analyte and the scintillator and permits light transfer from the scintillation cocktail to the photomultiplier tubes. Many materials have been used for scintillation vials. Low-potassium, borosilicate glass is the best material to meet the requirements for most LSC. It is nonpermeable and nonreactive with chemicals used in LSC. It has good optical clarity for light transmission. The glass should be selected for low radioactivity content, providing a low background counting rate. Polyethylene vials are popular because of their low cost. They show a low background counting rate and good counting efficiency. The opacity of polyethylene vials makes it difficult to detect phase separation and precipitation in the sample mixture. The major disadvantage of polyethylene vials is that they are permeable to aromatic hydrocarbons. As the primary solvent permeates into polyethylene, it carries the scintillators with it, changing the background counting rate and affecting the efficiency. The radioactivity may also be carried into the vial wall leading to geometry and self-absorption losses. Other materials for vials are quartz and Teflon. They are the best materials, but they are very expensive to make and used only in applications requiring very low backgrounds and long counting times. When one wants to measure low-level radionuclides with good precision and accuracy, one has to adjust the counter settings properly and to select the most suitable scintillators and vials to ensure the highest sensitivity and reproducibility. For these purposes, the “Figure of Merit” is adopted as a quantitative criterion: Figure of Merit ðFMÞ ¼ ðEMÞ2 =B,
(4.76)
where E is the counting efficiency (%), M is the mass (g) of a sample introduced in the vial and B is a background counting rate (cpm). The performances of several emulsifier scintillators prepared in a laboratory and available in the market show no significant difference in sensitivity. Some of the radionuclides commonly distributed in environmental samples include 3H, 14C, 6CO, 89Sr and 90Sr. Tritium and 14C emit low-energy βs which are efficiently counted by LSC. Here we present some details of sample preparation for LSC. Aqueous samples containing tritium are distilled to eliminate impurities and mixed with an emulsifier scintillator with the proportion of 40–50 vol% water. Tritium in biological samples exists as tissue free water tritium (TFWT) and organically bound tritium (OBT). Water containing TFWT is obtained by vacuum distillation (lyophilization) or azeotropic distillation using organic solvents such as toluene
4.3 Radiometric methods
and benzene. The OBT is converted into HTO for counting by combusting dried samples. The tritium concentration in the natural environment is less than 10 Bq L1 and sometimes less than 1 Bq L1. Only low-background-type LSCs detect these low tritium levels. To measure tritium below 10 Bq L1 with good accuracy, using the conventional type LSCs, it is necessary to enrich the tritium in a large water sample by electrolysis. This procedure can increase the tritium concentration 20–30-fold. Carbon-14 exists mostly as carbon dioxide (CO2) and as organic compounds in the environment. CO2 in the atmosphere or in the effluent gas from a combustion system for biological samples is first absorbed in alkaline solutions such as aqueous NaOH or NH4OH. If necessary, calcium chloride or barium chloride is added to the alkaline solution to precipitate CaCO3 or BaCO3, which are then purified and stored in a sealed bottle for future analysis. The alkaline solution containing CO2 or carbonate is acidified by titrating with a strong acid solution to generate CO2. The CO2 gas is then absorbed in a solution of organic amine absorber and liquid scintillator mixture (1:1). The procedure for 60Co measurement is as follows: after adding cobalt salt as carrier to seawater, 60Co is first precipitated as cobalt hydroxide in alkaline solution. The precipitate is separated from seawater and dissolved in HCl solution. 60Co is separated and recovered in the effluent from an anion-exchange column. The effluent containing 60Co is evaporated to dryness. The residue is redissolved in dil. HCl solution and transferred into a counting vial. The emulsifier scintillator is added into the vial and mixed well. The counting sample is counted by means of a low background LSC. This technique can be applied to 60Co in biological samples by using an ashing procedure beforehand. Strontium isotopes can be measured by utilizing the fact that high-energy β-rays in aqueous solution emit Cerenkov photons which can be counted by photomultipliers in a LSC without any phosphors. Strontium radioisotopes in a seawater sample are separated and purified by precipitation of strontium carbonate and subsequently strontium oxalate precipitate followed by the cation exchange. Yttrium-90, the daughter nuclide of 90Sr, is scavenged from the solution by Fe(OH)3 coprecipitation. The acidified aqueous solution containing strontium isotopes is transferred into a counting vial and counted after the total volume of the solution is adjusted by adding distilled water. By setting the lower level of the discriminator properly, the LSC can count only Cerenkov light signals produced by 1.49 MeV β-rays of 89Sr, without counting the 0.546 MeV β-rays of 90Sr. 90Sr is determined by counting Cerenkov photons generated by 2.28 MeV β-rays of 90Y which grows into the solution after separation.
4.3.6 Radiochemical analysis 4.3.6.1 Introduction There is a considerable amount of attention focused on long-lived artificial radionuclides such as 90Sr and 137Cs released into natural environments. Their inventories began to grow as a result of nuclear weapon testing programs since 1950 and the
203
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CHAPTER 4 Measurements of radioactivity
nuclear power reactor accident of Chernobyl in 1986. The new sources contaminating the environment with radionuclides are recently thought to be liquid effluents from nuclear power stations and nuclear fuel reprocessing facilities. The long-lived fission products and their characteristics are given in Table 4.11 (World Health Organisation, 1966; Sugihara, 1961; Norton, 1967; Chase and Rabinowitz, 1967; Harley, 1972; Greenberg et al., 1981). It is not practicable to present here all the methods used; we outline only the reliable methods used for 89Sr and 90Sr for tritium and for actinides (for details, see IAEA-295, 1989).
4.3.6.2 Analysis of strontium Strontium-90 is one of the most important fission products because of its relatively high yield (about 6%), long physical half-life (29 years) and its uptake and retention in biological systems. For assessing the integrated exposure to large populations, not only the direct measurements of biological materials must be done, but also the monitoring of the transportation of the nuclide in the environment, e.g., in oceans and streams. The assay of low-level strontium-90 in biological and radiotoxicological samples requires time-consuming and laborious techniques because both the nuclide 90Sr and its daughter 90Y are pure β-ray emitters. Therefore, the radiostrontium must be completely separated from other radionuclides prior to the β-ray counting. Table 4.11 Long-lived fission products Nuclides
Decay mode
Half-life (years)
79
β β (γ) β β β β β (γ) β β (γ) β (γ) β (γ) β (γ) β (γ) β (γ) β β (γ) β (γ) β (γ)
6.5 104 10.7 4.8 1010 28.8 1.5 106 2.14 105 1.00 6.5 106 14.6 55 2.7 1 105 1.57 107 2.06 3 106 30.17 2.62 4.9
Se Kr 97 Rb 90 Sr 93 Zr 99 Tc 106 RU 107 Pd 113m Cd 121m Sn 125 Sb 126 Sn 129 I 134 Cs 135 Cs 137 Cs 147 Pm 155 EU 95
4.3 Radiometric methods
The most commonly used method for separating strontium is by nitrate precipitation. With some modifications this method can be applied to all kinds of environmental samples and foods. The chemical yield varies according to the type of material. The use of 85Sr tracer to determine chemical yield is a general practice. When determining yield in this manner, it is important that the tracer is pure 85Sr, i.e., free from 89Sr and 90Sr. Although the method is time consuming, it is reliable and safe. Rapid methods for 90 Sr analysis exist, and it has been shown that they can be used after short-lived nuclides have decayed. In fresh fallout situations, the nitrate precipitation method has been shown to be more reliable, also, during periods of fresh fallout, the amount of 89Sr is of interest and the rapid methods can only analyse for 90Sr. In the case of higher contamination with 90Sr, the daughter 90Y can be separated without waiting for equilibrium. Within 10 h the activity concentration of 90Y will be approximately 10% of the equilibrium value and may be sufficient for a reliable 90Sr analysis (IAEA-295, 1989). A special application of liquid scintillation counters is in the measurement of Cerenkov radiation produced by β-emitters with β-energies >260 keV. This application can be used for screening samples for 90Sr (Carmon and Dyer, 1986; Eakins et al., 1986; IAEA-118, 1970; Johns et al., 1979; Kleinberg and Smith, 1982; Krieger and Whittaker, 1980; Regan and Tyler, 1976; Szabo, 1982). An outline of the method used for determination of radiostrontium in various materials by nitrate precipitation is as follows. The ashed material is dissolved in nitric acid in the presence of strontium and barium carriers. The nitric acid concentration is then increased to precipitate all the strontium and barium (and part of the calcium) as nitrates. After further nitric acid separations, barium chromate and iron hydroxide scavenges are carried out. The subsequent treatment depends somewhat on the circumstances but the following is normal practice. Yttrium carrier is added to the purified strontium solution and, after a delay of about 14 days for the growth of 90Y, the yttrium is separated, mounted and counted. The storage period for the growth of 90Y can be reduced if sufficient 90Sr is known to be present, and the appropriate growth factor applied. For samples of very low activity, as well as for measurement of 89Sr, strontium is precipitated from the solution remaining after the removal of yttrium and mounted for counting. In many cases the determination of the natural inactive strontium content of the material is required so that the strontium chemical yield can be corrected. In the case of milk, direct application of the nitric acid separation to a solution of the ash usually gives low strontium yields. The calcium, strontium and barium are therefore concentrated by an initial phosphate precipitation. The mixed phosphates are then dissolved in an acid and the general procedure continued from that point. In the case of cereals, and vegetation generally, the ash is very variable in composition and contains numerous elements other than calcium: a mixture of hydrofluoric and perchloric acids is necessary to decompose and dissolve the ash. After heating to remove the hydrofluoric and most of the perchloric acid, the residue is dissolved in dilute acid and the alkaline earth precipitated as phosphates. For the details of procedure, see report IAEA (1989) and references therein.
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CHAPTER 4 Measurements of radioactivity
Here is procedure designed to separate 90Y from a partially purified Sr fraction isolated from various marine samples. It has been used successfully in soil, sediment and ashed biological samples up to 300 g in weight; and, in fresh water and seawater samples, up to 600 L in volume. 90Sr levels as low as 1 dpm have been measured in these samples. Several important facts should be kept in mind when planning 90Sr analysis on marine samples. Average seawater contains 8 ppm of stable Sr, or about 8 mg/kg, while average carbonate soil is about 1% (10,000 mg/kg) Sr. Thus even modest size samples require some thought in sample handling and about chemical yield measurements, especially if stable Sr is to be used for a yield monitor. This procedure has been developed to accommodate a Sr carrier from 20 mg to 4 g. For most routine analysis, however, 85Sr, a γ-emitter readily measured by γ-spectrometry and free from 90Sr contamination, is the preferred yield monitor. If stable strontium is used for chemical recovery, the strontium content of the original sample has to be determined. In average seawater the 226Ra content is about 0.2 dpm/L and 238U (or 234U) is about 2 dpm/L. The uranium usually presents no problem, but a good separation of the final 90Y from Ra and its daughters is important, especially for deep ocean water. During the separation steps, this procedure effectively removes all radionuclides that are chemically similar to yttrium or rare earths from the Sr fraction; and, after the establishment of 90Sr/90Y radioactive equilibrium, the final purification steps further remove any interfering radionuclides in the ingrowth 90Y fraction before measurement by β-counting. A great number of analytical methods has been developed and applied for the determination of 90Sr and 89Sr in environmental and nuclear samples using various measuring techniques like β-counting, liquid scintillation spectrometry and mass spectrometry. Most techniques use radiochemical procedures for the separation of strontium and/or yttrium including the classical procedure based on a series of semi-selective precipitations, the ion exchange, solvent extraction and extraction chromatographic procedures. All of this has been discussed in details in the review article by Vajda and Kim (2010).
4.3.6.3 Analysis of tritium Tritium is measured by LSC of a portion of a distilled sample. Several reagents (such as sodium sulphite and silver iodide) can be added in the distillation to prevent interference by radioiodine. The allowed concentration of tritium in water for human consumption is relatively high; thus the method presented here is normally adequate for routine determinations. However, if required, lower concentrations of tritium in water can be determined by electrolytic enrichment. The principles of the tritium determination procedure are as follows. The water sample is distilled to remove nonvolatile quenching materials and nonvolatile radioactive materials. Prior to distillation, sodium carbonate (Na2CO3) and sodium thiosulphate (Na2S2O3) are added to the sample. The majority of the constituents that might interfere remain in the residue together with any radioactive iodide and bicarbonate that might be present. If the tritium content of nonaqueous biological
4.3 Radiometric methods
samples is required, the sample can be converted into water by oxidation. The distillation is carried out to dryness to ensure complete transfer of the tritium to the distillate. An aliquot of the distillate is mixed with a scintillation solution in a counting vial. The mixture is cooled and counted in a liquid scintillation spectrometer (coincidence type). In this sample (usually an emulsion) the kinetic energy of the tritium β-particles is partly converted into light photons. When certain boundary conditions are satisfied (e.g. simultaneous detection by two or more photomultiplier tubes connected in coincidence) these photons are counted as pulses. Standard tritium and background samples are prepared and counted identically to minimize errors produced by ageing of the scintillation medium or instrumental drift. The counting rate is a measure of the tritium activity concentration, the sensitivity (counting time 100 min) is generally of the order of 20–200 Bq L1. Details of the method are described in IAEA-295 (1989). For additional reading, see Krieger and Whittaker (1980), Budnitz (1974), Bush (1968), Fox (1976), Horrocks (1974), IAEA-246 (1981), Konno and Suguro (1986), Lieberman (1984), Lieberman and Moghissi (1970), Momoshima et al. (1986) and Volchok and de Planque (1983).
4.3.6.4 Caesium analysis For many samples, the concentration of 137Cs is determined directly by γ-spectrometry. Only water samples and a small percentage of the soil, sediment and biota samples require preconcentration of 137Cs and measurement by β-counting and/or in some cases in low background well-type Ge detectors. The radiochemical procedure for the determination of 137Cs in aqueous samples is based on the batch extraction of caesium onto a microcrystalline cation exchanger, ammonium molybdophosphate (AMP), and subsequent purification from potassium and rubidium activities by ion-exchange separation using a strongly acidic cation exchange resin (BIO-REX-40). Natural K and Rb have radioactive isotopes that interfere with the β-counting of 137Cs. The purification of caesium is also necessary to determine the chemical recovery. γ-Spectrometry is used when samples contain both 134Cs and 137Cs activities. In theory, using the proper absorbers, 134Cs can also be resolved from 137Cs activity because of the difference in β-energy. However, the absorbers greatly reduce the counting efficiency, thereby eliminating any gain in the β-measurements of samples containing both 134Cs and 137Cs. The following are steps to be followed for sample preparation and preconcentration in the case of aqueous samples. If other radionuclides (Pu, Am and Sr) are also analysed in the same sample, the MnO2 preconcentration steps are generally performed first to remove the transuranics, followed by Cs extraction with AMP; and then, the Sr fraction is removed last as the oxalate. After the completion of MnO2 separation, the water sample is transferred to the proper size polyethylene processing container. Adjust to pH 1–4 with nitric acid, add AMP as a slurry in water to extract the caesium (use 0.2 g AMP/L of sample), stir the sample thoroughly and let the AMP settle, filter or decant the supernatant (discard the
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CHAPTER 4 Measurements of radioactivity
supernatant or save for Sr analysis, if required), separate the AMP by centrifugation and purify the Cs for β- or γ-counting. The steps to be followed in the sample preparation procedures used for soil, sediment or ashed biota are: dissolve or leach the sample with conc. HNO3 and separate the residual materials. The acid sample is diluted with water and adjusted to pH 2–4 with NaOH. Add 1–2 of AMP to extract the Cs from solution. Use a minimum of 1 g AMP/L of sample. Larger amounts of AMP are required for soil, sediment or biota samples compared with seawater samples, because of the higher ionic strength in the acid-leached sample solution, which reduces the extraction efficiency of AMP for Cs. Separate the AMP as described earlier for the water sample and purify as described further. The steps required for 137Cs purification are described next. The amount of AMP collected from the preconcentration step, especially from large water samples, is invariably >1 g. The following steps reduce the amount of AMP to about 1 g in order to perform the ion-exchange procedure for the separation of Cs from K and Rb. 1.
2.
3.
4.
5.
6.
7.
Dissolve the AMP with a minimum amount of 10 M NaOH. Add an 0.5 mL excel of 10 M NaOH for each gram of AMP dissolved. Transfer the sample solution to a 400 mL beaker with a few millilitre of water, and heat the solution on a hot plate at a medium setting, without a cover glass, to decompose the AMP and to evaporate the ammonia. Periodically check the vapour phase (just above the hot beaker) with a wet pH paper or ammonium specific test paper to check for the presence of ammonia fumes. The sample solution must be kept strongly basic with NaOH for the AMP decomposition to be effective. When the vapour no longer shows the presence of ammonia, and the solution is strongly basic (pH >13+), stop the heating, cool the sample to room temperature, and add water to dissolve any salts that may be formed during cooling. Dilute the sample to about 200 mL with water, add 1–2 drops of methyl red indicator, adjust the pH to 1–4 with 8 N HNO3, add 1.0 g of AMP, let the AMP settle, decant the clear liquid, transfer the AMP slurry to a clean 50 mL C-tube with water, centrifuge and discard the supernatant. Dissolve the AMP with a minimum amount of 10 M NaOH and add 10 mL of 2% EDTA-07.5 NaOH solution. The sample solution should be clear. If any precipitate forms, centrifuge and decant the clear solution into another clean 50-mL C-tube. Discard the precipitate. Load a medium-size ion-exchange column with a goose-neck adapter (see Fig. 4.44) with 20 mL of BIO-REX-40 20–50 mesh cation exchange resin. Precondition the column with 100 mL of 3 N HCl followed by 150 mL of 5% NaCl solution and 50 mL of water. Using a Teflon-coated stirring rod, carefully pour the sample solution from Step 4 directly onto the top of the resin column. Try not to splatter any solution on the upper part of the reservoir. After the sample has drained to the top of the glass-wool plug, rinse the column walls and the reservoir three times with about 5–10 mL aliquots of water.
4.3 Radiometric methods
1.5 cm
15 cm
8 cm
5 cm cm
25 mL reservoir
250 mL reservoir
12 cm
Glass wool plugs top and bottom
1000 mL reservoir
20 cm
3 cm top resin 100–200 mesh
3 cm top resin 100–200 mesh
6 cm 50–100 mesh 1.0 cm
0.7 cm
21 cm 20–50 mesh Total resin bed = 24 cm
1.5 cm
21 cm 20–50 mesh Total resin bed = 24 cm
Goose-neck adapter for Cs analysis only
(A)
(B)
(C)
FIG. 4.44 Glass columns for anion-exchange separation.
8.
Rinse the column with another 40 mL of water (necessary to remove any dissolved AMP from the resin and preventing the possibility of AMP reforming in the column when the acid rinse is added in the next step). Total water rinses in Steps 7 and 8 are not critical but should not exceed about 60–70 mL. Discard the rinses. 9. Wash the column with 160 mL of 0.75 N HCl (to remove K and Rb). Discard the wash. 10. Elute the Cs with 125 mL of 3 N HCl and collect the sample in a 150 mL beaker. 11. Evaporate the caesium eluate to dryness and prepare the Cs for β-counting as described in the following section. Later Cs samples should be prepared for β-counting following the steps: 1. Dissolve the caesium salts from Step 11 with 1–2 drops of 8 N HNO3 and 2–3 mL of water. Transfer the solution to a 50 mL C-tube. Rinse the beaker twice with 2–3 mL of water. 2. Add 1 mL of 10 N NaOH, dilute the sample to about 10 mL with water, and add 2 mL of 0.12 chloroplatinic acid (H2PtCl6) to precipitate Cs2PtCl6. 3. Cool sample in a refrigerator or ice bath for 30–40 min. 4. Prepare a tarred glass-fibre filter paper. (a) Assemble a filtering apparatus with a 2.54 cm base. (b) Cut a 2.54 cm diameter glass-fibre filter paper disc. (c) With the vacuum off, centre a filter disc on the base of the filter holder, wet the filter with
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CHAPTER 4 Measurements of radioactivity
5. 6.
7. 8. 9.
water, apply vacuum, wash the filter with 2–3 mL of water, and 2–3 mL of acetone. (d) Dry the filter under a heat lamp. (e) Weigh the filter to 0.01 mg. Filter the sample through the tarred filter from Step 4. With the vacuum on, remove the filter chimney, wash the filter and the Cs2PtCl6 thoroughly with a few millilitre of cold acetone. Turn off vacuum, transfer the filter and precipitate to a petri dish. Dry the filter under a heat lamp and cool the filter to room temperature. Weigh the sample to constant weight ( 0.01 mg). Mount the filter containing the Cs2PtCl6 on the ring and disc holder as shown in Fig. 4.45. Count the sample using a low background β-detector or by γ-spectrometry.
4.3.6.5 Determination of actinides Actinides in the environment can be classified into two groups: (i) the uranium and thorium series of radionuclides in the natural environment and (ii) neptunium, plutonium, americium and curium which are formed in a nuclear reactor during the neutron bombardment of uranium through a series of neutron capture and radioactive decay reactions. Transuranics thus produced have been spread widely in the atmosphere, geosphere and aquatic environment on the earth, as a result of nuclear bomb Platinum electrode Wing nut
Anode Insulation tape
Stainless steel support ring Teflon top support
Glass chimney Constat current power supply Teflon washer Stainless steel polished disc, 2.45 cm Cathode
Stainless steel base and posts
FIG. 4.45 Electroplating cell.
4.3 Radiometric methods
Table 4.12 Nuclear data of actinide isotones Nuclide
α-Energy (MeV)
Yield (%)
Half-life (years)
244
5.81, 5.76 6.11, 6.07 5.28, 5.23 5.49, 5.44 4.90, 4.86 5.16, 5.12 5.16, 5.14, 5.10 5.50, 5.46 5.77, 5.72 4.79, 4.77 4.20, 4.15 4.40, 4.37 4.77, 4.72 5.32, 5.26 4.02, 3.96 4.69, 4.62 4.90, 4.85 5.42, 5.34
77, 74, 88, 86, 74, 76, 73, 71, 69, 51, 77, 57, 72, 69, 77, 76, 11, 73,
1.81 10 6.09 106 7.37 103 4.32 102 3.76 105 6.57 103 2.41 104 8.77 10 2.85 2.14 106 4.47 109 7.04 105 2.45 105 7.18 10 1.41 1010 8.03 104 7.34 103 1.91
Cm Cm 243 Am 241 Am 242 Pu 240 Pu 239 Pu 238 Pu 238 Pu 237 Np 238 U 235 U 234 U 232 U 232 Th 230 Th 229 Th 228 Th 242
23 26 11 13 26 24 15, 12 29 31 36 23 18 28 31 23 23 56 27
tests in the atmosphere, and accidental release from nuclear facilities (Sakanoue, 1987). Most of these radionuclide inventories have deposited in the northern hemisphere following the tests conducted by the United States and the Soviet Union. In actinide series, the elements of greatest interest as environmental contaminants are neptunium, plutonium, americium and curium, because their presence at relatively high concentrations in ecosystems would represent potential health problems (Katz et al., 1986). Nuclear data for actinide isotopes are presented in Table 4.12. As a result of nuclear bomb tests in the atmosphere, accidental release from nuclear facilities, and the accidental fall of artificial satellite SNAP 9A, plutonium isotopes have been spread widely in the earth’s atmosphere, geosphere and aquatic environment from the 1960s. They include 241Pu, 240Pu, 239Pu and 239Pu (Harley et al., 1973). These plutonium isotopes in the geosphere and the aquatic environment are incorporated metabolically into plants and ultimately find their way into man through the food chain; these radionuclides in the atmosphere are also incorporated into man directly by inhalation. From the standpoint of a dose assessment, these radionuclides are important because they are mostly α-emitting radionuclides and have half-lives of 13–2.4 104 years. From the standpoint of the safety assessment of the disposal of high-level radioactive waste, it is important to clarify the migration behaviour of plutonium in ground strata. Consequently, it is essential to know the plutonium concentrations in food and environmental samples for these studies.
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CHAPTER 4 Measurements of radioactivity
The commonly applied methods for determination of actinides in environmental samples may be classified as follows: 1. Preparation: (a) drying, (b) ashing and (c) scavenging. 2. Solubilization and equilibration: (a) fusion and (b) leaching. 3. Concentration and separation: (a) coprecipitation, (b) ion exchange and (c) solvent extraction. 4. Electrodeposition and α-spectrometry. For a description of a procedure for plutonium separation in large volumes of fresh and saline water by manganese dioxide coprecipitation, see Wong et al. (1975). Among ion-exchange separation methods for transuranics, strong base anion exchange in hydrochloric and nitric acids is important (Keder, 1962; Keder et al., 1960; Horwitz et al., 1990; Chu, 1971; Wong, 1971; Diamond et al., 1954; Korkisch, 1989). Among solvent-extraction reagents for transuranics, thenoyltrifluoroacetone (TTA) and trioctylamine (TOA) are important (Keder, 1962; Chieco et al., 1990). Each transuranic element has many valencies and their behaviour in aqueous solution is very complicated because of disproportionation reactions. As stated above, the ion-exchange and solvent-extraction behaviours of transuranics are dependent on their valency state. Therefore valency control is very important in their analysis (Katz et al., 1986). Additional references discussing this problem include Korkisch (1989), Diamond et al. (1954), Stevenson and Nervik (1961), Abuzwida et al. (1987), Bernabee et al. (1980), Budnitz (1973), Chu (1971), Fukai et al. (1976), Hampson and Tennant (1973), Hindman (1986), Holm et al. (1979), Irlweck and Veselsky (1975), Jiang et al. (1986), Johns (1975), Scott and Reynolds (1975) and Sekine et al. (1987). Procedures are described for the determination of plutonium and americium in environmental samples by anion exchange (HNO3). Procedures are also described for the determination of plutonium, americium and their sequential analysis by anion exchange (HNO3) and TOA extraction (Chieco et al., 1990). Livens and Singleton (1989) developed the method for the determination of Am in environmental samples (Fig. 4.46). Yamamoto et al. (1989) developed the method for the determination of Np in environmental samples (Fig. 4.47). There follows a description of plutonium separation by anion exchange. Plutonium as well as several other heavy elements (e.g. U, Th, Np, Am, Cm) may be separated and purified by anion-exchange chromatography. If only 239+240Pu activity is to be determined, a single anion-exchange column separation, done carefully, is
4.3 Radiometric methods
Sample leachate (HCl)
Fe solvent extraction (diisopropyl ether)
Anion exchange (HCl)
Pu fraction
Cation exchange (HCl)
BiPO4 or Fe(OH)3 precipitation Anion exchange (HNO3 – CH3OH)
Anion exchange (HNO3)
Am fraction
FIG. 4.46 Flowsheet for Pu/Am method. After Livens, F.R., Singleton, D.L., Evaluation of methods for the radiometric measurement of Americium-241 in environmental samples. Analyst 114 (1989) 1097.
usually sufficient. However, if the 238Pu activity is also required, then a second small column separation will be necessary to eliminate any trace of interfering activities from the naturally occurring elements, Ra and Th, present in all soil and sediment samples. The steps to be followed in plutonium purification are •
• •
•
Prepare anion-exchange columns as shown in Fig. 4.44. Use the medium-size column for sample volumes <500 mL or the large size column for samples with final volumes >500 mL. Precondition the columns by passing about 20 mL of 8 M HNO3 through each column. Check the sample solution for particulate material. If insoluble silicate (white, gelatinous materials) or other particles are noted, centrifuge or filter the solution through a glass-fibre filter before loading the sample onto the preconditioned column. This step is important in order to maintain a continuous flow of the sample solution through the column. Load the sample solution onto the column as fast as the column will allow. The plutonium is retained on the ion-exchange column. If insoluble material has been removed from the sample and there is no outgassing in the column, the average flow rate in the medium-size column is about 3–6 mL per minute, depending on the viscosity of the sample solution.
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CHAPTER 4 Measurements of radioactivity
Soil or sediment sample 239
Heat
Res.
Np tracer Aqua-regia
Sup. (Pa, U)
Sup.
Evaporate to dryness 10 M HCl Isopropyl ether extraction
ppt. (LaF3) 8M HNO3 saturated with Al(NO3)3 Anion exchange
Feed solution 8 M HNO3 10 M HCl 10 M HCl-0.1 MHI 4 M HCl
Dewox 1´8, 100–200 mesh (U) (Th) (Pu) (Np)
Evaporate to dryness Aq.
Org. (Fe) 0.05 M HI (Heat)
10% TOA-xylene extraction (x3)
Anion exchange
Feed solution 4M CH3COOH (Np)
Aq. (Th, Pu, Am)
Org.
Electroplate from 2 M NH4Cl colution
10 M HCI wash Plated Np
1 M HCl-0. 1 M HF strip Continued to Pu and Am separation procedure
NH2OH x HCl Conc. H2SO4 La carrier
r-Ray spectrometry α-Ray spectrometry
FIG. 4.47 Separation scheme of neptunium for environmental soil or sediment sample. After Yamamoto, M., Chatani, K., Komura, K., Ueno, K., Development of alpha-ray spectrometric techniques for the measurement of low-level 237Np in environmental soil and sediment. Radiochim. Acta 47 (1989) 63.
•
• •
• • •
After the sample has drained to the top glass-wool plug in the column, rinse the wall of the column reservoir thoroughly with 8 M HNO3. This can be done conveniently with a polyethylene wash bottle. Do this at least three times with about 5–10 mL of 8 M HNO3 and allow each rinse to drain to the top glasswool plug. After the last rinse, wash the column-volumes of 8 M HNO3 and discard the rinse (about 150 mL for the medium column or 350 mL for the large column). After the 8 M HNO3 rinse, wash the wall of the column reservoir with about 5–10 mL of conc. HCl. Repeat the conc. HCl wash 3 times, each time draining the liquid to the top of the glass-wool plug. Now wash the column-volumes of conc. HCl (about 150 mL for med. Column or 350 mL for the large column). Discard HCl rinse. Elute the Pu with five column-volumes of NH4I-HCl solution. Use 100 mL of NH4I-HCl for the medium-size column. Collect the Pu eluent in a 150 mL beaker. Add approximately 5 mL of conc. HNO3 to the Pu eluent, mix, and evaporate to dryness on a hot plate. A small residue may be visible on the bottom of the beaker after the evaporation, which appears to have no significant effect on the analysis.
4.3 Radiometric methods
This material is normally seen and probably results from the decomposition products of strong acids with the resin. The following is a description of plutonium purification by anion-exchange separation. 1.
Cool the sample to room temperature. Rinse the wall of the beaker with 8–10 mL of conc. HCl. Add 2–3 drops of 30% H2O2, 3–5 drops of 1 M NaNO2, and heat on a hot plate for about 5 min. Again, cool the sample to room temperature. 2. Prepare a small sample with Dowex 1 8.50–100 mesh and precondition it with 1 mL of conc. HCl. Collect the eluent in a 50 mL C-tube. 3. Load the sample from the beaker onto the column. 4. Rinse the beaker two times with 2–3 mL of conc. HCl and load the rinse onto the column. 5. When the liquid has drained to the top glass-wool plug of the column, rinse the wall of the reservoir with 2–3 mL of conc. HCl. 6. Rinse the column 2 more times with 2–3 mL of conc. HCl. 7. Wash the column with 10 mL of conc. HCl. 8. Discard the rinse from Step 2 through Step 7. 9. Elute the Pu with 20 mL of NH4I-HCl solution. Collect the Pu in a 50-mL glass beaker. 10. Add 2–3 mL of conc. HNO3 and evaporate the solution to dryness on a hot plate. The steps to be followed in the implementation of the electrodeposition of plutonium procedure are: 1. 2. 3. 4. 5.
Assemble the plating cell. Fill the cell with water to test for leakage. Add 1 mL of conc. Sulphuric acid to the sample. Heat the sample on a hot plate until copious white fumes evolve. Cool the sample to room temperature, carefully rinse the wall of the beaker with 2 mL of 1 N H2SO4. Add two drops of 0.1% methyl red indicator. 6. Transfer the sample to the electroplating cell. 7. Rinse the beaker with 2 mL of 1 N H2SO4. Add the rinse to the plating cell. 8. Repeat Step 7 three times. 9. Add conc. NH4OH dropwise until the colour of the sample changes from red (pH 4.4) to yellow (pH 6.2). Mix the solution by swirling the plating cell. 10. Add 1 N H2SO4 dropwise to the red end point (pH 4.4) and then add two drops excess. 11. Complete the assembly of the electroplating cell by attaching the platinum anode with plastic insulation tape. Position the platinum wire anode about 0.5 cm from the stainless steel disc (cathode). 12. Connect the anode and cathode of the electroplating cell to a constant current power supply.
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CHAPTER 4 Measurements of radioactivity
13. 14.
15.
Electroplate at 1.0 A for 60–70 min. Before turning off the power supply when electrodeposition is completed, add 1 mL of conc. NH4OH to the cell and continue plating for about 1 min then turn off the power supply and as quickly as possible: (a) disconnect the cell from the power supply, (b) discard the solution from the cell, (c) rinse the cell with diluted NH4OH from a wash bottle. (Make the diluted NH4OH solution by adding 0.5 mL conc. NH4OH to 500 mL of water), (d) disassemble the plating cell, (e) rinse the plated disc with diluted NH4OH and (f) rinse the plated disc with acetone and let the disc air dry on a clean paper tissue. Count the plated disc and determine the activity of plutonium isotopes by α-spectrometry.
4.3.7 Rapid methods The need for rapid methods is apparent in accident and similar situations. An impulse to the development of rapid methods is also provided through the so-called coordinated research program of IAEA, Vienna. Here we present some of the methods reported in a Research coordination meeting on Rapid Methods held in Vienna in 1991. As an example of rapid methods, we mention here the work by Brodzinski and Perkins (1992). They have described a completely portable instrument, operable by one man, which is capable of quantifying the radioactive content of drums. Eleven radioisotopes are measured simultaneously in just a few minutes. The assayer uses two measuring techniques: segmented γ-ray spectrometry and neutron counting. A drum (or other container) to be assayed is placed on a rotating turntable by a selfcontained electric hoist. A collimated high-resolution germanium γ-ray spectrometer vertically scans the rotating drum to measure the intensity of γ-rays as a function of the energy emanating from the drum. Most fission and activation products and some transuranic radionuclides emit measurable quantities of monochromatic photons that serve as “fingerprints” of those radioisotopes. Comparison with emission rate from known standards provides a quantitative measure of radioactivity from each γ-ray emitter in the drum. This germanium spectrometer is used to measure the bremsstrahlung radiation from 90Sr. By manipulating the software with the on-board computer, the intensity of the 90Sr bremsstrahlung in the assayed drum is also compared to that of standards and the 90Sr concentration is quantified. The reported sensitivity for transuranic radionuclides is approximately 1 nCi/g, while that for γ-emitters is of the order of 0.1–1 pCi/g. Also, based on the bremsstrahlung radiation measurement, 90 Sr can be determined at concentrations of 100 pCi/g.
4.3 Radiometric methods
4.3.7.1 Rapid determination of transuranic elements and plutonium The rapid methods are based on fast removal of the transuranic elements from interfering materials so that they can be electrodeposited as a group, and measured by α-energy analysis. The procedure involves the following basic steps (Thomas, 1991). 1. 2.
The sample is first brought into solution. Radioisotope tracers, including 242Pu, 243Am and 234Th (if appropriate), are added. 3. A small amount of Fe carrier (10 mg) plus sodium sulphite is added to this solution, which is subsequently made basic by addition of ammonium hydroxide to allow the formation of an iron hydroxide precipitate. This precipitate serves to carry the thorium and the transuranic elements. 4. The mixture is centrifuged and the solution discarded: the precipitate is dissolved in dilute hydrochloric acid then diluted with water. It is then made basic with ammonium hydroxide, which results in a second iron hydroxide precipitate forming. 5. Following centrifuging and discarding of the solution, the precipitate is dissolved in dilute hydrochloric acid, diluted with water and a small amount of sodium sulphite added to maintain the transuranic elements in their lower valence states. 6. Small amounts of calcium and oxalic acid are then added and the pH is adjusted to approximately three to allow formation of an oxalate precipitate. The iron forms a very soluble oxalate, thus remaining in solution. This and two subsequent oxalate precipitations serve to remove any remaining iron. 7. The final oxalate precipitate is then dissolved in a small amount of sulphuric acid (0.5 mL of concentrated H2SO4), and the pH adjusted using dilute ammonium hydroxide. 8. The solution is then placed in an electrodeposition cell, where the transuranic elements are electrodeposited on a 1 cm2 area of a 2.5 cm diameter stainless steel disc. 9. Electrodeposition is conducted for a 1-h period at a current of 1 A. 10. Immediately before turning off the current, 1 mL of concentrated NH4OH is added to the cell and the electrodeposition continued for an additional minute. The current is then turned off, the solution discarded, and the electrode washed with water, then ethanol and air dried. Following α-energy analysis, the radiochemical yield, as determined from the radioisotope tracer content and the concentrations of the radionuclides of interest, are calculated. Samples with a large amount of iron such as soil extracts or vegetation ash may require partial removal of iron prior to initiation of this procedure (Thomas, 1991).
4.3.7.2 Rapid determination of
90
Sr
Up to now, two different approaches have been used successfully in fast radiochemical separation procedures for the determination of strontium-90 in environmental sample:
217
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CHAPTER 4 Measurements of radioactivity
• •
investigations dealing with the extraction of yttrium-90 and investigations which tailor precipitation methods for the strontium-90 separation to the needs of special sample types.
We mention only some of the work in these fields. Vajda et al. (1991) have described a simple and rapid method for the separation and successive determination of total radiostrontium in soil by using a crown ether. The method consists of three basic steps: oxalate precipitation to remove bulk potassium, chromatographic separation of strontium from most inactive and radioactive interferences utilizing crown ether, oxalate precipitation of strontium to evaluate the chemical yield. Radiostrontium is then determined by LSC of the dissolved precipitate. When 10-g samples of soil are used, the sensitivity of the method is about 10 Bq/kg. The chemical yield is about 80%. The separation and determination of radiostrontium can be carried out in about 8 h. Another method for 90Sr determination in food and environmental samples has been described by Shuzhong et al. (1991). It is based on the use of a tributyl phosphate for extraction of 90Y, the daughter of 90Sr. The method is shown to be sensitive to 0.2 Bq/kg of dry grass and milk powder and 2 Bq/kg of soil. In the case of a nuclear accident, most of the radioisotopes in the environment and food can be reliably and quickly assayed by γ-ray spectroscopy. There is a problem with some important isotopes which are pure β- or α-emitters and which cannot be identified directly by γ-ray spectroscopy. The activity of the isotopes of the strontium group 89Sr, 90Sr and 91Y after a 3-year reactor fuel cycle can reach about 8% of the total in-core activity and one of them, 90Sr(Y), is important for the long-term health consequences. It has been shown (Vapirev and Hristova, 1991) that the Ba/Sr reactor core ratio can be used for estimation of the upper limit of strontium activity in the fallout immediately after an accident. The Cs/Sr ratio can be used for estimation of the strontium fallout in the late postaccident period.
4.4 Nonradiometric methods There is a variety of situations in which it is better to determine the concentration of a radionuclide by a mass measurement rather than by measuring the activity present. This approach is possible using a wide range of instrumental methods of nonradiometric elemental analysis; analytical measurements can be performed also by element-specific chemical methods, some of which are extremely sensitive. The most important criterion for selecting an analytical method is whether the technique is sufficiently sensitive to measure the amount of radionuclide present in the sample. This is a very different problem when considered from the viewpoint of analytical chemists who use radiometric methods and those who use nonradiometric methods. Limits of detection in radiometric methods can be as low as 104 Bq, although 1 mBq is a more generally attainable detection limit. For nonradiometric methods, the detection limit is expressed in terms of mass and the
4.4 Nonradiometric methods
Table 4.13 The mass of 1 mBq for a selection of radionuclides with a variety of half-lives Radionuclide
Half-life (years)
Mass of 1 mBq (g)
232
1.4 1010 4.5 109 1.7 107 2.1 105 2.4 104 5.8 103 30
2.5 107 8.1 108 1.6 1010 1.6 1012 4.3 1013 6.1 1015 3.1 1016
Th U 129 I 99 Tc 239 Pu 14 C 137 Cs 238
After McMahon, A.W., 1992. An intercomparison of non-radiometric methods for the measurements of low levels of radionuclides. Appl. Radiat. Isot. 43, 289–303.
relationship between radiometric and nonradiometric limits of detection will depend on the half-life of the radionuclide of interest. Table 4.13 (from McMahon, 1992) shows the mass corresponding to 1 mBq for a number of important radionuclides. Although 1 mBq of 232Th is readily measured by a number of nonradiometric methods, 1 mBq of 137Cs could only be detected by the most sensitive of methods and is probably best determined radiometrically. Nonradiometric methods offer a variety of features and their use may be favoured for reasons other than improved sensitivity or isotopic selectivity. They can, in some instances, be used to perform analyses with less sample preparation and greater speed or sample throughput, and allow remote analysis or provide elemental or isotopic maps or depth profiles (McMahon, 1992). The instrumental methods of elemental analysis can be conveniently grouped as follows: (i)
Methods based on X-ray fluorescence analysis X-ray fluorescence analysis (XRF) Total reflection X-ray fluorescence analysis (TXRF) Electron microprobe analysis (EMA) Particle-induced X-ray emission (PIXE) Synchrotron radiation-induced X-ray emission (SRIXE) (ii) Methods based on ultraviolet or visible spectroscopy Atomic absorption spectroscopy (AAS) Graphite furnace AAS (GFAAS) Atomic fluorescence spectroscopy (AFS) Inductively coupled-plasma optical-emission spectroscopy (ICPO-ES) Glow-discharge optical-emission spectroscopy (GC-OES) Laser-excited resonance-ionization spectroscopy (LERIS) Laser-excited atomic fluorescence spectroscopy (LEAFS) Laser-induced-breakdown spectroscopy (LIBS) Laser-induced photoacoustic spectroscopy (LIPAS)
219
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Table 4.14 Analytical techniques classified by amount of isotopic information and amount of sample required Bulk samples No isotopic information Minor isotope determination Trace isotope determination
ICP-OES ICP-MS, GDMS, SSMS
Small samples profiling
Imaging and depth
XRF, GFAAS, LEAFS, TXRF ETV-ICP-MS
PIXE, SRIXE SIMS, SNMS, LMS, SIRIS
TIMS, RIMS, AMS
After McMahon, A.W., 1992. An intercomparison of non-radiometric methods for the measurements of low levels of radionuclides. Appl. Radiat. Isot. 43, 289–303.
Resonance-ionization spectroscopy (RIS) (iii) Methods based on mass spectrometry Spark-source mass spectrometry (SSMS) Glow-discharge mass spectrometry (GDMS) Inductively coupled-plasma mass spectrometry (ICP-MS) Electrothermal vaporization-ICP-MS (ETV-ICP-MS) Thermal-ionization mass spectrometry (TIMS) Accelerator mass spectrometry (AMS) Secondary-ion mass spectrometry (SIMS) Secondary neutral mass spectrometry (SNMS) Laser mass spectrometry (LMS) Resonance-ionization mass spectrometry (RIMS) Sputter-initiated resonance-ionization spectroscopy (SIRIS) Laser-ablation resonance-ionization spectroscopy (LARIS) McMahon (1992) has reported on intercomparison of nonradiometric methods for the measurement of low levels of radionuclides. He has classified the earlier analytical techniques according to the amount of isotopic information and the amount of sample required. The conclusions are presented in Table 4.14.
4.4.1 Methods based on X-ray spectrometry The electronic transitions which give rise to X-ray emission spectra involve core electrons and are therefore relatively insensitive to the chemical and physical form of the determinant (Bertin, 1978). As a result, analyses can be performed with a minimum of sample preparation directly on materials in the condensed phase. This insensitivity of sample matrix applies to the wavelength of the emitted X-rays, not to their intensities and as quantitation is based on intensity measurement, closely matched standards are required. X-ray emission spectra can be excited by primary X-rays in a fluorescence experiment or by changed particles via collisional excitation.
4.4 Nonradiometric methods
The cross sections for excitation of X-ray emission are rather low and this is combined with the low efficiency of collection, collimation, diffraction and detection of the emitted X-rays. This low overall efficiency leads to a relatively low sensitivity in some cases and is compounded by high backgrounds either from scattered primary radiation in a fluorescence experiment or due to bremsstrahlung in the charged-particle-excitation methods. Methods based on X-ray spectrometry do not provide isotopic information about the sample. Nonetheless, there are a number of radio analytical problems which can be solved by methods based on X-ray spectrometry. The following instrumental methods of elemental analysis are based on X-ray spectrometry: X-ray fluorescence analysis (XRF) Total reflectance X-ray fluorescence analysis (TXRF) Electron microprobe analysis (EMA) Particle-induced X-ray emission (PIXE) Synchrotron radiation-induced X-ray emission (SRIXE) XRF is the simplest of these methods. It allows bulk analysis of solid or liquid samples with detection limits of approximately 0.1 μg. The method can thus only compete with radiometric methods for the longest lived of radionuclides. It has approximately the same sensitivity for 232Th as α-spectrometry but has the advantage that little sample preparation is required and that analysis is rapid and easily automated. XRF would be the method of choice for measurement of airborne thorium collected onto filter papers, for example. The more sophisticated methods address the problem of the low overall efficiency of generation and acquisition of the X-ray spectrum. The low-fluorescence cross section is addressed by using a highly intense X-ray source, a synchrotron in the SRIXE method. The high intensity of synchrotron X-rays allows the beam to be focused and collimated while retaining significant intensity. The method can therefore be used in a microprobe mode and by moving the sample in a raster pattern across the incident X-ray beam, elemental images can be generated with micron spatial resolution. The scattered primary radiation background can be reduced by using the total reflectance technique in TXRF (Knoth and Schwenke, 1978). The instrumental geometry limits scattering of primary X-rays in the direction of the detector, however this is at the expense of increased sample preparation. The gains in sensitivity achieved by each of these methods may be compounded in a method which uses a total reflectance sample geometry in combination with a synchrotron X-ray source. The charged-particle-beam methods EMA and PIXE also allow elemental imaging within the sample. These methods generally require that the sample be enclosed in a vacuum. The approximately 15 keV electrons used in an EMA instrument, penetrate only 1–2 μm into the sample. This rapid declaration of the charged particles generates bremsstrahlung X-rays which generate a strong background signal in the spectral region of interest. EMA thus has relatively poor detection limits. The method can be used for analysis of electrodeposits such as sources prepared for
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α-particle spectrometry where the element of interest is present at high concentration in a very thin surface layer. The approximately 2.5-MeV-proton beam used in PIXE analysis penetrated much deeper into the sample than the EMA electron beam. The resulting proton bremsstrahlung is less intense and backgrounds are therefore reduced. PIXE can thus achieve much lower detection limits. PIXE ( Johansson and Campbell, 1988) and SRIXE ( Jones and Gordon, 1989) have similar imaging capabilities and detection limits but both suffer from the drawback that they rely on major pieces of hardware, an accelerator in the PIXE experiment and a synchrotron X-ray source for SRIXE.
4.4.2 Methods based on ultraviolet-visible spectroscopy Atomic spectroscopy in the ultraviolet-visible region involves transitions of valence shell electrons and the spectra are thus sensitive to the chemical and physical form of the element of interest. For sensitive quantitative work the sample is normally converted to free atoms in the gas phase. This can be achieved by vaporization from a furnace, by aspiration of a solution into a flame or inductively coupled plasma, by sputtering in a glow discharge or by laser ablation. Producing free, gas-phase atoms is a particular problem for thorium, uranium and plutonium as these elements react with traces of oxygen, even in a high vacuum system, to give oxides and dioxides. The following methods are based on types of atomic spectroscopy in the ultravioletvisible region: Atomic absorption spectroscopy (AAS) Graphic furnace AAS (GFAAS) Atomic fluorescence spectroscopy (AFS) Inductively coupled-plasma optical-emission spectroscopy (ICPOES) Glow-discharge optical-emission spectroscopy (GDEOS) Laser-excited atomic fluorescence spectroscopy (LEAFS) Laser-induced-breakdown spectroscopy (LIBS) Resonance-ionization spectroscopy (RIS) The methods range from simple, inexpensive absorption spectroscopy to sophisticated tunable-laser-excited fluorescence and ionization spectroscopies. AAS has been used routinely for uranium and thorium determinations (see for example Pollard et al., 1986). The technique is based on the measurement of absorption of light by the sample. The incident light is normally the emission spectrum of the element of interest, generated in a hollow-cathode lamp. For isotopes with a shorter half-life than 238U and 232Th, this requires construction of a hollow-cathode lamp with significant quantities of radioactive material. Measurement of technetium has been demonstrated in this way by Pollard et al. (1986). Lawrenz and Niemax (1989) have demonstrated that tunable lasers can be used to replace hollow-cathode lamps. This avoids the safety problems involved in the construction and use of active hollow-cathode lamps. Tunable semiconductor lasers were used as these are low-cost
4.4 Nonradiometric methods
devices. They do not, however, provide complete coverage of the spectral range useful for AAS and the method has, so far, only been demonstrated for a few elements, none of which were radionuclides. Absorption spectroscopy measures the difference in intensity between an incident and transmitted signal. Lower detection limits can be potentially obtained by monitoring a single low-intensity signal, as in emission or fluorescence spectroscopy. LEAFS uses tunable lasers to efficiently excite fluorescence and, by passing the sample atoms repeatedly through excitation–fluorescence cycles, very high sensitivities can be obtained. Again, LEAFS has been demonstrated for only a limited number of elements, none of which were radionuclides. A particularly sensitive approach is to excite fluorescence by a two photon process. In this way the wavelength of the fluorescent light is much shorter than that used to excite fluorescence and scattered primary radiation can easily be discriminated from the fluorescent signal. As an alternative to observing the fluorescent signal in a LEAFS experiment, the state which has been resonantly excited by a tunable laser can be further excited by further laser photons to produce an ion. Ions can be collected and detected with electron multipliers with high-efficiency leading to the extremely high sensitivity of ionization spectroscopies. LEAFS and RIS combine the high selectivity of laser spectroscopy with high sensitivities. Both these components are required to give low detection limits. Isotope effects are observable in high-resolution ultraviolet-visible spectroscopy. At very low atomic numbers and at very high atomic numbers (especially for the actinides) isotope shifts can be observed and coupling of the electron spin with the nuclear spin in odd-mass-number elements. Ultraviolet-visible spectroscopy can thus potentially provide isotopic information in these regions of the periodic table although routine methods are not yet available. For those elements where isotopes cannot be distinguished from their simple atomic spectra an isotope-specific resonance-ionization method has been suggested by Lethokov (1987). An instrument is being developed, based on this suggestion, for the determination of 90Sr in which strontium ions are accelerated to an energy of about 50 keV and neutralized collisionally. At this kinetic energy, the different strontium isotopes are travelling at sufficiently different velocities that collinear resonance-ionization spectroscopy can differentiate between isotopes on the basis of their different Doppler shifts (Monz et al., 1993). Thus some methods of atomic spectroscopy can provide isotopic information (Hurst and Payne, 1988).
4.4.2.1 Inductively coupled-plasma-optical emission spectrometry (ICP-OES) Atomic spectroscopy is widely used in inorganic chemistry to determine total element concentrations in many sample types, and generally allows rapid sample throughput. The optical techniques allow determination of atomic concentrations down to subnanogram/millilitre levels (108 M and below) in samples of a few
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millilitres or less. The recent introduction of a new mass spectrometric technique allows isotope-specific measurements to be made with the ease of use and sample throughput of the atomic spectroscopic techniques. The inductively coupled plasma (ICP) is stable argon plasma heated by inductive coupling of argon cations and free electrons, but is perhaps best thought of as simply a hot flame. Temperature measurement indicates that the plasma has a temperature approaching 6–7000 K. Samples in solution are nebulized (at about 0.4 mL min1 solution consumption) to produce an aerosol of fine droplets. A spray chamber is used to select only the smallest droplets for analysis, in practice those below approximately 5 μm. The selected droplets are swept into the centre of the plasma by an argon stream. In the plasma, droplets undergo rapid heating causing firstly desolvation of droplets, and then breakage of molecular bonds. The resulting free atoms are electronically excited; many are ionized. As atoms leave the plasma and cool, they relax leading to emission of light. Detection of this light is the basis of ICP-optical emission spectrometry (OES). The wavelengths emitted are characteristic of the elements present and the intensity proportional to their concentrations. ICP-OES limits of detection for many elements lie in the range 1–100 ng/mL (ppb) in solution. A few elements, notably Li, Be, Mg, Ca, Sc, Ti, Mn, Cu, Sr, Y and Ba, have limits <1 ng/mL. ICP-MS on the other hand generally is more sensitive (by 1–3 orders of magnitude) and gives isotope-specific information; ICP-OES to a first approximation only gives total element concentrations. The principal advantages of the technique are that it is multielement and that data acquisition takes approximately 1 min with a changeover time between samples of a similar order. The technique has drawbacks: spectral interference is possible, depending on other elements present; therefore in ICP-OES a high-resolution optical spectrometer may be required. The technique is best suited to solution, although direct solid sampling techniques are being developed.
4.4.2.2 Laser-excited resonance-ionization spectroscopy (LERIS) To achieve both high isotopic selectivity and high sensitivity at the same time, collinear laser spectroscopy is combined with resonance ionization. The principle of resonance-ionization spectroscopy is the following: the atoms are excited by one or several resonant optical excitation steps into an energetically high-lying state. Subsequently the atoms are ionized either by absorption of another photon, by collisions, or by field ionization. The photo-ions produced in this process can then be detected with high efficiency. This technique has proved to be extremely useful and sensitive in numerous applications. For the combination of collinear fast-beam laser spectroscopy with resonance-ionization detection, the excitation into highlying Rydberg states with subsequent field ionization is best suited because of the effective suppression of background. This technique has already been successfully applied for trace analysis of 3He in environmental samples as well as for the
4.4 Nonradiometric methods
Laser system
Ion source
Deflector Ion filter Separator magnet
Ion filter Neutralization Optical excitation
Energy selection
Ion detector
Field ionization
FIG. 4.48 Simplified scheme of the experimental set-up for detector of 89,90Sr by collinear fast-beam laser excitation and resonance-ionization detection.
sensitive study of radioactive Yb-isotopes at the online mass separator facility ISOLDE at CERN. Monz et al. (1993) have described the use of laser-excited resonance-ionization spectroscopy (LERIS) for low-level detection of 90Sr and 89Sr in environmental samples. The experimental set-up is shown schematically in Fig. 4.48. After chemical separation from the environmental sample, the Sr is inserted into the ion source. The ions are accelerated to an energy of 30 keV and pass through the mass separator, where the stable isotopes are strongly suppressed. The 89Sr or 90Sr ions enter the apparatus for resonance ionization in collinear geometry and are deflected by 10 degree to enable collinear superimposition of the laser beam. Neutralization takes place inside a charge exchange cell filled with caesium vapour. Remaining ions are removed afterwards from the resulting fast atomic beam by different electrostatic deflectors. Subsequently, the selective excitation into high-lying Rydberg states is induced by the laser light. The Rydberg atoms are field-ionized in a longitudinal electric field and the resulting ions are deflected out of the atomic beam for counting with a particle detector. A total isotope selectivity of 88Sr/90Sr > 1010 and an overall efficiency of 5 106 have been achieved. With these values, a detection limit of 1 108 atoms of 90Sr in the presence of >1017 atoms of stable isotopes has been demonstrated. The trace determination of such a contamination can be carried out with an accuracy of 30% within one working day including all chemical extraction steps. The chemical procedure of separation of strontium from air filters is carried out without the usual addition of strontium carrier to keep the content of stable strontium low. Such a chemical procedure has been worked out. Water and soil samples may have higher contents of stable strontium and thus require still higher values for the selectivity of the method.
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The performance of the technique might be increased by an optimization of the ion source efficiency and higher optical excitation probability affecting both the overall efficiency and the selectivity. These improvements should enable us to lower the detection limit for 90Sr and extend the measurements to 89Sr (see also Bolshov et al., 1988; Lethokov, 1987).
4.4.3 Methods based on mass spectrometry The electric and magnetic fields, used for the analysis of ions, provide only information about the two quantities E/q and M/q where E, M, and q are the energy, mass, and charge of the ion respectively. There are four ways in which the quantities E/q and M/ q may be determined: 1. 2. 3. 4.
magnetic selection, (Bρ)2 ¼ 2(M/q)(E/q) electrostatic selection, Eρ ¼ 2(E/q) cyclotron selection, 1/f ¼ (π/B)(M/q) velocity selection, v2 ¼ 2(E/q)(M/q)
where B is the magnetic field, ρ the radius of the ion path, E the electric field, f the cyclotron frequency and v the ion velocity. Low-resolution measurements that separate neighbouring isotopes as their final output have the quantities M/q and E/q. Since M can be regarded as an integer, ambiguities can arise if M and q have common factors. It is for this reason that some flexibility in the choice of q is desirable. If it is possible to measure the energy of the ion also, then it is possible to determine q and so determine the mass from the ratio M/q. The use of energy, mass and charge signatures, at energies such that charge state 3 + or higher is dominant, is the basis for the accelerator mass spectrometry (AMS) of almost all stable isotopes. The methods listed in the following are based on mass spectrometry, differing mainly in the design of the ion source used: Spark-source mass spectrometry (SSMS) Glow-discharge mass spectrometry (GDMS) Inductively coupled-plasma mass spectrometry (ICP-MS) Electrothermal vaporization-ICP-MS (ETV-ICP-MS) Thermal-ionization mass spectrometry (TIMS) Accelerator mass spectrometry (AMS) Secondary-ion mass spectrometry (SIMS) Secondary neutral mass spectrometry (SNMS) Laser mass spectrometry (LMS) Resonance-ionization mass spectrometry (RIMS) Sputter-initiated resonance-ionization spectroscopy (SIRIS) Laser-ablation resonance-ionization spectroscopy (LARIS) Let us briefly discuss again the limitations of radioactive decay measurement. The observation of the radioactive decay of a single atom is possible, consequently, with efficient apparatus for the detection of the decay particles and a radioactive species
4.4 Nonradiometric methods
with a half-life of seconds and minutes, it is possible to detect all or nearly all of a small number of radioactive atoms in the presence of a large number of nonradioactive atoms with radiation detection techniques. However, as the half-life increases, the time taken to carry out an experiment with a small number of radioactive atoms naturally increase, for half-lives, of say, 106 years efficient detection of the radioactive decay products becomes impossible unless the measurement can be continued for 106 years. Therefore, studies of long-lived radioactive isotopes invariably use very large numbers of atoms and the apparatus detects the decay of only a small fraction of the total during the experiment. In this situation the mass spectrometric detection sensitivity surpasses by far the sensitivity of radioactive counting methods. An important example is the study of 14C (half-life ¼ 5730 years), generated in the atmosphere by cosmic rays, in connection with radiocarbon dating. The observed β-ray counting rate from 1 g of contemporary carbon of biological origin is about 15/ min/g. However, this low counting rate is supported by the presence of 6.5 1010 atoms of 14C in the 1-g sample. If 14C atoms could be counted efficiently by accelerator mass spectrometry, it would be possible to determine the 14C content of very small quantities of carbon. This has now been accomplished for milligram carbon samples even though the ratio 14C/12C is near or below 1012. A comparison of measurements of long-lived radioisotopes at natural levels with β-ray counting and AMS (after Elmore and Phillips, 1987) is shown in Table 4.15. Atomic mass spectrometry is inherently sensitive and by its nature provides isotopic information. The goal of methods of elemental analysis based on mass spectrometry is to produce a spectrum of singly charged atomic species. Again this can be a problem for elements, such as uranium, which readily form oxides. If molecular or multiply charged species “contaminate” the atomic mass spectrum, they can give rise to background signals at the mass of interest, or if these molecular ions contain the element of interest, then the signal due to that element is distributed between the atomic singly charged ion, the multiply charged ions and any molecular species formed. Atomic mass spectra are simple to interpret, however great care must be taken to avoid molecular interferences especially at very low concentrations. SIMS analysis of electrodeposited 232Th α-particle sources gives rise to higher signals for ThO+ and ThO+2 than for Th+. This leads to difficulties in quantitation as the oxide to atomic ion ratios will be sensitive to local oxygen concentrations. Isobaric atomic interferences also present a problem: 99Ru and 99Tc, for example, have the same nominal mass and cannot be discriminated between on the basis of mass except by high-resolution mass spectrometers. Even with high mass resolution, if the interfering isobar is present in excess then discrimination at high mass resolution will be difficult and in any spectrometric method there is a trade-off between resolution and sensitivity. If there is a vast excess of an isotope of adjacent mass, even this may interfere with the signal of interest. The ability for a mass spectrometer to discriminate against such an interference is termed the “abundance sensitivity.” Methods such as ICP-MS and TIMS must discriminate against isobaric interferences by chemical separation methods prior to instrumental analysis.
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Table 4.15 A comparison of measurements of long-lived radioisotopes at natural levels with β-ray counting and AMS Radioisotope 10
Half-life (years) Stable isotopes Stable isobar Chemical form Sample size (mg) Atom per sample AMS run time (min) Decaycounting interval (years)
14
B
C
26
5730
7.05 10
9
13
12
10
14
1.6 10
6
Be
C N
36
Al 5
129
Cl
4.05 10
5
I
1.57 107
C
27
26
36
Ar AgCl
S AgI
Al
35
Cl
36
B BeO
C
Mg Al2O3
0.2
0.25
3
2
2
2 105
2 105
4 105
5 105
2 106
10
7
40
30
20
110
3
250
86
1130
After Elmore and Phillips (1987).
AMS, which is most commonly used for radiocarbon dating, discriminates against interferences in a number of ways. The 14N interference in 14C measurement is removed by generating a beam of anions and relying on the instability of the nitrogen anion. Molecular interferences are removed by high-energy (several MeV) collisions in a gas cell or thin foil. Further discrimination can be achieved by charge stripping in the same collisional processes to produce highly charged ions or even ions in their maximum charge state. The combination of discrimination processes used depends on the isotope of interest and the potential interferences. RIMS approaches the same problem by selectively ionizing only the element of interest prior to mass spectrometric separation. A selectivity of approximately 1 in 105 can be achieved per resonant excitation step in the ionization process and two or three such steps are frequently used. In combination with the selectivity of the mass spectrometer the method should potentially offer elemental selectivities in excess of 1015.
4.4.3.1 Inductively coupled-plasma mass spectrometry ICP-MS uses inductively coupled plasma as an ion source for a mass spectrometer. The basic units of an ICP-MS system, in the order used, are the sample introduction device, the plasma, the plasma/mass spectrometer interface, the ion focusing/ion filtering system, the detector and the data acquisition/data handling system (Fig. 4.49). Beauchemin (1992) gives a helpful comparison between ICP-MS and ICP-OES. The
4.4 Nonradiometric methods
Computer
Ion lenses Quadrupole
MCS
To vacuum
Plasma
Ion filtration
Interface
Detector
Data acquisition and handling
Sample introduction
FIG. 4.49 Component of a typical ICP-MS system.
sample introduction device introduces liquid samples as either a dry vapour or a fine mist into the plasma, with several options available. These include pneumatic nebulization (the most common), ultrasonic nebulization, electrothermal vaporization (ETV, which uses a graphite furnace), flow injection (Denoyer and Stroh, 1992) and direct injection (Wiederin et al., 1991). The transport efficiencies of sample into the plasma for pneumatic nebulization, ultrasonic nebulization and ETV are about 1%, and 100%, respectively. Laser ablation is a common method for introducing solid samples into the plasma. This and other sampling methods for solids are reviewed by Baumann (1992). Inductively coupled argon plasma is used most frequently in ICP-MS, with argon as the cooling, carrier and auxiliary gas. The high-temperature plasma (5000–8000 K) is sustained by radio frequency fields at the tip of a quartz torch. The plasma desolvates (if necessary), atomizes, and ionizes the sample. Horlick (1992) notes the use of helium or nitrogen-based microwave-induced plasmas to eliminate interference from argon-based background species. Smith et al. (1991) and Lam and McLaren (1990) have used mixed carrier gases to reduce argon-based background ions. The plasma/mass spectrometer interface allows the import of a stream of ions from the plasma, at atmospheric pressure, into the mass spectrometer, which is under vacuum. The interface has a sample cone and a skimmer cone, usually of nickel and typically with 1.0 mm and 0.8 mm apertures, respectively. The turbomolecular pump is presently the most widely used type in the vacuum system. The extraction lens, held at a negative voltage, attracts the positive ions as they emerge from the skimmer cone. The negative ions are repelled and the neutral species diffuse away. The accelerated positive ions are focused by the ion-lens stack and then enter the quadrupole region. Positioned between the extraction lens and the first ion lens, the photon stop prevents stray photons from reaching the detector. Voltages
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applied to the quadrupole, four metal rods mounted in a square array, produce an electric field that affects the trajectories. For any specific applied voltage, only ions of a very narrow range of mass/charge (M/Z) ratios have stable trajectories and reach the detector. A single M/Z ratio may be continuously focused onto the detector, or a selection of M/Z ratios can be sequentially focused onto the detector. A channel electron multiplier, the commonly used detector, responds to each incoming positive ion and produces a measurable pulse. The number of pulses measured is proportional to the number of ions of the selected M/Z ratio reaching the detector. In the pulse counting mode, maximum gain is obtained by applying a high voltage to the multiplier so that individual ion arrivals at the detector are recorded and ultimate detection limits are obtained. The analogue mode uses lower voltages, which produce lower gains. The use of lower gains extends the useful analytical range but results in higher detection limits. A Faraday cup replaces the analogue mode in some instruments. The data acquisition/data handling system consists of a multichannel scalar (MCS) and a computer system. Signal pulses from the detector accumulate into memory channels of the MCS according to their M/Z ratios. Signal pulses from replicate scans are sorted into the appropriate channels and accumulated until the analysis of that sample is complete. A computer program retrieves the totals from the MCS memory and stores the data for later manipulation or display. As mass spectrometry has continued to gain sensitivity and reliability, inductively coupled-plasma/mass spectrometry (ICP-MS) has become increasingly useful in the measurement of radionuclides. The optimization of ICP-MS is improving our ability to use the atomic detection of radionuclides in that it allows the near-complete isotopic analysis of any form of sample. Aqueous samples are generally introduced into the plasma source, and solids or individual particles, and organic solutions, may be atomized and continuously introduced into the plasma source. ICP-MS sensitivity, which is currently 8 109 atoms, can be improved by • • •
the use of more efficient sample introduction techniques, understanding of the basic principles of ion and gas dynamics in the ICP-MS interface and the use of high-resolution mass spectrometers with high-ion transmission.
The ultimate sensitivity could approach 107 atoms, which would result in a superior detection capability for all radionuclides with half-lives >1 year. For radionuclides with half-lives of thousands of years and longer, ICP-MS has two principal advantages over radiation counting, which are speed of measurement and sensitivity. Most radiation counting times range from 50 to 2500 min for most samples and most backgrounds. In contrast, an ICP-MS analysis requires only a few minutes per sample or blank, whether it is introduced via nebulizer, ETV unit or other device. The analysis time is independent of the half-life or decay scheme of the radionuclide. The analysis time is also not greatly lengthened by a lower required MDA. Indeed, this advantage of ICP-MS over radiation counting becomes greater with increasing half-life and decreasing MDA. Analysis by ICP-MS may be the preferred
4.4 Nonradiometric methods
method even when its sensitivity is not as great as the one obtained by radiation counting, because of its speed. Using ICP-MS quantitatively becomes feasible for radionuclides with half-lives greater than about 1 103 years. However, the sensitivity that is routinely achievable with ICP-MS is not as high as that achieved with radiation counting for radionuclides with half-lives less than about 1 104 years, unless the decay scheme is unfavourable for radiation counting. For present-day instruments, at least 107–108 atoms are necessary to qualify a nuclide by an ICP-MS with an ETV unit (Smith et al., 1992). To use an ultrasonic nebulizer requires at least 108–109 atoms; while a pneumatic nebulizer requires at least 109–1010 atoms. The corresponding masses vary according to the atomic weight of the nuclide. (For 239Pu, 2 107 atoms equals 8 fg). Under favourable operating conditions, an instrument with an ETV unit should just meet an 8 fg detection limit for 239Pu (t1/2 ¼ 2.41 104 years). This equals an MDA of 0.001 dpm (1.7 105 Bq), which is 5 times lower than the contractual MDA given earlier. This corresponding MDA for 2 107 atoms of 240Pu (t1/2 ¼ 6.57 103 years) is 0.004 dpm (6.7 105 Bq). The sensitivity of ICP-MS for heavier elements is better than for lighter ones, due to lower background in the higher mass region and more stable trajectories for more massive ions ( Jarvis et al., 1992). This is an advantage when one is interested in analysing for long-lived radionuclides of the rare earth and heavier elements. Essentially the entire inert sample matrix needs to be removed when performing radiation counting of all α-particle emitters and low-energy β-particle emitters because of sample self-absorption. Complete matrix removal may not be necessary for analyses by ICP-MS, depending on the elemental composition of the sample, the analytes, the sampling device and the required sensitivity. Partial or less-complex matrix decompositions and separations of analytes may suffice. For example, Hursthouse et al. (1992) compare the extent of chemical purification necessary to obtain good results for 237Np via ICP-MS, α-particle spectrometry and neutron activation analysis. Depending on the dissolved solids content, natural waters may need filtration only and/or treatment with acid. A chemical separation of a group of elements may be satisfactory. Preparation of purified samples for α-particle spectrometry is usually by either electrodeposition or micro-coprecipitation. Either technique takes at least an hour. Many β-particle emitters are precipitated with several milligrams of carrier and weighted for determining the chemical yield prior to counting, which also takes time. In contrast, a few millilitres of solution are satisfactory for ICP-MS. If the analyte concentration should exceed the linear part of the calibration curve, a simple dilution overcomes the problem. Mixtures of β-particle-emitting nuclides of more than one element usually have overlapping spectra. This is also often true for α-particle-emitting nuclides. Some of these mixtures can be analysed by ICP-MS without internal interferences. For example, the α-particle spectra of 237Np and 242Pu partially overlap even under the best of conditions, whereas ICP-MS is appropriate for analysing the long-lived radionuclides
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in a NpdPu mixture. Only X-and γ-ray counting are comparable to ICP-MS in the number of radionuclides that can be measured simultaneously. As with all analytical techniques, ICP-MS has its problem areas (Olesik, 1991). Although the mass spectrum of a sample is usually much simpler than an atomic emission spectrum for the same sample, spectral interferences from isobaric interferences and peak overlap can still be a problem. Isobaric interferences result from two situations. The first occurs when two elements in a sample have nearly identical M/Z values. An example of this is the presence of 113In+ interfering with the analysis of 113 Cd+. The second situation occurs due to the formation of “background species.” These are ions, usually polyatomic, formed from the plasma gas alone or in combination with elements from the solvent used in the sample preparation. Examples, together with the ions for which they cause the greatest interference, are Ar+2 (80Se+), ArO+ (56Fe+), Ar+ (40Ca+ and 40K+) and O+2 (32S+). Tables of common background species for an argon plasma are readily found in the literature. Most of these interfering species have m/Z < 81. Fortunately, most long-lived radionuclides have masses >81 amu. Peak overlap occurs when major constituents in the sample have such massive peaks at particular M/Z value channels. Examples of this type of interference would be a massive peak for uranium at mass 238 that tails into the 237 and 239 mass channels, thus complicating the analysis of 237Np+ or 239Pu+. Other problems may be caused by the matrix of the sample itself. If chloride is present, a series of polyatomic chloride-containing species may cause major interferences. As an example, 40Ar35Cl+ is an intense peak that interferes with 75As+ Arsenic is monoisotopic; therefore appreciable levels of chloride in the sample will seriously compromise the precise determination of arsenic. Components of the sample matrix may also contribute to oxide formation. Oxides of the form MO+ give rise to peaks at the (M/Z) + 16 position. One or more of these may interfere with nuclides of interest. An example of this is 48Ti16O+ interfering with the analysis of 64Zn+. The four other naturally occurring titanium isotopes would, of course, also give interference at their respective (M/Z) + 16 values to any analytes with these masses. Formation of the oxide of the analyte also reduces the signal measured at M/Z. The sample matrix may also induce changes in the analyte signal intensity. High concentrations of contaminant elements generally cause suppression of an analyte signal, although under certain conditions signal enhancement has been observed. In general, the lower mass elements are more subject to suppression than higher mass elements, and higher mass elements are more likely to cause signal suppression of lower mass elements than the reverse ( Jarvis et al., 1992). An ICP-MS instrument will not tolerate dissolved solids at concentrations that can be run with an ICP-atomic emission spectrometer. In addition to increasing the probability of interelement (isobaric) interferences and signal suppression, high levels of dissolved solids condense on the sample cone orifice. This deposition degrades the sensitivity and stability of the analytical signal. Typically, a maximum of 0.1% dissolved solids is recommended for continuous nebulization with a pneumatic nebulizer. Dissolved solids should be kept below about 0.01% with an
4.4 Nonradiometric methods
ultrasonic nebulizer, due to its desolvation effect. Liquid samples containing up to about 1% dissolved solids can be run with ETV and flow injection. The sensitivity achievable for an element is inversely related to its ionization potential (I.P.). Thus, for example, the ICP-MS sensitivity for iodine (I.P. ¼ 10.34 eV) is not as good as it is for nearby caesium (I.P. ¼ 3.89 eV). Finally, the sample aliquant is consumed during an ICP-MS measurement, whereas with radiation counting the sample aliquant usually can be retained and can be remeasured. Some of the reported ISP-MS applications to radionuclide determination are presented in the following. Kim et al. (1989a) measured the 240Pu/239Pu ratio in two soils and an estuary silt after performing radiochemistry. They also measured this ratio in these soils by the fission track-etch technique and found that ICP-MS gave better precision. They measured the 239Pu concentration on separate aliquants by α-spectrometry, with 3.76 105-year 242Pu tracer as the chemical yield monitor. However, at the 239Pu concentrations in their samples, ICP-MS could have measured these directly. Determination of low levels of 99Tc in environmental samples by ICP-MS was reported by Nicholson et al. (1991). In earlier work Nicholson et al. (1989) used ICP-MS with the β-particle-emitting nuclide 95Tc, in salt marsh soil, seaweed and seawater. They employed 95mTc as a yield tracer, and radiochemically isolated Tc from the matrix. They confirmed the chemical removal of any interfering isobaric 99 Ru by monitoring for other stable Ru isotopes. The chemical yield was measured by γ-ray spectrometry. Beals (1992) used 2.6-million year 97Tc as the yield monitor in ICP-MS measurements of 99Tc in river water, thereby eliminating the need for a separate yielding measurement. In purified water, they have reliably detected 0.05 ng/mL of very low specific-activity 113Cd by ICP-MS with pneumatic nebulization. For environmental waters, better sensitivities could be achieved by sample concentration and a clean-up that includes the removal of interfering 113In. It is entirely impractical to detect 0.05 ng of 113Cd by β-particle counting. James et al. (1989) demonstrated, with diluted aqueous standards, an MDA of approximately 8 fg of 239Pu (1.7 105 Bq) with ETV-ICP-MS in the peak dwelling mode. Comparison measurements by ICP-MS and α-spectrometry on radiochemically processed urine with moderately higher activity gave good agreement, considering the very low amount present. Plutonium-244 (t1/2 ¼ 8.3 107 years) is also amenable to analysis by ICP-MS, whereas 238Pu is not. ICP-MS was found to be compatible with LC for the trace metal speciation. The role of ICP-MS in trace element speciation studies at the FSL was described (Crews et al., 1987). The characteristics of LC-ICP-MS for the study of metalloprotein species were assessed and the chromatographic efficiency of ICP-MS was found to be similar to that obtained with a UV detector (Dean et al., 1987a, 1987b). Information about the chemical nature of trace elements from food can be obtained by first treating the foods in vitro with enzymes to broadly simulate the action of enzymes in the gastrointestinal tract (Crews et al., 1988). The soluble components can be separated by size-exclusion chromatography (SEC) and an estimate of their molecular size obtained. By coupling SEC directly to ICP-MS, the trace element content of the
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chromatographic fractions can also be measured. This approach has been used at the FSL to investigate the speciation of cadmium in raw and cooked pig kidney (Crews et al., 1989). The sensitivity of ICP-MS enabled the researchers to study retail samples in which the levels of multielement data obtained indicated that, while the feeds were contaminated with a number of elements, only lead presented a serious problem in parts of the rest of the food chain. For example, while samples from affected cattle were not allowed to enter the food chain, experiments with meat on contaminated bones showed that lead did not migrate significantly from the bone under a variety of cooking conditions (Baxter et al., 1992). Kim et al. (1991) have reported the measurement of some long-lived radionuclides, such as 99Tc, 226Ra, 232Th, 237Np, 238U, 239Pu and 240Pu using high-resolution inductively coupled-plasma mass spectrometry (HR-ICP-MS). By using HR-ICPMS with an ultrasonic nebulizer, the detection limits of these nuclides were 0.002–0.02 pg mL1 and the sensitivities were 10 times better than those obtained using HR-ICP-MS without the ultrasonic nebulizer. More accurate isotopic data were also obtained using HR-ICP-MS than with quadrupole ICP-MS at lower concentrations of the analyte because of improvement in counting statistics that can be obtained with HR-ICP-MS due to the greater efficiency of ion transmission. Morita et al. (1993) have applied ICP-MS to the determination of technetium-99 in environmental samples. The determination of eliminating the interfering element (Ru) before the ICP-MS measurements is made. Technetium-95m is used as the chemical recovery tracer. Compared with conventional methods, the method sensitivity is 10–100 times higher and the counting time is 300–10,000 times shorter. Tye and Mennie (1994) have reported development of the performance of a new interface for the Plasma Quad ICP-MS, which enhances the signal-to-noise performance of the standard instrument by a factor of 10. In order to maintain the flexibility of the instrument, the new interface is designed such that the enhanced performance can be easily switched on and off, offering the benefit of routine performance plus the high sensitivity mode when required. These improvements in signal to noise make it possible for the routine monitoring of many actinide elements directly by ICP-MS, potentially shortening the analytical cycle from days to hours. A further improvement on these impressive limits of detection is possible if a high-efficiency nebulizer is used to introduce samples into the new instrument, giving the capability of single figure ppq detection limits in an analysis which takes minutes, not hours.
4.4.3.2 Accelerator mass spectrometry For isotopes with long lifetimes (>1 year), it may often be more advantageous to use atom-counting techniques rather than traditional decay-counting methods. This is especially true for measurements where efficiency is a criterion, as for small samples, or if high precision is required. While atom counting has a counting rate that is essentially independent of decay lifetime and sample size, the decaycounting rates are comparable only if the isotopic half-life is less than 1 year for a sample size of the order of 1 mg. Of course, if sufficient material is available,
4.4 Nonradiometric methods
Table 4.16 Long-lived cosmogenic isotopes detected with accelerator mass spectrometry Interfering stable Isotope 10
Be C 26 A1 36 Cl 41 Ca 129 I
14
a
Half-life (years) 1.5 106 5.7 103 7.2 105 3.1 105 1.3 105 15.9 106
Isotopes
Isobars
9
Be
10
12, 13
14
C
27
Al
35,37 40,42 127
I
Cl Ca
B N 26 Mg 36 S 41 K 129 Xe
AMS detection limita
Range of terrestrial concentration a
7 1015 0.3 1015 10 1015 36 Ar 500 1015 100 1015
108–1014 1012–1016 1014 0.2 1015 1015–1016 1016
Compared to the stable isotope of the same element.
the decay-counting rate can always be improved by using more material (Litherland, 1987; Kilins et al., 1992). AMS extends the capabilities of atom counting using conventional mass spectrometry, by removing whole-mass molecular interferences without the need for a mass resolution very much better than the mass difference between the atom and its molecular isobar. This technique has been used with great success for the routine measurement of 14C, 10Be, 26Al, 36Cl and, recently, 129I (see Table 4.16). Analysis of 14C by AMS can, for example, generate dates with a precision that is at least equal to the best conventional β-particle-counting facility. In many cases, where small sample analysis is required, the AMS method has proved superior (Beukens, 1990). A complete description of AMS can be found in review articles (Litherland et al., 1987; Elmore and Philips, 1987) or recent conference publications. The application of AMS to 129I measurement has been discussed in detail in Kilins et al. (1992). AMS is an analytical technique that uses an ion accelerator and its beam transport system as an ultrasensitive mass spectrometer. AMS was introduced in Muller (1977), who suggested that a cyclotron could be used for detecting 14C, 10Be and other long-lived radioisotopes, and independently by the Rochester group, who demonstrated that 14C could be separated from the isobar 14N by relying on the instability of negative ion 14N. Presently AMS measurements are being made at about 30 accelerator laboratories around the world, and half of these are dedicated to AMS measurements of long-lived radioisotopes. Six long-lived radionuclides beyond uranium exist which have half-lives >100 ka (236Np, 237Np, 242Pu, 244Pu and 248Cm). The first two are natural byproducts of the nuclear industry. Nuclear weapons tests will generate the plutonium and curium isotopes although attempts have been made to detect presolar system 244 Pu in ores (Hoffman et al., 1971) or 244Pu from more recent supernova debris. The detection of these isotopes is still in the development stage. Unlike the natural elements, isobaric interferences are not a major problem as all isotopes will be
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Stripper AT +HV terminal
Negative ion source
Velocity selector
ES analyzer
Magnetic analyzer
Preaccelaration Injector
Tandem accelerator
Magnetic analyzers
Gas-filled ionization chamber Time-of-flight Gas-filled magnet
ES analyzer Detector system
Positive ion Analysis
FIG. 4.50 Components of a typical AMS system.
equally rare or nonexistent because of their very short decay half-lives compared to the lifetime of the solar system. The components of accelerator mass spectrometry (AMS) system are shown in Fig. 4.50 and they include ion source, injector, tandem accelerator, positive ion analysis and detection system. A caesium sputter ion source is used for most AMS work. This is essentially a secondary-ion mass spectrometry (SIMS) instrument that has been refined to produce high current of negative ions. Generally, solid samples are used; gas samples can give intense beams, but the problem of contamination from the previous sample (memory) is difficult to overcome. For radioisotope studies, sample sizes are 1–10 mg of processed material and beam currents of 1–50 μm are typical, depending on the element and ion source model. Some sort of multiple sample changing system is used at most AMS installations. For example, the main features of the 846B model high-intensity sputter source (High-Voltage Engineering Europa) include: a hemispherical ionizer giving a focused Cs beam spot of <0.5 mm, an x–y scanning stage to limit cratering effects and a 60-sample carousel with automated remote loading for throughput work. Currents of up to μA 12C have been quoted for this source from graphite targets. Mass analysis of the negative ion beam with a resolution sufficient to separate isotopes of heavy elements is needed prior to acceleration. For example, an electrostatic analyser is used at the University of Toronto to sharpen the energy distribution of ions produced from a caesium sputter ion source. A preacceleration of the negative
4.4 Nonradiometric methods
ion beam to 100–400 keV is used with large tandem accelerators to ensure that the injected ion beam is focused at the central terminal where the stripper canal is located. The name “tandem” refers to a dual acceleration design. The negative ions are accelerated to the terminal of the accelerator, which is held at a constant positive voltage, typically in the range 2–10 MV. The electron stripper at the terminal removes several electrons while energetic negative ions pass through; positive ions are then accelerated from the terminal to the end of the accelerator (ground potential). Tandetrons operate reliably below 3 MV using a solid-state power supply, and tandem Van de Graaff accelerator use a rotating belt or chain to charge the terminal up to 25 MV in some models. Tandem accelerators have the following characteristics: (i) the ion source and detector are located at ground potential for tandetrons; (ii) they do not require pulsed beam; (iii) the electron-stripping step need to eliminate molecules is an integral step in the operation of the accelerator and (iv) transmission through the accelerator and subsequent analysers can be made independent of small changes in the terminal voltage. Analysers positioned after the accelerator remove scattered particles accepted by the injector analyser, molecular fragments and unwanted charge states. Magnetic analysers alone are not sufficient. An electrostatic analyser or velocity selector is necessary to remove particles that have different mass but would otherwise have the correct mass-energy product to pass through the magnetic analysers. At 1 MeV/amu energies, the dE/dx and total energy measurements are made with either gas ionization detectors or silicon surface-barrier detectors or a combination of these. The time-of-flight detector serves as an additional positive ion mass analysis stage. It is most useful for the heaviest (slowest) ion such as 129I and consists of two time-pickoff detectors with time resolution of a few hundred picoseconds. Isotope ratios are obtained by alternately selecting each stable isotope and measuring its beam current in a removable or offset Faraday cup and then by measuring the radioisotope (rare nuclide) counting rate in the detector. Standards (samples with a known isotope ratio) are periodically measured for normalization, and blanks (samples containing no detectable nuclides to be measured) are used to measure background. Ratios are corrected for time-varying linear mass fractionation when more than one stable isotope is measured and for nonlinear fractionation, which arises from the stripping process and from stray magnetic fields in the accelerator, by comparison to the standard. The precision of ratios ranges from 1% to 10%, in the AMS measurement, depending on the value of ratios and counting time (if background is low enough). The long-lived radioisotopes 10Be, 14C, 26A1, 36C1, 41Ca and 129I can be measured now in small (mg) natural samples having isotopic abundances in the range 1012–1015 and as few as 105 atoms. At elevated energies (>1 MeV/amu), ions can pass through thin or equivalent gas with virtually no attenuation of the particle beam and little energy loss. As a result of electron capture and loss interaction, an ion passing through matter isPcharacterized by the fraction of the total ions (Fq) in a given charge state (q) where Fq ¼ 1. The resulting charge-state distribution is determined by the electron capture and loss
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cross section of an ion in a gas or solid. An equilibrium distribution will be established, the character of which depends only on the ion velocity and the target material. This equilibrium distribution is independent of the initial ionic charge or the target thickness, and approximation is valid as long as the energy loss remains insignificant. The passage of an ion through matter with the subsequent removal of electrons from molecules will decrease the bond strength among the constituents. Generally, after a reduction of two electrons in ionically bound molecules, no bond is possible and the molecule is broken up by Coulomb force. A sufficient number may remain to leave the charged molecule in a stable or metastable configuration. To avoid the possibility of long-lived (>1 ms) metastable molecules, at least three electrons must be removed. At present no 3 + molecules are known to exist. For light ions (Z < 20) an energy of at least 3 MeV is needed to maximize the production of charge state +3 ions in a gas cell. As the mean ionic charge rises approximately linearly with energy, the higher charge states will dominate at the 8 MV accelerating potentials. Usually one of these charge states was chosen to provide the highest conversion efficiency and no molecular interference. Separation of isobars can be accomplished by using different approaches: (i)
By chemical separation
No two isobars will have the same atomic number and rarely will they belong to the same chemical group of elements. Consequently, an initial reduction of the isobaric interference is always possible by some form of chemical processing. The degree to which this is effective depends on the specific chemical differences among the isobars and the level of isobaric contamination. Because the mass ambiguity is always formed from stable elements which have had the benefit of the last 4–5 109 years to achieve some level of contamination in all materials, chemistry cannot usually eliminate this type of ambiguity significantly below 1 ppb. One part in 106 (1 ppm) is the typical level of purity for most reagents. At this level of refinement, the interference is still at least many orders of magnitude greater in concentration than the rare long-lived radioisotopes. For some very rare isobars, specifically 36S (0.017%) in the case of 36Cl, careful chemical processing is the primary means of isobaric reduction. (ii)
Using negative ions
Some isobars can be eliminated by exploiting the instability of negative ions. For example, noble gas negative ions are known to be metastable or unstable, thereby removing 36Ar and 129Xe as a source of interference in the measurement of 36Cl and 129I. Others such as 14N and 26Mg do not form readily or have metastable states which decay in a time scale ≪ 1 ms. Since the lifetimes of these metastable states are small in comparison to the transit time (>1 ms) through the ion analysis system, attenuation factors in excess of 106 were possible with negative ions for the detection of 14C and 26Al. In contrast to the high probability for scattering and multiple charge changes for positive ions used at low energies (keV), the scattered negative ions from the more intense isobaric beams are greatly reduced in intensity after an interaction with the
4.4 Nonradiometric methods
residual gas in the ion source. This can be attributed to the low binding energy of the negative ions. (iii)
By fully stripping the electrons
At sufficiently high energies and for cases where the radioisotope has a higher Z than the stable isobar, separation by fully stripping with subsequent magnetic analysis can be accomplished. This method has been investigated for the systems 3He2+–3H1+, 26 Al13+–26Mg12+, 36Cl17+–36S16+, 41Ca20+–41K19+, 53Mn25+–53Cr24+ and 59Ni25 + 59 – Co24+. Isobar separation by fully stripping actually allows the simultaneous acceleration of the radioactive and stable isobar, with the latter of sufficient intensity to provide beam feedback signals. (iv)
By energy loss measurement
When ions with energies about 1 MeV/amu are passing through matter, the energy loss per unit path length (dE/dx) of an ion (Z) traversing a solid or gas with velocity (V) is governed by the Bethe–Block equation: dE/dx ¼ k (Z/V)2 where k is a constant. The effectiveness of energy loss measurements decreases as the atomic number increases. The percentage difference in energy loss between 14C and 14N is 30% at 40 MeV in isobutane gas. This rapidly decreases to 6.8% between 36Cl and 36S. Beyond calcium, very high energies (>100 MeV) are required. (v)
By gas-filled magnet
The gas-filled magnet is a powerful isobar separation instrument developed only recently for AMS. An ion passing through a gas changes its charge frequently by electron capture and loss. If this charge-changing occurs frequently enough in a magnetic field region, the trajectory is determined by the average charge state of the ion, which depends on Z. Development of the gas-filled magnet should make possible AMS of 36Cl at lower energies and aid in AMS detection of other heavier isotopes. (vi)
By lasers
Lasers are very powerful instruments to separate elements. Since the separation of isobars from different elements is the most difficult task in AMS, the use of lasers in connection with AMS could provide a very effective clean-up of background. The basic idea in a recent proof-of-principle experiment at the Rehovot AMS facility was to clean a negative ion beam from unwanted isobaric background ions by selective electrons detachment. 32S ions which have an electron affinity of 2.08 eV were effectively neutralized by interaction with 2.33 eV photons from a pulsed Nd: YAg laser. The same photons did not affect 37Cl ions whose electron affinity is 3.62 eV. This clearly demonstrated that a laser depletion of 36S background in 36 Cl measurements is feasible, opening up the possibility for sensitive 36Cl measurement at small AMS facilities where the ion energy is too low to perform isobar separation. However, for actual applications in AMS measurements, a substantial improvement in overall efficiency of the laser depletion process is necessary. The developments in instrumentation include production of dedicated machines. For example, with the introduction of the “Attomole 2000,” HVEE (High Voltage
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Engineering Europa B.V., P.O. Box 99, 3800 AB Amersfoort, The Netherlands) has made available a compact 14C isotope ratio mass spectrometer (14C IRMS) for biomedical applications. The system provides 14C/12C ratios down to 1013 from submilligram samples in typically a few minutes. Both solid samples (carbon) as well as CO2 can be analysed. The Attomole 2000 combines a compact instruments package with the extreme sensitivity of large tandem accelerators that are normally found in big research centres. The main application of the instrument will be for tracer kinetic and pharmacological measurements in biomedical studies. Essentially, 14C IRMS is an alternative technique to 13C Isotope Ratio Mass Spectrometry (13C IRMS), which is widely accepted within the biomedical community. 14C IRMS is at least 100–1000 times more sensitive than 13C IRMS. This allows, e.g., that the kinetics of 14C-labelled enzymes, aminoacids or carcinogens can be studied in the human body at levels that are comparable to actual environmental exposure. However, a widespread acceptance of 14C IRMS in the biomedical community is presently handicapped by the size of the existing 14C IRMS systems and their need for expensive support personnel. These shortcomings of the present instrumentation are overcome with the introduction of the Attomole 2000. The system is a compact, turnkey instrument that is user friendly. Its characteristics are reflected in the following specifications: System footprint Sample medium Required sample size Output Accuracy Counting rate Background Detection efficiency Detection limit Throughput
2.22 1.25 m2 Solid graphite or CO2 Solid graphite minimum 100 μg; CO2 sample 0.2 μmol CO2 (minimum 10 modern) 14 C/12C ratios Better than 3% > 250 cnts/s. for 10 modern samples < 1013 for 14C/12C ratios (0.1 modern) 1%–2% Approx. 4 attomole (1018 mol) 14C for 3% accuracy 400 sample/day (based on samples with an average of 4 modern 14 C content)
Furthermore the system is fully automated, self-tuning and needs little or no maintenance. The operator will consider the instrument as an analytic tool; the fact that an accelerator is involved is incidental. Up to 50 solid graphite samples can be loaded in a carousel prior to analysis. CO2 samples can be admitted on line to the ion source. The ion source uses a primary caesium beam to sputter the sample under investigation to form a negative carbon ion beam. The ion beam is accelerated through the system to reach the detector with an energy of 2.5 MeV. A detailed description of the system can be found in Mous et al. (1997).
4.4 Nonradiometric methods
4.4.4 Laser-induced photoacoustic spectroscopy Growing interest has been recently directed to the application of photoacoustic sensing techniques to the spectroscopic analysis of various optical absorbers in very dilute concentrations. For this purpose a laser is commonly used as a light source. Since the discovery of the photoacoustic effect by Bell, 1880, its application has a long history of development. Renewed interest in photoacoustics has emerged, starting with the work of Kreuzer (1971) who analysed trace amounts of gas molecules by laser-induced photoacoustic spectroscopy (LPAS) generation. The theory, instrumentation and application of laser-induced photoacoustic generation developed in recent years have been thoroughly reviewed by Patel and Tam (1981) and more recently by Tam (1983, 1986). Other reviews are also available in the literature from different authors: Pao (1977), Somoano (1978), Rosencwaig (1978), Colles et al. (1979), Kirkbright and Castleden (1980), Lyamshev and Sedov (1981), Kinney and Staley (1982), West et al. (1983) and Zharov (1985). Because of difficulties involved in handling radioactive preparations, the photoacoustic sensing technique had not been applied until some years ago to the spectroscopy of aqueous actinide ions. A relatively simple detection apparatus of photoacoustic spectroscopy for the spectral work of actinide ions using pulsed laser as a light source has been developed. This detection apparatus can be used for radioactive α-emitting aqueous samples without restriction to corrosive solutions and
Modulated light source, e.g., Laser pulse
Ion-specific absorption
Generation of heat by nonradiative relaxation
Modulated volume expansion
Generation and propagation of acoustic wave
Detection of compression wave by piezoelectric detector
FIG. 4.51 Generation and detection of photoacoustic signals.
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Radial intensity distribution
Sample cell
Temporal intensity distribution
Laser beam Halfwidth t P
Halfwidth R
Detector
FIG. 4.52 Geometry of laser beam and photoacoustic signal generation.
facilitates the spectroscopic investigation of actinide solutions, particularly transuranic ions, in very dilute concentrations. The creation of the photoacoustic signal in the solution is based on the conversion of absorbed optical radiation into heat by nonradiative relaxation processes. This principle is illustrated in Fig. 4.51. The geometry of laser beam and photoacoustic signal generation is schematically presented in Fig. 4.52. The spectroscopic system has been introduced to different nuclear chemical laboratories and further developed for a variety of purposes. Most of these developments are confined primarily to the spectroscopic investigation (i.e. speciation) of actinides in very dilute solutions or natural aquatic systems in which the solubility of actinides is, in general, very low (<106 mol L1). Optical spectroscopy of high sensitivity is an indispensable tool for the study of the chemical behaviour of actinides in natural aquatic systems, which has a newly developing research field in connection with nuclear waste disposal in the geosphere. For this reason, not only is photoacoustic spectroscopy attracting great attention but also thermal lensing spectroscopy and fluorescence spectroscopy, all using laser light sources, are in growing use for the same purpose. Actinides have particular spectroscopic properties which are characterized primarily by the f ! f transitions within the partially filled 5f shell and thus by a number of relatively weak, sharp absorption bands. The optical spectra of actinides are characteristic for their oxidation states, and to a lesser degree dependent on the chemical environment of the ion. Thus spectroscopic investigation provides information on the oxidation state of an actinide element and also serves to characterize the chemical states, such as hydrolysis products, various complexes and colloids. Hence, laserinduced photoacoustic spectroscopy with its high sensitivity can be conveniently used for the speciation of aqueous actinides in very dilute concentrations. For a summary of the present knowledge of laser-induced photoacoustic spectroscopy, as regards theoretical backgrounds, instrumentation and radiochemical applications to particular problems in aquatic actinide chemistry, see Kim et al. (1990). Since there is no other radiochemical application known in the literature,
4.5 QA/QC procedures
except the measurement of tritium decay by an acoustic sensing technique, the present discussion is limited to application to actinide chemistry, particularly in aquatic systems. The most interesting field of application is and will be the geochemical study of long-lived radionuclides, namely man-made elements (transuraniums). The main importance for such a study is not only the detection of a migrational quantity of radioactivity but also the characterization of their chemical states and hence their chemical behaviour in given aquifer systems. Knowledge of this kind will facilitate a better prediction of the environmental impact of transuranic elements which are being produced in ever-growing quantities and will be disposed of in the geosphere. Since LPAS application to actinide chemistry is in its infancy, only a limited number of works are available in the published literature. Experiments hitherto performed are confined to either hydrolysis, complexation reactions with carbonate, EDTA and humate ligands and a variety of speciation works for Am(III) and too much lesser extent for U(IV), U(VI); Np(IV), Np(V), Np(VI); Pu(IV) and Pu(VI). Of considerable interest is the LPAS application to the direct speciation of actinides in natural aquifer systems, where the solubility of actinides is in general very low and multicomponent constituent elements as well as compounds are in much higher concentrations than actinide solubilities. The study of the chemical behaviour of actinides in such natural systems requires a selective spectroscopic method of high sensitivity. LPAS is an invaluable method for this purpose but its application to the problem is only just beginning.
4.5 QA/QC procedures Quality assurance to determine radionuclides in food and environmental samples ensures that the quality of data obtained is maintained at an adequate confidence level, and is objectively evaluated. Quality assurance includes quality control, which involves all those actions by which the adequacy of equipment, instruments and procedures are assessed against established requirements. For the purpose of quality assurance, the following items must be ensured: (1) equipment and instruments function correctly, (2) procedures are correctly established and implemented, (3) analysis are correctly performed, (4) errors are limited, (5) records are correctly and promptly maintained, (6) the required accuracy of measurements is maintained and (7) systematic errors do not arise. In general, the design of a quality assurance program should take the following factors into account: a. quality of equipment and instruments, b. training and experience of personnel, c. verification of procedures by the routine analysis of control samples and the use of standard methods for analysis, d. frequency of calibration and maintenance of equipment and instruments (variability in the measuring system is an important aspect of this),
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e. the need for traceability of the results of determinations to a national standard and f. the degree of documentation needed to demonstrate that the required quality has been achieved and is maintained. It is important to have each item of the quality assurance program established. Intercomparison is also necessary to generally evaluate the quality assurance of the determinations. By this process, it is possible for data to be compared between laboratories or within a laboratory at different times. The concept of “quality control” should be discussed in general comparison with that of the concept of “quality assurance.” The basic concept of quality assurance is that quality should be assured comprehensively and wholly from the beginning to the end of a fixed volume of successive procedures. It assures that whole data acquired by using the fixed volume come to have a signification result to meet intended objectives. On the other hand, quality control is related only to definite and practical control of respective procedures, and is limited to only some portions of those procedures. Its main objective is being maintenance of quality of results within a specific limit. The scope of quality assurance is extended from one laboratory, to a group of laboratories in a region, then on to those in a country, and then to international groups of laboratories. The wider the scope of subjects to be assured the more effective the quality assurance. A smaller scope can be utilized as part of a larger one. This means that more effects are found in scopes of quality assurance in ascending order from a single laboratory to a region, a country, a continent, and to the whole world. The causes of errors which are treated as problems in quality assurance for radioactivity analysis and measurements are (1) collection of samples; (2) sampling; (3) transportation of samples; (4) labelling of samples; (5) storage of samples; (6) pretreatment of samples; (7) procedures for measurements; (8) measuring instruments; (9) human errors; (10) erroneous conversion; (11) reporting and notification; (12) environmental changes and (13) misinterpretation of data. The causes from items (1) collection of samples to (5) storage of samples are related to procedures to handle them. Quality assurance is established by comprehensively quantifying the partial uncertainty of errors which are to be generated from these causes. There are a few items for implementation: a. Organization must be implemented to aid in establishing quality assurance. First, a laboratory or research institute to play the central role has to be set up or appointed in a region or a country. It should act as a centre, play a leading role and handle the clerical work in that country. A committee may also be set up in the central laboratory for specialist members to offer guidance and advice. Ideally, a network should be formed to cover all the laboratories concerned, with the central laboratory acting as leader. In some smaller operations, a network will be composed only of laboratories who agree to join. The items to be executed for QA involve: • To form a network of quality assurance covering all laboratories concerned, with a central laboratory acting as leader. • For the central laboratory to check, compare and analyse the work of all the laboratories including itself.
4.5 QA/QC procedures
•
b.
c.
d.
e.
For the central laboratory and/or all the laboratories to conduct periodic calibration and stability checks of instruments. • For the central laboratory and other related laboratories to make comparative measurement and analyses either continually or periodically. • To carry out exercises related to the network. It is easier to maintain technological levels if instruments are subjected to periodic calibration and stability checks. These can be performed by the respective laboratories. Needless to say, checks should be done whenever operators are changed, instruments are installed, replaced, or moved, or environmental conditions are changed. Intercomparison and comparative measurements have to be conducted continually and periodically for assuring quality. Irregular and/or short-term checks never represent real assurance and have the least effect as a quality assurance system. Quality cannot be assured completely from the very beginning. Quality assurance will take time before it is refined. To improve the level of quality assurance, the following steps must be carried out: • unify the subjects to be sampled and the sampling methods; • standardize measuring procedures; • standardize specifications of measuring instruments; • have specialists to operate the measuring instruments; • standardize forms to make results accessible to all concerned and • understand regional characteristics. The procedures to keep up quality assurance levels must be incorporated into the quality assurance system by all laboratories. The list includes: • secure (a) operators with full expertise, (b) appropriate methods and (c) appropriate, well organized locations and space; • supply standard samples; • examine materials prior to application; • calibrate and adjust instruments; • make use of reference and standard samples which have appropriate records; • check quality assurance procedures; • effect continuous review of related data; • check whether objectives are met or not; • use divided samples; • compare data with those of other laboratories; • correctly handle requests; • review results in an organized manner and • correct errors, if any, by means of continuous measurements.
Quality control measures are necessary to provide documentation to show that the analytical results are reliable. This is very important since analytical results can form a basis on which economic, administrative, medical and/or legal decisions are made.
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It is essential to develop a quality assurance (QA) programme that covers sample collection, sample handling and methods for on-site and laboratory analysis, data handling and record keeping. The QA programme should address the variety of different scenarios likely to be encountered. Appropriate calibration and analytical standards and a variety of reference materials will be needed. To keep costs down, one should carefully design a QA programme that recognizes that for some signatures high-precision data are not required. If, for example, one analyses for a typical short-lived radionuclide which does not exist in nature, background measurements are unnecessary, however low the reported concentration. In other cases where one looks for faint anomalies in certain isotope ratios, the QA programme should demand knowledge of background values and their variability; this would be much more expensive. The protocols should include “blank” samples as well as “background” samples. In the case that an attempt is made to find an undeclared facility adjacent to a declared one, the analyst should try to take “background” samples from a plant somewhere else, which is similar to that part of the installation which is being examined. When attempting to find an undeclared nuclear facility at a declared site, the optimum background samples would be from similar facilities which are a part of the declared installation. Reliability of results is a function of precision (reproducibility) and accuracy (true value). The precision of results can easily be determined by internal measurement. The determination of accuracy in most cases, however, requires more detailed procedures such as the following: • • •
•
analysis by as many different methods, analysts and techniques as possible. In cases where agreement is good, the results are assumed to be accurate. control by as many different methods, analysts and techniques as possible. In cases where agreement is good, the results are assumed to be accurate. control analysis with reference materials that are as similar as possible to the materials to be analysed. Agreement between certified and observed values is then a direct measure of accuracy for that particular determination. Participation in an interlaboratory comparison. Samples used in such an intercomparison should be, as far as possible, similar in composition and concentration to the samples to be analysed on a routine basis. The agreement between the results received from a particular laboratory and the most probable mean value obtained from statistical evaluations of all the results will be a measure of the accuracy for that particular determination.
4.5.1 Intercomparison For practical reasons, most analytical laboratories are not in a position to check accuracy internally, without an external source of reference materials. To overcome some of the difficulties in checking the accuracy of analytical results, the IAEA provides the Analytical Quality Control Services (AQCS) programme to assist laboratories in
4.5 QA/QC procedures
Table 4.17 Analytical quality control services (AQCS) Year
Intercomparison
Reference available
Materials distributed
1986 1987 1988 1989 1990
24 24 33 27 19
39 38 46 50 58
1450 1680 2700 1800 1850
assessing the quality of their work. AQCS coordinates intercomparison studies and supplies reference materials. Participation is on a voluntary basis and at minimum cost. Information supplied by laboratories taking part in the intercomparisons is treated as confidential. The IAEA has traditionally played an important role in the development and testing of analytical methodology for the determination of radionuclides and through the AQCS programme provides a service by offering laboratories the option of determining their accuracy by distributing reference and intercomparison materials containing radionuclides in different types of materials. The analytes of interest in these samples include naturally occurring radionuclides and radionuclides of fission and activation products. The activities of the IAEA AQCS programme are shown in Table 4.17. Currently the orders for reference and intercomparison materials are running at the level of about 3000 units per year for the whole AQCS programme. The distribution of reference and intercomparison materials is coordinated by the Chemistry Unit of the Agency’s Laboratories at Seibersdorf, but it also receives input from other Sections of the IAEA, including the Hydrology Section, the Nutrition and Health Related Environmental Studies Section, the Safeguards Analytical Laboratory, Monaco. Intercomparison studies organized over the last 20 years are generally based on recommendations of consultants’ group meetings, and in response to the demands of many of the IAEA Member States for assistance in developing methodologies for the measurement of radioactivity. The Chemistry Unit distributes every 4 years a questionnaire concerning the need for organizing intercomparison tests and the preparation of reference materials. Using this data the AQCS programme collects different kinds of environmental and foodstuff bulk samples, some of which are affected by fallout radioactivity following the Chernobyl nuclear reactor accident. The general policy is to organize intercomparisons with those materials which are in most demand and have various levels of activity. Collection of a sufficient quantity of the raw materials (typically of the order of 2–400 kg) is first organized. The samples obtained by a sampling operation are generally dried, ground and homogenized. Aliquots are then taken at this stage and analysed to check the homogeneity of the bulk materials. Other preparation steps include aliquoting into bottles in amounts of about 25–100 g per bottle. To ensure long-term stability of the material, the sealed bottles are sterilized by γ-ray irradiation (Co-60 at a dose of 2.5 Mrad). A further control of
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homogeneity takes place after the materials have been distributed into bottles. Within-bottle and between bottle homogeneity is determined separately, usually by determining 40K, 137Cs, 90Sr and U. When this has been done, the material is announced in the AQCS Catalogue as an intercomparison material. Participants in such intercomparisons are provided with information about the material and special forms on which they are requested to report, for each element, up to six individual net results on a dry-weight basis, the sample weights used, information about the analytical method, and various other items. To preserve anonymity, each participant is assigned a code number, known only to him and the AQCS programme, by which he is identified in the report that is subsequently prepared on the results of the intercomparisons. The number of participants in each intercomparison varies but at present it is around 50. A chronological list of materials for intercomparisons which have been organized by the AQCS programme during a 9-year period is given in Table 4.18. Table 4.18 IAEA intercomparison exercises involving radionuclides during the 9 years: 1983–92 Matrix
Level
Alga, marine Milk powder Sediment, marine
Environmental Environmental Environmental
Soil Sediment marine Fish flesh
Environmental Environmental Environmental
Sediment, deep sea Sediment, like Air-filter, simulated
Environmental Environmental Artificial
Milk powder
Elevateda
Whey powder
Elevateda
Soil
Environmental
Sediment, stream
Environmental
Sediment, stream
Environmental
Milk powder
Environmental
IAEA code AG-B-1 A-14 SD-N-1/ 2 Soil-6 SD-N-2 MA-B-3/ RM SD-A-1 SL-2 IAEA083 IAEA152 IAEA154 IAEA312 IAEA313 IAEA314 IAEA321
Year
Certified as RM
1983 1983 1983
+ + +
1983 1983 1986
+ + +
1986 1986 1986
+ + +
1987
+
1987
+
1988
+
1988
+
1988
+
1988
+
4.5 QA/QC procedures
Table 4.18 IAEA intercomparison exercises involving radionuclides during the 9 years: 1983–92 Continued Matrix
Level
Clover
Elevateda
Seaweeds, Mediterranean Sediment, Baltic Sea
Elevateda
Sea-plant, Posidonia oceanica Uranium ore, phosphate
Elevateda Elevateda Environmental
Tuna homogenate, Mediterranean Sediment, Pacific Ocean
Elevateda
Soil
Elevateda
Grass
Elevateda
Cockle flesh
Environmental
Sediment, marine
Environmental
a
Natural
IAEA code IAEA156 IAEA308 IAEA306 IAEA307 IAEA364 IAEA352 IAEA368 IAEA375 IAEA373 IAEA134 IAEA135
Year
Certified as RM
1988
+
1988
+
1988
+
1988
+
1989
+
1989
+
1990
+
1991– 92 1991– 92 1992
1992
Contaminated with radioactive fallout from Chernobyl.
The results submitted by the participants are in all cases evaluated by the AQCS programme. A specific feature of any intercomparison is that gross errors occur quite frequently and results differing by as much as two or three orders of magnitude may be reported by participating laboratories. Various approaches and criteria for the detection and rejection of the highest and the lowest values or outliers have been discussed in the literature. The analytical data received in intercomparison exercises by the AQCS programme are treated using two different methods in order to derive a consensus value, which is considered to be a reliable estimate of the true value. Applying the first method, four different criteria, namely Dixon’s test, Grubbs’ test, the coefficient of Dewness test and the coefficient of kurtosis test are used at a significance level of α ¼ 0.05. If a laboratory mean for each element as single unweighted value was declared to be an outlier by any criterion, it is rejected and the whole procedure repeated until no more outliers could be identified. The remaining laboratory means are then combined in the usual way to provide estimates of the overall mean (consensus value) and its associated standard deviation, standard error and 95% confidence interval.
249
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CHAPTER 4 Measurements of radioactivity
The consensus values cannot automatically be accepted as recommended to certified values because their analytical validity usually requires a re-assessment in the light of additional analytical information such as concentration level, number of different analytical methods used, per cent of outliers and other criteria. In practice, certified or recommended values are always based on the following requirements: data should be available from a certain number of participants and two or more different analytical methods; there should be no significant differences between the groups of accepted results; outliers should not exceed 20%–30% of the submitted results. Depending on the extent to which the data satisfy such acceptance criteria, the consensus values are then assigned to one of the following conclusions: certified or recommended concentration, information value or not recommended. The Agency’s AQCS programme provides mainly four types of materials as follows: •
•
• •
Materials which can be used in analytical laboratories working in the fields of nuclear technology and isotope hydrology. These include uranium ore reference materials and other substances of interest for nuclear fuel technology as well as stable isotope reference materials for mass spectrometric determination of isotope ratios in natural waters. Materials with a known content of uranium, thorium and/or transuranium elements or fission products for the determination of environmental radioactivity or control of nuclear safety. Materials for use in the determination of stable trace elements in environment, biomedical and marine research. Materials which can be used in analytical laboratories working in the fields of monitoring organic microcontaminants in the marine environment.
Many countries practise national intercomparison programmes. For example, the Japanese nationwide intercomparison programme is based on the following: a. Comparison method Two methods of comparison, the “sample dividing method” and the “reference sample method,” were adopted for comparing the results of radionuclide analysis. b. Item for analysis and measurement method γ-Ray spectrometry is used. Participating laboratories are requested to determine artificial radionuclides as 40K, 54Mn, 59Fe, 60Co, 131I, 137Cs and 144 Ce, for the “reference sample method,” but as 40K and 137Cs for the “sample dividing method.” c. Samples and materials for intercomparison The environmental samples are soil, milk and crops. The reference samples are agar gel, alumina powder and liquid milk, which are all spiked with known radioisotopes.
4.5 QA/QC procedures
Table 4.19 IAEA reference materials for measurements of natural and fallout radioactivity in environmental and food samples Matrix
Analytes
Sediment, lake Sediment, stream
K-40, CS-137 Ra-226, Th, U
Soil Soil Bone, animal Clover Milk powder Milk powder Milk powder Whey powder Fish, flesh
Ra-226, Th, U Sr-90, Cs-137, Ra-226, Pu-239 Sr-90, Ra-226 K-40, Sr-90, CS-134, Cs-137 K-40, Sr-90, Cs-137 K-40, Sr-90, Cs-134, Cs.137 K-40, Sr-90, Cs-134, Cs-137 K-40, Sr-90, CS-134, Cs-137 K-40, Cs-137
Seaweeds, Mediterranean Sea-plant, posidonia oceanica Sediment, marine Sediment, Pacific Ocean Sediment, Pacific Ocean Tuna homogenate, Mediterranean Water, Pacific Ocean
K-40, Ru-106, Ag-110m, Cs-134, Cs-137, Pb210, Th-228, Pu-238, Pu-239 + 240, Am-241 K-40, Ru-106, Ag-110m, Cs-134, Cs-137, Ra226, Pu-238, Pu-230 + 240, Am-241
IAEA code SL-2 IAEA-313, IAEA, 314 IAEA-312 Soil 6 A-12 IAEA-156 A-14 IAEA-152 IAEA-154 IAEA-154 MA-B-3/ RM IAEA-308 IAEA-307
K-40, Cs-137, Th-232, Pu-239 + 240 Co-60, Sr-90, Cs-137, Pu-239 + 240
SD-N-2 IAEA-367
Co-60, Eu-155, Pb-210, Ra-226, Pu-238, U-238, Pu-239-240 K-40, Cs-137, Pb-210, Po-210
IAEA-368
Sr-90, Cs-137, Pb-210, Po-210
IAEA-352
IAEA-352
4.5.2 Reference materials All of the IAEA reference materials which are currently available have been certified on the basis of previously conducted intercomparison exercises. Natural matrix reference materials with certified values for the activities of various radionuclides are listed in Table 4.18. Some of the materials listed in Table 4.19 are the first “post-Chernobyl” natural matrix radionuclides reference materials that are internationally available. Including those reference materials available before Chernobyl, activities range for 137Cs from 0.8 (marine sediment, IAEA SD-N-2) to 3.7 kBq/kg (whey powder, IAEA-154).
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CHAPTER 4 Measurements of radioactivity
Table 4.20 Reference materials of a similar matrix with different levels of analytes Analytes
Activity (Bq/kg)
Reference date
Matrix
Code
Ra-226
342
30.01.88
IAEA-313
732
30.01.88
1.5 3.3 7.7 1.79 72.6 2159
31.08.87 01.01.90 31.08.87 31.08.87 01.01.90 31.08.87
Stream sediment Stream sediment Milk powder Milk powder Milk powder Milk powder Milk powder Milk powder
Sr-90
Cs-137
IAEA-314 A-14 IAEA-321 IAEA-152 A-14 IAEA-321 IAEA-152
Ideally, there is a need for several reference materials which have a similar matrix type to the samples being analysed and which contain a concentration of the analyte representative of the whole working range that is of interest. Table 4.20 lists stream sediments and milk powder reference materials which reflect the fact that such materials have a different level of activity with practically the same matrix type. Reference materials for radioactivity measurements can also be obtained from the following specialized international or national organizations. 1. Central Bureau for Nuclear Measurements, Commission of the European Communities, Joint Research Centre, Geel, Belgium. 2. Office des Rayonnements Ionisants Commissariat a` l’Energie Atomique BP 21, 91910, Gif-Sur-Yvette, France. 3. Commission d’Etablissement des Methodes d’Analyse Commissariat a` l’Energie Atomique BP 6, 92265, Fontenay aux Roses, France. 4. AEA Fuel Services, Chemistry Division, Harwell Laboratory, Oxfordshire OX11 0EA, UK. 5. New Brunswick Laboratory, US Department of Energy 9800 South Cass Avenue, Argonne, IL 60439-4899, USA. 6. All Union Foreign Economic Association “Techsnabexport,” Staromonetniy Per. 26, 109180, Moscow, Russia. The IAEA AQCS Programme provides the following three main types of material: •
Materials that can be used in analytical laboratories working in the fields of nuclear technology and isotope hydrology. These include uranium ore
4.5 QA/QC procedures
•
•
reference materials and other substances relevant to nuclear fuel technology as well as stable isotope reference materials for mass spectrometric determination of isotope ratios in natural waters. Materials with known contents of uranium, thorium and/or transuranic elements or fission products for the determination of environmental radioactivity or control of nuclear safety. Materials for use in the determination of stable trace elements in environmental or biomedical research. Radiochemical methods such as neutron activation or isotope dilution analysis are often used in the determination of such trace elements and constitute an important contribution of nuclear techniques to applied science (Strachnov et al., 1993).
Table 4.20 lists the radionuclides referenced by IAEA, their activity, matrix and sample code. Table 4.21 includes also materials of marine origin (Ballestra et al., 1992). The intercomparison samples cover a range of materials and contain radionuclides with very different levels. IAEA intercalibration exercises are conducted with the involvement of many laboratories. As an example, Fig. 4.53 shows the results of an intercomparison run for 137 Cs determination in milk powder. Some laboratories had difficulties in determining the activity level. This situation is rapidly improving with time. Table 4.22 shows the improvements in the quality of the work at the participating laboratories. This is also seen in Table 4.23 where the mean values and relative standard deviations of three intercomparison runs for the 90Sr determination in simulated air filters are presented. The intercomparison exercises show a need for greater standardization of the analytical techniques used for radionuclide determination. This is indicated in McGee et al. (1992), where the bias and measurement errors in radioactivity data from four European radiation research laboratories were reported. Within the framework of the International Chernobyl Project, the IAEA’s Seibersdorf Laboratories organized an intercalibration exercise (Cooper et al., 1992) among some of the laboratories which were involved in assessing the environmental contamination in the former USSR caused by the accident. The objective was to assess the reliability of the radioanalytical data for food and environmental samples, which were used to assess the doses. The initial study reference materials from the stocks of the IAEA’s Analytical Quality Control Service (AQCS) were relabelled and submitted to 71 laboratories as blind samples in June and July of 1990. These natural matrix materials included samples of milk (containing two different levels of radioactivity), soil, air filters and clover. The concentrations of radionuclides (137Cs, 134Cs, 40K, 90Sr, 239Pu, 226Ra, 60Co, 133Ba and 210Pb) in these samples were known from previous intercalibration exercises. The overall range in performance was broad, which is as observed in previous international intercomparisons. This is illustrated in Table 4.24 where the results
253
254
Table 4.21 Radionuclides referenced by IAEA (Bq/kg)a
3
H C
14
60 40
Co K
Activity or conc.
Confidence (Bq/ kg)
37.2 15.03 pMC 0.12 pMC 2.50 391
0.5
0.17 pMC 0.21 pMC 2.40–2.60 379–405
527 150 220 240 272 539 552 657 1381 1575 103 11.4 607 6.8 1188 1080 72.1 660 373 481 234
510–543 141–161 189–226 211–269 252–299 510–574 563–569 637–676 1320–1456 1511–1644
0.4 604–612 6.5–7.1
30
56
3.2 626–671 360–380 470–486
12
Matrix
Reference date
Sample code
Irish Seawater Oxalic acid Wood Fangataufa Sediment Tuna homogenate, Mediterranean Milk powder Sea-plant, Posidonia oceanica Sediment, marine Sediment, lake Fish flesh Milk powder Milk powder Clover Seaweeds, Mediterranean Hay powder Irish Seawater Irish Sea Sediment Fangataufa Sediment Spinach Grass Brown rice Baltic Sea Seaweed Mussel from Mediterranean Sea Fish Soil from oil field
1 1 1 1 1
IAEA-443 IAEA-C8 IAEA-C9 IAEA-384 IAEA-352
Jan 2007 Jan 1998 Jan 2004 Aug 1996 Jan 1989
31 Aug 1987 1 Jan 1988 1 Jan 1985 31 Jan 1986 1 Jan 1986 31 Aug 1987 1 Jan 1990 1 Aug 1986 1 Jan 1988 31 Aug 1987 1 Jan 2007 1 Jan 1996 1 Aug 1996 15 Oct 2007 1 Jun 2006 1 Jan 2015 1 Aug 2008 1 Nov 2003 1 Jan 1997 1 Jan 2009
A-14 IAEA-307 SD-N-2 SL2 MA-B-3/RN IAEA-152 IAEA-321 IAEA-156 IAEA-308 IAEA-154 IAEA-443 IAEA-385 IAEA-384 IAEA-330 IAEA-372 IAEA-464 IAEA-446 IAEA-437 IAEA-414 IAEA-448
CHAPTER 4 Measurements of radioactivity
Ref. analyte
89 90
Sr Sr
106
Ru
110m
134
Ag
77 1.33–1.57 3.16–3.44 6.0–8.0 7.0–8.3 13.4–16.3 24.2–31.67 46.3–59.2
5
2.1
0.005 22–25 30.0–36.5 1–2.27 4.8–5.5 1.5–1.8 1.5–1.9 14.8–16.2 126–138 722–802 1295–14 17
0.4
7
Milk powder Milk powder Milk powder Hay powder Milk powder Clover Soil Bone, animal Milk powder Spinach Irish Seawater Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Milk powder Clover Milk powder Whey powder Brown Rice Milk powder
1 Nov 2014 31 Aug 1987 1 Jan 1990 31 Aug 1987 31 Aug 1987 1 Aug 1986 30 Jan 1983 15 Dec 1981 1 Nov 2014 15 Oct 2007 1 Jan 2007 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1990 1 Aug 1986 31 Aug 1987 31 Aug 1987 1 Jan 2015 1 Nov 2014
IAEA-473 A-14 IAEA-321 IAEA-154 IAEA-152 IAEA-156 SOIL-6 A-12 IAEA-473 IAEA-330 IAEA-443 IAEA-308 IAEA-307 IAEA-308 IAEA-307 IAEA-308 IAEA-307 IAEA-321 IAEA-156 IAEA-152 IAEA-154 IAEA-464 IAEA-473 Continued
4.5 QA/QC procedures
Cs
2405 1.5 3.3 6.9 7.7 14.8 30.34 54.8 209 20.1 0.110 23 33.5 20 5.1 1.6 1.6 15.5 132 764 1355 12.0 357
255
256
Ref. analyte 137
155
Activity or conc.
Confidence (Bq/ kg)
Cs
2.7
2.5–2.8
Eu
0.8 1.79 2.4 4.9 5.6 14.2 53.65 72.6 264 2159 3749 11,320 38.6 0.36 33.0 1235 18.8 5.18 425 224 7.0
0.5–1.0 1.62–1.97 22–2.6 4.5–5.2 5.3–6.0 13.7–15.3 51.43–57.91 71.1–74.2 254–274 2503–22 09 3613–38 87
360
0.8
0.01 32.7–33.6
35 18.2–19.2 5.12–5.22
10
5 6.7–7.3
Matrix
Reference date
Sample code
Tuna homogenate, Mediterranean Sediment, marine Milk powder Sediment lake Sea-plant, Posidonia oceanica Seaweeds, Mediterranean Fish flesh Soil Milk powder Clover Milk powder Hay powder Grass Brown Rice Irish Seawater Irish Sea Sediment Spinach Baltic Sea Seaweed Fish Moss-soil Milk powder Fangataufa Sediment
1 Jan 1989
IAEA-352
1 Jan 1985 31 Aug 1987 31 Aug 1986 1 Jan 1988 1 Jan 1988 1 Jan 1986 30 Jan 1983 1 Jan 1990 1 Aug 1986 31 Aug 1987 31 Aug 1987 1 Jun 2006 1 Jan 2015 1 Jan 2007 1 Jan 1996 15 Oct 2007 1 Aug 2006 1 Jan 1997 15 Nov 2009 1 Nov 2014 1 Aug 1996
SD-N-2 A-14 SL-2 IAEA-307 IAEA-308 MA-B-3/RN SOIL-6 IAEA-321 IAEA-156 IAEA-152 IAEA-154 IAEA-372 IAEA-464 IAEA-443 IAEA-385 IAEA-330 IAEA-446 IAEA-414 IAEA-447 IAEA-473 IAEA-384
CHAPTER 4 Measurements of radioactivity
Table 4.21 Radionuclides referenced by IAEA (Bq/kg)a Continued
210
0.6
0.36–1.0
210
Po
73 680 0.42 103 2.2
66–75
58
0.02 103 1.7–27
226
Ra
228
Ra
228
Th Th
423 3.1 5.2 79.92 269 342 732 21.9 19.05 103 25.1 780 32.0 25 2.50 211 30.6
10 21–4.4 4.4–6.7 69.56–93-43 250–287 307–379 678–787 21.6–22.4
0.26 103
2.0
62 31.3–32.5 2.2–3.6 2.38–2.61
9 30.0–33.6
230
Tuna homogenate, Mediterranean Seaweeds, Mediterranean Phosphogypsum Moss-soil Tuna homogenate, Mediterranean Moss-soil Sea-plant, Posidonia oceanica Bone, animal Soil Soil Sediment, stream Sediment, stream Irish Sea Sediment Soil from oil field Moss-soil Phosphogypsum Irish Sea Sediment Seaweeds, Mediterranean Fangataufa Sediment Phosphogypsum Irish Sea Sediment
1 Jan 1989
IAEA-352
1 Jan 1988 1 Jan 2008 15 Nov 2009 1 Jan 1989
IAEA-308 IAEA-434 IAEA-447 IAEA-352
15 Nov 2009 1 Jan 1988 15 Dec 1981 30 Jan 1983 30 Jan 1988 30 Jan 1988 30 Jan 1988 1 Jan 1996 1 Jan 2009 15 Nov 2009 1 Jan 2008 1 Jan 1996 1 Jan 1988 1 Aug 1996 1 Jan 2008 1 Jan 1996
IAEA-447 IAEA-307 A-12 SOIL-6 IAEA-312 IAEA-313 IAEA-314 IAEA-385 IAEA-448 IAEA-447 IAEA-434 IAEA-385 IAEA-308 IAEA-384 IAEA-434 IAEA-385 Continued
4.5 QA/QC procedures
Pb
257
258
Ref. analyte 232
Th
234
U
235
U
238
U
Activity or conc.
Confidence (Bq/ kg)
Matrix
Reference date
Sample code
4.9 0.028 33.7 1.22 2.3 1.02 10.5 120 27 21.8 0.044 0.050 0.00185 1.11 35.5 1.87 0.95 120 29 22.2 0.039
4.5–5.4 0.025–0.031 32.8–33.9 1.15–1.26 2.2–2.4
0.07 10.0–11.0
9 26–28
0.8
0.002 0.045–0.055
0.00010 1.07–1.15 33.4–36.8 1.80–1.92
0.05
11 28–30
0.8
0.002
Sediment, marine Fish Irish Sea Sediment Fish Mussel from Mediterranean Sea Spinach Baltic Sea Seaweed Phosphogypsum Irish Sea Sediment Moss-soil Irish Seawater Fish Irish Seawater Fish Fangataufa Sediment Mussel from Mediterranean Sea Spinach Phosphogypsum Irish Sea Sediment Moss-soil Irish Seawater
1 Jan1985 1 Jan 1997 1 Jan 1996 1 Jan 1997 1 Nov 2003 15 Oct 2007 1 Aug 2006 1 Jan 2008 1 Jan 1996 15 Nov 2009 1 Jan 2007 1 Jan 1997 1 Jan 2007 1 Jan 1997 1 Aug 1996 1 Nov 2003 15 Oct 2007 1 Jan 2008 1 Jan 1996 15 Nov 2009 1 Jan 2007
SD-N-2 IAEA-414 IAEA-385 IAEA-414 IAEA-437 IAEA-330 IAEA-446 IAEA-434 IAEA-385 IAEA-447 IAEA-443 IAEA-414 IAEA-443 IAEA-414 IAEA-384 IAEA-437 IAEA-330 IAEA-434 IAEA-385 IAEA-447 IAEA-443
CHAPTER 4 Measurements of radioactivity
Table 4.21 Radionuclides referenced by IAEA (Bq/kg)a Continued
238
Pu
239
Pu Pu
239
241
a
0.017 0.025 0.0230 39.0 0.44 1.04 8.8 0.50 0.72 0.036 0.17 0.0197 3.84 7.1 0.197
Am
The
232
Th is in equilibrium with 228Ra and
0.016–0.023 0.022–0.028 0.0221–0.0250 38.6–39.6 0.42–0.48 0.962–1.11 6.51–4.0 0.46–0.52 0.66–0.79 0.030–0.050 0.16–0.25
0.0002 3.78–4.01 6.7–7.4 0.193–0.204
Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Fish Fangataufa Sediment Irish Sea Sediment Soil Sediment, marine Seaweeds, Mediterranean Sea-plant, Posidonia oceanica Sea-plant, Posidonia oceanica Seaweeds, Mediterranean Irish Seawater Irish Sea Sediment Fangataufa Sediment Fish
1 Jan 1988 1 Jan 1988 1 Jan 1997 1 Aug 1996 1 Jan 1996 30 Jan 1983 1 Jan 1985 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 1988 1 Jan 2007 1 Jan 1996 1 Aug 1996 1 Jan 1997
IAEA-308 IAEA-307 IAEA-414 IAEA-384 IAEA-385 SOIL-6 SD-N-2 IAEA-307 IAEA-307 IAEA-307 IAEA-308 IAEA-443 IAEA-385 IAEA-384 IAEA-414
228
Th.
4.5 QA/QC procedures 259
CHAPTER 4 Measurements of radioactivity
2800
2400 2000 Bq (kg)
260
1600
1200
800
400
36
29
5
37
16
4
13
34 398
6
10
28
12
35
14
6
30
33
27
32
26
20
17
21
31
2 25
9 18A 22 11 1 39A 23 15
19 3
Lab. Code No.
FIG. 4.53 Results of IAEA-152 intercomparison 137Cs determinations. Recommended value—2065 Bq/kg; confidence interval—1991–2143 Bq/kg.
of the original IAEA intercomparison run (worldwide) and former Soviet Union laboratories, for high-level (H) milk are presented. The Central Service for Protection against Ionizing Radiation (SCPRI), a service of the French Ministry of Public Health, National Institute of Health and Medical Research, was nominated at the end of 1969 as the International Reference Centre (IRC) of the World Health Organisation for Radioactivity measurements. Four laboratories in the world have been officially designated as WHO collaborating laboratories. These laboratories are: • • • •
Radiation Protection Bureau in Ottawa, Canada. National Institute of Radiation Protection in Stockholm, Sweden, Environmental Monitoring and Support Laboratory (EPA) in Las Vegas, USA and National Radiation Laboratory in Christchurch, New Zealand.
At the present time, 28 laboratories from 17 countries are interested in the WHO-IRC Intercomparisons. Its programme of intercomparison shows the following characteristics: 1. the radioactivity of the samples is the present environmental monitoring level; 2. generally, the samples present real radioactivity due to the fallout or releases of nuclear facilities; 3. the amount of the product provided allows several tests to be carried out; 4. standard materials can be provided and 5. a preliminary study of the results of each intercomparison is given to the participants as soon as possible.
4.5 QA/QC procedures
Table 4.22 Determination of 137Cs in the same milk powder during intercomparisons in 1983 and 1989
1983 1989
Mean value (Bq/kg)
Rel. SD (%)
% outliers/lab.
2.08 1.70
45 19
10 0
Table 4.23 Determination of Sr-90 in simulated air filters during intercomparison runs in 1973, 1976, 1988
1973 1976 1988
Mean value
Rel. SD (%)
178 Bq/filter 179.5 Bq/filter 231 Bq/filter
27 17 3
Table 4.24 Comparison of performance of the two groups of laboratories: worldwide vs Soviet Union Range of reported values for milk(H) (Bq/kg) Radionuclide
Worldwide
USSR
137
469.3–2491.3 58.0–652.5 103.6–3650.0 5.53–8.54
175–3070 184.7–542.5 429–4959 1.43–68.8
Cs Cs 40 K 90 Sr 134
Table 4.24 shows concrete contents of intercomparisons which WHO has carried out in the period 1970–91. The first column of Table 4.25 indicates periods when samples were sent to participating laboratories, the second their nature, the third their numbers, the fourth nuclides and stable elements to be measured and determined for the intercomparison purposes, and the fifth and last column, the scope of radioactivity levels in the samples. As demonstrated by this table, a wide variety of samples has been adopted since 1970, among which are liquid milk, animal bones, human bones, foods, low-level radioactive liquid waste, ground water, mineral water, river sediment, seaweed, pond water, fresh water fish, cereals, seawater, rain water, drinking water, soil and vegetation. Concerning the general conditions of the intercomparison programmes in progress and the results obtained, it can be noted that: •
The IRC has diversified its programme by introducing new categories of samples (waters from various origins, sediments, fish, seaweed, liquid waste, cereals, soil, etc.) in which laboratories involved in environmental monitoring of nuclear power plants are interested.
261
262
Period of dispatch June 1970 Feb 1971 June 1971 Feb 1972 Nov 1972
Nature of the sample
No. sample
Proposed determinations
Radioactivity level
Sr,
Cs, Ca, K
90
Sr, 137Cs, Ca, K Sr, Sr, Ca
90
Liquid milk
A010
90
Liquid milk Animal bones
A338 A504 A505 A806 B078
90
Human bones Dried total diet
90
90 90
137
Sr, Sr, Ca Sr, 137Cs, Sr, Ca, K
B845 C140 C141
90
C617 C618 D256
3
River water
D402
3
Sept 1976 Dec 1976
Liquid milk
D499
90
River sediment
D601
March 1977 July 1977
Animal bones
D736 D737 D925
Activation and fission products (54Mn, 58Co, 60 Co, 90Sr, 134Cs, 137Cs, ...) 90Sr, Sr, Ca
Feb1975 March 1976 June 1976
Sea fish
3
Sr, H
Sr 20 pCi/L, 137Cs 20 pCi/L Sr 3 pCi/g-ash 90 Sr 3 pCi/g-ash 90 Sr < 0.5 pCi/g-ash 90 Sr 40 pCi/kg-dry 137 Cs 90 pCi/kg-dry 90 Sr 20 pCi/L, 137Cs 30 pCi/L 3 H 10,000 pCi/L 3 H 150,000 pCi/L 90
Liquid milk Ground water low-level liquid waste Rain water spiked ground water Mineral water
July 1973 April 1974
Sr 35 pCi/L, 137Cs 120 pCi/L
137
Cs, Ca, K
H 1000 pCi/L H 5000 pCi/L Nat. U 20 μg/L,
3
H
3
Gross β, K nat. U, H,
90
90
Sr,
Sr,
Sr,
106
226
Ra
Ru + 106Rh, 125Sb, 137Cs...
137
Cs, Ca, K
134
Cs, 137Cs, Sr, Ca, K
226
Ra 15 pCi/L
H < 5000 pCi/L 106Ru + 106Rh y 1000 pCi/L 90Sr, 125Sb, 137Cs < 50 pCi/ L 90 Sr 10 pCi/L, 137Cs 15 pCi/L 3
5000–20,000 pCi per kg-dry for each radionuclide 90 Sr 7 pCi/g-ash 90 Sr 3 pCi/g-ash 90 Sr 150 pCi/kg-dry 134 Cs 400 pCi/kg-dry 137 Cs 4000 pCi/kg-dry
CHAPTER 4 Measurements of radioactivity
Table 4.25 WHO-IRC intercomparisons
Dec 1977
Marine sediment
E114
90
Sr, 95Zr + 95Nb, 103Ru, 106Ru + 106Ru, Ag, 125Sb, 137Cs, ...
110m
March 1978 June 1978 Oct 1978
Liquid milk
E414
90
Seaweed
E468
Fission products
Pond waters
3
H
March 1979
Low-level liquid waste
E888 E889 F100
3
H,
June 1979 Nov 1979
Liquid milk
F140
90
Fresh water fish
F290
U,
Sr,
37
Cs, Ca, K
54
Sr,
Mn, 58Co,
60
Co
137
Cs
226
Ra,
90
Sr,
137
Cs
106 Ru + 106Ru, 144Ce + 144Pr < 10,000 pCi/kg-dry other nuclides <500 pCi/kg-dry 90 Sr 15 pCi/L, 137Cs 30 pCi/L
20–10,000 pCi/kg-dry according to the radionuclides 3 H 5000 pCi/L 3 H 20,000 pCi/L 3 H 2.5 μCi/L, 54Mn 800 pCi/L 58 Co 50,000 pCi/L 58 Co 2500 pCi/L 90 Sr 40 pCi/L, 137Cs 250 pCi/L U 1500 μg/kg-dry Ra 3500 pCi/kg-dry 90 Sr 6000 pCi/kg-dry 137 Cs 1500 pCi/kg-dry 90 Sr, 137Cs 20 pCi/kg-dry 226
F553
90
Seawater
F712
Fission products
Total activity 5000 pCi/L
Rainwater Liquid milk Drinking water
F856 G041 G336
3
3
90
90
Oct 1981
Soil
G477
90
Sr,
137
Feb1982
Total diet
G660
90
Sr,
137
Sr,
137
Cs
H Sr, 137Cs, Ca, K 3 H, 90Sr, 106Ru
Cs natural radionuclides Cs, Ca, K, nat. U
H 2500 pCi/L Sr 10 pCi/L, 137Cs 300 pCi/L 3 H < 10,000 pCi/L 106 Ru 75 pCi/L 90 Sr 3 pCi/L 90 Sr 200 pCi/kg-dry 137 Cs 500 pCi/kg-dry 90 Sr 0.4 Bq/kg-dry (10 pCi/kg-dry) 137 Cs 0.4 Bq/kg-dry (10 pCi/kg-dry)
263
Continued
4.5 QA/QC procedures
Cereals (wheat)
March 1980 June 1980 Oct 1980 Feb 1981 June 1981
264
Period of dispatch
Nature of the sample
No. sample
Proposed determinations
Radioactivity level Be 200 Bq/kg-dry (5000 pCi/kgdry) 90 Sr 20 Bq/kg-dry (500 pCi/kg-dry) 137 Cs 5 Bq/kg-dry (100 pCi/kg-dry) U 50–100 μg/L 226 Ra 0.4 Bq L1 (10 pCi/L) 90 Sr 0.3 Bq L1 (pCi/L) 137 Sr 0.3 Bq L1 (pCi/L) 90 Sr 2 Bq/kg-dry (50 pCi/kg-dry) 137 Cs 20 Bq/kg-dry (500 pCi/kg-dry) 5–500 Bq/kg dry, according to the radionuclides 200 Bq L1 10,000 Bg L1
June 1982
Vegetation
G730
7
Dec 1982
Mineral water
nat U,
Feb 1983
Liquid milk
G972 G973 H071
90
Sr,
137
June 1983 Feb 1984
Sea fish
H264
90
Sr,
137
River sediment
H519
June 1984
Rain water Ground water
Feb 1985
Liquid milk
40 PM 300 41 P 300 42 L 300
Activation and fission products (54Mn, 58Co, 60Co, 90Sr, 134Cs, 137Cs, 106Ru) 3 H 3 H
3
Be,
90
Sr,
137
9
Cs, Ca, K
226
Ra
Cs, Ca, K Cs, Ca, K, Sr
100 Bq L1
H
90
Sr Cs 3 H γ-emitters (54Mn, 58Co, 137
May 1985
Low-level radioactive liquid effluent
43 E 300
Feb 1986
Urine
44 UR 300
3
H
60
Co,
134
Cs,
137
Cs)
0.3 Bg L1 0.6 Bq l1 3 H 5 105 Bq L1 54 Mn 1 102 Bq L1 58 Co 2 103 Bq L1 60 Co 2 103 Bq L1 134 Cs 1 102 Bq L1 137 Cs 1 102 Bq L1 500,000 Bq L1
CHAPTER 4 Measurements of radioactivity
Table 4.25 WHO-IRC intercomparisons Continued
May 1986
Aquatic plants
45 V 300
April 1987
Liquid milk
46 L 3000
Activation and fission products (54Mn, 58Co, 60 Co, 90Sr, 137Cs, 106Ru) natural radionuclides 90 Sr Cs Ca 134 Cs 137 Cs 134 Cs 137 Cs 226 Ra
April 1988 Oct 1988
Dry may blossom
April 1989 Nov 1989
Radioactive effluent Herba Drosera longifolia
Aug 1990
River sediments
Dec 1990
Seaweeds
May 1991
River sediments
Mineral water
47 V 300 48 V 300 49 SH 300 50 E 300 51 V 300 52 V 300 53 SD 300 54 AL 300 55 SR 300
3
H Mn, 58Co, 60Co, 124Sb, 228 Th, 228Ra, 232Th 54
134
Cs, 137Cs
<2 Bq L1 5–100 μg L1 103–106 Bq L1 0–104 Bq L1 for each nuclide 5 102 Bq/kg dry for each nuclide
Activation and fission products
300 Bq/kg dry
Transuranium elements
5 Bq L1
Natural radionuclides
4.5 QA/QC procedures
Shelled hazelnuts
0–1 Bq L1 0–50 Bq L1 0–100 Bq L1 0–200 Bq/kg, according to the radionuclides 0–5000 Bq L1
134
137
Oct 1987
10–2000 Bq/kg dry, according to the radionuclides
265
266
CHAPTER 4 Measurements of radioactivity
• •
With regard to the quality of the analyses, the situation presented in the preceding annual report has not evolved much. The regular participation of different laboratories in the intercomparisons provides a comprehensive view of their technical capabilities and of the quality of their analytical work.
According to the Articles 35/36 of the Euroatom Treaty (2012) and the Commission Recommendation (EC, 2000) derived from the Treaty, the Member states of European Union have the legal obligation to inform the European Commission on a regular basis on the radioactivity levels in their environment (drinking water, soil air and food staff). In order to obtain information on the quality of reported values an International Comparison Scheme for Radioactivity Environmental Monitoring (ICS-REM) has been established by European Commission, EC (W€atjen, 2008). Several EC Interlaboratory Comparisons have been organized the Institute for Reference Materials and Measurements in Geel, Belgium, most notably for the determination of 137Cs in air filters (W€atjen et al., 2007; Ma´te et al., 2016). In the last case spiked air filters were prepared for each 76 participating laboratory from Europe and 93.4% of the Laboratories Reported within 33% Interval of the Reference Value. The ALMERA network (Analytical Laboratories for the Measurement of Environmental Radioactivity) was established by the IAEA in 1995, and is a cooperative effort of analytical laboratories worldwide. Members of the network are nominated by their respective IAEA. Member States as those laboratories which would be expected to provide reliable and timely analysis of environmental samples in the event of an accidental or intentional release of radioactivity. ALMERA currently (December 2017) consists of 166 laboratories representing 87 countries. The Agency’s Environment Laboratories in Seibersdorf and Monaco are additional members of the network. The International Atomic Energy Agency’s Environment Laboratories are the central coordinator of the ALMERA network’s activities. IAEA proficiency tests and interlaboratory comparison exercises are organized on a regular basis, specifically for the members of the ALMERA network. These exercises are designed to monitor and demonstrate the performance and analytical capabilities of the network members, and to identify gaps and problem areas where further development is needed. At least one exercise is organized per year by the IAEA for the ALMERA network (see IAEA, 2017; Osvath et al., 2016).
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