Maximum permissible fluxes of intermediate energy neutrons and their measurements

MAXIMUM

PERMISSTBLE FLUXES OF INTERMEDIATE NEUTRONS AND THEIR MEASUREMENTS* A. G. ISTOMINA and

ENERGY

I. B. KEIRIM-MARKUS

(Received 31 March 1959)

Abstract-The maximum and mean doses of neutrons absorbed by human tissue are calculated from data given in the literature over the range from thermal energies to 1 MeV. The results are averaged for the typical I/,!? spectrum and for various conditions of irradiation. The maximum permissible dose of intermediate neutrons is 680 n:cm2/sec. The application of known methods of neutron detection to the dosimetry of intermediate neutrons is considered and it is shown that, with certain reservations, shielded thermal neutron detectors are suitable for this purpose.

T13r: term ‘intermediate energy neutrons’ is applied to neutrons of energies ranging from 0.2-I eV to 0.51 MeV. From the point of view of dosimetry, intermediate neutrons have the following characteristics. 1. They make up a substantial proportion of the absorbed d.ose of neutrons slowed down in the human body. Thus, at a neutron energy of 0.5 MeV, more than 10 per cent of the mean tissue dose (rem) is accounted for by neutrons slowed down to thermal Expressing the same dose (rad), more energies”‘. than 50 per cent of the absorbed dose is attributable to capture of the y-radiation emitted by the sloweddown neutrons’“). Slowed-down neutrons with an energy of less than 0.5 MeV represent an even higher proportion of the absorbed dose. 2. Because of the appreciable contribution from the y-component the relative biological effectiveness (RBE) of intermediate neutrons in contrast to the RBE of faster neutrons depends strongly on the volume of the body, decreasing with depthc3). 3. In interaction with the tissue the ionization produced by their recoil nuclei is less significant than that produced by the recoils froIn fast neutrons. Thus for neutrons with energies less than 1 MeV recoil nuclei other than protons gradually cease to ionize the tissue’“). Hence one energy transfer process in the tissue, biologically the most effective, is cut out. Below 20 kcV the recoil protons also gradually cease to ionize the medium as a result of electron capture.(4,5) The energy of such protons is partly expended in rearrangement collisions between the molecules of the medium, n process of which the mechanism, relative importance and RBE are as yet unknown. Neverthelesss we may assume that the contribution of this mechanism to the total effect of neutrons on the

organism is not large because a neutron with an energy of less than 20 keV spends almost the whole of its life within the organism as a thermal neutron and the energy released on capture is many times greater than the kinetic energy of an intermediate neutron. 4. Intermediate neutrons are mostly obtained by slowing down fast neutrons. Thus in weakly absorbing Inedia interlnediate electrons have the characteristic spectrum p(E) dE - dE/E, where v(E) : flux of neutrons with energy E. 5. Lastly a notable, though not predominant, characteristic of intermediate neutrons is that they are hard to record. This is one of the reasons why intermediate neutrons have not so far been considered in dosimetry although they often form a substantial proportion of the total neutron flux. Thus in the beaIns of radiation emitted from the active zone of a thermal reactor the flux of intermediate neutrons is of the same order as the flux of thermal neutrons’“). ln neutron sources intermediate neutrons make up as much as 40 per cent of the total fluxf7). The flux reaching the earth from an aerial nuclear explosion contains Inore intermediate than fast neutrons because the fission neutrons are slowed down in the casing of the device and in the air whilst the thermal neutrons are absorbed by the nitrogen in the atmosphere@). Accordingly the contribution of intermediate neutrons to the absorbed dose is normally considerable, the more so since the influence on the organism of a flux of intermediate neutrons is stronger than that of an equivalent current of thermal neutrons. The distribution of absorbed doses of secondary radiation in a tissue-equivalent layer of 30 cm thickness has been worked out.(lU3) These results are plotted in Fig. 1 after conversion. 1n calculating the absorbed doses in rem the authors of papers(l-“) used extremely 151

152

A. G. ISTOMINA and I. B. KEIRIM-MARKUS

500 20 I00 % O.i-

0.1

g 30

1

Dtstance

from

Surface

of layer.

Cm

FIG. I.-Absorbed dose from unit neutron flux incident normally on a layer of tissue 30 cm thick12). Numbers adjoining curves = neutron energies in keV; recoil protons and protons produced by the reaction N (n, p); y-radiation from the reaction H(n, y)D; -----heavyrecoilnuclei.

10 for protons and 20 for heavy large RBE values: recoil nuclei. This latter figure is too high for the range of neutron energies studied although it does not greatly affect the final values. At the Institute of Chemical Physics of the Academy of Sciences of the USSR in 1956, P. A. Iampol’skii, L. A. Chudov, G. G. Petrov and A. M. Kogan performed similar calculations for a neutron flux incident on a semi-infinite block of paraffin. The contribution of heavy recoil nuclei to the absorbed dose was not taken into account. The maximum absorbed dose was computed from the RBE values of the protons, i.e. 2, 4.5 and 10. The converted results are plotted in Fig. 2. However the maximum absorbed dose does not

always determine the biological effect of exposure to radiation. The RBE of radiation also varies under different conditions of exposure and is governed by the reaction of the organism. A number of examples is cited below. 1. Chronic whole-body exposure to small doses and remote after-effects of exposure are normally characteristic of a working environment where radiation constitutes an occupational hazard. Here the RBE of protons with a mean energy close to 10 and the biological effect of the radiation are determined by the absorbed dose in the critical organ or, in a first approximation, by the maximum absorbed dose. This is represented by curves 10 of Fig. 2, which should clearly be used for calculating the maximum permissible levels of exposure and for the dosimetric monitoring of working environments. 2. In studying the distribution in depth of the absorbed dose it is important to know the absorbed dose from the first impact for the real neutron spectrum and from the secondary y-rays traversing a given element of volume. 3. In experiments designed to ascertain the RBE of radiation, the mean (or maximum) tissue dose (rad) (curve 1, Fig. 2) must be known. 4. In therapeutic applications of neutrons and also in accid.ents and illness resulting from nuclear weapons large doses may produce a short-term effect. Here the mean tissue dose is substantial and the RBE of the protons for acute reactions of the organism appears to lie between 2 and 4.5. We calculated the mean absorbed dose for a layer of tissue 30 cm thick using different RBE values for protons (the results used are set out in Fig. 1c2)). The results are listed in Table 1 and plotted in Fig. 3. The plot of the absorbed dose against the neutron energy differs from all the plots in Figs. 2 and 3. From this, incidentally, it follows that the efforts of of authors(g,lO) to achieve the best tissuea number equivalent pick-up for dosimeters are not always successful. From the data in Figs. 2 and 3 the mean absorbed dose may be calculated for a unit flux of intermediate neutrons with an arbitrary energy spectrum. It is equal to

s s E2

D

W)p(E) dE

= E1

E,

(1)

P(E)dE

n1

where D(E) is the absorbed dose for a unit neutron flux with energy E; p(E) dE is a neutron flux in the

Maximum

permissible

energy

153

neutrons and their measurements

keV

1,

O-I

fluxes of intermediate

IO

I.0 NeUtrOn

energy,

100

1000

keV

FIG. 2.-Maximum absorbed dose for a layer of tissue irradiated with unit neutron flux. Numbers adjoining curves ~ assumed RBE of protons: ~ results reported in”): - - - results obtained at the Institute of Chemical Physics of the Academy of Sciences of the U.S.S.R. The top curves are normalized to their value at 1 keV.

interval E to E : LEE; E1 and E2 are the limits of the intermediate neutron spectrum. From this formula, inserting E1 := 0.4 eV, E2 = 0.5 MeV, we worked out the mean absorbed dose for a unit flux of intermediate neutrons with a spectrum -l/E (Table 2 and Fig. 4). The mean absorbed dose remains almost independent of the lower limit of integration when this lies between 0.2 and 1 eV, and it is not affected to any large extent by divergences from the l/E law that occur in the intermediate neutron spectrum at low neutron energies owing to absorption by the nitrogen or hydrogen atoms in normal moderating media. The results obtained, therefore, may be of quite wide validity. Two considerations govern the choice of the upper limit of integration: first, in the range of energies beyond 0.5 MeV considerable overlapping between the intermediate neutron spectrum by the primary fission neutron spectrum occurs and the I/E law does not hold good : secondly, neutrons with energies higher than 0.5 MeV may be recorded by fast neutron dosimeters’“). With an upper energy limit of 1 MeV the error in the mean absorbed dose from intermediate neutrons is still sufficiently small for many practical purposes.

Table 1 also gives the absorbed dose and RBE for thermal neutrons and neutrons with an energy of I MeV. It will be seen that the intermediate neutrons actually have intermediate characteristics. Intermediate neutrons contribute two to four times more than the mean absorbed dose. The maximum permissible flux of intermediate neutrons for a six-hour working day is 680 n/cm”/sec,” whereas for thermal neutrons the flux is 2600 n/cm?/sec (on the basis of other dataor) 1650 n/cmZ/sec). Hence the importance of monitoring intermediate neutron fluxes becomes clear.

The efhciency of a dosimeter in the intermediate neutron region should correspond to the curves in Fig. 2 or 3 which characterize a given dosimetei (medical, military. for the monitoring of protective arrangements, etc.) and in the fast neutron range it should correspond to the continuation of the dose curve or decrease sharply. Such dosimeters do not exist at present. Thus it is sometimes convenient to divide the complete neutron spectrum into three

* The authors’“” data on the maximum permissible flux of intermediate energy neutrons are not officially confirmed. standards.

151

A. G. ISTOMINA and I. B. KEIRIM-MARKUS keV

6

0

IO

I.0

0.1

Neutron

energy.

100

IO00

keV

FIG. 3.-Mean dose absorbed by a layer of tissue irradiated by unit neutron flux. Numbers adjoining curves assumed RBE of protons. The absorbed dose of recoil components from secondary radiation is expressed rads; y = y-radiation; p = protons; c = heavy recoil nuclei. The top curves are normalized to their value at 1 keV.

regions with a known energy distribution: thermalMaxwellian distribution; intermediate-l/E spectrum; fast-fission neutron spectrum. Once the absorbed doses from neutron fluxes confined to parts of the spectrum where these partial spectra are known not to overlap have been measured, the corresponding doses for the integrated fluxes between the conventional boundaries of each partial spectrum can be determined, and from these can be obtained a reasonTABLE l.-MEAN

RBE of protons 7 y-Radiation

capture

-

Recoil protons

1

Heavy recoil nuclei

-

Total dose

1

DOSE

ably accurate idea of the dose absorbed from the neutron spectrum as a whole. We shall now consider the suitability of existing methods of recording thermal and fast neutrons for use in the intermediate energy region. The activation method can be used to record both slow and fast neutrons. In measuring thermal neutrons intermediate neutrons are commonly separated out by the cadmium ratio method. A cadmium

ABSORBEDBY TISSUE (M~REM/N/CM~)

Neutron

I

-iTiIzLJ-heavy

-1nuclei ;

5 keV

0,091

i 0’“‘”

0.20 0””

30

CM

energy

20 keV

‘__ 0.24 0:s

THICK(~)

0.25 OF4

I

( 100keV

0.28 025

_____ Total dose

2

Total dose

4.5 10

Total

dose

: 10 20

I ;:;;f I 0.114 i 0.143

0.215 0.23 0.28 0.35

0.25 0.27 0.32 0.42

500 keV

1 MeV

0.30 w!;9

0.28 @I3

0.90 0.57

0.78 1.48 2.90 5.95

~_

! ~

0.27 0.30 0.36 0.49

0.41 0.345

Maximum permissible fluxes of intermediate

the ratio of the intermediate neutron flux may be derived:

Ft

\

D,O.

SC

(R,,

-

’

1) l;Jx lo5 $

where o’t = cross section for thermal neutron activation; j = activation resonance integral. This we can illustrate by considering the chronic Here we must effect of small doses of neutrons. assume that the maximum absorbed dose, for which the RBE of the protons is 10, characterizes the effect. The ratio of the absorbed dose of intermediate (DJ and thermal (D,) neutrons in this case will be as follows (cf. Table 2)

380

36C

(F,)

(2)

IO

=

375

0 ; 0 -

(FJ to the thermai

13.90,

_Fi_ 7

155

energy neutrons and their measurements

7’1

D,.X 36

2:

2,;

3.7 F-

D,

4’

Then

Equal absorbed doses of thermal and intermediate neutrons will produce the same effect (R,:, = 2) in a detector for which the following relation holds

FIG. 4.-Absorbed dose per unit neutron flux for a number of proton RBE values q@‘. &v -mean dose absorbed by tisssue, and Drnax - maximum dose absorbed by tissue for thermal neutrons (t); for the spectrum of intermediate neutrons -l/E(i) and fast neutrons (f) with enegies of 1 MeV. At the bottom is shown the neutron RBE. The broken line is obtained when the intermediate neutron spectrum is extended up to 1 MeV.

s 0.5

x 10” d &

E

0.4

N_ 50.

(4)

CT

foil of 0.5 mm thickness absorbs practically all neutrons with energies below O-4 eV(“). After the cadmium ratio Rc, has been ascertained TABLE 2.-ABSORBED

DOSE FROM

D

In this case the induced activity is proportional to the sum of the absorbed doses from thermal and interSuch a detector is the isotope mediate neutrons.

(MFREM/N/CM~)

0.4 TO 0.5 MEV

AVERAGED (OR

0.4

TO 1

OVER THE SPECTRUM

-l/E

MEV)

RBE of protons (ri)

7,

1 M%ltl tissue

* Institute

of Chemical

0.096

0.32

1

1 0.70 1

2.4 3.4

Results obtained

1at IKhF*

1

Mean, :issue ___

I-

Results repor ted inlE) ____

Results obtained at IKhF’

Mean 1 :issue

-._.

Maximum .._~~~_

0.30

I

’ 0.57 1

y.91

~

y.63

3.6

~

2.7

:, 2.2

/

::;

Physics of the Academy

4-5 Maximum

Maximum

0.23 1 7.25

l]

1

Results reported in(‘J

_____ For thermal neutwns dose CD,) RBE For intermediate nemr~ns up to 0.5 MeV dose (DJ RBE For intermediate ne”trOns up to 1 MeV dose (0;) RBE For fast ne”trOns up to 1 MeV dose CD,) RBE D,lDt DilDi

=

0.31 1.2

of Sciences of the U.S.S.R.

1.0 1.6 22.0

A.

156

G. ISTOMINA and I. B. KEIRIM-MARKUS

123Sb(12). Unfortunately when neutrons are captured by 123Sb a number of isomers with different half-lives are formed; moreover in natural antimony the isotope 12’Sb is also activated for which the ratio of the crosssections is 21*6(3). No reference can be found in the current literature to other isotopes for which the cross-section ratio given by the expression (4) is in the region of 50. The activation method gives the value of the absorbed dose averaged over the activation time. For accurate measurements the activation time should be less than the half-life of the induced activity by a factor of three to five. For one-day dosimetric monitoring, gold foil is suitable (T,,, = 2.7 days, ct = 98 barns, J = 1558 barnso3)); moreover Fi -= Ft

0.87 R,,l’

3

_

3.2

D, -R,,--1

D=D,fDi=D, Thus a gold detector is almost equally sensitive to Indium intermediate and thermal neutron fluxes. detectors have a similar property. Pieces of gold foil may be used to measure the maximum permissible dose of neutrons: some 0.5 PC of activity is induced in 100 mg of gold after 6 hr. Expressions (2) and (3) are also suitable for measuring the absorbed dose or flux. In the thermal energy region exothermal nuclear reactions on rOB, 6Li and 14N, the cross sections of which are inversely proportional to the velocity t’ of the neutrons, are used. For l/v detectors expression (4) is equal to 0.5 (6), hence 100 27.8 LJ _ Fi _ _p* --Fl; D, Rm - 1 Ft In other words, l/v detectors are a hundred times more sensitive to thermal neutrons than to intermediate energy neutrons having the same biological effect. Ionization chambers or counters filled with boron trifluoride detect thermal neutrons with an efficiency of a few per cent. c4*14)Thus a boron counter covered with a protective layer of cadmium 0.5-l mm thick may be used for the dosimetry of small intermediate neutron fluxes. A standard counter of this type records about 200 counts/min in the maximum permissible flux of intermediate neutrons. Scintillation counters with a ZnS-Ag, B phosphor detect thermal neutrons with an efficiency of a few per cent.05*16) A counter sealed with cadmium and with a 6 cm2 screen records about 100 counts/min in the maximum permissible dose of intermediate neutrons.

Existing methods of personnel dosimetry for monitoring thermal neutrons(17-20) cannot suitably be adapted for the dosimetry of intermediate neutrons. Fission chambers may also be utilized to record intermediate neutrons. For example, in the case of 235U (Go = 549 barns, j = 400 barnst2u) 18.7 Fi- --; l”t Rc, - 1

Et _ 70 _p. D, Rc, -

1

As with boron detectors it is convenient to have two chambers containing different amounts of 235U for use with thermal and intermediate neutrons’22). Fission chambers have uranium coatings up to 10 mg/cm2 thick.(14,23) The maximum permissible flux of intermediate neutrons produces up to 20 fissions per minute from 1 cm2 of the uranium coating. The sensitivity of the fission chamber to fast neutrons is much smaller. Chambers using 233U or 23gPu will have analogous characteristics. Detectors with thresholds higher than 0.5 MeV (238U, sulphur etc.(24)) are obviously unsuitable for intermediate neutrons. We shall now consider the behaviour of detectors such as ionization chambers and proportional counters which have tissue-equivalent walls and filling, and record neutrons by means of recoil protons(g). Assume the proportional counter is connected to a back-biased amplifier which passes pulses corresponding to an energy release greater than B. The ionization current attributable to the recoil protons(4,25) is i(E) = n~(E)o(E)(E

- B)K

(E, B) CY E112(1 - ;I K(E, B),

(5)

where E = energy of the neutrons; n = number of hydrogen atoms contained in the counter; 9 = neutron flux; G = proton scattering cross section; K = correction coefficient for sub-threshold ionization losses’26) (K normally has a value 0.7-0.9 which slowly decreases as E increases). As E increases the ionization current grows, quickly at first then somewhat more slowly than 1/E. The counter fails to respond to neutrons with energies below B so that its mean efficiency for intermediate neutrons is much lower than for fast neutrons. The measurement of intermediate neutrons in the presence of fast neutrons is difficult. An ionization chamber is analogous to a proportional counter with a low discrimination threshold (B N 20 keV). As this threshold falls the detection

Maximum permissible fluxes of intermediate energy neutrons and their measurements

efficiency for intermediate neutrons grows relative to that for fast neutrons. But since for y-radiation the threshold is almost zero, the ionization chamber measures the accompanying y-radiation simultaneously, i.e. its selectivity is poor. Thus in practice ionization chambers are unsuitable. A higher selectivity may be achieved for fast neutrons provided they are counted with a proportional counter. The efficiency of the counter’“5’ F(E) y I?’ .‘? 1 - ; K(E, B) t )

(6)

increases sharply above the threshold, attains a maximum at E e 3B, and then decreases uniformly with increasing E. However the mean efficiency for an intermediate neutron flux with a l/E spectrum is several times less than that for fast neutrons with energies of l-2 MeV (at B - 0.1 MeV). Counters for penetrating radiation are even less suitable for measuring intermediate neutrons; they are characterized by their directivity, and this is frequently undesirable. Thus, of all the types of detector discussed above, those suitable for recording intermediate neutrons are gold, 1°B or 235U thermal neutron detectors shielded with cadmium (or boron). At the present time there is no acceptable method of designing a practical neutron dosimeter that can be used in all energy regions of practical importance. Devices equivalent to a long counter may possibly constitute an exception.(4.2”.25,S7J Perhaps by choosing an appropriate configuration of the retarding and absorbing media it may be possible to establish a relationship between efficiency and energy of the kind illustrated in Figs. 2 and 3. REFERENCES

I. 2. 3. 4. 5. 6.

SNYUER W. and N~UFEI,L> J., Brit. J. Rrrdiol. 28, No.

330, 342 (1955). SNYDERW. and NEUFELI)J., ORNL-LR-DW 11192~11205. SNYDERW. and NEUFELUJ., ORNL-LR-DW 11546; 19164. Rossr H., Rndintion Dosimrtry. (Edited by HINE G. J. and BROWNELLG. L.) New York (1956). BErt(E G. and ASHKIN J., Experimental Ntrclenr Physics. (Edited by SECREE.) Vol. 1, New York (1953). HUGHES D., Pile Neutron Resenrch, Cambridge (Mass.) (1953).

157

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