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A secular technetium–molybdenum generator
Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46
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Nuclear Instruments and Methods in Physics Research A journal homepage: www.elsevier.com/locate/nima
A secular technetium–molybdenum generator Wagner L. Araujo, Tarcisio P.R. Campos Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Belo Horizonte, 31.270-901, Minas Gerais, Brazil
art ic l e i nf o
a b s t r a c t
Article history: Received 5 November 2014 Received in revised form 23 January 2015 Accepted 24 January 2015 Available online 7 February 2015
A compact secular molybdenium generator is subject of this paper. This generator represents a nuclear system that comprises a hydrogen-isotopes fusor, moderator, reflector and shield. Deuterium fusion reactions in a tritiated or deuterated target provide the neutron source. A moderation fluid slowdown the neutron energy which increases 98Mo(n,γ)99Mo capture reaction rates. Neutron reflection minimizes the neutron escape and the radiation shield encloses the device. The neutron yield calculation along with electromagnetic and nuclear simulations were addressed. Results revealed the accelerator equipotential surfaces ranging from 30 to 150 kV, the ion trajectories and the energy beam profile define a deuteron current of 1 A with energy of 180 keV at the target, the spatial distribution of the neutron flux, and the 99 Mo and 99mTc activities in function of transmuter operation time. The kinetics of the 99mTc correlated to its precursor activity demonstrates a secular equilibrium providing 2 Ci in a operational time of 150 h. As conclusion, the investigated nuclear and electromagnetic features have demonstrated that such generator shall have a notable potential for feeding the 99mTc clinical application. & 2015 Elsevier B.V. All rights reserved.
Keywords: Tc-99m Radioisotope generator Molybdenum Deuteron accelerator dd fusor
1. Introduction Radiopharmaceuticals based on 99mTc become, over the last 50 years, important tools in the diagnosis of various diseases and organ disorders. The main reasons of such success is based on its prompt availability by a radionuclide generator associated with its plurality of chemical valences and relative low cost [1]. Several complexes labeled with technetium turn over to diagnosis routine in nuclear medicine, yielding 80% of the clinical protocols [2–5]. The 99mTc radionuclide is generally produced by 99Mo decayment enclosed in a column inside a radionuclide generator. The column is filled with adsorbed material made of alumina or zirconia, in which the 99Mo irradiated isotope was adsorbed. The 99mTc daughter radionuclide grows as the result of the decayment of an initial mass of 99Mo radionuclide until a transient equilibrium is reached. The activity of the daughter is eluted by a saline solvent, leaving the residual 99Mo radionuclide and the 98Mo–Al or Zr columns. Ideal chemical and physical 99mTc properties, like physical half-life of 6.01 h, decay by isomeric transition, emission of a 140 keV gamma ray, quickly elution from a 99Mo/99mTc generator system and the diversity of oxidation states with possible large number of coordination links held by the 99mTc ion giving rise to different radiopharmaceuticals based on a simple reconstitution of lyophilized reagent sets, justify its high rate of application. Modern portable gel-column generators have been produced [6–8]. In these devices, the radiochemical purity
E-mail addresses: [email protected] (W.L. Araujo), [email protected] (T.P.R. Campos). http://dx.doi.org/10.1016/j.nima.2015.01.095 0168-9002/& 2015 Elsevier B.V. All rights reserved.
of 99mTc is higher than 99%, and the final product contains less than 0.02%) of 99Mo. Generally the technology for preparing radioactive columns for radionuclide generators, such as 99Mo/99mTc, involves large and complex installations. The 99Mo can be generated by neutron activation of 98Mo in a fission reactor, based on the 98Mo neutron capture reaction. However, the fission route in nuclear reactor is the common choice to the commercial production of 99Mo. Two types of fission target are often used: high-enriched uranium (HEU) and low-enriched uranium (LEU), based on aluminate dispersion targets, are used. 99Mo, 133Xe and 131I are all produced together. Another possibility is 99Mo production based on nuclear reactor, operating in low power, whose fuel is a uranium solution. In this case, the 235U target are not solid as HEU or LEU-targets, but in a homogeneous fluid fuel. Thus, during the operation, 99Mo is produced and removed continually [9–12]. Fig. 1 provides the microscopic cross-sections of the neutron nuclear reactions addressed to produced 99Mo. Proton accelerators can also be used to produce 99Mo and 99mTc. It is produced by 100Mo(p,2n)99mTc reaction. A proton beam of 20 MeV accelerated in a isochronal cyclotron can bombard a molybdenum metallic target, which can be highly enriched in 100Mo (4 99%)). Indeed, this type of device has been applied to produce 18F, 15O, 11 C, 13N and now 99mTc through the 100Mo(p,2n)99mTc reaction. An estimated activity is about 16 Ci mA 1h 1(Ci of 99mTc EOB) at 24 MeV [9]. The corresponding saturated 99mTc activity is 141 Ci mA 1(Ci of 99mTc EOB) for a beam of protons (6 mm diameter) with 25 MeV and metallic Mo targets (density of 10.3 g cm 3), and a total stopping power of about 13 MeV mm 1 [9].
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46
Fig. 1. Microscopic cross-sections of 99 Mo and 235U fission [13].
100
Mo(p,2n)
99m
Tc,
100
Mo(n, 2n),99Mo,
98
Mo(n,g)
The 99Mo can also be production by photo fission route. Two reactions can be used based on U or Mo target, as 100Mo(γ,n) 99Mo or 238U(γ,f) 99Mo. In this case, intense photon flux can be produced by linear electron accelerators [9]. Spallation neutron sources provide high intense neutron flux, whose neutron fission can be used to produce 99Mo. Particularly , 100 Mo(n, 2n) 99Mo reaction are of interest. The cross-section of 100 Mo(n,2n) 99Mo reaction is close to 1.5 b at the 10–17 MeV energy interval. A 40 MeV deuteron beam bombarding a natural carbon converter can generate intense 14 MeV neutrons, which can be interact via 100Mo(n,2n) 99Mo reaction [9]. The spallation neutron sources can also support the 235U(n,f) 99Mo reactions [14]. In summary, uranium fission in research reactors (HEU and LEUalumine dispersion targets), liquid-fuel reactor technology and neutron activation of 98Mo in nuclear reactor are based on various technological scenarios together with direct 99mTc production with isochronal cyclotrons are the methods of 99Mo/ 99mTc production at the present time. Spallation neutron source by 100Mo(n,2n) have been investigated to produced 99Mo, but it has not been applied commercially yet. Nevertheless, 99Mo from fission fragments of 235U generated into nuclear reactor is the most common method for assembling 99Mo/99mTc alumina-based column [15]. Nowadays, the world demand for 99Mo is about 450,000 GBq per week and the annual demand for 99Mo is considered to have an 8–12% growth over this decade [16]. Currently, five main nuclear research reactors located in Canada, Belgium, France, Holland and South Africa [17] produce 99Mo at commercial scale. Factors associated mainly to the cost, complexity and security involving nuclear reactor technology can make difficult the application of the 99m Tc, considering that all other nonproducer countries import the 99 Mo radioactive columns from countries that holds technology of production. Such radioactive columns have been used to assemble the radioisotope generator which shall be weekly transferred to the nuclear medical clinics. The 99Mo radionuclide has a half-life of 65.94 h and decays by beta minus emission; however, 87%) goes to the metastable state of technetium, generating 99mTc; while 13%) to the ground state of technetium which has a half-life of 2.1 105 year. The technetium decay to 99Ru isotope by beta emission. On this nuclear process, photons of 740 and 780 keV are emitted. The daughter nuclide 99m Tc decays by isomeric transition; however, 10% decays by internal conversion. A typical decay-growth kinetics of 99mTc and 99 Mo nuclide shows activities in transient equilibrium due to the fact that half-live of 99Mo is about 10 times greater than that of 99m Tc, which is 6.01 h, where the yield of 99mTc is maximum close to 24 h after elution.
41
In order of transmutation of 98Mo to 99Mo, the following three major processing can be addressed: nuclear reactors, radioisotope sources and particle accelerators. Nuclear reactors can provide high level of neutron flux; however, reactors are complex, expensive and has large dimensions. Sealed radioisotope source emits radiation whose strengths decay with time. Hence, these sources are associated to continuum radiation protection requirements. Another method of neutron generation is represented by particle accelerators. This device may provide advantages over the other two available neutron sources, since it is able to turn off. The essence of a modern and compact neutron generator based on deuteron fusion should comprise the design of a gas-control reservoir, a plasma and ion source to generate and gather the ions in a beam shape, a set of electrodes, including a metal target electrode loaded with deuterium (2H) or tritium (3H) hydrides. The plasma source produces the ions generally through magnet and electrode configurations or radio frequency antenna. Subsequently, the deuterium or tritium ions, deuterons (d) or tritons (t), are accelerated by an electrode system toward a hydride target loaded with deuterium, tritium, or a mixture of both; in which the fusion reaction occur generating neutrons, that are emitted with energy of 2.45 MeV from d-d and 14.1 MeV from d–t reactions [18]. Although the d–t reaction is more prolific in terms of neutron generation, tritium is a radioisotope while deuterium is stable. In this case, a generator can be assembled in compact dimensions due to the appreciable isotope hydrogen fusion cross-section at relative low-energy acceleration, depicted on Fig. 2. In an attempt to improve the commercial technetium generation method, a generator based on neutron activation of the precursor is presented, whose technology includes a potential route to 99m Tc and other radioisotopes production. The main issues related to a fusor generator for radioisotope production can be summarized as: (i) supplement of radioisotopes rich in neutrons of short physical half-life; (ii) independence of the high flux reactor technology in the activation of 99Mo, available only in few industrialized countries; (iii) replacement of complex technologies represented by reactors or particle accelerators; (iv) supporting the philosophy of compact radioisotope generators, without dependence on preexistent radioactive parent nuclide; (v) improvement of radiological safety, allowing full facility shutdown and consequently interruption of the radiation produced in the device, similar to a x-ray machine with minimum radioactive internal contamination. Herein, a device named Hemispherical Fusor Technetium Generator (HFTG) [20] is presented based on a particle accelerator for Energy
d-t reaction d-d reaction
Energy Fig. 2. Cross-sections of neutron production for 2H and 3H target bombarded by deuterons [19].
42
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46
neutron generation by means of hydrogen isotope fusion whose neutron particles are collimated, slowed down and reflected toward 98 Mo nuclides dispersed in columns placed onto the moderator. This article addressed the design of the secular molybdenum generator, showing its main structures, the equipotential surfaces and the radiation shield. Also, it presents electromagnetic simulations of the accelerator head and the deuterium beamlet trajectories. From this, deuteron current and neutron yield is estimated. And, a nuclear simulation supported the information on the molybdenum-technetium balance onto the moderator.
2. Design HFTG, depicted in Fig. 3(a), consists essentially of a plasma chamber, a set of electrodes designed as hemi-spherical shells, a receptacle for the molybdenum target, a neutron moderator and structures for shielding and reflecting the radiation. The device geometry shown here reproduces the present level of the apparatus expertise. The electrodes are shown in Fig. 3(b). The plasma electrode (PE), with internal and external radius of 12.5 and 13.0 cm, possesses 34 cylindrical apertures with 2.48 mm in radius respectively. The hemispherical target electrode (TE) is defined with 5.0 cm in radius. The plasma housing (PH) with 16.0 cm in internal radius, between PE and PH, is a container structure where the plasma generation take place. Cylindrical shielding structure (SS) possesses radial thickness of 40.0 cm and length of 1.2 m, and neutron reflector (NR) with 10 cm in radius.
3. Methods 3.1. Electromagnetic simulation Simulation of the electromagnetic interactions into the acceleration site was performed on the CST package [21]. It possess a 3D
PH
code, in the module CST Particle Studio, which computes the particles path through a pre-calculated electromagnetic field. Electric and magnetic fields are evaluated on a computation grid, in which the code interpolates the fields to the particle position following a linear interpolation scheme. Particle trajectory equations are based on updates of time (Eq. (1)) and position (Eq. (2)), as follows: !n þ 1=2 !n þ 1=2 !n þ 1=2 !n þ 1 !n mn þ 1 v ¼ mn v þ qΔtð E þv B Þ
!n þ 3=2 !n þ 1=2 !n þ 1 r ¼ r þv Δt ð2Þ ! ! where B is the magnetic field, q is the particle charge, r is the particle position, and Δt interval of time. The ions emerge from the plasma by the electric potentials of the electrodes, traveling from each plasma electrode hole toward the target electrode. The beamlet space-charge produced by a set of ions emerging from each hole was simulated simultaneously; whose the sum of beamlet currents produces the total deuteron current (id). Accelerator geometry was optimized by simulations, considering the beam current levels extracted by the electrodes and its intensity at the electrode target. 3.2. d–d reaction rate After consolidate the electrodes geometry, the neutron yield (Υ ) at the target was evaluated based on the current provided by electromagnetic simulations. It was considered a deuteron beam composed of monoatomic and molecular species impinging on a titanium target loaded with deuterium. Thus the neutron yield of the generator, could be appraised in according to Eq. (3) [22], as follows.
Υ¼
2 ηd id X
e
k¼1
Z kf k
E 0
σ dd dE=dx
dE;
ð3Þ
in which ηd is the number of deuterons per cm3 in the target, σ dd ðEÞ is the neutron production cross-section of the fusion reaction d-d in function of deuteron beam energy. Ion species are weighted by their fraction fk and their number of nuclei k per ion. dE=dxðEÞ is the molecular stopping power of the target loaded with deuterium. The Bragg's law of additivity [23] was applied to determine the stopping power in the target, as shown in the following equation:
TE
dE dE dE ¼ þ ηd ; dx dx M dx d
PE
ð1Þ
MV
RS
ð4Þ
in which the index M and d represents the metal and deuterium on the target. The stopping power values was obtained by means of SRIM code [24]. Eq. (3) was solved by numerical integration, discretizing the energy in intervals of 30 keV and considering the deuterium density in the target ρd equals to 3.76 g cm 3 [25]. 3.3. Nuclear simulation
PH
PE
TE
Fig. 3. (a) General representation of the HFTG presenting the plasma housing (PH), plasma electrode (PE), target (TE), neutron moderation volume (MV) and sample container (RS) and (b) plasma electrode (PE), target (TE) and plasma housing (PH).
The deuteron flux decreases as the deuteron beam penetrates the target, this point is provided by the stopping power calculation obtained by SRIM code. The thickness of the plasma electrode is defined slit larger than the range of the deuterium in the target. The overall neutrons generated were evaluated by a general deuterium current and not by an individual current produced by beamlet from each hole. However, after find the neutron yield, the spatial distribution of the neutron source, established on MCNP, was defined as a set points inside and three-dimensional distributed over a semi-sphere defined by the target electrode, in number equivalent to the number of the holes in the plasma electrode.
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46
The neutronic of the HFTC was simulated in the MCNP5 [26] code incorporating the neutron reflectors and shield structures, and 98Mo dispersed nuclide samples. The rate of neutron emission from d–d fusion reactions at the target electrode was introduced in the MCNP code as particle source. Spatial and spectral neutron flux and 98Mo(n,γ) 99Mo absorbed capture reaction rate were investigated on the moderator volume. After neutron yield calculations with the helping of the CST code, the distribution of neutron source on the target electrode was considered on the MCNP. The spatial distribution of neutrons and its spectrum in the moderator volume were evaluated by voxelaposs distribution. The MCNP executed the transport of monoenergetic neutrons of 2.45 MeV investigating its slowing down to thermalized energy into the moderator. The macroscopic cross-Sections of the 98Mo material were available in ENDF-VIIB.8. At each voxel, the transmutation rate in atoms per cm3 was evaluated multiplying the microscopic transmutation cross-section, by the neutron flux, normalized by the number of neutrons emitted at the source, and by the atomic density per volume unit of the 98Mo sample.
Later, the volume transmutation rate of (Eq. (5)), was evaluated as follows: N 〈σ Z ðEÞϕðE; rÞ〉 ; Rv ¼ ρω A p A
43 98
Mo, defined as Rv
ð5Þ
where ρ is the sample density, NA is Avogadro's number, A is the sample atomic mass, ω is chemical fraction multiplied by the isotope fraction of the sample in the compound, σ Z ðEÞ is the neutron ! absorption cross-section and ϕðE; r Þ is the neutron flux at the ! position r in the moderator, both with energy dependence. The ! term σ Z ðEÞϕðE; r Þ per particle p is evaluated by MCNP code in a ! given position r from the moderator and integrated on energy spectrum in the interval between the thermal energy and 2.5 MeV. Afterwards, the geometric parameters and materials of the reflectors and moderator were adjusted in order to maximize the rate of 99Mo production achieving the present suitable HFTG setup. The transmutation rate was multiplied by Υ and by the 98Mo volume V and inserted into Eq. (6), in order to determine the radioisotope
Fig. 4. Electromagnetic results: (a) potential distribution and (b) deuteron beam paths.
44
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46
activity as function of irradiation time, as follows: AðtÞ ¼ YRv Vð1 e in which
λt
Þ
R1
ð6Þ
EI PE
λ is the decay constant of the radioisotope product.
4. Results NM 1
TE
4.1. Electromagnetic and Nuclear Evaluation An electric potential of 30 kV in the plasma electrode and 150 kV in the target was assumed. The electrodes geometries were adjusted in order to correct improprieties in deuterium trajectory, optimizing the energy and current of the beam at the target electrode. Electrons impinged upon the titanium target were not simulated. However, a collector in a ring shape around the target loaded with a slightly higher potential than the target potential is installed for the purpose of collecting electrons and negative ions produced in collisions of the accelerated deuterium beam at the target. The fraction of the negatively ionized deuterium molecules are collected by the ring and the fraction of positively charged deuterium molecules are incorporated into the target, increasing the density of the deuterons at the target and thereby the neutron yield estimated by Eq. (3). A map of equipotential surfaces in the vicinity of the electrodes is shown in Fig. 4(a), and the Fig. 4(b) presents the path of the deuteron beams. This electrode configuration allowed to achieve a deuteron current of 1 A at the targets, finding in on CST simulations. According to numerical integration in Eq. (3) the device reaches a yield of 1013 n s 1. Several arrangements of HFTG, based on distinct geometry, material and elemental composition were designed in MCNP-5 code in order to allow us achieving highest rate of isotope transmutation. Fig. 5 shows some essentials components of the device generated on MCNP 5 code and materials in Table 1. Neutron source was distributed on both target electrodes. Axial neutron flux at the moderator, revealing a flux in the order of 1012 n cm 3 s 1, as shown in Fig. 6. Neutron fluency was multiplied by the atomic density of 98Mo and by the (n,γ) cross-section, determining the 98Mo transmutation volume rate, Rv. The term ρωðNA =AÞ, in the Eq. (5) is equal to 4:3 10 3 at. cm 1 b 1. This yield was multiplied by γ and inserted in Eq. (6) in order to determine the 99Mo and 99mTc activities as a function of irradiation time at the central position of the sample receptacle.
MD, NM2
EI
R1 R2 SH
PE
Fig. 5. HFTG (a) internal and (b) general components: electric insulator (EI), plasma electrode (PE), target (TE), neutron moderator (NM1) and the mesh data (MD) in the neutron moderation site (NM2), internal reflector (R1), external reflector and radiation shield (SH).
Table 1 Some materials of HFTG. Volume
Description
Material
SH R1 EI PE TE R2 NM2 NM1
Radiation shield (SH) Reflector Insulator Plasma electrode Target Reflector Moderator Moderador
Borated polyethylene Chromel A Kaolinite Copper Copper/titanium/deuterium Lead–bismuth alloy Graphite Water
5. Discussion and conclusion It is considered that the HFTG space-charge is a dynamic system that will find a state-stable mode. As different from usual neutron generator which operates to a limited number of hours until its gascamera does not fill more the ions; the HFTG has a cold plasma in very low pressure, ionized by electromagnetic field, that will be feed by a deuterium gas inlet to keep the pressure, temperature and plasma density constant as possible in function of long time operation. The ions will be generated on plasma camera. The ion charges on the plasma are kept balanced, despite of large number of electrons and positive ions. Although some positive ions will be extracted by the electrode system toward the target; electrons from plasma and impinged from target will pollute the environment. However the influence of those electrons can be easily limited by collectors placed near target, with slight different potential. The HFTG simulation with the present configurations has demonstrated that such generator is able to elute 99mTc, after 150 h of operation, with activities up to 2 Ci, equivalent to a commercial
Fig. 6. Axial neutron flux at the moderator.
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46 99 Mo/99mTc transient generator (Fig. 7 and Fig. 8). However, the production of 99mTc on the secular generator, based on the HFTG technology, is constant along the time, producing 99Mo continuously in locus; whereas conventionally 99Mo/99mTc generators are often used for five days and replaced weekly. The technology of HFTG is compact, simpler when compared with a nuclear reactor or particle accelerator. The set up presented here is powered by electric potential difference in the order of 200 kV and a deuterium gas. When turned off, the device ceases
10.0
5
45
the radiation, equivalent to a conventional X-ray device. Tritium confined on the target electrode and radioactive nuclides on the metal structures of the HFTG shall be the contaminants present on this device, however all possible contaminant radionuclide would be confined into the device. Such contamination could be minimized by the choice of low neutron absorption materials to assemble the device, as polymers and zirconium alloys, preferable, or aluminum and titanium. The simulation findings have revealed the capacity of the HFTG to produce 99mTc from 98Mo activation with suitable activities for clinical use. HFTG may compete with the production in locus by commercial generators where there is need of loading an amount of radioactive precursor, said radioactive parent, previously produced in high-flux nuclear reactors. The management of 99Mo/99mTc generator is taken weekly since its transient activity decreases proportionality to the parent half-life. Therefore, the major feature of 99Mo/99mTc generator is the a transient regime presented by the balance of activities between 99Mo and 99mTc. Technology presented in the HFTG can continuously generate the parent, whose activity on the device becomes stable, and consequently the balance between the parent and daughter follows a secular regime. A more efficient radiological protection is found with a HFTG running with dd fusion reactions. However, dt reaction is also possible. Fig. 2 shows that he cross-section of dt interaction, at the deuteron kinetic energies of 100 keV, is about 250 times lager than the dd cross-section evaluated in the present investigations. In this case, a neutron yield with two orders of magnitude above those findings with a HFTG running dd reactions can be reached. Under strict conditions of radiological protection, HFTG based on dt fusion could be implemented in hospitals facilities, reducing the operation time for activation in proportion to the increase of the fusion crosssection. New theoretical investigations should be conducted on this matter. Electromagnetic and nuclear evaluations along with the present calculations highlighted the potential of the HFTG. We expect in the near future that the investigations will drive toward experimental level.
7.5
References 0 5.0
_5
2.5
_ 10 98
0
_5
0
5
10
Mo transmutation volume rate in (a) 3d graph and (b) transverse view.
Molybdenum-99 Technetium-99m Eluted Technetium-99m
A(t) (Ci)
Fig. 7.
[1] S. Hesslewood, D. Keeling, European Journal of Nuclear Medicine 24 (1997) 1179. [2] M.K. Dewanjee, Radiochemistry: A Current Status Report I. Seminars in Nuclear Medicine 20 (1990) 5. [3] S. Jurisson, D. Berning, W. Jia, D. Ma, Chemical Reviews 93 (1993) 1137. [4] J. Knapp, F.F. Russ, S. Mirzadeh, European Journal of Nuclear Medicine 21 (1994) 1151. [5] F.L.N. Marques, M.R.Y. Okamoto, C.A. Buchpiguel, Radiologia Brasileira 34 (2001) 233. [6] P. Saraswathy, A.C. Dey, S.K. Sarkar, C. Kothalkar, P. Naskar, G. Arjun, S.S. Arora, A.K. Kohli, Tc-99m generators for clinical use based on zirconium molybdate
t Fig. 8. The
99
Mo and
99m
Tc activities, in mCi, as a function of time.
46
[7]
[8]
[9]
[10]
[11]
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Contents lists available at ScienceDirect
Nuclear Instruments and Methods in Physics Research A journal homepage: www.elsevier.com/locate/nima
A secular technetium–molybdenum generator Wagner L. Araujo, Tarcisio P.R. Campos Departamento de Engenharia Nuclear, Universidade Federal de Minas Gerais, Belo Horizonte, 31.270-901, Minas Gerais, Brazil
art ic l e i nf o
a b s t r a c t
Article history: Received 5 November 2014 Received in revised form 23 January 2015 Accepted 24 January 2015 Available online 7 February 2015
A compact secular molybdenium generator is subject of this paper. This generator represents a nuclear system that comprises a hydrogen-isotopes fusor, moderator, reflector and shield. Deuterium fusion reactions in a tritiated or deuterated target provide the neutron source. A moderation fluid slowdown the neutron energy which increases 98Mo(n,γ)99Mo capture reaction rates. Neutron reflection minimizes the neutron escape and the radiation shield encloses the device. The neutron yield calculation along with electromagnetic and nuclear simulations were addressed. Results revealed the accelerator equipotential surfaces ranging from 30 to 150 kV, the ion trajectories and the energy beam profile define a deuteron current of 1 A with energy of 180 keV at the target, the spatial distribution of the neutron flux, and the 99 Mo and 99mTc activities in function of transmuter operation time. The kinetics of the 99mTc correlated to its precursor activity demonstrates a secular equilibrium providing 2 Ci in a operational time of 150 h. As conclusion, the investigated nuclear and electromagnetic features have demonstrated that such generator shall have a notable potential for feeding the 99mTc clinical application. & 2015 Elsevier B.V. All rights reserved.
Keywords: Tc-99m Radioisotope generator Molybdenum Deuteron accelerator dd fusor
1. Introduction Radiopharmaceuticals based on 99mTc become, over the last 50 years, important tools in the diagnosis of various diseases and organ disorders. The main reasons of such success is based on its prompt availability by a radionuclide generator associated with its plurality of chemical valences and relative low cost [1]. Several complexes labeled with technetium turn over to diagnosis routine in nuclear medicine, yielding 80% of the clinical protocols [2–5]. The 99mTc radionuclide is generally produced by 99Mo decayment enclosed in a column inside a radionuclide generator. The column is filled with adsorbed material made of alumina or zirconia, in which the 99Mo irradiated isotope was adsorbed. The 99mTc daughter radionuclide grows as the result of the decayment of an initial mass of 99Mo radionuclide until a transient equilibrium is reached. The activity of the daughter is eluted by a saline solvent, leaving the residual 99Mo radionuclide and the 98Mo–Al or Zr columns. Ideal chemical and physical 99mTc properties, like physical half-life of 6.01 h, decay by isomeric transition, emission of a 140 keV gamma ray, quickly elution from a 99Mo/99mTc generator system and the diversity of oxidation states with possible large number of coordination links held by the 99mTc ion giving rise to different radiopharmaceuticals based on a simple reconstitution of lyophilized reagent sets, justify its high rate of application. Modern portable gel-column generators have been produced [6–8]. In these devices, the radiochemical purity
E-mail addresses: [email protected] (W.L. Araujo), [email protected] (T.P.R. Campos). http://dx.doi.org/10.1016/j.nima.2015.01.095 0168-9002/& 2015 Elsevier B.V. All rights reserved.
of 99mTc is higher than 99%, and the final product contains less than 0.02%) of 99Mo. Generally the technology for preparing radioactive columns for radionuclide generators, such as 99Mo/99mTc, involves large and complex installations. The 99Mo can be generated by neutron activation of 98Mo in a fission reactor, based on the 98Mo neutron capture reaction. However, the fission route in nuclear reactor is the common choice to the commercial production of 99Mo. Two types of fission target are often used: high-enriched uranium (HEU) and low-enriched uranium (LEU), based on aluminate dispersion targets, are used. 99Mo, 133Xe and 131I are all produced together. Another possibility is 99Mo production based on nuclear reactor, operating in low power, whose fuel is a uranium solution. In this case, the 235U target are not solid as HEU or LEU-targets, but in a homogeneous fluid fuel. Thus, during the operation, 99Mo is produced and removed continually [9–12]. Fig. 1 provides the microscopic cross-sections of the neutron nuclear reactions addressed to produced 99Mo. Proton accelerators can also be used to produce 99Mo and 99mTc. It is produced by 100Mo(p,2n)99mTc reaction. A proton beam of 20 MeV accelerated in a isochronal cyclotron can bombard a molybdenum metallic target, which can be highly enriched in 100Mo (4 99%)). Indeed, this type of device has been applied to produce 18F, 15O, 11 C, 13N and now 99mTc through the 100Mo(p,2n)99mTc reaction. An estimated activity is about 16 Ci mA 1h 1(Ci of 99mTc EOB) at 24 MeV [9]. The corresponding saturated 99mTc activity is 141 Ci mA 1(Ci of 99mTc EOB) for a beam of protons (6 mm diameter) with 25 MeV and metallic Mo targets (density of 10.3 g cm 3), and a total stopping power of about 13 MeV mm 1 [9].
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46
Fig. 1. Microscopic cross-sections of 99 Mo and 235U fission [13].
100
Mo(p,2n)
99m
Tc,
100
Mo(n, 2n),99Mo,
98
Mo(n,g)
The 99Mo can also be production by photo fission route. Two reactions can be used based on U or Mo target, as 100Mo(γ,n) 99Mo or 238U(γ,f) 99Mo. In this case, intense photon flux can be produced by linear electron accelerators [9]. Spallation neutron sources provide high intense neutron flux, whose neutron fission can be used to produce 99Mo. Particularly , 100 Mo(n, 2n) 99Mo reaction are of interest. The cross-section of 100 Mo(n,2n) 99Mo reaction is close to 1.5 b at the 10–17 MeV energy interval. A 40 MeV deuteron beam bombarding a natural carbon converter can generate intense 14 MeV neutrons, which can be interact via 100Mo(n,2n) 99Mo reaction [9]. The spallation neutron sources can also support the 235U(n,f) 99Mo reactions [14]. In summary, uranium fission in research reactors (HEU and LEUalumine dispersion targets), liquid-fuel reactor technology and neutron activation of 98Mo in nuclear reactor are based on various technological scenarios together with direct 99mTc production with isochronal cyclotrons are the methods of 99Mo/ 99mTc production at the present time. Spallation neutron source by 100Mo(n,2n) have been investigated to produced 99Mo, but it has not been applied commercially yet. Nevertheless, 99Mo from fission fragments of 235U generated into nuclear reactor is the most common method for assembling 99Mo/99mTc alumina-based column [15]. Nowadays, the world demand for 99Mo is about 450,000 GBq per week and the annual demand for 99Mo is considered to have an 8–12% growth over this decade [16]. Currently, five main nuclear research reactors located in Canada, Belgium, France, Holland and South Africa [17] produce 99Mo at commercial scale. Factors associated mainly to the cost, complexity and security involving nuclear reactor technology can make difficult the application of the 99m Tc, considering that all other nonproducer countries import the 99 Mo radioactive columns from countries that holds technology of production. Such radioactive columns have been used to assemble the radioisotope generator which shall be weekly transferred to the nuclear medical clinics. The 99Mo radionuclide has a half-life of 65.94 h and decays by beta minus emission; however, 87%) goes to the metastable state of technetium, generating 99mTc; while 13%) to the ground state of technetium which has a half-life of 2.1 105 year. The technetium decay to 99Ru isotope by beta emission. On this nuclear process, photons of 740 and 780 keV are emitted. The daughter nuclide 99m Tc decays by isomeric transition; however, 10% decays by internal conversion. A typical decay-growth kinetics of 99mTc and 99 Mo nuclide shows activities in transient equilibrium due to the fact that half-live of 99Mo is about 10 times greater than that of 99m Tc, which is 6.01 h, where the yield of 99mTc is maximum close to 24 h after elution.
41
In order of transmutation of 98Mo to 99Mo, the following three major processing can be addressed: nuclear reactors, radioisotope sources and particle accelerators. Nuclear reactors can provide high level of neutron flux; however, reactors are complex, expensive and has large dimensions. Sealed radioisotope source emits radiation whose strengths decay with time. Hence, these sources are associated to continuum radiation protection requirements. Another method of neutron generation is represented by particle accelerators. This device may provide advantages over the other two available neutron sources, since it is able to turn off. The essence of a modern and compact neutron generator based on deuteron fusion should comprise the design of a gas-control reservoir, a plasma and ion source to generate and gather the ions in a beam shape, a set of electrodes, including a metal target electrode loaded with deuterium (2H) or tritium (3H) hydrides. The plasma source produces the ions generally through magnet and electrode configurations or radio frequency antenna. Subsequently, the deuterium or tritium ions, deuterons (d) or tritons (t), are accelerated by an electrode system toward a hydride target loaded with deuterium, tritium, or a mixture of both; in which the fusion reaction occur generating neutrons, that are emitted with energy of 2.45 MeV from d-d and 14.1 MeV from d–t reactions [18]. Although the d–t reaction is more prolific in terms of neutron generation, tritium is a radioisotope while deuterium is stable. In this case, a generator can be assembled in compact dimensions due to the appreciable isotope hydrogen fusion cross-section at relative low-energy acceleration, depicted on Fig. 2. In an attempt to improve the commercial technetium generation method, a generator based on neutron activation of the precursor is presented, whose technology includes a potential route to 99m Tc and other radioisotopes production. The main issues related to a fusor generator for radioisotope production can be summarized as: (i) supplement of radioisotopes rich in neutrons of short physical half-life; (ii) independence of the high flux reactor technology in the activation of 99Mo, available only in few industrialized countries; (iii) replacement of complex technologies represented by reactors or particle accelerators; (iv) supporting the philosophy of compact radioisotope generators, without dependence on preexistent radioactive parent nuclide; (v) improvement of radiological safety, allowing full facility shutdown and consequently interruption of the radiation produced in the device, similar to a x-ray machine with minimum radioactive internal contamination. Herein, a device named Hemispherical Fusor Technetium Generator (HFTG) [20] is presented based on a particle accelerator for Energy
d-t reaction d-d reaction
Energy Fig. 2. Cross-sections of neutron production for 2H and 3H target bombarded by deuterons [19].
42
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46
neutron generation by means of hydrogen isotope fusion whose neutron particles are collimated, slowed down and reflected toward 98 Mo nuclides dispersed in columns placed onto the moderator. This article addressed the design of the secular molybdenum generator, showing its main structures, the equipotential surfaces and the radiation shield. Also, it presents electromagnetic simulations of the accelerator head and the deuterium beamlet trajectories. From this, deuteron current and neutron yield is estimated. And, a nuclear simulation supported the information on the molybdenum-technetium balance onto the moderator.
2. Design HFTG, depicted in Fig. 3(a), consists essentially of a plasma chamber, a set of electrodes designed as hemi-spherical shells, a receptacle for the molybdenum target, a neutron moderator and structures for shielding and reflecting the radiation. The device geometry shown here reproduces the present level of the apparatus expertise. The electrodes are shown in Fig. 3(b). The plasma electrode (PE), with internal and external radius of 12.5 and 13.0 cm, possesses 34 cylindrical apertures with 2.48 mm in radius respectively. The hemispherical target electrode (TE) is defined with 5.0 cm in radius. The plasma housing (PH) with 16.0 cm in internal radius, between PE and PH, is a container structure where the plasma generation take place. Cylindrical shielding structure (SS) possesses radial thickness of 40.0 cm and length of 1.2 m, and neutron reflector (NR) with 10 cm in radius.
3. Methods 3.1. Electromagnetic simulation Simulation of the electromagnetic interactions into the acceleration site was performed on the CST package [21]. It possess a 3D
PH
code, in the module CST Particle Studio, which computes the particles path through a pre-calculated electromagnetic field. Electric and magnetic fields are evaluated on a computation grid, in which the code interpolates the fields to the particle position following a linear interpolation scheme. Particle trajectory equations are based on updates of time (Eq. (1)) and position (Eq. (2)), as follows: !n þ 1=2 !n þ 1=2 !n þ 1=2 !n þ 1 !n mn þ 1 v ¼ mn v þ qΔtð E þv B Þ
!n þ 3=2 !n þ 1=2 !n þ 1 r ¼ r þv Δt ð2Þ ! ! where B is the magnetic field, q is the particle charge, r is the particle position, and Δt interval of time. The ions emerge from the plasma by the electric potentials of the electrodes, traveling from each plasma electrode hole toward the target electrode. The beamlet space-charge produced by a set of ions emerging from each hole was simulated simultaneously; whose the sum of beamlet currents produces the total deuteron current (id). Accelerator geometry was optimized by simulations, considering the beam current levels extracted by the electrodes and its intensity at the electrode target. 3.2. d–d reaction rate After consolidate the electrodes geometry, the neutron yield (Υ ) at the target was evaluated based on the current provided by electromagnetic simulations. It was considered a deuteron beam composed of monoatomic and molecular species impinging on a titanium target loaded with deuterium. Thus the neutron yield of the generator, could be appraised in according to Eq. (3) [22], as follows.
Υ¼
2 ηd id X
e
k¼1
Z kf k
E 0
σ dd dE=dx
dE;
ð3Þ
in which ηd is the number of deuterons per cm3 in the target, σ dd ðEÞ is the neutron production cross-section of the fusion reaction d-d in function of deuteron beam energy. Ion species are weighted by their fraction fk and their number of nuclei k per ion. dE=dxðEÞ is the molecular stopping power of the target loaded with deuterium. The Bragg's law of additivity [23] was applied to determine the stopping power in the target, as shown in the following equation:
TE
dE dE dE ¼ þ ηd ; dx dx M dx d
PE
ð1Þ
MV
RS
ð4Þ
in which the index M and d represents the metal and deuterium on the target. The stopping power values was obtained by means of SRIM code [24]. Eq. (3) was solved by numerical integration, discretizing the energy in intervals of 30 keV and considering the deuterium density in the target ρd equals to 3.76 g cm 3 [25]. 3.3. Nuclear simulation
PH
PE
TE
Fig. 3. (a) General representation of the HFTG presenting the plasma housing (PH), plasma electrode (PE), target (TE), neutron moderation volume (MV) and sample container (RS) and (b) plasma electrode (PE), target (TE) and plasma housing (PH).
The deuteron flux decreases as the deuteron beam penetrates the target, this point is provided by the stopping power calculation obtained by SRIM code. The thickness of the plasma electrode is defined slit larger than the range of the deuterium in the target. The overall neutrons generated were evaluated by a general deuterium current and not by an individual current produced by beamlet from each hole. However, after find the neutron yield, the spatial distribution of the neutron source, established on MCNP, was defined as a set points inside and three-dimensional distributed over a semi-sphere defined by the target electrode, in number equivalent to the number of the holes in the plasma electrode.
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46
The neutronic of the HFTC was simulated in the MCNP5 [26] code incorporating the neutron reflectors and shield structures, and 98Mo dispersed nuclide samples. The rate of neutron emission from d–d fusion reactions at the target electrode was introduced in the MCNP code as particle source. Spatial and spectral neutron flux and 98Mo(n,γ) 99Mo absorbed capture reaction rate were investigated on the moderator volume. After neutron yield calculations with the helping of the CST code, the distribution of neutron source on the target electrode was considered on the MCNP. The spatial distribution of neutrons and its spectrum in the moderator volume were evaluated by voxelaposs distribution. The MCNP executed the transport of monoenergetic neutrons of 2.45 MeV investigating its slowing down to thermalized energy into the moderator. The macroscopic cross-Sections of the 98Mo material were available in ENDF-VIIB.8. At each voxel, the transmutation rate in atoms per cm3 was evaluated multiplying the microscopic transmutation cross-section, by the neutron flux, normalized by the number of neutrons emitted at the source, and by the atomic density per volume unit of the 98Mo sample.
Later, the volume transmutation rate of (Eq. (5)), was evaluated as follows: N 〈σ Z ðEÞϕðE; rÞ〉 ; Rv ¼ ρω A p A
43 98
Mo, defined as Rv
ð5Þ
where ρ is the sample density, NA is Avogadro's number, A is the sample atomic mass, ω is chemical fraction multiplied by the isotope fraction of the sample in the compound, σ Z ðEÞ is the neutron ! absorption cross-section and ϕðE; r Þ is the neutron flux at the ! position r in the moderator, both with energy dependence. The ! term σ Z ðEÞϕðE; r Þ per particle p is evaluated by MCNP code in a ! given position r from the moderator and integrated on energy spectrum in the interval between the thermal energy and 2.5 MeV. Afterwards, the geometric parameters and materials of the reflectors and moderator were adjusted in order to maximize the rate of 99Mo production achieving the present suitable HFTG setup. The transmutation rate was multiplied by Υ and by the 98Mo volume V and inserted into Eq. (6), in order to determine the radioisotope
Fig. 4. Electromagnetic results: (a) potential distribution and (b) deuteron beam paths.
44
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46
activity as function of irradiation time, as follows: AðtÞ ¼ YRv Vð1 e in which
λt
Þ
R1
ð6Þ
EI PE
λ is the decay constant of the radioisotope product.
4. Results NM 1
TE
4.1. Electromagnetic and Nuclear Evaluation An electric potential of 30 kV in the plasma electrode and 150 kV in the target was assumed. The electrodes geometries were adjusted in order to correct improprieties in deuterium trajectory, optimizing the energy and current of the beam at the target electrode. Electrons impinged upon the titanium target were not simulated. However, a collector in a ring shape around the target loaded with a slightly higher potential than the target potential is installed for the purpose of collecting electrons and negative ions produced in collisions of the accelerated deuterium beam at the target. The fraction of the negatively ionized deuterium molecules are collected by the ring and the fraction of positively charged deuterium molecules are incorporated into the target, increasing the density of the deuterons at the target and thereby the neutron yield estimated by Eq. (3). A map of equipotential surfaces in the vicinity of the electrodes is shown in Fig. 4(a), and the Fig. 4(b) presents the path of the deuteron beams. This electrode configuration allowed to achieve a deuteron current of 1 A at the targets, finding in on CST simulations. According to numerical integration in Eq. (3) the device reaches a yield of 1013 n s 1. Several arrangements of HFTG, based on distinct geometry, material and elemental composition were designed in MCNP-5 code in order to allow us achieving highest rate of isotope transmutation. Fig. 5 shows some essentials components of the device generated on MCNP 5 code and materials in Table 1. Neutron source was distributed on both target electrodes. Axial neutron flux at the moderator, revealing a flux in the order of 1012 n cm 3 s 1, as shown in Fig. 6. Neutron fluency was multiplied by the atomic density of 98Mo and by the (n,γ) cross-section, determining the 98Mo transmutation volume rate, Rv. The term ρωðNA =AÞ, in the Eq. (5) is equal to 4:3 10 3 at. cm 1 b 1. This yield was multiplied by γ and inserted in Eq. (6) in order to determine the 99Mo and 99mTc activities as a function of irradiation time at the central position of the sample receptacle.
MD, NM2
EI
R1 R2 SH
PE
Fig. 5. HFTG (a) internal and (b) general components: electric insulator (EI), plasma electrode (PE), target (TE), neutron moderator (NM1) and the mesh data (MD) in the neutron moderation site (NM2), internal reflector (R1), external reflector and radiation shield (SH).
Table 1 Some materials of HFTG. Volume
Description
Material
SH R1 EI PE TE R2 NM2 NM1
Radiation shield (SH) Reflector Insulator Plasma electrode Target Reflector Moderator Moderador
Borated polyethylene Chromel A Kaolinite Copper Copper/titanium/deuterium Lead–bismuth alloy Graphite Water
5. Discussion and conclusion It is considered that the HFTG space-charge is a dynamic system that will find a state-stable mode. As different from usual neutron generator which operates to a limited number of hours until its gascamera does not fill more the ions; the HFTG has a cold plasma in very low pressure, ionized by electromagnetic field, that will be feed by a deuterium gas inlet to keep the pressure, temperature and plasma density constant as possible in function of long time operation. The ions will be generated on plasma camera. The ion charges on the plasma are kept balanced, despite of large number of electrons and positive ions. Although some positive ions will be extracted by the electrode system toward the target; electrons from plasma and impinged from target will pollute the environment. However the influence of those electrons can be easily limited by collectors placed near target, with slight different potential. The HFTG simulation with the present configurations has demonstrated that such generator is able to elute 99mTc, after 150 h of operation, with activities up to 2 Ci, equivalent to a commercial
Fig. 6. Axial neutron flux at the moderator.
W.L. Araujo, T.P.R. Campos / Nuclear Instruments and Methods in Physics Research A 782 (2015) 40–46 99 Mo/99mTc transient generator (Fig. 7 and Fig. 8). However, the production of 99mTc on the secular generator, based on the HFTG technology, is constant along the time, producing 99Mo continuously in locus; whereas conventionally 99Mo/99mTc generators are often used for five days and replaced weekly. The technology of HFTG is compact, simpler when compared with a nuclear reactor or particle accelerator. The set up presented here is powered by electric potential difference in the order of 200 kV and a deuterium gas. When turned off, the device ceases
10.0
5
45
the radiation, equivalent to a conventional X-ray device. Tritium confined on the target electrode and radioactive nuclides on the metal structures of the HFTG shall be the contaminants present on this device, however all possible contaminant radionuclide would be confined into the device. Such contamination could be minimized by the choice of low neutron absorption materials to assemble the device, as polymers and zirconium alloys, preferable, or aluminum and titanium. The simulation findings have revealed the capacity of the HFTG to produce 99mTc from 98Mo activation with suitable activities for clinical use. HFTG may compete with the production in locus by commercial generators where there is need of loading an amount of radioactive precursor, said radioactive parent, previously produced in high-flux nuclear reactors. The management of 99Mo/99mTc generator is taken weekly since its transient activity decreases proportionality to the parent half-life. Therefore, the major feature of 99Mo/99mTc generator is the a transient regime presented by the balance of activities between 99Mo and 99mTc. Technology presented in the HFTG can continuously generate the parent, whose activity on the device becomes stable, and consequently the balance between the parent and daughter follows a secular regime. A more efficient radiological protection is found with a HFTG running with dd fusion reactions. However, dt reaction is also possible. Fig. 2 shows that he cross-section of dt interaction, at the deuteron kinetic energies of 100 keV, is about 250 times lager than the dd cross-section evaluated in the present investigations. In this case, a neutron yield with two orders of magnitude above those findings with a HFTG running dd reactions can be reached. Under strict conditions of radiological protection, HFTG based on dt fusion could be implemented in hospitals facilities, reducing the operation time for activation in proportion to the increase of the fusion crosssection. New theoretical investigations should be conducted on this matter. Electromagnetic and nuclear evaluations along with the present calculations highlighted the potential of the HFTG. We expect in the near future that the investigations will drive toward experimental level.
7.5
References 0 5.0
_5
2.5
_ 10 98
0
_5
0
5
10
Mo transmutation volume rate in (a) 3d graph and (b) transverse view.
Molybdenum-99 Technetium-99m Eluted Technetium-99m
A(t) (Ci)
Fig. 7.
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t Fig. 8. The
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Mo and
99m
Tc activities, in mCi, as a function of time.
46
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