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Feasibility of sealed D–T neutron generator as neutron source for liver BNCT and its beam shaping assembly
Applied Radiation and Isotopes 86 (2014) 1–6
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Feasibility of sealed D–T neutron generator as neutron source for liver BNCT and its beam shaping assembly Zheng Liu, Gang Li, Linmao Liu n Institute of Radiation Technology, Northeast Normal University, Jingyuestreet, Changchun, China
H I G H L I G H T S
The feasibility of sealed neutron generator as neutron source for liver BNCT. Using natural uranium and low enrichment uranium as neutron multiplier for D–T generator is examined. A beam shaping assembly is designed to optimize the output neutron beam. The output of the assembly can fulfill the beam port recommended quality parameters by IAEA.
art ic l e i nf o
a b s t r a c t
Article history: Received 15 September 2013 Received in revised form 26 December 2013 Accepted 26 December 2013 Available online 4 January 2014
This paper involves the feasibility of boron neutron capture therapy (BNCT) for liver tumor with four sealed neutron generators as neutron source. Two generators are placed on each side of the liver. The high energy of these emitted neutrons should be reduced by designing a beam shaping assembly (BSA) to make them useable for BNCT. However, the neutron flux decreases as neutrons pass through different materials of BSA. Therefore, it is essential to find ways to increase the neutron flux. In this paper, the feasibility of using low enrichment uranium as a neutron multiplier is investigated to increase the number of neutrons emitted from D-T neutron generators. The neutron spectrum related to our system has a proper epithermal flux, and the fast and thermal neutron fluxes comply with the IAEA recommended values. & 2014 Elsevier Ltd. All rights reserved.
Keywords: BNCT Sealed neutron generator Beam shaping assembly Liver metastases
1. Introduction Boron neutron capture therapy (BNCT) is a kind of physical therapy for deeply seated cancer tumors when surgery is impossible. In this therapy, compounds containing 10B, are injected into the tumor tissues. 10B is a stable isotope of boron with a large absorption cross section for thermal neutron as high as 3840 barn (Gibbons and Macklin, 1959). The reaction equations are as follows: 10
B þ n-4 Heð1:47 MeVÞ þ 7 Lið0:84 MeVÞ þ γð0:48 MeVÞ ð93:7%Þ
10
B þ n-4 Heð1:47 MeVÞ þ 7 Lið1:01 MeVÞ ð6:3%Þ
The two fragments α and 7Li particles have high Linear Energy Transfer (LET) and high Relative Biological Effectiveness (RBE). The average free paths for the fragments are 7.7 and 4.1 μm for α and 7 Li particles, respectively, almost in the same order of magnitude n
Corresponding author. Tel.: +86 13843000265. E-mail address: [email protected] (L. Liu).
0969-8043/$ - see front matter & 2014 Elsevier Ltd. All rights reserved. http://dx.doi.org/10.1016/j.apradiso.2013.12.031
as the cell dimensions (Cerullo et al., 2004). 10B concentration in the tumor tissue will be higher than that in the healthy tissue, thus these particles deposit their energy over the range of reaction region to destroy tumors as well as to reduce the damage to the normal tissues (Yamamoto et al., 2008). Although the 10B reacts with thermal neutron, the deeply seated tumor requires beams of epithermal neutrons (0.5 eV oEn o10 keV) which can penetrate deeply into tissues and thermalize into the proximity of the tumor tissue. The liver is the most common target of metastases from many primary tumors (Vitale et al., 1986). The success of the first extracorporeal application of BNCT performed on liver metastases in Pavia, Italy in 2001 (Zonta et al., 2006, 2009).The research reactor of type TRIGA Mark II at the University Mainz can be used for the extracorporeal treatment of organs (Hampel et al., 2009). BNCT of explanted organs is also being studied at RA-3 reactor, Argentina (Bortolussi et al., 2011). In Finland, FiR 1 TRIGA reactor with an epithermal neutron beam is especially designed for brain tumors BNCT (Serén et al., 2001; Auterinen et al., 2004) now has also been used to irradiate patients with liver tumors (Auterinen et al., 2004).
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Z. Liu et al. / Applied Radiation and Isotopes 86 (2014) 1–6
BNCT requires epithermal neutron flux, which can penetrate deeply into tissues and thermalize into thermal neutrons to react with 10B compounds, which are injected into the tumor tissues. Therefore, epithermal neutron flux is one of the most important characteristics of the neutron source for BNCT. The flux must be high enough so that the therapy procedures could be given in a reasonable time. At present, a nuclear reactor is the only neutron source, which is capable of providing the required high epithermal neutron flux to treat liver tumors (Tahara et al., 2006). Considering the huge cost of nuclear reactor and the public acceptability, a neutron source that can be used in the hospital and with lower cost should be taken into account. The neutron generator could be an excellent choice because of the high neutron flux, the lower cost, the safety, and the public acceptability (Koivunoro et al., 2004; Kononov et al., 2004; Ghassoun et al., 2009). In this paper, the feasibility of neutron generation for BNCT using 3H(d, n)4He (D–T) reaction has been investigated. Compared with other kinds of neutron sources, the sealed tube generator is safer and more economical. The neutron flux is lower than the nuclear reactor; however, owing to the considerable size of the sealed tube generator, there is enough room for accommodating more than one tube, for instance 4–6. So the sealed tube generator which has a high neutron yield is likely to provide a feasible option (Montagnini et al., 2002). The cost of the whole instrument is considered about two million dollars. Furthermore, parameter studies (Wortmann and Knorr, 2012) have shown that rotating the liver 1801 once half way through the irradiation time results a better dose distribution (Wortmann and Knorr, 2012). We can take full advantage of the smaller size of the sealed neutron generator to place the generators on both sides of the liver to achieve the same effect as rotating the liver 1801. In our work, we aimed to produce an epithermal neutron flux, which is high enough for BNCT with four sealed tube generators to irradiate both sides of the liver at the same time. The energy of the proposed neutrons for this source is 14.1 MeV, as these neutrons cannot be used directly in BNCT for the high energy, it is necessary to moderate them into epithermal energy range. For this purpose, neutrons must pass through a system that contains different materials. Such a system is called beam shaping assembly (BSA). The major components of BSA are the moderator, the reflector and the gamma shielding (Montagnini et al., 2002). The aim of designing such a BSA is to moderate the high-energy neutrons and to remove fast and thermal neutrons, as well as gamma contaminations to fulfill the BNCT in air beam port recommended quality parameters by IEAE in Table 1. In this table, epithermal neutron flux, thermal neutron flux, fast neutron dose and gamma ray dose are denoted by ϕepi, ϕth, Dfast and Dγ, respectively. A Monte Carlo simulation was used in our study to find out the appropriate BSA option. A feasible choice of BSA materials and geometries can fulfill all the BNCT requirements. However, the neutron flux decreases
Table 1 BNCT in air beam port recommended quality parameters and corresponding neutron beam energy limit by IAEA (IAEA-TECDDOC-1223, 2001). BNCT beam port parameters
Limit
ϕepi[n/cm2 s] ϕepi/ϕth Dfast/ϕepi [Gy cm2/n] Dγ/ϕepi[Gy cm2/n] Fast energy group Epithermal energy group Thermal energy group
45 108 420 o 2 10 13 o 2 10 13 E 410 keV 0.5 eV oE o 10 keV E o0.5 eV
Fig. 1. Proposed BNCT facility on sealed neutron generator and perspective beam shaping assembly. Phantom is placed in the middle of the facility to be irradiated from both sides. The BSA is cylindrical and the symbols of the numbers are as follows: (1) Pb reflector, (2) 20 cm AlF3 moderator, (3) 22 cm Fluental moderator, (4) phantom, (5) neutron generator, (6) 2 cm hemisphere natural uranium, (7) 17 cm 235U and (CH2)n moderator, (8) 5 cm Fe.
when neutrons pass through different components of BSA. To fulfill the in air beam port recommended quality parameters, using uranium as neutron multiplier was adopted to increase the intensity of the epithermal neutron flux. 10 B concentrations in tumor tissues and normal tissues are important to the efficacy of BNCT. The success of BNCT bases on maximizing the neutron dose to tumor tissues and minimizing the neutron dose to healthy tissues at the same time. Furthermore, 10B concentration in tumor tissues must be high enough to allow a sufficient number of 10B(n, α)7Li reactions to occur to produce enough α and 7Li particles. Treatment using boronophenylalanine (BPA) or sodium mercaptoundecahydrododecaborane (BSH) to transport 10B has been performed in the USA, Europe, Argentina, and Japan. The results show a potential therapeutic advantage for this method and leave room for improvement (Coderre and Morris, 1999; Garabalino et al., 2010) (Fig. 1).
2. Methods The sealed generator tube is a neutron generator, which uses a microwave ion source to produce deuterium and tritium ions from plasma. The microwave ion source yields a high fraction of monoatomic ion species in the ion beam. The ions are accelerated to impinge the beam on a target where 14.1 MeV D–T neutrons are generated by fusion reactions. The yield of the neutrons can be controlled under the automatic control system and the yield is stable (with a variation o2%). Previous work was to develop high yield sealed generator tube for BNCT. In this paper, we mainly focus on the feasibility of a sealed generator tube with yield of 1 1011 n/s to use as neutron source for liver tumor BNCT. We used four neutron tubes simultaneously to form a neutron yield of 4 1011 n/s and placed two tubes on each side of the liver, taking full advantage of the smaller size of the neutron tube to decrease the therapy procedure time. As the energy of the neutron from this source is 14.1 MeV, it is assumed that the neutrons are emitted isotropically and monoenergetically from a sphere of 2 cm radius. As these high energy neutrons cannot be used directly in liver tumor BNCT, it is essential to moderate them into epithermal energy range. Neutrons must pass through a BSA system that contains different moderator materials to provide an appropriate neutron energy spectrum to fulfill the BNCT beam port recommended quality. In addition, the neutrons need to be guided to the beam aperture direction with the reflector materials. In this study, lead was
Z. Liu et al. / Applied Radiation and Isotopes 86 (2014) 1–6
chosen as a reflector material and the radius of the beam aperture was 10 cm considering the size of human liver. Monte Carlo method was used in our study to find the suitable materials for the moderators for the emitted neutrons. However, these moderator materials often conflict with the intensity of the epithermal neutron output of the beam port. So it was crucial to find ways to increase the neutron flux even if four neutron tubes were used (Martin and Abrahantes, 2004).
3. Materials 3.1. Multiplier As mentioned above, the neutron number will decrease when neutrons pass through different materials of BSA, so the proper multiplier materials can be used to increase the number of neutrons, which are emitted from neutron source. First, natural uranium was used as a neutron multiplier to increase the number of neutrons via fission reaction (Eskandari and Kashian, 2009). The large cross section of 238U for (n, 2n) reaction is used to compensate for the neutron losses by absorption during moderation. It shows that the number of neutron will be increased by 2.8 times reaches its maximum when neutrons pass through a natural uranium sphere with a radius of 14 cm (Rasouli et al., 2012). However, we want to use four generator tubes at the same time to form a large neutron yield, there is no room to accommodate the four tubes each one is surrounded by a natural uranium sphere with a radius of 14 cm. To solve the problem, we surrounded each neutron source with a 2 cm thick natural uranium hemisphere and the number of the neutron increased by 1.7 times. Furthermore, using uranium as neutron multiplier could decrease ϕepi/ϕfast and ϕepi/ϕth to meet the free beam parameters more closely. Therefore, the BSA design is necessary to achieve an optimal result. The result of Fig. 2 demonstrates the neutron energy spectrum when neutrons passed through a 2 cm thick natural uranium hemisphere. In our simulation, the number of the neutrons is considered so large that the average statistical uncertainty is less than 1%. The energy of each neutron is 14.1 MeV when emitted isotropically and monoenergetically from a sphere with a 2 cm radius. Fig. 2 demonstrates that the energy decreases after passing through the 2 cm thick natural uranium hemisphere, but there are still too many fast neutrons. Then, we adopted a low enrichment of 235U as neutron multiplier. Due to the large fission cross section of 235U for low energy neutrons, we chose an appropriate material to moderate the neutrons that came from the 2 cm thick natural
Fig. 2. Neutron flux versus 2 cm thicknesses of natural uranium hemisphere.
3
uranium hemispheres. Fe has a high inelastic scattering cross section above 860 keV (Verbeke et al., 2001). Thus, it can decrease the fast neutron flux in the range of 1–14.1 MeV (Rasouli et al., 2012). We tested different thicknesses of Fe moderator to find the best thickness corresponding to the maximum of neutron flux. Fe cylinder with a radius of 10 cm was placed near the natural uranium hemisphere. Fig. 3 shows the ϕepi/ϕfast corresponding to Fe as moderator versus its different thickness. Fig. 4 shows the epithermal neutron flux versus different thickness of Fe. The average statistical uncertainty is 0.4%. As Fig. 3 shows, the ϕepi/ ϕfast value goes up with the increase of Fe thickness, however, the epithermal flux decreases. The proper thickness of Fe corresponds to the larger value of ϕepi/ϕfast and a peak value when the thickness of Fe is 5 cm. A 5 cm thickness of Fe corresponded to the largest value of ϕepi/ϕfast and a peak value of neutron flux. Therefore, a Fe cylinder with a radius of 10 cm and 5 cm thickness was chosen. Neutrons passing through the Fe cylinder with a lower energy enter the multiplier facility. The fuel rod is a uranium dioxide cylinder with a height of 1 cm and a radius of 0.5 cm. We designed a cylinder with low enrichment uranium and (CH2)n with a radius of 10 cm and a height of 17 cm as the simulation multiplier facility where the (CH2)n acted as solid moderator. We used a 25% enrichment of 235U as multiplier with a total quality of 5.75 kg, and the quality of 235U on each side of the BSA was less than 1.5 kg. The neutron flux increases as the fission reactions occur. Due to the low core power, no refueling should be necessary throughout the whole life of the facility or at least for a period of 5–10 years (Montagnini et al., 2002). Fig. 5
Fig. 3. The ratio of epithermal neutron flux to fast neutron flux for different thicknesses of Fe as fast neutron filter.
Fig. 4. Epithermal neutron flux for different thicknesses of Fe as fast neutron filter.
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Fig. 5. Neutron spectrum at the beam port with and without using neutron multiplier.
Fig. 6. The ratio of epithermal neutron flux to fast neutron flux for different thicknesses of moderators.
demonstrates the neutron fluxes with or without multiplier at the beam port. 3.2. Moderator The energy of the neutrons from the multiplier facility is still too high for BNCT. Therefore, the neutrons still need to be moderated to epithermal energy range. These moderator materials can be selected because they must have a low scattering cross section at the desirable epithermal energy, but a high scattering cross section at higher energy. In addition, it must be considered that the moderator materials with high absorption cross section absorb too many neutrons before they reach to the patient, which leads to high gamma ray contamination. Materials containing fluorine are presenting better performance in terms of neutron accumulation in epithermal energy range during the slowing down from source energies. If ∑Sf/epi is fast to epithermal slowing down macroscopic cross section and ∑γ is the macroscopic absorption cross section, materials containing fluorine have a maximum fast to epithermal over macroscopic absorption cross sections (∑Sf/epi/∑γ). Materials containing fluorine have large ratio of ∑Sf/epi/∑γ and for this reason are tested as moderators. And aluminum is also an option to function as an appropriate moderator (Tanaka et al., 2006). We tested some materials containing fluorine or aluminum such as Al, Al2O3, AlF3, (CF2)n and Fluental to function as neutron moderators. Fluental consists of 69% AlF3, 30% Al, and 1% of LiF (by weight). The moderator effect was also strongly dependent on the geometries of the materials. The geometries of the materials were cylinders with the same radius of 35 cm and we mainly focused on the thickness of the moderator. The moderator was arranged after the simulation multiplier facility, surrounded by lead as reflector. Figs. 6–9 show the epithermal neutron flux and ϕepi/ϕfast for different materials chosen as the moderators versus their thicknesses. It comes from the fact that fluorine has a strong resonance structure in elastic scattering cross section for fast neutrons; moreover, fluorine has a high cross section of inelastic neutron scattering with low levels (197 keV), and also the element has a small atomic mass (Kononov et al., 2004). So due to the fact and Figs. 6 and 7, 20 cm of AlF3 was selected as the first moderator. However, the ϕepi/ϕfast does not satisfy the IAEA BNCT in air beam port recommended quality parameters; we need to choose proper material to serve as second moderator. To increase ϕepi/ϕfast, a second moderator can be used to reduce the number of fast neutrons. After selecting 20 cm of AlF3 as first moderator, Figs. 8 and 9 demonstrate the epithermal neutron flux
Fig. 7. Epithermal neutron flux for different thicknesses and compositions of moderators.
Fig. 8. The ratio of epithermal neutron flux to fast neutron flux where. 20 cm AlF3 is selected as the first moderator.
and ϕepi/ϕfast respectively when different materials act as second moderators. Table 2 gives the in air beam port output of four different materials at proper thicknesses. Limited by the yield of the sealed neutron generator, the Dfast/ϕepi of our assembly is higher than the IAEA in air beam port quality parameters by IAEA; however, ϕepi/ϕth and Dγ/ϕepi complied with the recommended parameters, especially epithermal neutron flux, the most important parameter.
Z. Liu et al. / Applied Radiation and Isotopes 86 (2014) 1–6
Fig. 9. Epithermal neutron flux versus the second moderator thickness where. 20 cm AlF3 is selected as the first moderator.
Table 2 Epithermal neutron flux versus different second moderator where 20 cm AlF3 is selected as the first moderator. Moderator
ϕepi [n/cm2s]
ϕepi/ϕth
Dfast/ϕepi [Gy cm2/n]
Dγ/ϕepi [Gy cm2/n]
Fluental (22 cm) AlF3 (20 cm) CF2(15 cm) Al2O3(20 cm)
5.48 108 4.88 108 5.28 108 4.20 108
21.17 19.27 18.46 19.44
7.66 10 13 6.89 10 13 18.56 10 13 12.94 10 13
0.58 10 13 1.42 10 13 2.01 10 13 0.75 10 13
5
Fig. 10. Epithermal neutron flux for different thickness of Pb as reflector.
If we merely want to increase the number of epithermal neutrons regardless of the increase of fast neutrons, we can choose Pb, which is thick enough. As the Pb was used as reflector, Fig. 10 shows the epithermal neutron flux corresponding to different thicknesses of Pb. When the thickness reaches 65 cm, the increasing of the epithermal flux is not evident. Therefore, a thickness of 65 cm Pb was used to surround the whole BSA and worked as dose shielding to the surrounding environment.
4. Results and discussion Table 3 Albedoes of some reflectors [15]. Material
Albedoes
Slab thickness 5 (cm)
10 (cm)
100 (cm)
Ni
βf-f βepi-epi
0.400 0.744
0.510 0.759
0.676 0.760
V
βf-f βepi-epi
0.367 0.550
0.500 0.643
0.691 0.682
Bi
βf-f βepi-epi
0.312 0.413
0.486 0.588
0.873 0.895
Pb
βf-f βepi-epi
0.352 0.487
0.521 0.652
0.862 0.867
The number of epithermal neutrons in Table 2 is a little lower than that of Fig. 9. It is because a layer of thin 6LiF, which has an exceedingly high absorption, was placed along the moderator to reduce the number of the thermal neutrons. Cd, which has a large absorption cross section for thermal neutrons, was not chosen because Cd produces a high energy (7.8 MeV) of capture gamma ray, which is difficult to control, and cadmium oxide represents a health hazard. Bi is proper for shielding gamma rays, thus, a layer of 2 cm Bi was placed at the beam port to reduce the γ contamination. 3.3. Reflector The thermalized neutrons need to be guided to the beam port with the reflector material (Koivunoro et al., 2004). A proper reflector can contribute to the increase of the epithermal neutron flux. An appropriate reflector material is supposed to have a high elastic scattering cross section and a low absorption cross section to reenter the neutrons to the moderator. Table 3 presents the albedoes of several reflector materials in different thickness.
In our BSA, when the second moderator was chosen as a 22 cm thickness of Fluental, the epithermal neutron flux of the beam reached 5.48 108 n cm 2 s 1, and ϕepi/ϕth and Dγ/ϕepi comply with the recommended parameters well. However, the Dfast/ϕepi reached 7.66 10 13 Gy cm2/n, which was higher than the in air port parameter recommended for BNCT which is less than 2 10 13 Gy cm2/n. If we could choose a 42 cm thickness of Fluental as the second moderator, the Dfast/ϕepi reaches 2.18 10 13 Gy cm2/n, almost satisfies the recommended parameters. However, this results an epithermal neutrons flux to 3 108 n cm 2 s 1. As motioned above, a limited epithermal flux requires a long treatment time. To solve the problem, increasing the yield of sealed neutron generator seems to be a feasible option. If we could increase the yield of our generator to 2 1011 n/s, then the problem could be solved.
5. Conclusion We discuss the possibility of using four sealed neutron generators as neutron source for BNCT for liver tumor. Four generators are used at the same time to form a yield of 4 1011 n/s and two generators are placed on each side of the liver. A beam shaping assembly for BNCT is designed based on the use of low enrichment uranium as multiplier system for D–T neutron source. According to our results, a 20 cm thickness of AlF3 as the first moderator and a 22 cm thickness of Fluental as the second moderator is the best option for the moderator to achieve a proper epithermal neutron flux. The study of the beam quality shows that using a layer of thin 6 LiF filter to remove thermal neutrons and Bi to reduce gamma contamination is necessary to meet the free beam port parameters by IAEA. The epithermal neutron flux of our facility at the beam port is 5.48 108 n/cm2 s and the values of ϕepi/ϕth and Dγ/ϕepi
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fulfill the free beam port parameters. The value of Dfast/ϕepi does not fulfill the parameters because of the limited neutron yield.
References Auterinen, I., Serén, T., et al., 2004. Measurement of free beam neutron spectra at eight BNCT facilities worldwide. Appl. Radiat. Isot. 61 (5), 1021–1026. Bortolussi, S., Pinto, J.M., et al., 2011. Simulation of the neutron flux in the irradiation facility at RA-3 reactor. Appl. Radiat. Isot. 69 (12), 1924–1927. Cerullo, N., Esposito, J., et al., 2004. Spectrum shaping assessment of acceleratorbased fusion neutron sources to be used in BNCT treatment. Nucl. Instrum. Methods Phys. Res., Sect. B 213, 641–645. Coderre, J.A., Morris, G.M., 1999. The radiation biology of boron neutron capture therapy. Radiat. Res. 151 (1), 1–18. Eskandari, M.R., Kashian, S., 2009. Design of moderator and multiplier systems for D–T neutron source in BNCT. Ann. Nucl. Energy 36 (8), 1100–1102. Garabalino, M.A., Hughes, A. Monti, et al., 2010. Boron neutron capture therapy (BNCT) for the treatment of liver metastases: biodistribution studies of boron compounds in an experimental model. Radiat. Environ. Biophys. 50 (1), 199–207. Ghassoun, J., Chkillou, B., et al., 2009. Spatial and spectral characteristics of a compact system neutron beam designed for BNCT facility. Appl. Radiat. Isot. 67 (4), 560–564. Gibbons, J., Macklin, R., 1959. Total neutron yields from light elements under proton and alpha bombardment. Phys. Rev. 114 (2), 571–580. Hampel, G., Wortmann, B., et al., 2009. Irradiation facility at the TRIGA Mainz for treatment of liver metastases. Appl. Radiat. Isot. 67 (7-8), S238–S241. IAEA-TECDOC-1223, 2001. Current Status of Neutron Capture Therapy. International Atomic Energy Agency. IAEA, Vienna, Austria. Koivunoro, H., Bleuel, D.L., et al., 2004. BNCT dose distribution in liver with epithermal D–D and D–T fusion-based neutron beams. Appl. Radiat. Isot. 61 (5), 853–859.
Kononov, O.E., Kononov, V.N., et al., 2004. Optimization of an accelerator-based epithermal neutron source for neutron capture therapy. Appl. Radiat. Isot. 61 (5), 1009–1013. Martin, G., Abrahantes, A., 2004. A conceptual design of a beam-shaping assembly for boron neutron capture therapy based on deuterium–tritium neutron generators. Med. Phys. 31 (5), 1116–1122. Montagnini, B., Cerullo, N., et al., 2002. Spectrum shaping of accelerator-based neutron beams for BNCT. Nucl. Instrum. Methods Phys. Res., Sect. A 476 (1–2), 90–98. Rasouli, F.S., Masoudi, S. Farhad, et al., 2012. Design of a model for BSA to meet free beam parameters for BNCT based on multiplier system for D–T neutron source. Ann. Nucl. Energy 39 (1), 18–25. Serén, T., Auterinen I., et al. (2001). Neutron Beam Characterization at the Finnish BNCT Facility—Measurements and Calculations. Tahara, Y., Oda, Y., et al., 2006. Engineering design of a spallation reaction-based neutron generator for boron neutron capture therapy. J. Nucl. Sci. Technol. 43 (1), 9–19. Tanaka, K., Kobayashi, T., et al., 2006. Characterization of moderator assembly dimension for accelerator boron neutron capture therapy of brain tumors using 7Li(p,n) neutrons at proton energy of 2.5 MeV. Med. Phys. 33 (6), 1688–1694. Verbeke, J.M., Chen, A.S., et al., 2001. Optimization of beam-shaping assemblies for BNCS using the high-energy neutron sources D–D and D–T. Nucl. Technol. 134 (3), 278–293. Vitale, G.C., Heuser, L.S., et al., 1986. Malignant tumors of the liver. Surg. Clin. North Am. 66 (4), 723–741. Wortmann, B., Knorr, J., 2012. Basic requirements and parameter optimization for boron neutron capture therapy of extracorporeal irradiated and autotransplanted organs. Appl. Radiat. Isot. 70 (8), 1709–1717. Yamamoto, T., Nakai, K., et al., 2008. Boron neutron capture therapy for glioblastoma. Cancer Lett. 262 (2), 143–152. Zonta, A., Pinelli, T., et al., 2009. Extra-corporeal liver BNCT for the treatment of diffuse metastases: what was learned and what is still to be learned. Appl. Radiat. Isot. 67 (7-8), S67–S75. Zonta, A., Prati, U., et al., 2006. Clinical lessons from the first applications of BNCT on unresectable liver metastases. J. Phys. Conf. Ser. 41, 484–495.
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Feasibility of sealed D–T neutron generator as neutron source for liver BNCT and its beam shaping assembly Zheng Liu, Gang Li, Linmao Liu n Institute of Radiation Technology, Northeast Normal University, Jingyuestreet, Changchun, China
H I G H L I G H T S
The feasibility of sealed neutron generator as neutron source for liver BNCT. Using natural uranium and low enrichment uranium as neutron multiplier for D–T generator is examined. A beam shaping assembly is designed to optimize the output neutron beam. The output of the assembly can fulfill the beam port recommended quality parameters by IAEA.
art ic l e i nf o
a b s t r a c t
Article history: Received 15 September 2013 Received in revised form 26 December 2013 Accepted 26 December 2013 Available online 4 January 2014
This paper involves the feasibility of boron neutron capture therapy (BNCT) for liver tumor with four sealed neutron generators as neutron source. Two generators are placed on each side of the liver. The high energy of these emitted neutrons should be reduced by designing a beam shaping assembly (BSA) to make them useable for BNCT. However, the neutron flux decreases as neutrons pass through different materials of BSA. Therefore, it is essential to find ways to increase the neutron flux. In this paper, the feasibility of using low enrichment uranium as a neutron multiplier is investigated to increase the number of neutrons emitted from D-T neutron generators. The neutron spectrum related to our system has a proper epithermal flux, and the fast and thermal neutron fluxes comply with the IAEA recommended values. & 2014 Elsevier Ltd. All rights reserved.
Keywords: BNCT Sealed neutron generator Beam shaping assembly Liver metastases
1. Introduction Boron neutron capture therapy (BNCT) is a kind of physical therapy for deeply seated cancer tumors when surgery is impossible. In this therapy, compounds containing 10B, are injected into the tumor tissues. 10B is a stable isotope of boron with a large absorption cross section for thermal neutron as high as 3840 barn (Gibbons and Macklin, 1959). The reaction equations are as follows: 10
B þ n-4 Heð1:47 MeVÞ þ 7 Lið0:84 MeVÞ þ γð0:48 MeVÞ ð93:7%Þ
10
B þ n-4 Heð1:47 MeVÞ þ 7 Lið1:01 MeVÞ ð6:3%Þ
The two fragments α and 7Li particles have high Linear Energy Transfer (LET) and high Relative Biological Effectiveness (RBE). The average free paths for the fragments are 7.7 and 4.1 μm for α and 7 Li particles, respectively, almost in the same order of magnitude n
Corresponding author. Tel.: +86 13843000265. E-mail address: [email protected] (L. Liu).
0969-8043/$ - see front matter & 2014 Elsevier Ltd. All rights reserved. http://dx.doi.org/10.1016/j.apradiso.2013.12.031
as the cell dimensions (Cerullo et al., 2004). 10B concentration in the tumor tissue will be higher than that in the healthy tissue, thus these particles deposit their energy over the range of reaction region to destroy tumors as well as to reduce the damage to the normal tissues (Yamamoto et al., 2008). Although the 10B reacts with thermal neutron, the deeply seated tumor requires beams of epithermal neutrons (0.5 eV oEn o10 keV) which can penetrate deeply into tissues and thermalize into the proximity of the tumor tissue. The liver is the most common target of metastases from many primary tumors (Vitale et al., 1986). The success of the first extracorporeal application of BNCT performed on liver metastases in Pavia, Italy in 2001 (Zonta et al., 2006, 2009).The research reactor of type TRIGA Mark II at the University Mainz can be used for the extracorporeal treatment of organs (Hampel et al., 2009). BNCT of explanted organs is also being studied at RA-3 reactor, Argentina (Bortolussi et al., 2011). In Finland, FiR 1 TRIGA reactor with an epithermal neutron beam is especially designed for brain tumors BNCT (Serén et al., 2001; Auterinen et al., 2004) now has also been used to irradiate patients with liver tumors (Auterinen et al., 2004).
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BNCT requires epithermal neutron flux, which can penetrate deeply into tissues and thermalize into thermal neutrons to react with 10B compounds, which are injected into the tumor tissues. Therefore, epithermal neutron flux is one of the most important characteristics of the neutron source for BNCT. The flux must be high enough so that the therapy procedures could be given in a reasonable time. At present, a nuclear reactor is the only neutron source, which is capable of providing the required high epithermal neutron flux to treat liver tumors (Tahara et al., 2006). Considering the huge cost of nuclear reactor and the public acceptability, a neutron source that can be used in the hospital and with lower cost should be taken into account. The neutron generator could be an excellent choice because of the high neutron flux, the lower cost, the safety, and the public acceptability (Koivunoro et al., 2004; Kononov et al., 2004; Ghassoun et al., 2009). In this paper, the feasibility of neutron generation for BNCT using 3H(d, n)4He (D–T) reaction has been investigated. Compared with other kinds of neutron sources, the sealed tube generator is safer and more economical. The neutron flux is lower than the nuclear reactor; however, owing to the considerable size of the sealed tube generator, there is enough room for accommodating more than one tube, for instance 4–6. So the sealed tube generator which has a high neutron yield is likely to provide a feasible option (Montagnini et al., 2002). The cost of the whole instrument is considered about two million dollars. Furthermore, parameter studies (Wortmann and Knorr, 2012) have shown that rotating the liver 1801 once half way through the irradiation time results a better dose distribution (Wortmann and Knorr, 2012). We can take full advantage of the smaller size of the sealed neutron generator to place the generators on both sides of the liver to achieve the same effect as rotating the liver 1801. In our work, we aimed to produce an epithermal neutron flux, which is high enough for BNCT with four sealed tube generators to irradiate both sides of the liver at the same time. The energy of the proposed neutrons for this source is 14.1 MeV, as these neutrons cannot be used directly in BNCT for the high energy, it is necessary to moderate them into epithermal energy range. For this purpose, neutrons must pass through a system that contains different materials. Such a system is called beam shaping assembly (BSA). The major components of BSA are the moderator, the reflector and the gamma shielding (Montagnini et al., 2002). The aim of designing such a BSA is to moderate the high-energy neutrons and to remove fast and thermal neutrons, as well as gamma contaminations to fulfill the BNCT in air beam port recommended quality parameters by IEAE in Table 1. In this table, epithermal neutron flux, thermal neutron flux, fast neutron dose and gamma ray dose are denoted by ϕepi, ϕth, Dfast and Dγ, respectively. A Monte Carlo simulation was used in our study to find out the appropriate BSA option. A feasible choice of BSA materials and geometries can fulfill all the BNCT requirements. However, the neutron flux decreases
Table 1 BNCT in air beam port recommended quality parameters and corresponding neutron beam energy limit by IAEA (IAEA-TECDDOC-1223, 2001). BNCT beam port parameters
Limit
ϕepi[n/cm2 s] ϕepi/ϕth Dfast/ϕepi [Gy cm2/n] Dγ/ϕepi[Gy cm2/n] Fast energy group Epithermal energy group Thermal energy group
45 108 420 o 2 10 13 o 2 10 13 E 410 keV 0.5 eV oE o 10 keV E o0.5 eV
Fig. 1. Proposed BNCT facility on sealed neutron generator and perspective beam shaping assembly. Phantom is placed in the middle of the facility to be irradiated from both sides. The BSA is cylindrical and the symbols of the numbers are as follows: (1) Pb reflector, (2) 20 cm AlF3 moderator, (3) 22 cm Fluental moderator, (4) phantom, (5) neutron generator, (6) 2 cm hemisphere natural uranium, (7) 17 cm 235U and (CH2)n moderator, (8) 5 cm Fe.
when neutrons pass through different components of BSA. To fulfill the in air beam port recommended quality parameters, using uranium as neutron multiplier was adopted to increase the intensity of the epithermal neutron flux. 10 B concentrations in tumor tissues and normal tissues are important to the efficacy of BNCT. The success of BNCT bases on maximizing the neutron dose to tumor tissues and minimizing the neutron dose to healthy tissues at the same time. Furthermore, 10B concentration in tumor tissues must be high enough to allow a sufficient number of 10B(n, α)7Li reactions to occur to produce enough α and 7Li particles. Treatment using boronophenylalanine (BPA) or sodium mercaptoundecahydrododecaborane (BSH) to transport 10B has been performed in the USA, Europe, Argentina, and Japan. The results show a potential therapeutic advantage for this method and leave room for improvement (Coderre and Morris, 1999; Garabalino et al., 2010) (Fig. 1).
2. Methods The sealed generator tube is a neutron generator, which uses a microwave ion source to produce deuterium and tritium ions from plasma. The microwave ion source yields a high fraction of monoatomic ion species in the ion beam. The ions are accelerated to impinge the beam on a target where 14.1 MeV D–T neutrons are generated by fusion reactions. The yield of the neutrons can be controlled under the automatic control system and the yield is stable (with a variation o2%). Previous work was to develop high yield sealed generator tube for BNCT. In this paper, we mainly focus on the feasibility of a sealed generator tube with yield of 1 1011 n/s to use as neutron source for liver tumor BNCT. We used four neutron tubes simultaneously to form a neutron yield of 4 1011 n/s and placed two tubes on each side of the liver, taking full advantage of the smaller size of the neutron tube to decrease the therapy procedure time. As the energy of the neutron from this source is 14.1 MeV, it is assumed that the neutrons are emitted isotropically and monoenergetically from a sphere of 2 cm radius. As these high energy neutrons cannot be used directly in liver tumor BNCT, it is essential to moderate them into epithermal energy range. Neutrons must pass through a BSA system that contains different moderator materials to provide an appropriate neutron energy spectrum to fulfill the BNCT beam port recommended quality. In addition, the neutrons need to be guided to the beam aperture direction with the reflector materials. In this study, lead was
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chosen as a reflector material and the radius of the beam aperture was 10 cm considering the size of human liver. Monte Carlo method was used in our study to find the suitable materials for the moderators for the emitted neutrons. However, these moderator materials often conflict with the intensity of the epithermal neutron output of the beam port. So it was crucial to find ways to increase the neutron flux even if four neutron tubes were used (Martin and Abrahantes, 2004).
3. Materials 3.1. Multiplier As mentioned above, the neutron number will decrease when neutrons pass through different materials of BSA, so the proper multiplier materials can be used to increase the number of neutrons, which are emitted from neutron source. First, natural uranium was used as a neutron multiplier to increase the number of neutrons via fission reaction (Eskandari and Kashian, 2009). The large cross section of 238U for (n, 2n) reaction is used to compensate for the neutron losses by absorption during moderation. It shows that the number of neutron will be increased by 2.8 times reaches its maximum when neutrons pass through a natural uranium sphere with a radius of 14 cm (Rasouli et al., 2012). However, we want to use four generator tubes at the same time to form a large neutron yield, there is no room to accommodate the four tubes each one is surrounded by a natural uranium sphere with a radius of 14 cm. To solve the problem, we surrounded each neutron source with a 2 cm thick natural uranium hemisphere and the number of the neutron increased by 1.7 times. Furthermore, using uranium as neutron multiplier could decrease ϕepi/ϕfast and ϕepi/ϕth to meet the free beam parameters more closely. Therefore, the BSA design is necessary to achieve an optimal result. The result of Fig. 2 demonstrates the neutron energy spectrum when neutrons passed through a 2 cm thick natural uranium hemisphere. In our simulation, the number of the neutrons is considered so large that the average statistical uncertainty is less than 1%. The energy of each neutron is 14.1 MeV when emitted isotropically and monoenergetically from a sphere with a 2 cm radius. Fig. 2 demonstrates that the energy decreases after passing through the 2 cm thick natural uranium hemisphere, but there are still too many fast neutrons. Then, we adopted a low enrichment of 235U as neutron multiplier. Due to the large fission cross section of 235U for low energy neutrons, we chose an appropriate material to moderate the neutrons that came from the 2 cm thick natural
Fig. 2. Neutron flux versus 2 cm thicknesses of natural uranium hemisphere.
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uranium hemispheres. Fe has a high inelastic scattering cross section above 860 keV (Verbeke et al., 2001). Thus, it can decrease the fast neutron flux in the range of 1–14.1 MeV (Rasouli et al., 2012). We tested different thicknesses of Fe moderator to find the best thickness corresponding to the maximum of neutron flux. Fe cylinder with a radius of 10 cm was placed near the natural uranium hemisphere. Fig. 3 shows the ϕepi/ϕfast corresponding to Fe as moderator versus its different thickness. Fig. 4 shows the epithermal neutron flux versus different thickness of Fe. The average statistical uncertainty is 0.4%. As Fig. 3 shows, the ϕepi/ ϕfast value goes up with the increase of Fe thickness, however, the epithermal flux decreases. The proper thickness of Fe corresponds to the larger value of ϕepi/ϕfast and a peak value when the thickness of Fe is 5 cm. A 5 cm thickness of Fe corresponded to the largest value of ϕepi/ϕfast and a peak value of neutron flux. Therefore, a Fe cylinder with a radius of 10 cm and 5 cm thickness was chosen. Neutrons passing through the Fe cylinder with a lower energy enter the multiplier facility. The fuel rod is a uranium dioxide cylinder with a height of 1 cm and a radius of 0.5 cm. We designed a cylinder with low enrichment uranium and (CH2)n with a radius of 10 cm and a height of 17 cm as the simulation multiplier facility where the (CH2)n acted as solid moderator. We used a 25% enrichment of 235U as multiplier with a total quality of 5.75 kg, and the quality of 235U on each side of the BSA was less than 1.5 kg. The neutron flux increases as the fission reactions occur. Due to the low core power, no refueling should be necessary throughout the whole life of the facility or at least for a period of 5–10 years (Montagnini et al., 2002). Fig. 5
Fig. 3. The ratio of epithermal neutron flux to fast neutron flux for different thicknesses of Fe as fast neutron filter.
Fig. 4. Epithermal neutron flux for different thicknesses of Fe as fast neutron filter.
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Fig. 5. Neutron spectrum at the beam port with and without using neutron multiplier.
Fig. 6. The ratio of epithermal neutron flux to fast neutron flux for different thicknesses of moderators.
demonstrates the neutron fluxes with or without multiplier at the beam port. 3.2. Moderator The energy of the neutrons from the multiplier facility is still too high for BNCT. Therefore, the neutrons still need to be moderated to epithermal energy range. These moderator materials can be selected because they must have a low scattering cross section at the desirable epithermal energy, but a high scattering cross section at higher energy. In addition, it must be considered that the moderator materials with high absorption cross section absorb too many neutrons before they reach to the patient, which leads to high gamma ray contamination. Materials containing fluorine are presenting better performance in terms of neutron accumulation in epithermal energy range during the slowing down from source energies. If ∑Sf/epi is fast to epithermal slowing down macroscopic cross section and ∑γ is the macroscopic absorption cross section, materials containing fluorine have a maximum fast to epithermal over macroscopic absorption cross sections (∑Sf/epi/∑γ). Materials containing fluorine have large ratio of ∑Sf/epi/∑γ and for this reason are tested as moderators. And aluminum is also an option to function as an appropriate moderator (Tanaka et al., 2006). We tested some materials containing fluorine or aluminum such as Al, Al2O3, AlF3, (CF2)n and Fluental to function as neutron moderators. Fluental consists of 69% AlF3, 30% Al, and 1% of LiF (by weight). The moderator effect was also strongly dependent on the geometries of the materials. The geometries of the materials were cylinders with the same radius of 35 cm and we mainly focused on the thickness of the moderator. The moderator was arranged after the simulation multiplier facility, surrounded by lead as reflector. Figs. 6–9 show the epithermal neutron flux and ϕepi/ϕfast for different materials chosen as the moderators versus their thicknesses. It comes from the fact that fluorine has a strong resonance structure in elastic scattering cross section for fast neutrons; moreover, fluorine has a high cross section of inelastic neutron scattering with low levels (197 keV), and also the element has a small atomic mass (Kononov et al., 2004). So due to the fact and Figs. 6 and 7, 20 cm of AlF3 was selected as the first moderator. However, the ϕepi/ϕfast does not satisfy the IAEA BNCT in air beam port recommended quality parameters; we need to choose proper material to serve as second moderator. To increase ϕepi/ϕfast, a second moderator can be used to reduce the number of fast neutrons. After selecting 20 cm of AlF3 as first moderator, Figs. 8 and 9 demonstrate the epithermal neutron flux
Fig. 7. Epithermal neutron flux for different thicknesses and compositions of moderators.
Fig. 8. The ratio of epithermal neutron flux to fast neutron flux where. 20 cm AlF3 is selected as the first moderator.
and ϕepi/ϕfast respectively when different materials act as second moderators. Table 2 gives the in air beam port output of four different materials at proper thicknesses. Limited by the yield of the sealed neutron generator, the Dfast/ϕepi of our assembly is higher than the IAEA in air beam port quality parameters by IAEA; however, ϕepi/ϕth and Dγ/ϕepi complied with the recommended parameters, especially epithermal neutron flux, the most important parameter.
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Fig. 9. Epithermal neutron flux versus the second moderator thickness where. 20 cm AlF3 is selected as the first moderator.
Table 2 Epithermal neutron flux versus different second moderator where 20 cm AlF3 is selected as the first moderator. Moderator
ϕepi [n/cm2s]
ϕepi/ϕth
Dfast/ϕepi [Gy cm2/n]
Dγ/ϕepi [Gy cm2/n]
Fluental (22 cm) AlF3 (20 cm) CF2(15 cm) Al2O3(20 cm)
5.48 108 4.88 108 5.28 108 4.20 108
21.17 19.27 18.46 19.44
7.66 10 13 6.89 10 13 18.56 10 13 12.94 10 13
0.58 10 13 1.42 10 13 2.01 10 13 0.75 10 13
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Fig. 10. Epithermal neutron flux for different thickness of Pb as reflector.
If we merely want to increase the number of epithermal neutrons regardless of the increase of fast neutrons, we can choose Pb, which is thick enough. As the Pb was used as reflector, Fig. 10 shows the epithermal neutron flux corresponding to different thicknesses of Pb. When the thickness reaches 65 cm, the increasing of the epithermal flux is not evident. Therefore, a thickness of 65 cm Pb was used to surround the whole BSA and worked as dose shielding to the surrounding environment.
4. Results and discussion Table 3 Albedoes of some reflectors [15]. Material
Albedoes
Slab thickness 5 (cm)
10 (cm)
100 (cm)
Ni
βf-f βepi-epi
0.400 0.744
0.510 0.759
0.676 0.760
V
βf-f βepi-epi
0.367 0.550
0.500 0.643
0.691 0.682
Bi
βf-f βepi-epi
0.312 0.413
0.486 0.588
0.873 0.895
Pb
βf-f βepi-epi
0.352 0.487
0.521 0.652
0.862 0.867
The number of epithermal neutrons in Table 2 is a little lower than that of Fig. 9. It is because a layer of thin 6LiF, which has an exceedingly high absorption, was placed along the moderator to reduce the number of the thermal neutrons. Cd, which has a large absorption cross section for thermal neutrons, was not chosen because Cd produces a high energy (7.8 MeV) of capture gamma ray, which is difficult to control, and cadmium oxide represents a health hazard. Bi is proper for shielding gamma rays, thus, a layer of 2 cm Bi was placed at the beam port to reduce the γ contamination. 3.3. Reflector The thermalized neutrons need to be guided to the beam port with the reflector material (Koivunoro et al., 2004). A proper reflector can contribute to the increase of the epithermal neutron flux. An appropriate reflector material is supposed to have a high elastic scattering cross section and a low absorption cross section to reenter the neutrons to the moderator. Table 3 presents the albedoes of several reflector materials in different thickness.
In our BSA, when the second moderator was chosen as a 22 cm thickness of Fluental, the epithermal neutron flux of the beam reached 5.48 108 n cm 2 s 1, and ϕepi/ϕth and Dγ/ϕepi comply with the recommended parameters well. However, the Dfast/ϕepi reached 7.66 10 13 Gy cm2/n, which was higher than the in air port parameter recommended for BNCT which is less than 2 10 13 Gy cm2/n. If we could choose a 42 cm thickness of Fluental as the second moderator, the Dfast/ϕepi reaches 2.18 10 13 Gy cm2/n, almost satisfies the recommended parameters. However, this results an epithermal neutrons flux to 3 108 n cm 2 s 1. As motioned above, a limited epithermal flux requires a long treatment time. To solve the problem, increasing the yield of sealed neutron generator seems to be a feasible option. If we could increase the yield of our generator to 2 1011 n/s, then the problem could be solved.
5. Conclusion We discuss the possibility of using four sealed neutron generators as neutron source for BNCT for liver tumor. Four generators are used at the same time to form a yield of 4 1011 n/s and two generators are placed on each side of the liver. A beam shaping assembly for BNCT is designed based on the use of low enrichment uranium as multiplier system for D–T neutron source. According to our results, a 20 cm thickness of AlF3 as the first moderator and a 22 cm thickness of Fluental as the second moderator is the best option for the moderator to achieve a proper epithermal neutron flux. The study of the beam quality shows that using a layer of thin 6 LiF filter to remove thermal neutrons and Bi to reduce gamma contamination is necessary to meet the free beam port parameters by IAEA. The epithermal neutron flux of our facility at the beam port is 5.48 108 n/cm2 s and the values of ϕepi/ϕth and Dγ/ϕepi
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fulfill the free beam port parameters. The value of Dfast/ϕepi does not fulfill the parameters because of the limited neutron yield.
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