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Conceptual design of prompt gamma neutron activation analysis facility at Tehran Research Reactor for BNCT application
Nuclear Inst. and Methods in Physics Research, A 935 (2019) 185–190
Contents lists available at ScienceDirect
Nuclear Inst. and Methods in Physics Research, A journal homepage: www.elsevier.com/locate/nima
Conceptual design of prompt gamma neutron activation analysis facility at Tehran Research Reactor for BNCT application K. Babaeian a , F. Rahmani a ,∗, Y. Kasesaz b a b
Department of Physics, K. N. Toosi University of Technology, P.O. Box 15875-4416, Tehran, Iran Nuclear Science and Technology Research Institute (NSTRI), Tehran, Iran
ARTICLE
INFO
Keywords: PGNAA BNCT Monte Carlo MCNPX2.6 code Tehran Research Reactor
ABSTRACT Since Tehran Research Reactor (TRR) is the only available neutron source, which can be used for Boron Neutron Capture Therapy (BNCT) in Iran, it is essential to provide an appropriate analysis system at TRR in order to determine 10 B concentration in sample tissue or blood. Based on the advantages, Prompt Gamma Neutron Activation Analysis (PGNAA) facility has been selected. Irradiation channels in TRR have been studied using Monte Carlo MCNPX2.6 code and evaluated to provide the appropriate neutron beam for PGNAA. Based on the results of neutron intensity and spectrum, F-beam tube is considered to be the most favorable for PGNAA, and has therefore been selected for further investigation and MCNP simulation. Bismuth blocks for decreasing gamma radiation, and Fluental™ as a filter for fast neutron have been used in F-beam tube. The convergent collimator with iron and polyethylene has been considered to shape the neutron beam and to reduce the neutron and gamma-ray fluxes at the position of the HPGe gamma-ray detector. The sensitivity of the designed system is determined, from MCNP simulation, to be 7.35 cps/μg.
1. Introduction Boron Neutron Capture Therapy (BNCT) is a binary form of radiation therapy, which is based on the capture of thermal neutrons by 10 B. This nuclear reaction produces two high linear energy transfer (LET) particles (4 He and 7 Li) with ranges comparable to a cell diameter; this offers the possibility of destroying tumor cells with high efficiency and fewer side effects on healthy tissue [1]. The efficacy of BNCT depends on 10 B concentration ratio between the tumor and healthy tissue: if this ratio is high enough, the tumor cells are killed with a substantial sparing of the healthy ones. For this reason, the knowledge of the boron concentration by measurement in tumor as well as in healthy tissue is a crucial issue for BNCT treatment [1,2]. There are different methods for measuring the 10 B concentration in tissue, blood and urine, such as prompt gamma ray spectroscopy [2, 3], inductively coupled plasma atomic emission spectroscopy (ICPAES) [4], high resolution alpha autoradiography, alpha spectroscopy [5], and neutron capture radiography [6]. The selected method should be suitable for accurate measuring of 10 B concentration in samples. Prompt gamma neutron activation analysis (PGNAA) is a rapid and non-destructive analytical technique with the best balance between accuracy of results and measurement time [3,7–10]. The PGNAA method for 10 B concentration measurement is based on the detection of prompt gamma ray (478 keV, with a branching ratio of 93.9%) emitted from ∗
thermal neutron capture by the 10
B + nth →4 He +7 Li
10
4
7
10 B
B + nth → He + Li + 𝛾 (478 keV)
(Eqs. (1) and (2)) [2]. 6.1%
(1)
93.9%
(2)
The emission rate of the 478 keV is proportional to the reaction rate of thermal neutron capture reactions related to the 10 B concentration. For the design of PGNAA facility, different parameters such as sensitivity (S) and detection limit (DL) must be calculated and measured. Sensitivity is the net count rate under the measured photopeak (478 keV) divided by the mass of the 10 B in the sample. It could be expressed as [10]: 𝑁A ⟨𝜎 𝜙 ⟩𝜀 (𝐸) (3) 𝑀 th th where, 𝜀 is the gamma-ray absolute detection efficiency, 𝜎th is the thermal neutron capture cross section with emission of 478 keV photons, 𝜙th is the thermal neutron flux, so < 𝜎th 𝜙th > is the (n, 𝛼𝛾) reaction rate per 10 B atomic density, 𝑁A is Avogadro number and M is the molar mass of the 10 B. Another parameter is the detection limit, which relates to the minimum concentration of 10 B in the sample and relates to the detectable count [11]: √ 𝑁𝐷 = 4.653 𝑅b ⋅ 𝑡 + 2.706 (4) 𝑆=
Where, 𝑅b is the background count rate under the photopeak of the 478 keV, and t is the counting time [9]. The precise 𝑅b determination is one
Corresponding author. E-mail address: [email protected] (F. Rahmani).
https://doi.org/10.1016/j.nima.2019.05.040 Received 8 January 2019; Received in revised form 30 April 2019; Accepted 11 May 2019 Available online 15 May 2019 0168-9002/© 2019 Elsevier B.V. All rights reserved.
K. Babaeian, F. Rahmani and Y. Kasesaz
Nuclear Inst. and Methods in Physics Research, A 935 (2019) 185–190
Fig. 1. A schematic view of the TRR pools and its irradiation facilities, A, D, G, and E: 6′′ diameter beam tubes, C: 6′′ diameter through tube, B: 12′′ beam tube, F: 8′′ diameter beam tube. 1: Core, 2: Stall-end pool, 3: Open pool, 4: Medical room, 5: Gamma room, 6: Vertical thermal column, 7: Horizontal thermal column, 8: Hole in the wall.
of the most important issues in the measurement of 10 B concentration, because it has an effect on detection limit as well as on the accuracy and the precision of the analysis. According to this, gamma background should be kept as low as possible for best performance of the PGNAA system. Currently, the BNCT project is being conducted at Tehran Research Reactor (TRR) [12–14], so it is necessary to provide a system to measure the 10 B concentration. Therefore, in this work the conceptual design for PGNAA facility at TRR has been proposed.
3. Surface source PGNAA facility are located far from the core, so for evaluating and designing neutron and gamma shaping, including filters and collimators, the sufficient number of neutrons should be transported to obtain appropriate statistics at the end of F-beam tube in non-analog Monte Carlo method using MCNPX2.6. Also, neutron transport in each simulation program for each design is time consuming. Therefore, instead of reactor core, two new mutual surface sources (neutron and photon) with radial distribution as well as with angular and energy distribution have been defined near the end of the selected beam. Surface sources in the F-beam tube (185 cm far from the center of the reactor core) are shown in Fig. 2. Different classification of energy, angular distribution and position have been tested. Based on the results, the angular distribution of the neutron emission has been considered in 3 bins (90◦ to 10◦ , 10◦ to 5◦ , and 5◦ to 0◦ ) with related probability as well as its energy spectrum for surface sources. The neutron energy in the range of 0 to 12 MeV and the photon energy in the range of 0.01 keV to 14 MeV have been divided by 27 and 33 different energy bins, respectively. In addition, probability of radial distribution of surface sources in various radii (2 cm, 4 cm, 6 cm, 8 cm, 10 cm, and 12.77 cm) have been considered. Neutron and photon fluxes have been calculated at the end of Fbeam tube (in ‘cell 1’ at 298.4 cm from the center of reactor core, Fig. 2) with reactor core as well as with new defined surface sources. This comparison shows the validity of the defined surface source approximation.
2. Materials and methods TRR is a 5MW Material Test Reactor (MTR), pool type research reactor. Its fuel assemblies contain low enriched uranium fuel plates in the form of U3 O8 Al alloy. The reactor pool has two major parts, a stall-end and an open pool. The reactor core can operate in both parts of the pool for different purposes [12]. As shown in Fig. 1, TRR contains seven beam tubes (labeled A–G) with different sizes and shapes and a thermal column, which is filled with removable graphite blocks. As the first step for conceptual design of PGNAA facility at TRR, all accessible irradiation channels of TRR, which can be considered for PGNAA, have been investigated. The investigation has been performed using simulation of TRR core and facilities by MCNPX2.6 code [15]. The C-beam tube has been reserved for neutron radiography purposes, so this beam tube has not been studied for PGNAA facility [10]. In order to select the most suitable irradiation channel, the neutron flux at the end of each beam tubes have been calculated. The choice of the most suitable beam tube has been selected based on the maximum thermal neutron flux at the end of beam tubes, therefore, thermal to fast as well as thermal to total neutron ratio has been considered as criteria. According to Table 1, F-beam tube is considered to be the most favorable for PGNAA, and has therefore been selected for further investigation and MCNP simulation. Beside the maximum thermal neutron flux, undesirable fast neutron flux should be minimized to protect HPGe structure from neutron damage. Photon flux should be also as low as possible to protect operational staff. All this spectrum shaping can be performed using sets of appropriate collimators and filters.
4. Neutron and Gamma filter In order to achieve the suitable neutron beam for PGNAA facility, the fast and epithermal neutrons as well as gamma rays should be decreased as possible. Therefore, the conditions of the gamma and neutron filters have been studied. Bismuth and lead have been investigated as proper materials for attenuation of gamma rays with different thicknesses at the different distances from the reactor core in the selected beam tube. In order to decrease fast and epithermal neutron fluxes, Fluental™ and aluminum oxide as proper materials have been investigated. Fluental™ is a composite which manufactured of Al+AlF3+Li and it is one of the best neutron moderator and is commonly used in neutron filters [16]. 186
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Nuclear Inst. and Methods in Physics Research, A 935 (2019) 185–190
Table 1 Neutron fluxes at the end of beam tubes in the TRR. Beam tubes
Thermal flux (𝜙th ) (n/cm2 s) ± Abs. error
Epithermal flux (𝜙epi ) (n/cm2 s) ± Abs. error
Fast flux (𝜙F ) (n/cm2 s) ± Abs. error
Total flux (𝜙T ) (n/cm2s) ± Abs. error
𝜙th∕ 𝜙f ± Abs. error
𝜙th∕ 𝜙T ± Abs. error
B F G E A D
2.35E+10 ± 1.79E+8 1.62E+10 ± 2.16E+8 7.70E+9 ± 1.81E+8 5.14E+9 ± 1.84E+8 5.48E+9 ± 1.62E+8 3.59E+9 ± 1.23E+8
9.61E+9 5.62E+9 4.93E+9 3.08E+9 3.94E+9 2.77E+9
2.56E+10 ± 2.44E+8 1.58E+10 ± 2.70E+8 1.22E+10 ± 2.91E+8 9.11E+9 ± 2.69E+8 9.56E+9 ± 2.54E+8 6.58E+9 ± 2.15E+8
5.87E+10 3.77E+10 2.48E+10 1.73E+10 1.90E+10 1.29E+10
0.90 1.02 0.63 0.56 0.57 0.55
0.40 0.43 0.31 0.30 0.29 0.28
± ± ± ± ± ±
1.58E+8 1.86E+8 1.79E+8 1.62E+8 1.87E+8 1.50E+8
± ± ± ± ± ±
3.41E+8 3.93E+8 3.85E+8 3.64E+8 3.55E+8 2.89E+8
± ± ± ± ± ±
0.011 0.022 0.021 0.026 0.022 0.026
± ± ± ± ± ±
0.003 0007 0.008 0.012 10E-3 0.011
Fig. 2. The Schematic view of reactor core and the defined surface source in the F-beam tube.
5. Collimator The collimator should be considered to reduce the diameter of the beam at the sample position. Also, a smaller beam diameter can help to shield the facility and to decrease the radiation background during the measurement. Different kinds of collimators have been investigated and finally the suitable design has been selected. The convergent design of collimator has been proposed to reduce the size of irradiated beam by using separate parts. 6. High Purity Germanium detector Since the gamma spectrum of sample in PGNAA has variety of peaks, a high energy resolution gamma ray detector is required. Hence, the High Purity Germanium (HPGe) with good energy resolution detector has been simulated. The geometrical information of GMX series HPGe coaxial detector, model GMX40P4-83, has been considered according to the data provided by manufacturer [17]. The crystal has a diameter of 60.6 mm, length of 66.9 mm and 0.3 μm Ge-B dead layer as well as 700 μm Ge-Li dead layer.
Fig. 3. Comparison of (a) the neutron and (b) photon fluxes in cells 1, using reactor core and surface sources.
7. Results and discussion The results of the simulations for the TRR beam tubes are shown in Table 1; all calculated results have relative error less than 8%. As shown in Table 1, F-beam tube with maximum value of 𝜙th ∕𝜙f and 𝜙th ∕𝜙T (𝜙th neutron with energies under 0.4 eV) is considered to be the most favorable for PGNAA.
As mentioned in the materials and methods section, TRR is 5 MW which can produce 3.75E+17 neutron per second, so all of simulation results have been multiplied by this value. Eq. (5) shows the calculation to convert the 5 MW reactor power to number of neutrons emission from reactor core in one second [18]. joule 1 MeV 1 fission 2.4 neutron × × × s fission 1.6 × 10−13 joule 200 MeV 17 neutron = 3.75 × 10 s
Fig. 2 shows an illustration of the defined surface source, cell 1, and reactor core. Fig. 3 compares the results between reactor core as an initial radiation source and two surface sources. As seen in Fig. 3, there
𝑆 = 5 × 106
(5)
187
K. Babaeian, F. Rahmani and Y. Kasesaz
Nuclear Inst. and Methods in Physics Research, A 935 (2019) 185–190
Fig. 4. Comparison of photon flux attenuation for bismuth and lead filters. Fig. 6. Comparison of neutron flux attenuation for Fluental and aluminum oxide filters.
been calculated in cell 1. The simulations confirmed that the Fluental™ filter should be placed as far downstream as possible. Therefore, the farthest possible distance from the reactor core has been selected. To select the optimal thickness of gamma filter, various thicknesses of bismuth have been considered separately in the F-beam tube with its reactor-side surface at 195 cm from the core. According to Fig. 7, by increasing the bismuth thickness, the photon flux decreases in cell 1, but with filters thicker than 14 cm, no significant gamma attenuation for energy lower than 1 MeV can be seen. In addition, thicker filters cause higher reduction of the thermal neutron flux. So, bismuth with 14 cm in thickness has been chosen as a gamma filter. To absorb the undesired components of the neutron beam and to avoid of high attenuation of thermal neutron, as well as to select the optimal thickness of neutron filter, the various thicknesses of Fluental™ have been tested. As shown in Table 2, for filters thicker than 11 cm, no significant attenuation is occurred for non-thermal neutron flux specially for fast neutron. Therefore, 11 cm of Fluental™ has been selected as a neutron filter. As a convergent collimator, 15 cm thick of iron and 25 cm thick of polyethylene in conical shape have been placed at the end of F-beam tube as desirable materials to collimate the neutron beam. The changes of neutron flux with increase of distance from the surface source are shown in Fig. 8. As shown, the neutron beam has been shaped and the diameter of the beam at the end of F-beam tube has been reduced. Maximum internal diameter of collimator is 14 cm which reduces to 4 cm at the exit. For more shielding of detector from gamma rays produced via neutron interactions with the structure and decreasing the detection limit, it is necessary to decrease gamma rays (Rb as a rate of gamma background) as low as possible in the position of detector, so lead shield with thickness of 7 cm has been used around HPGe. In order to investigate of the absolute efficiency of simulated HPGe detector as a function of distance from the gamma source, the detector, perpendicularly located with respect to the neutron beam direction and at different distances (11 to 25 cm) from the irradiated boron sample (478 keV gamma source), and gamma count rate under the photopeak of the 478 keV as well as (n, 𝛼𝛾) reaction rate has been calculated. As shown in Eq. (3), the absolute efficiency is the net count rate under photopeak of the 478 keV, divided by (n, 𝛼𝛾) reaction rate in the boron sample. As shown in Fig. 9, the blood sample contain 10 B has been placed at 347 cm from the reactor core center. The thermal neutron flux in sample position has been calculated about 3.21e+7 n/cm2 .s, and the
Fig. 5. Comparison of neutron flux attenuation for bismuth and lead filters.
are good agreement between neutron and photon flux from surface sources and reactor core as a neutron source. As mentioned above, lead and bismuth have been studied as a suitable gamma filter for PGNNA facility. To compare between them, a 10 cm thick block of each material has been placed separately at 195 cm from the reactor core, and the fluxes of neutron and photon in cell 1 (Fig. 2) has been investigated. As shown in Figs. 4 and 5, the gamma attenuation for both of bismuth and lead are high, but based on the lower thermal neutron absorption cross section, bismuth has been selected as gamma filter. To determine the best location of gamma filter, the 10 cm thick of bismuth block has been inserted in the F-beam tube at different positions from the reactor core (195, 215, 235, 255, and 275 cm) and the neutron and gamma flux have been studied in cell 1. The simulations confirmed that the bismuth filter should be placed as far upstream as possible, at 195 cm from the reactor core center (10 cm downstream from the defined surface source). For neutron filter, the 10 cm thick of Fluental™ and aluminumoxide powder blocks have been simulated separately in F-beam tube. As shown in Fig. 6, Fluental™ with lower thermal neutron absorption cross section has been selected. In order to select the suitable position for Fluental™, 10 cm thick of Fluental™ block has been located at 225, 245, 265, and 285 cm from the reactor core and the neutron flux has 188
K. Babaeian, F. Rahmani and Y. Kasesaz
Nuclear Inst. and Methods in Physics Research, A 935 (2019) 185–190
Table 2 Neutron flux in cell 1 for different thicknesses of Fluental™, as well as the attenuation factor due to the implementation of the Fluental™.
Empty beam tube 5 cm Fluental™ 8 cm Fluental™ 11 cm Fluental™ 14 cm Fluental™ 17 cm Fluental™ 20 cm Fluental™
Thermal neutron flux (n/cm2 s)
Attenuation factor
Epithermal neutron flux (n/cm2 s)
Attenuation factor
Fast neutron flux (n/cm2 s)
Attenuation factor
1.83E+10 8.23E+9 5.89E+9 4.51E+9 3.61E+9 3.00E+9 2.57E+9
– 0.45 0.32 0.25 0.20 0.16 0.14
6.52E+9 3.60E+9 2.91E+9 2.85E+9 2.23E+9 1.56E+9 1.10E+9
– 0.55 0.45 0.44 0.34 0.24 0.17
1.88E+10 9.90E+9 7.89E+9 6.21E+9 5.77E+9 5.56E+9 5.37E+9
0.53 0.42 0.33 0.31 0.30 0.29
Fig. 7. Comparison of photon flux in the cell 1 for bismuth with 5, 8, 11, 14, 17 and 20 cm in thicknesses. Top-Right shows photon flux with energy lower than 1 MeV.
absolute detection efficiency of HPGe when sample located at 15 cm far from detector has been calculated about 5.03e−3. From MCNP simulations, the sensitivity of the facility for 10 B in blood samples is determined to be 7.35 cps/μg, which is appropriate for 10 B measurement in BNCT application [10]. The schematic view of the designed PGNAA facility as well as the characteristics of materials (and related mass densities) has been illustrated in Fig. 9 and Table 3, respectively. 8. Conclusion In this paper, the conceptual design of PGNAA facility at TRR in the F-beam tube has been presented. The equipment in F-beam tube consists of 14 cm thick bismuth block and 11 cm thick of Fluental™, which respectively located at 195 and 285 cm from the reactor core. Convergent collimator includes 15 cm thick of iron and 25 cm thick of polyethylene with 14 cm initial internal diameter and 4 cm final internal diameter at the end of F-beam tube have been used. The simulated HPGe detector with appropriate lead shield (gamma shield) perpendicularly located with respect to the neutron beam direction at 15 cm from irradiated boron sample. For this facility, the sensitivity of 10 B in blood sample has been determined to be 7.35 cps/μg, which is appropriate for 10 B measurement in BNCT application.
Fig. 8. 2D pseudo-color plot of neutron flux over the region of the convergent collimator and sample . (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
189
K. Babaeian, F. Rahmani and Y. Kasesaz
Nuclear Inst. and Methods in Physics Research, A 935 (2019) 185–190
Fig. 9. Scheme of designed PGNAA setup. 1: Bi filter, 2: Al2 O3 filter, 3: Iron collimator, 4: Polyethylene collimator, 5: Lead shield. 6: HPGe detector, 7: Boron contain sample position, 8: Boron dead layer, 9: High Purity Germanium crystal, 10: Lithium drifted, 11: Aluminum, 12: Beryllium. Table 3 Material that used in this study and their mass density. Material
Density (g/cm3 )
Al2 O3 Fluental™ Polyethylene Iron Bismuth Lead Beryllium Germanium, High Purity Boron Lithium Aluminum
3.97 2.99 0.93 5.9 9.78 11.34 1.85 5.32 2.34 0.53 2.70
[7] W.A.G. Sauerwein, A. Wittig, R. Moss, Y. Nakagawa, Neutron Capture Therapy Principles and Application, Springer, 2012. [8] T. Belgya, Prompt gamma activation analysis at the Budapest research reactor, Phys. Procedia 31 (2012) 99–109. [9] T. Schmitz, K. Appelman, P. Kudejova, C. Shutz, J.V. Kratz, R. Moss, G. Otto, G. Hampel, Determination of boron concentration in blood and tissue samples from patients with liver metastases of colorectal carcinoma using prompt gamma ray activation analysis (PGAA), Appl. Radiat. Isot. 69 (2011) 936–941. [10] S. Stella, Design of PGNAA System for Boron Neutron Capture Therapy (BNCT) using Monte Carlo Code (Ph.D. thesis), University of Pavia, 2011. [11] G.E. Knoll, Radiation Detection and Measurement, third ed., John Wiley & Sons, New York, 2000. [12] Y. Kasesaz, H. Khalafi, F. Rahmani, A. Ezati, M. Keyvani, A. Hossnirokh, M. Azizi Shamami, M. Monshizadeh, A feasibility study of the tehran research reactor as a neutron source for BNCT, Appl. Radiat. Isot. 90 (2014) 132–137. [13] Y. Kasesaz, H. Khalafi, F. Rahmani, Design of an epithermal neutron beam for BNCT in thermal column of Tehran research reactor, Ann. Nucl. Energy 68 (2014) 234–238. [14] Y. Kasesaz, H. Khalafi, F. Rahmani, et al., Design and construction of a thermal neutron beam for BNCT at Tehran research reactor, Appl. Radiat. Isot. 94 (2014) 149–151. [15] MCNP—A general Monte Carlo N-particle transport code, Los Alamos National Laboratory Monte Carlo Team, Version 2.6, LA-CP-07-1473, 2008. [16] I. Auterinen, S.E.J. Salmenhaara, FIR 1 reactor in service for boron neutron capture therapy (BNCT) and isotope production, in: Research Reactor Utilization, Safety, Decommissioning, Fuel and Waste Management, STI/PUB/1212 IAEA, ss. 313-324, 2005. [17] GMX series coaxial HPGe detector product configuration guide, Ortec-Online, Available at: https://www.ortec-online.com/-/media/ametekortec/brochures/ gamma-x.pdf. (Last accessed on 7th 2019). [18] E. Bavarnegin, A. sadremomtaz, H. Khalafi, Y. Kasesaz, M. Golshanian, H. Ghods, A. Ezzati, M. Keyvani, M. Haddadi, Measurement and simulation of the TRR BNCT beam parameters, Nucl. Instrum. Methods Phys. Res. A 830 (2016) 53–58.
References [1] I.A.E.A, TECDOC-1223, Current Status of Neutron Capture Therapy, International Atomic Energy Agency, Vienna, 2001. [2] K.J. Riley, Improved Boron 10 Quantification Via PGNAA and ICP-AES, Massachusetts Institute of Technology, 1997, pp. 20–24. [3] C.P.J. Raaijmakers, M.W. Koninenberg, L. Dewit, D. Haritz, R. Huiskamp, K. Philip, et al., Monitoring of blood-10b concentration for boron neutron capture therapy using prompt gamma-ray analysis, Acta Oncol. 34 (1995) 517–523. [4] S. Evans, U. Krahenbuhl, Boron analysis in biological material: microwave digestion procedure and determination by different methods, Fresenius Z. Anal. Chem. 349 (1994) 349–459. [5] G.R. Solares, R.G. Zamenhof, G. Cano, Microdosimetry and Compound Factors for Neutron Capture Therapy, Plenum Press, New York, 1993. [6] Z.B. Alfassi, T.Y. Probst, On the calibration curve for determination of boron in tissue by quantitative neutron capture therapy radiography, Nucl. Instrum. Methods A 428 (1999) 502–507.
190
Contents lists available at ScienceDirect
Nuclear Inst. and Methods in Physics Research, A journal homepage: www.elsevier.com/locate/nima
Conceptual design of prompt gamma neutron activation analysis facility at Tehran Research Reactor for BNCT application K. Babaeian a , F. Rahmani a ,∗, Y. Kasesaz b a b
Department of Physics, K. N. Toosi University of Technology, P.O. Box 15875-4416, Tehran, Iran Nuclear Science and Technology Research Institute (NSTRI), Tehran, Iran
ARTICLE
INFO
Keywords: PGNAA BNCT Monte Carlo MCNPX2.6 code Tehran Research Reactor
ABSTRACT Since Tehran Research Reactor (TRR) is the only available neutron source, which can be used for Boron Neutron Capture Therapy (BNCT) in Iran, it is essential to provide an appropriate analysis system at TRR in order to determine 10 B concentration in sample tissue or blood. Based on the advantages, Prompt Gamma Neutron Activation Analysis (PGNAA) facility has been selected. Irradiation channels in TRR have been studied using Monte Carlo MCNPX2.6 code and evaluated to provide the appropriate neutron beam for PGNAA. Based on the results of neutron intensity and spectrum, F-beam tube is considered to be the most favorable for PGNAA, and has therefore been selected for further investigation and MCNP simulation. Bismuth blocks for decreasing gamma radiation, and Fluental™ as a filter for fast neutron have been used in F-beam tube. The convergent collimator with iron and polyethylene has been considered to shape the neutron beam and to reduce the neutron and gamma-ray fluxes at the position of the HPGe gamma-ray detector. The sensitivity of the designed system is determined, from MCNP simulation, to be 7.35 cps/μg.
1. Introduction Boron Neutron Capture Therapy (BNCT) is a binary form of radiation therapy, which is based on the capture of thermal neutrons by 10 B. This nuclear reaction produces two high linear energy transfer (LET) particles (4 He and 7 Li) with ranges comparable to a cell diameter; this offers the possibility of destroying tumor cells with high efficiency and fewer side effects on healthy tissue [1]. The efficacy of BNCT depends on 10 B concentration ratio between the tumor and healthy tissue: if this ratio is high enough, the tumor cells are killed with a substantial sparing of the healthy ones. For this reason, the knowledge of the boron concentration by measurement in tumor as well as in healthy tissue is a crucial issue for BNCT treatment [1,2]. There are different methods for measuring the 10 B concentration in tissue, blood and urine, such as prompt gamma ray spectroscopy [2, 3], inductively coupled plasma atomic emission spectroscopy (ICPAES) [4], high resolution alpha autoradiography, alpha spectroscopy [5], and neutron capture radiography [6]. The selected method should be suitable for accurate measuring of 10 B concentration in samples. Prompt gamma neutron activation analysis (PGNAA) is a rapid and non-destructive analytical technique with the best balance between accuracy of results and measurement time [3,7–10]. The PGNAA method for 10 B concentration measurement is based on the detection of prompt gamma ray (478 keV, with a branching ratio of 93.9%) emitted from ∗
thermal neutron capture by the 10
B + nth →4 He +7 Li
10
4
7
10 B
B + nth → He + Li + 𝛾 (478 keV)
(Eqs. (1) and (2)) [2]. 6.1%
(1)
93.9%
(2)
The emission rate of the 478 keV is proportional to the reaction rate of thermal neutron capture reactions related to the 10 B concentration. For the design of PGNAA facility, different parameters such as sensitivity (S) and detection limit (DL) must be calculated and measured. Sensitivity is the net count rate under the measured photopeak (478 keV) divided by the mass of the 10 B in the sample. It could be expressed as [10]: 𝑁A ⟨𝜎 𝜙 ⟩𝜀 (𝐸) (3) 𝑀 th th where, 𝜀 is the gamma-ray absolute detection efficiency, 𝜎th is the thermal neutron capture cross section with emission of 478 keV photons, 𝜙th is the thermal neutron flux, so < 𝜎th 𝜙th > is the (n, 𝛼𝛾) reaction rate per 10 B atomic density, 𝑁A is Avogadro number and M is the molar mass of the 10 B. Another parameter is the detection limit, which relates to the minimum concentration of 10 B in the sample and relates to the detectable count [11]: √ 𝑁𝐷 = 4.653 𝑅b ⋅ 𝑡 + 2.706 (4) 𝑆=
Where, 𝑅b is the background count rate under the photopeak of the 478 keV, and t is the counting time [9]. The precise 𝑅b determination is one
Corresponding author. E-mail address: [email protected] (F. Rahmani).
https://doi.org/10.1016/j.nima.2019.05.040 Received 8 January 2019; Received in revised form 30 April 2019; Accepted 11 May 2019 Available online 15 May 2019 0168-9002/© 2019 Elsevier B.V. All rights reserved.
K. Babaeian, F. Rahmani and Y. Kasesaz
Nuclear Inst. and Methods in Physics Research, A 935 (2019) 185–190
Fig. 1. A schematic view of the TRR pools and its irradiation facilities, A, D, G, and E: 6′′ diameter beam tubes, C: 6′′ diameter through tube, B: 12′′ beam tube, F: 8′′ diameter beam tube. 1: Core, 2: Stall-end pool, 3: Open pool, 4: Medical room, 5: Gamma room, 6: Vertical thermal column, 7: Horizontal thermal column, 8: Hole in the wall.
of the most important issues in the measurement of 10 B concentration, because it has an effect on detection limit as well as on the accuracy and the precision of the analysis. According to this, gamma background should be kept as low as possible for best performance of the PGNAA system. Currently, the BNCT project is being conducted at Tehran Research Reactor (TRR) [12–14], so it is necessary to provide a system to measure the 10 B concentration. Therefore, in this work the conceptual design for PGNAA facility at TRR has been proposed.
3. Surface source PGNAA facility are located far from the core, so for evaluating and designing neutron and gamma shaping, including filters and collimators, the sufficient number of neutrons should be transported to obtain appropriate statistics at the end of F-beam tube in non-analog Monte Carlo method using MCNPX2.6. Also, neutron transport in each simulation program for each design is time consuming. Therefore, instead of reactor core, two new mutual surface sources (neutron and photon) with radial distribution as well as with angular and energy distribution have been defined near the end of the selected beam. Surface sources in the F-beam tube (185 cm far from the center of the reactor core) are shown in Fig. 2. Different classification of energy, angular distribution and position have been tested. Based on the results, the angular distribution of the neutron emission has been considered in 3 bins (90◦ to 10◦ , 10◦ to 5◦ , and 5◦ to 0◦ ) with related probability as well as its energy spectrum for surface sources. The neutron energy in the range of 0 to 12 MeV and the photon energy in the range of 0.01 keV to 14 MeV have been divided by 27 and 33 different energy bins, respectively. In addition, probability of radial distribution of surface sources in various radii (2 cm, 4 cm, 6 cm, 8 cm, 10 cm, and 12.77 cm) have been considered. Neutron and photon fluxes have been calculated at the end of Fbeam tube (in ‘cell 1’ at 298.4 cm from the center of reactor core, Fig. 2) with reactor core as well as with new defined surface sources. This comparison shows the validity of the defined surface source approximation.
2. Materials and methods TRR is a 5MW Material Test Reactor (MTR), pool type research reactor. Its fuel assemblies contain low enriched uranium fuel plates in the form of U3 O8 Al alloy. The reactor pool has two major parts, a stall-end and an open pool. The reactor core can operate in both parts of the pool for different purposes [12]. As shown in Fig. 1, TRR contains seven beam tubes (labeled A–G) with different sizes and shapes and a thermal column, which is filled with removable graphite blocks. As the first step for conceptual design of PGNAA facility at TRR, all accessible irradiation channels of TRR, which can be considered for PGNAA, have been investigated. The investigation has been performed using simulation of TRR core and facilities by MCNPX2.6 code [15]. The C-beam tube has been reserved for neutron radiography purposes, so this beam tube has not been studied for PGNAA facility [10]. In order to select the most suitable irradiation channel, the neutron flux at the end of each beam tubes have been calculated. The choice of the most suitable beam tube has been selected based on the maximum thermal neutron flux at the end of beam tubes, therefore, thermal to fast as well as thermal to total neutron ratio has been considered as criteria. According to Table 1, F-beam tube is considered to be the most favorable for PGNAA, and has therefore been selected for further investigation and MCNP simulation. Beside the maximum thermal neutron flux, undesirable fast neutron flux should be minimized to protect HPGe structure from neutron damage. Photon flux should be also as low as possible to protect operational staff. All this spectrum shaping can be performed using sets of appropriate collimators and filters.
4. Neutron and Gamma filter In order to achieve the suitable neutron beam for PGNAA facility, the fast and epithermal neutrons as well as gamma rays should be decreased as possible. Therefore, the conditions of the gamma and neutron filters have been studied. Bismuth and lead have been investigated as proper materials for attenuation of gamma rays with different thicknesses at the different distances from the reactor core in the selected beam tube. In order to decrease fast and epithermal neutron fluxes, Fluental™ and aluminum oxide as proper materials have been investigated. Fluental™ is a composite which manufactured of Al+AlF3+Li and it is one of the best neutron moderator and is commonly used in neutron filters [16]. 186
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Table 1 Neutron fluxes at the end of beam tubes in the TRR. Beam tubes
Thermal flux (𝜙th ) (n/cm2 s) ± Abs. error
Epithermal flux (𝜙epi ) (n/cm2 s) ± Abs. error
Fast flux (𝜙F ) (n/cm2 s) ± Abs. error
Total flux (𝜙T ) (n/cm2s) ± Abs. error
𝜙th∕ 𝜙f ± Abs. error
𝜙th∕ 𝜙T ± Abs. error
B F G E A D
2.35E+10 ± 1.79E+8 1.62E+10 ± 2.16E+8 7.70E+9 ± 1.81E+8 5.14E+9 ± 1.84E+8 5.48E+9 ± 1.62E+8 3.59E+9 ± 1.23E+8
9.61E+9 5.62E+9 4.93E+9 3.08E+9 3.94E+9 2.77E+9
2.56E+10 ± 2.44E+8 1.58E+10 ± 2.70E+8 1.22E+10 ± 2.91E+8 9.11E+9 ± 2.69E+8 9.56E+9 ± 2.54E+8 6.58E+9 ± 2.15E+8
5.87E+10 3.77E+10 2.48E+10 1.73E+10 1.90E+10 1.29E+10
0.90 1.02 0.63 0.56 0.57 0.55
0.40 0.43 0.31 0.30 0.29 0.28
± ± ± ± ± ±
1.58E+8 1.86E+8 1.79E+8 1.62E+8 1.87E+8 1.50E+8
± ± ± ± ± ±
3.41E+8 3.93E+8 3.85E+8 3.64E+8 3.55E+8 2.89E+8
± ± ± ± ± ±
0.011 0.022 0.021 0.026 0.022 0.026
± ± ± ± ± ±
0.003 0007 0.008 0.012 10E-3 0.011
Fig. 2. The Schematic view of reactor core and the defined surface source in the F-beam tube.
5. Collimator The collimator should be considered to reduce the diameter of the beam at the sample position. Also, a smaller beam diameter can help to shield the facility and to decrease the radiation background during the measurement. Different kinds of collimators have been investigated and finally the suitable design has been selected. The convergent design of collimator has been proposed to reduce the size of irradiated beam by using separate parts. 6. High Purity Germanium detector Since the gamma spectrum of sample in PGNAA has variety of peaks, a high energy resolution gamma ray detector is required. Hence, the High Purity Germanium (HPGe) with good energy resolution detector has been simulated. The geometrical information of GMX series HPGe coaxial detector, model GMX40P4-83, has been considered according to the data provided by manufacturer [17]. The crystal has a diameter of 60.6 mm, length of 66.9 mm and 0.3 μm Ge-B dead layer as well as 700 μm Ge-Li dead layer.
Fig. 3. Comparison of (a) the neutron and (b) photon fluxes in cells 1, using reactor core and surface sources.
7. Results and discussion The results of the simulations for the TRR beam tubes are shown in Table 1; all calculated results have relative error less than 8%. As shown in Table 1, F-beam tube with maximum value of 𝜙th ∕𝜙f and 𝜙th ∕𝜙T (𝜙th neutron with energies under 0.4 eV) is considered to be the most favorable for PGNAA.
As mentioned in the materials and methods section, TRR is 5 MW which can produce 3.75E+17 neutron per second, so all of simulation results have been multiplied by this value. Eq. (5) shows the calculation to convert the 5 MW reactor power to number of neutrons emission from reactor core in one second [18]. joule 1 MeV 1 fission 2.4 neutron × × × s fission 1.6 × 10−13 joule 200 MeV 17 neutron = 3.75 × 10 s
Fig. 2 shows an illustration of the defined surface source, cell 1, and reactor core. Fig. 3 compares the results between reactor core as an initial radiation source and two surface sources. As seen in Fig. 3, there
𝑆 = 5 × 106
(5)
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Fig. 4. Comparison of photon flux attenuation for bismuth and lead filters. Fig. 6. Comparison of neutron flux attenuation for Fluental and aluminum oxide filters.
been calculated in cell 1. The simulations confirmed that the Fluental™ filter should be placed as far downstream as possible. Therefore, the farthest possible distance from the reactor core has been selected. To select the optimal thickness of gamma filter, various thicknesses of bismuth have been considered separately in the F-beam tube with its reactor-side surface at 195 cm from the core. According to Fig. 7, by increasing the bismuth thickness, the photon flux decreases in cell 1, but with filters thicker than 14 cm, no significant gamma attenuation for energy lower than 1 MeV can be seen. In addition, thicker filters cause higher reduction of the thermal neutron flux. So, bismuth with 14 cm in thickness has been chosen as a gamma filter. To absorb the undesired components of the neutron beam and to avoid of high attenuation of thermal neutron, as well as to select the optimal thickness of neutron filter, the various thicknesses of Fluental™ have been tested. As shown in Table 2, for filters thicker than 11 cm, no significant attenuation is occurred for non-thermal neutron flux specially for fast neutron. Therefore, 11 cm of Fluental™ has been selected as a neutron filter. As a convergent collimator, 15 cm thick of iron and 25 cm thick of polyethylene in conical shape have been placed at the end of F-beam tube as desirable materials to collimate the neutron beam. The changes of neutron flux with increase of distance from the surface source are shown in Fig. 8. As shown, the neutron beam has been shaped and the diameter of the beam at the end of F-beam tube has been reduced. Maximum internal diameter of collimator is 14 cm which reduces to 4 cm at the exit. For more shielding of detector from gamma rays produced via neutron interactions with the structure and decreasing the detection limit, it is necessary to decrease gamma rays (Rb as a rate of gamma background) as low as possible in the position of detector, so lead shield with thickness of 7 cm has been used around HPGe. In order to investigate of the absolute efficiency of simulated HPGe detector as a function of distance from the gamma source, the detector, perpendicularly located with respect to the neutron beam direction and at different distances (11 to 25 cm) from the irradiated boron sample (478 keV gamma source), and gamma count rate under the photopeak of the 478 keV as well as (n, 𝛼𝛾) reaction rate has been calculated. As shown in Eq. (3), the absolute efficiency is the net count rate under photopeak of the 478 keV, divided by (n, 𝛼𝛾) reaction rate in the boron sample. As shown in Fig. 9, the blood sample contain 10 B has been placed at 347 cm from the reactor core center. The thermal neutron flux in sample position has been calculated about 3.21e+7 n/cm2 .s, and the
Fig. 5. Comparison of neutron flux attenuation for bismuth and lead filters.
are good agreement between neutron and photon flux from surface sources and reactor core as a neutron source. As mentioned above, lead and bismuth have been studied as a suitable gamma filter for PGNNA facility. To compare between them, a 10 cm thick block of each material has been placed separately at 195 cm from the reactor core, and the fluxes of neutron and photon in cell 1 (Fig. 2) has been investigated. As shown in Figs. 4 and 5, the gamma attenuation for both of bismuth and lead are high, but based on the lower thermal neutron absorption cross section, bismuth has been selected as gamma filter. To determine the best location of gamma filter, the 10 cm thick of bismuth block has been inserted in the F-beam tube at different positions from the reactor core (195, 215, 235, 255, and 275 cm) and the neutron and gamma flux have been studied in cell 1. The simulations confirmed that the bismuth filter should be placed as far upstream as possible, at 195 cm from the reactor core center (10 cm downstream from the defined surface source). For neutron filter, the 10 cm thick of Fluental™ and aluminumoxide powder blocks have been simulated separately in F-beam tube. As shown in Fig. 6, Fluental™ with lower thermal neutron absorption cross section has been selected. In order to select the suitable position for Fluental™, 10 cm thick of Fluental™ block has been located at 225, 245, 265, and 285 cm from the reactor core and the neutron flux has 188
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Table 2 Neutron flux in cell 1 for different thicknesses of Fluental™, as well as the attenuation factor due to the implementation of the Fluental™.
Empty beam tube 5 cm Fluental™ 8 cm Fluental™ 11 cm Fluental™ 14 cm Fluental™ 17 cm Fluental™ 20 cm Fluental™
Thermal neutron flux (n/cm2 s)
Attenuation factor
Epithermal neutron flux (n/cm2 s)
Attenuation factor
Fast neutron flux (n/cm2 s)
Attenuation factor
1.83E+10 8.23E+9 5.89E+9 4.51E+9 3.61E+9 3.00E+9 2.57E+9
– 0.45 0.32 0.25 0.20 0.16 0.14
6.52E+9 3.60E+9 2.91E+9 2.85E+9 2.23E+9 1.56E+9 1.10E+9
– 0.55 0.45 0.44 0.34 0.24 0.17
1.88E+10 9.90E+9 7.89E+9 6.21E+9 5.77E+9 5.56E+9 5.37E+9
0.53 0.42 0.33 0.31 0.30 0.29
Fig. 7. Comparison of photon flux in the cell 1 for bismuth with 5, 8, 11, 14, 17 and 20 cm in thicknesses. Top-Right shows photon flux with energy lower than 1 MeV.
absolute detection efficiency of HPGe when sample located at 15 cm far from detector has been calculated about 5.03e−3. From MCNP simulations, the sensitivity of the facility for 10 B in blood samples is determined to be 7.35 cps/μg, which is appropriate for 10 B measurement in BNCT application [10]. The schematic view of the designed PGNAA facility as well as the characteristics of materials (and related mass densities) has been illustrated in Fig. 9 and Table 3, respectively. 8. Conclusion In this paper, the conceptual design of PGNAA facility at TRR in the F-beam tube has been presented. The equipment in F-beam tube consists of 14 cm thick bismuth block and 11 cm thick of Fluental™, which respectively located at 195 and 285 cm from the reactor core. Convergent collimator includes 15 cm thick of iron and 25 cm thick of polyethylene with 14 cm initial internal diameter and 4 cm final internal diameter at the end of F-beam tube have been used. The simulated HPGe detector with appropriate lead shield (gamma shield) perpendicularly located with respect to the neutron beam direction at 15 cm from irradiated boron sample. For this facility, the sensitivity of 10 B in blood sample has been determined to be 7.35 cps/μg, which is appropriate for 10 B measurement in BNCT application.
Fig. 8. 2D pseudo-color plot of neutron flux over the region of the convergent collimator and sample . (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)
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Fig. 9. Scheme of designed PGNAA setup. 1: Bi filter, 2: Al2 O3 filter, 3: Iron collimator, 4: Polyethylene collimator, 5: Lead shield. 6: HPGe detector, 7: Boron contain sample position, 8: Boron dead layer, 9: High Purity Germanium crystal, 10: Lithium drifted, 11: Aluminum, 12: Beryllium. Table 3 Material that used in this study and their mass density. Material
Density (g/cm3 )
Al2 O3 Fluental™ Polyethylene Iron Bismuth Lead Beryllium Germanium, High Purity Boron Lithium Aluminum
3.97 2.99 0.93 5.9 9.78 11.34 1.85 5.32 2.34 0.53 2.70
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