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A shielding design for an accelerator-based neutron source for boron neutron capture therapy
ARTICLE IN PRESS
Applied Radiation and Isotopes 61 (2004) 1027–1031
A shielding design for an accelerator-based neutron source for boron neutron capture therapy A.E. Hawk, T.E. Blue*, J.E. Woollard Nuclear Engineering Program, The Ohio State University, 206 West, 18th Avenue, Columbus, Oh 43210-1189, USA
Abstract Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a 7Li target, a target heat removal system (HRS), a moderator assembly, and a treatment room. The intent of this paper is to demonstrate the advantages of a shielded moderator assembly design, in terms of material requirements necessary to adequately protect radiation personnel located outside a treatment room for BNCT, over an unshielded moderator assembly design. r 2004 Elsevier Ltd. All rights reserved. Keywords: Boron neutron capture therapy (BNCT); Treatment room; Shielding
1. Introduction/background Preliminary shielding studies for an accelerator based neutron source (ABNS) for boron neutron capture therapy (BNCT) were conducted at The Ohio State University by Qu et al. (1993), Evans (1994) and Evans and Blue (1996). Qu et al. (1993) used the Monte Carlo N-Particle Transport Code (MCNP) (Briesmeister, 2000) to accurately model a 6 MV linear accelerator Xray irradiation room located at The Arthur James Cancer Hospital and evaluate whether this design provided adequate personnel protection. He concluded that it did not. Evans (1994) used the computer code MCNP to calculate the amount of ordinary concrete that would be necessary to provide adequate protection for radiation personnel standing outside a treatment room for an ABNS for BNCT. A radiation worker, according to Evans, was adequately protected when the transmitted effective dose to that worker is at least one order of magnitude less than the occupational effective *Corresponding author. E-mail address: [email protected] (T.E. Blue).
dose limit as recommended by the International Commission on Radiological Protection (ICRP) (1990). Both studies assumed that neutrons were produced in a 7Li target, and were moderated in energy by a moderator assembly. However, the moderator modeled by Qu et al. (1993) and assumed by Evans (1994) has evolved with time to consist of a fluental/lead fluoride moderator/reflector moderator assembly. Hence, the shielding requirements for the present moderator/reflector assembly differ from those of past moderator assemblies. Furthermore, Evans (1994) concluded that the primary component of the effective dose to radiation personnel located outside the treatment room was due to photons produced from inelastic neutron scatter and neutron absorption in the concrete walls. Therefore, the most efficient method to adequately protect radiation personnel, in terms of material requirements, is to prevent neutrons from entering the treatment room walls. There are three ways to prevent neutrons from entering the treatment room walls. Option 1 is to line the inner surface of the treatment room with a neutron absorbing material. Option 2 is to line the outer surface
0969-8043/$ - see front matter r 2004 Elsevier Ltd. All rights reserved. doi:10.1016/j.apradiso.2004.05.038
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of the moderator assembly with a neutron absorbing material, and Option 3 is to use options 1 and 2 in combination. Evans and Blue (1996) evaluated option 1 and concluded that a significant reduction in the amount of concrete, which is necessary to protect radiation personnel located outside the treatment room, could be achieved by lining the inner surface of the treatment room walls with a thin layer of gypsum wall board, borated flex panel and lead. The lead was included in the treatment room model to prevent gamma rays produced in the concrete walls from entering the treatment room. Furthermore, Evans and Blue (1996) found that if the primary shielding material, concrete, was changed to a denser material (to better attenuate the photon fluence produced in the treatment room walls), specifically barytes concrete, a further reduction in the thickness of the walls could be achieved. This paper evaluates shielding options 2 and 3. For option 2, the outer surface of the moderator assembly is surrounded by borated and lithium-doped polyethylene. It is assumed for option 3 (a combination of options 1 and 2), that for option 1, the inner surface of the treatment room is lined with gypsum wallboard, borated flex panel, and a thin (1.58 mm) layer of lead.
2. Methods The whole body effective dose to radiation personnel located directly outside the treatment room during patient treatment, is calculated for four moderator assembly/treatment room configurations: (Case 1) a basis treatment room with an unshielded moderator assembly; (Case 2) a revised treatment room with an unshielded moderator assembly; (Case 3) a basis treatment room with a shielded moderator assembly; and (Case 4) a revised treatment room with a shielded moderator assembly. The whole body effective dose to radiation personnel is calculated by modeling the appropriate moderator assembly/treatment room configuration using the MCNP version 4C computer code. Each MCNP model consists of a geometric representation of the treatment room, the moderator assembly and heat removal system (HRS)/target assembly, and the patient phantom developed by Cristy and Eckerman (1987). The patient phantom is included in all four configurations to act as a particle scatterer/absorber. Each model also includes a neutron source term, neutron and photon tallies, and a variance reduction scheme to allow for accurate, as well as timely results. 2.1. Treatment room geometry Fig. 1 is a cross-sectional view of the basis treatment room evaluated in the MCNP calculations. The room is
4.27 m (14 ft) high with 3.05 m 3.66 m (100 120 ) floor dimensions. These dimensions were based on those of a previous shielding study of a diagnostic X-ray room (Metzger et al., 1993). The treatment room geometry is relatively simple; it is essentially a large rectangular parallelopiped. Doors, windows, mazes and ducts are not included in the model. Support structures for the moderator assembly as well as support structures in the walls, are also not included in the model. In the basis treatment room design, concrete and lead are the primary materials used to attenuate the neutron and photon flux in the walls of the room. Gypsum wallboard, which is commonly used in radiotherapy rooms, was used on the head (with respect to the patient phantom), right, foot, and left walls. It was assumed that the treatment room was located on the first floor of a hospital facility and, as a result, soil was included in the model below the floor. Gypsum wallboard and lead sheet were not used on either the ceiling or the floor of the basis treatment room. In the revised treatment room design, concrete and lead again were the primary shielding materials, and gypsum wallboard lined the interior (including the ceiling) of the treatment room. A thin (69.9 mm) layer of borated flex panel was placed behind the gypsum wallboard. The borated flex panel reduces the thermal neutron flux entering the concrete walls, thereby reducing the number of thermal neutron capture gammas produced in the concrete. The flex panel also reduces the backscattered thermal neutron flux (from the concrete walls) in the treatment room. 2.2. Moderator assembly geometry The moderator/reflector assembly used in all the calculations is the 30 cm fluental/lead fluoride moderator/reflector assembly (Fig. 2) designed at The Ohio State University (Orr et al., 2001). The long axis of the moderator assembly is positioned 1 m above the floor of the treatment room. In order to maximize the shielding in the upstream (with respect to the direction of the proton beam) direction, the moderator assembly is placed in direct contact with the head wall. The patient is assumed to be lying on a table, facing the ceiling, approximately 1 m above the floor. The long axis of the patient’s body is parallel to the left and right walls, and is centered between the left and right walls. In the unshielded moderator/reflector assembly design, the moderator was modeled as a solid right circular cylinder of fluental, 31 cm in diameter and 30 cm in length. A 31 cm thick lead fluoride reflector surrounds the moderator on its radial surface and a right circular cylindrical annulus of lead fluoride with inner diameter 26 cm, outer diameter 93 cm and length 31 cm surrounds the beam port directly upstream of the moderator. A beam delimiter surrounding the patient treatment port
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Fig. 1. Cross-sectional view of basis treatment room design.
was modeled as a right circular cylindrical annulus of Li2CO3 with inner diameter 25 cm, outer diameter 93 cm, and length 11 cm. Finally, a 0.42 cm thick, 6Li enriched flex panel covers the moderator and beam delimiter on the patient side of the moderator assembly. Fig. 3 is a cross-sectional view of the shielded moderator/reflector assembly design evaluated in the MCNP calculations. The shielded moderator assembly design is similar to the unshielded moderator assembly except that the radial surface of the moderator assembly, from the front of the moderator to the treatment room head wall, is surrounded by 10 cm of 30% borated polyethylene. The production of 480 keV gamma rays from the 10B(n,a)7Li reaction is a significant disadvantage when using boron as a shielding material for a treatment room for BNCT. These gammas not only increase patient dose, but also may preclude the use of gamma spectroscopy for evaluating the effectiveness of
the delivery of the boron compound used in BNCT, to the tumor location. For these reasons, the front face of the moderator assembly, excluding the opening of the patient treatment port, is lined with 10 cm of lithiumdoped polyethylene, instead of 30% borated polyethylene. Unlike 10B, 6Li does not produce any capture gamma rays and can therefore be used in close proximity to the patient without contributing significantly to patient dose, and without contributing to 480 keV gamma contamination that would interfere with the gamma spectroscopy measurements. 2.3. MCNP source term The production of neutrons in a 7Li target was characterized by a FORTRAN code that calculated the neutron yield as a function of energy and emission angle from 2.5 MeV protons perpendicularly incident on a
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Fig. 2. Cross-sectional view of unshielded moderator/HRS/target assembly.
Fig. 3. Cross-sectional view of the shielded moderator/reflector assembly design.
ARTICLE IN PRESS A.E. Hawk et al. / Applied Radiation and Isotopes 61 (2004) 1027–1031
thick 7Li target (Wang et al., 1989). The code was validated through experiments performed at the OSU Van de Graaff accelerator (Wang et al., 1989). The neutron source was modeled as a uniformly distributed disk source with diameter 25 cm, centered on the moderator assembly long axis. The source was placed in direct contact with the Glidcop HRS, 1.25 cm upstream of the moderator. It was assumed that neutrons were produced in the 7Li target 2000 h a year, and that a radiation worker was located a distance of 1 m from the outer walls of the treatment room for the entire 2000 h.
3. Results/conclusions The thicknesses of concrete necessary to protect radiation personnel located outside the treatment room, at the location of the maximum dose rate, are presented in this section. Fig. 4 is a plot of the effective dose rate at the location of the maximum dose rate versus concrete thickness for each case evaluated. The highest dose rate (for any wall thickness) to radiation personnel standing outside the treatment room, occurs for the basis treatment room design with an unshielded moderator/ reflector assembly design (Case 1); while the lowest dose rate (for any wall thickness) to radiation personnel standing outside the treatment room during a BNCT treatment, occurs for the revised treatment room with a shielded moderator/reflector assembly design (Case 4).
Fig. 4. Effective dose rate outside left wall versus concrete thickness for 4 evaluated cases: (Case 1) a basis treatment room with an unshielded moderator assembly; (Case 2) a revised treatment room with an unshielded moderator assembly; (Case 3) a basis treatment room with a shielded moderator assembly; and (Case 4) a revised treatment room with a shielded moderator assembly.
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From Fig. 4, one observes that a shielded moderator/ reflector assembly design (Case 2), provides a significant reduction in dose to radiation personnel located outside a BNCT treatment room as compared with an unshielded moderator/reflector assembly design (Case 1). Furthermore, the shielded moderator/reflector assembly design (Case 2) also provides greater personnel protection than the basis moderator/reflector assembly design located in a shielded treatment room (Case 3). The additional reduction in personnel dose achieved by shielding the treatment room as well as the moderator/ reflector assembly design (Case 4), while certainly advantageous, is less than the personnel dose reduction observed when comparing the shielded moderator/ reflector assembly in an unshielded treatment room (Case 2) with either the basis moderator/reflector assembly in an unshielded treatment room (Case 1), or the basis moderator/reflector assembly in a shielded treatment room (Case 3).
References Briesmeister, J.F., (Ed.), 2000. MCNP—A general Monte Carlo N-particle transport code, Version 4C. LA-13709-M, Los Alamos National Laboratory. Cristy, M., Eckerman, K.F., 1987. Specific absorbed fractions of energy at various ages from internal photon sources. I. Methods, Oak Ridge National Laboratory, ORNL/TM8381/V1. Evans, J.F., 1994. Shielding design of the treatment room for an accelerator neutron source for BNCT. Thesis, The Ohio State University. Evans, J., Blue, T.E., 1996. Shielding design of a treatment room for an accelerator-based epithermal neutron irradiation facility for BNCT. Health Phys. 71, 692–699. International Commission on Radiological Protection, 1991. 1990 Recommendations of the International Commission of Radiological Protection, ICRP Publication 60, Pergamon Press, Oxford. Metzger, R., Richardson, R., Van Riper, K.A., 1993. A Monte Carlo model for retrospective analysis of shield design in a diagnostic X-ray room. Health Phys. 65 (2), 164–171. Orr, M.T., Blue, T.E., Woollard, J.E., 2001. Using DORT to improve the moderator assembly design for the OSU accelerator-based neutron source for boron neutron capture therapy. Proceedings of the Embedded Topl. Mtg. Accelerator Applications/Accelerator Driven Transmutation Technology and Applications 01, Reno, NV, November 11–15, 2001, American Nuclear Society (CD-ROM). Qu, Tanxia, Blue, T.E., Herminghuysen, K., Kanellitsas, C., Gahbauer, R., 1993. Monte Carlo shielding analysis for an accelerator epithermal neutron irradiation facility (AENIF) in a conventional X-ray irradiation room. In: Soloway A.H., et al. (Eds.), Advances in Neutron Capture Therapy. Plenum Press, New York. Wang, C-K.C., Blue, T.E., Gahbauer, R.A., 1989. A neutronic study of an accelerator-based neutron irradiation facility for boron neutron capture therapy. Nucl. Tech. 84, 93–107.
Applied Radiation and Isotopes 61 (2004) 1027–1031
A shielding design for an accelerator-based neutron source for boron neutron capture therapy A.E. Hawk, T.E. Blue*, J.E. Woollard Nuclear Engineering Program, The Ohio State University, 206 West, 18th Avenue, Columbus, Oh 43210-1189, USA
Abstract Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a 7Li target, a target heat removal system (HRS), a moderator assembly, and a treatment room. The intent of this paper is to demonstrate the advantages of a shielded moderator assembly design, in terms of material requirements necessary to adequately protect radiation personnel located outside a treatment room for BNCT, over an unshielded moderator assembly design. r 2004 Elsevier Ltd. All rights reserved. Keywords: Boron neutron capture therapy (BNCT); Treatment room; Shielding
1. Introduction/background Preliminary shielding studies for an accelerator based neutron source (ABNS) for boron neutron capture therapy (BNCT) were conducted at The Ohio State University by Qu et al. (1993), Evans (1994) and Evans and Blue (1996). Qu et al. (1993) used the Monte Carlo N-Particle Transport Code (MCNP) (Briesmeister, 2000) to accurately model a 6 MV linear accelerator Xray irradiation room located at The Arthur James Cancer Hospital and evaluate whether this design provided adequate personnel protection. He concluded that it did not. Evans (1994) used the computer code MCNP to calculate the amount of ordinary concrete that would be necessary to provide adequate protection for radiation personnel standing outside a treatment room for an ABNS for BNCT. A radiation worker, according to Evans, was adequately protected when the transmitted effective dose to that worker is at least one order of magnitude less than the occupational effective *Corresponding author. E-mail address: [email protected] (T.E. Blue).
dose limit as recommended by the International Commission on Radiological Protection (ICRP) (1990). Both studies assumed that neutrons were produced in a 7Li target, and were moderated in energy by a moderator assembly. However, the moderator modeled by Qu et al. (1993) and assumed by Evans (1994) has evolved with time to consist of a fluental/lead fluoride moderator/reflector moderator assembly. Hence, the shielding requirements for the present moderator/reflector assembly differ from those of past moderator assemblies. Furthermore, Evans (1994) concluded that the primary component of the effective dose to radiation personnel located outside the treatment room was due to photons produced from inelastic neutron scatter and neutron absorption in the concrete walls. Therefore, the most efficient method to adequately protect radiation personnel, in terms of material requirements, is to prevent neutrons from entering the treatment room walls. There are three ways to prevent neutrons from entering the treatment room walls. Option 1 is to line the inner surface of the treatment room with a neutron absorbing material. Option 2 is to line the outer surface
0969-8043/$ - see front matter r 2004 Elsevier Ltd. All rights reserved. doi:10.1016/j.apradiso.2004.05.038
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of the moderator assembly with a neutron absorbing material, and Option 3 is to use options 1 and 2 in combination. Evans and Blue (1996) evaluated option 1 and concluded that a significant reduction in the amount of concrete, which is necessary to protect radiation personnel located outside the treatment room, could be achieved by lining the inner surface of the treatment room walls with a thin layer of gypsum wall board, borated flex panel and lead. The lead was included in the treatment room model to prevent gamma rays produced in the concrete walls from entering the treatment room. Furthermore, Evans and Blue (1996) found that if the primary shielding material, concrete, was changed to a denser material (to better attenuate the photon fluence produced in the treatment room walls), specifically barytes concrete, a further reduction in the thickness of the walls could be achieved. This paper evaluates shielding options 2 and 3. For option 2, the outer surface of the moderator assembly is surrounded by borated and lithium-doped polyethylene. It is assumed for option 3 (a combination of options 1 and 2), that for option 1, the inner surface of the treatment room is lined with gypsum wallboard, borated flex panel, and a thin (1.58 mm) layer of lead.
2. Methods The whole body effective dose to radiation personnel located directly outside the treatment room during patient treatment, is calculated for four moderator assembly/treatment room configurations: (Case 1) a basis treatment room with an unshielded moderator assembly; (Case 2) a revised treatment room with an unshielded moderator assembly; (Case 3) a basis treatment room with a shielded moderator assembly; and (Case 4) a revised treatment room with a shielded moderator assembly. The whole body effective dose to radiation personnel is calculated by modeling the appropriate moderator assembly/treatment room configuration using the MCNP version 4C computer code. Each MCNP model consists of a geometric representation of the treatment room, the moderator assembly and heat removal system (HRS)/target assembly, and the patient phantom developed by Cristy and Eckerman (1987). The patient phantom is included in all four configurations to act as a particle scatterer/absorber. Each model also includes a neutron source term, neutron and photon tallies, and a variance reduction scheme to allow for accurate, as well as timely results. 2.1. Treatment room geometry Fig. 1 is a cross-sectional view of the basis treatment room evaluated in the MCNP calculations. The room is
4.27 m (14 ft) high with 3.05 m 3.66 m (100 120 ) floor dimensions. These dimensions were based on those of a previous shielding study of a diagnostic X-ray room (Metzger et al., 1993). The treatment room geometry is relatively simple; it is essentially a large rectangular parallelopiped. Doors, windows, mazes and ducts are not included in the model. Support structures for the moderator assembly as well as support structures in the walls, are also not included in the model. In the basis treatment room design, concrete and lead are the primary materials used to attenuate the neutron and photon flux in the walls of the room. Gypsum wallboard, which is commonly used in radiotherapy rooms, was used on the head (with respect to the patient phantom), right, foot, and left walls. It was assumed that the treatment room was located on the first floor of a hospital facility and, as a result, soil was included in the model below the floor. Gypsum wallboard and lead sheet were not used on either the ceiling or the floor of the basis treatment room. In the revised treatment room design, concrete and lead again were the primary shielding materials, and gypsum wallboard lined the interior (including the ceiling) of the treatment room. A thin (69.9 mm) layer of borated flex panel was placed behind the gypsum wallboard. The borated flex panel reduces the thermal neutron flux entering the concrete walls, thereby reducing the number of thermal neutron capture gammas produced in the concrete. The flex panel also reduces the backscattered thermal neutron flux (from the concrete walls) in the treatment room. 2.2. Moderator assembly geometry The moderator/reflector assembly used in all the calculations is the 30 cm fluental/lead fluoride moderator/reflector assembly (Fig. 2) designed at The Ohio State University (Orr et al., 2001). The long axis of the moderator assembly is positioned 1 m above the floor of the treatment room. In order to maximize the shielding in the upstream (with respect to the direction of the proton beam) direction, the moderator assembly is placed in direct contact with the head wall. The patient is assumed to be lying on a table, facing the ceiling, approximately 1 m above the floor. The long axis of the patient’s body is parallel to the left and right walls, and is centered between the left and right walls. In the unshielded moderator/reflector assembly design, the moderator was modeled as a solid right circular cylinder of fluental, 31 cm in diameter and 30 cm in length. A 31 cm thick lead fluoride reflector surrounds the moderator on its radial surface and a right circular cylindrical annulus of lead fluoride with inner diameter 26 cm, outer diameter 93 cm and length 31 cm surrounds the beam port directly upstream of the moderator. A beam delimiter surrounding the patient treatment port
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Fig. 1. Cross-sectional view of basis treatment room design.
was modeled as a right circular cylindrical annulus of Li2CO3 with inner diameter 25 cm, outer diameter 93 cm, and length 11 cm. Finally, a 0.42 cm thick, 6Li enriched flex panel covers the moderator and beam delimiter on the patient side of the moderator assembly. Fig. 3 is a cross-sectional view of the shielded moderator/reflector assembly design evaluated in the MCNP calculations. The shielded moderator assembly design is similar to the unshielded moderator assembly except that the radial surface of the moderator assembly, from the front of the moderator to the treatment room head wall, is surrounded by 10 cm of 30% borated polyethylene. The production of 480 keV gamma rays from the 10B(n,a)7Li reaction is a significant disadvantage when using boron as a shielding material for a treatment room for BNCT. These gammas not only increase patient dose, but also may preclude the use of gamma spectroscopy for evaluating the effectiveness of
the delivery of the boron compound used in BNCT, to the tumor location. For these reasons, the front face of the moderator assembly, excluding the opening of the patient treatment port, is lined with 10 cm of lithiumdoped polyethylene, instead of 30% borated polyethylene. Unlike 10B, 6Li does not produce any capture gamma rays and can therefore be used in close proximity to the patient without contributing significantly to patient dose, and without contributing to 480 keV gamma contamination that would interfere with the gamma spectroscopy measurements. 2.3. MCNP source term The production of neutrons in a 7Li target was characterized by a FORTRAN code that calculated the neutron yield as a function of energy and emission angle from 2.5 MeV protons perpendicularly incident on a
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Fig. 2. Cross-sectional view of unshielded moderator/HRS/target assembly.
Fig. 3. Cross-sectional view of the shielded moderator/reflector assembly design.
ARTICLE IN PRESS A.E. Hawk et al. / Applied Radiation and Isotopes 61 (2004) 1027–1031
thick 7Li target (Wang et al., 1989). The code was validated through experiments performed at the OSU Van de Graaff accelerator (Wang et al., 1989). The neutron source was modeled as a uniformly distributed disk source with diameter 25 cm, centered on the moderator assembly long axis. The source was placed in direct contact with the Glidcop HRS, 1.25 cm upstream of the moderator. It was assumed that neutrons were produced in the 7Li target 2000 h a year, and that a radiation worker was located a distance of 1 m from the outer walls of the treatment room for the entire 2000 h.
3. Results/conclusions The thicknesses of concrete necessary to protect radiation personnel located outside the treatment room, at the location of the maximum dose rate, are presented in this section. Fig. 4 is a plot of the effective dose rate at the location of the maximum dose rate versus concrete thickness for each case evaluated. The highest dose rate (for any wall thickness) to radiation personnel standing outside the treatment room, occurs for the basis treatment room design with an unshielded moderator/ reflector assembly design (Case 1); while the lowest dose rate (for any wall thickness) to radiation personnel standing outside the treatment room during a BNCT treatment, occurs for the revised treatment room with a shielded moderator/reflector assembly design (Case 4).
Fig. 4. Effective dose rate outside left wall versus concrete thickness for 4 evaluated cases: (Case 1) a basis treatment room with an unshielded moderator assembly; (Case 2) a revised treatment room with an unshielded moderator assembly; (Case 3) a basis treatment room with a shielded moderator assembly; and (Case 4) a revised treatment room with a shielded moderator assembly.
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From Fig. 4, one observes that a shielded moderator/ reflector assembly design (Case 2), provides a significant reduction in dose to radiation personnel located outside a BNCT treatment room as compared with an unshielded moderator/reflector assembly design (Case 1). Furthermore, the shielded moderator/reflector assembly design (Case 2) also provides greater personnel protection than the basis moderator/reflector assembly design located in a shielded treatment room (Case 3). The additional reduction in personnel dose achieved by shielding the treatment room as well as the moderator/ reflector assembly design (Case 4), while certainly advantageous, is less than the personnel dose reduction observed when comparing the shielded moderator/ reflector assembly in an unshielded treatment room (Case 2) with either the basis moderator/reflector assembly in an unshielded treatment room (Case 1), or the basis moderator/reflector assembly in a shielded treatment room (Case 3).
References Briesmeister, J.F., (Ed.), 2000. MCNP—A general Monte Carlo N-particle transport code, Version 4C. LA-13709-M, Los Alamos National Laboratory. Cristy, M., Eckerman, K.F., 1987. Specific absorbed fractions of energy at various ages from internal photon sources. I. Methods, Oak Ridge National Laboratory, ORNL/TM8381/V1. Evans, J.F., 1994. Shielding design of the treatment room for an accelerator neutron source for BNCT. Thesis, The Ohio State University. Evans, J., Blue, T.E., 1996. Shielding design of a treatment room for an accelerator-based epithermal neutron irradiation facility for BNCT. Health Phys. 71, 692–699. International Commission on Radiological Protection, 1991. 1990 Recommendations of the International Commission of Radiological Protection, ICRP Publication 60, Pergamon Press, Oxford. Metzger, R., Richardson, R., Van Riper, K.A., 1993. A Monte Carlo model for retrospective analysis of shield design in a diagnostic X-ray room. Health Phys. 65 (2), 164–171. Orr, M.T., Blue, T.E., Woollard, J.E., 2001. Using DORT to improve the moderator assembly design for the OSU accelerator-based neutron source for boron neutron capture therapy. Proceedings of the Embedded Topl. Mtg. Accelerator Applications/Accelerator Driven Transmutation Technology and Applications 01, Reno, NV, November 11–15, 2001, American Nuclear Society (CD-ROM). Qu, Tanxia, Blue, T.E., Herminghuysen, K., Kanellitsas, C., Gahbauer, R., 1993. Monte Carlo shielding analysis for an accelerator epithermal neutron irradiation facility (AENIF) in a conventional X-ray irradiation room. In: Soloway A.H., et al. (Eds.), Advances in Neutron Capture Therapy. Plenum Press, New York. Wang, C-K.C., Blue, T.E., Gahbauer, R.A., 1989. A neutronic study of an accelerator-based neutron irradiation facility for boron neutron capture therapy. Nucl. Tech. 84, 93–107.