Neutronics Design of a Beam Shaping Assembly in an Accelerator-Based Neutron Source for the Boron Neutron Capture Therapy System at the University of Osaka

Featured Application

The newly designed beam shaping assembly (BSA) achieved in the present study will be applied in accelerator-based BNCT machines employing a p-Li reaction-based neutron source at the University of Osaka, Japan. It could become the standard BSA design for the p-Li-based accelerator-based BNCT system in the future.

Abstract

An accelerator-based neutron source (ABNS) for boron neutron capture therapy (BNCT) is under development at the University of Osaka, Japan. It can supply a sufficiently intense epi-thermal neutron flux of ~8 × 10⁸ n/s/cm² and, at the same time, enable the whole-body dose to be substantially suppressed (~0.26 Sv/irradiation) by employing a p-Li reaction as a neutron emission reaction. At present, we are designing all the equipment to be implemented in the BNCT machine. Previously, prior to the design of the real machine, we carried out mock-up experiments to demonstrate epi-thermal neutron production and check our neutronics design code system with a prototype beam shaping assembly (BSA). The series of experiments was conducted with a Dynamitron accelerator at Birmingham University, UK. Based on the experimental results, the first neutronics design of the real machine was completed. The design result was described, together with details of the neutronics design goal and design code system. Now, our p-liquid Li-based ABNS-BNCT system is planned to be constructed in the Institute of Free Electron Laser at the University of Osaka, Osaka, Japan.

1. Introduction

Boron neutron capture therapy (BNCT) is a promising and up-to-date cancer therapy that is currently under development worldwide. BNCT is a kind of particle therapy employing the following ¹⁰B(n,α)⁷Li reaction:

$$\begin{array}{lll}{}^{10}\mathrm{B}+\mathrm{n}& \to {}^7\mathrm{Li}+\mathrm{\alpha }+2.79\mathrm{MeV}& (6.3\%)\\ & \to {}^7\mathrm{Li}^{*}+\mathrm{\alpha }+2.31\mathrm{MeV}& (93.7\%)\end{array}$$

(1)

After a drug containing ¹⁰B is administered to accumulate ¹⁰B in tumor cells, neutrons are irradiated from outside the human body. If the ¹⁰B-containing drug is accumulated only in tumor cells, the tumor cells can be killed. Two medicines are known as such a drug: L-p-boronophenylalanine (BPA) and disodium mercaptoundecahydrododecaborate (BSH). Using these drugs, tumor-cell-selective treatment can be administered uniquely among radiation therapies. In the past, BNCT was carried out in experimental nuclear reactors because an intense neutron flux of around 1 × 10⁹ n/s/cm² was required. However, it was commonly recognized that the use of BNCT could not increase if it relied only on a nuclear reactor as the neutron source because the construction of nuclear reactors inside or close to hospitals is limited or prohibited in many countries, including Japan. Under these circumstances, new accelerator-based neutron sources (ABNSs) have been developed worldwide [1]. The primary goal is to obtain a strong epi-thermal (0.5~10 keV) neutron source, the intensity of which is ~1 × 10⁹ n/s/cm² at the human body surface near the tumor [2,3]. At the same time, unwanted background radiation should be suppressed to avoid the normal tissue dose of a patient. This should finally be at a level similar to that of nuclear reactor-based BNCT. As neutron source targets of the ABNS, beryllium and lithium are commonly considered as the most promising materials. In Japan, several projects are underway that aim to produce a real ABNS-BNCT system. At present, nuclear reactions of p-Be, p-Li, and p-liquid Li are being examined in several institutes, including the University of Osaka. Recently, a Kyoto University group successfully acquired Japanese governmental approval for their own ABNS-BNCT [4,5]. Their system is based on a cyclotron accelerator using a p-Be reaction [6].

At the University of Osaka, the authors’ group is carrying out research and development of their own ABNS for BNCT. The project was press-released in September 2013 in Japan. The project was supported initially by Sumitomo Corporation, Tokyo, Japan, and Mitsubishi Heavy Industries Mechatronics Systems, Ltd., Kobe, Hyogo, Japan. The outline of the project was presented at ICNCT-16, held in Helsinki, Finland, in 2014 [7]. The main specification of the ABNS-BNCT system at the University of Osaka is shown in

Table 1. Main specification of the ABNS-BNCT system at the University of Osaka.

Epi-thermal neutron flux	~1 × 109 n/cm2/s
Accelerator	Electrostatic type
Proton energy	2.65 MeV
Beam current	30 mA
Target	Liquid lithium
Velocity	30 m/s at maximum
Temperature	450 °C at maximum

. The system at the University of Osaka employs an electrostatic accelerator that produces low-energy neutrons of ~10¹³ n/s via a proton liquid lithium reaction, ⁷Li(p,n)⁷Be. We specifically use a liquid lithium target in order to avoid severe problems commonly expected when using a solid lithium target, i.e., seriously high ⁷Be and tritium accumulation in the target and difficulty in heat removal. Also, the emitted neutron energy is low, especially compared to the beryllium target. As a result, in the case of the p-Li reaction, the intensity of neutron-induced secondary gamma-rays can also be kept low, and the whole-body dose can thus be suppressed. For tritium production in the target, very few problems are expected during the final disposal of the machine because the lithium inventory of the liquid lithium target is very much smaller compared to the solid lithium target, meaning that, as a result, the specific radioactivity can be suppressed to be very low. At present, the liquid lithium target is also being studied at the Tokyo Institute of Technology, Japan [8], and the Hebrew University, Israel [9]. The latter carried out a validation experiment on the liquid lithium target, which gives hope to our project.

In the present paper, the neutronics design goal and design code system for the ABNS-BNCT system at the University of Osaka are described. In addition, the summarized results of the mock-up experiments carried out at Birmingham University, Birmingham, UK, for the verification of the design code system are discussed. Finally, the neutronics design result of the system is detailed.

2. Materials and Methods

2.1. Neutronics Design Goal and the Code System

2.1.1. Design Goal

The neutronics design goal of the ABNS-BNCT system at the University of Osaka is essentially based on the well-known criteria proposed by IAEA [2,3], as follows:

• Epi-thermal neutron flux: >0.5 × 10⁹ n/s/cm².
  Based on this value, it is known that irradiation in BNCT will be completed in about 30 min to 1 h. As shown in Table 1, in our design, a beam current of 30 mA is necessary to obtain a neutron flux intensity of around 1 × 10⁹ epi-thermal neutrons/s/cm².
• Epi-thermal-to-thermal neutron flux ratio: >20.
  There is no clear reason for determining this value in the IAEA document. However, this should be as low as possible to reduce the dose of normal tissues that exist in positions closer to the neutron source when viewed from the tumor cells.
• Fast neutron and secondary gamma-ray contribution: <2 × 10⁻¹³ Gy·cm².
  It is normally difficult to achieve this value because the epi-thermal neutron flux may usually decrease when trying to meet this criterion. Few facilities meet both of these values perfectly. For these reasons, in the latest document for BNCT design compiled by IAEA, as described a little later [3], this value was relaxed to be 7 × 10⁻¹³ Gy·cm².
• Neutron current to flux ratio: >0.7.
  This is thought to be a simple ratio of neutron current to flux at the exit of the epi-thermal neutron source, though it is not obviously defined in the IAEA document [2,3]. In the case of the ABNS, it may become harder to meet this at the flat surface of the assembly. However, this may be a rather important value to really suppress the normal tissue dose.

The above criteria were originally established for nuclear reactor-based BNCT; however, their application in ABNS-based BNCT was recently questioned. For ABNS-BNCT, a precise discussion was carried out in 2020 [1], and the new TECDOC for BNCT was written by many BNCT researchers and finally released by the IAEA [3]. However, in the present study, we took into consideration the former criteria in the design because it is set to be conservative (safer) to realize the good performance of nuclear reactor-based BNCT.

In addition, our own targets were set, especially for decreasing the whole-body dose and realizing the safe maintenance and final disposal of the machine, as follows:

• Whole-body dose: <0.25 Sv/irradiation.
  This is quite an important factor that many medical doctors using BNCT request. However, in the past, this was neither strictly considered nor fixed. In the case of ABNS, it is difficult to shape the neutron beam appropriately, and wrap-around radiations are normally not negligible. The whole-body dose, as an important design criterion, was thus introduced. This is our own target value; however, it would also be valuable information for other ABNS-BNCT facilities.
• Tritium production: <100 Bq/g-Li/year.
  A fairly large amount of tritium can be produced and accumulated not only in p-Li but also in p-Be targets. This value may be difficult to achieve, especially for solid targets. In the case of a liquid lithium target, tritium can be diluted in a large amount of liquid lithium, and the concentration can thus be suppressed and, finally, kept low. This value is the same as the clearance level of tritium, 100 Bq/g-Li.
• Material activation: <Clearance level of each radioisotope.
  This is set to realize the final disposal of the ABNS-BNCT facility. In our system, this design value is considered for all used materials except the structural materials near the target. This is quite crucial, especially for a commercially available ABNS-BNCT. Unlike nuclear reactors, we can account for only radioisotopes with half-lives of several tens to one hundred years at the longest when disposing of it because the operating period is not so long and the total fluence is not so large. Consequently, the amount of such long half-lived radioisotopes that are created can be suppressed to fairly few, i.e., hopefully less than the clearance level, even at the end of life.

2.1.2. Design Code System

The neutron and gamma-ray transport calculations are performed with a general-purpose Monte Carlo code, MCNP5 [10]. In the present design, the calculation model is made manually. In the design of neutron and gamma-ray fluxes, their doses and the activities of materials are evaluated by tallies of F2, F6, and F4, respectively. The base nuclear data library is JENDL-5 [11]. The neutron source term of the ⁷Li(p,n) reaction is basically taken from DROSG-2000 [12], in which the absolute intensity and angular distribution of emitted neutrons are partly obtained experimentally, as shown later. For material activations, produced radioactivities, including tritium, are evaluated by taking into account the position-dependent neutron spectra to examine the feasibility of the final disposal of the machine. In addition, we specifically estimate the whole-body dose, even considering the equipment in the irradiation room.

2.2. Validation of the Neutronics Design Code System with a Mock-Up Assembly

To realize liquid lithium target-based neutron sources for BNCT, precise experimental tests are required before the development of the real machine to confirm the validity of the neutronics design code system for the BSA. For other equipment, like a particle accelerator and a liquid lithium loop, the accelerator is commercially available, meeting the required performance of our ABNS-BNCT [13]. For the lithium loop, the University of Osaka has a proven technology to treat liquid lithium originally developed for fusion reactor development [7]. For the neutronics design of the BSA, the code system should be examined carefully because it directly affects important performance metrics, like the exposure dose to the patient.

After the conceptual design of the ABNS-BNCT system, source term measurements were carried out with a solid lithium target at Tohoku University, Japan [14] because the absolute intensity and angular distribution of emitted neutrons are critically important for the design. Thereafter, we constructed a mock-up system of the BSA and conducted experiments at Birmingham University, UK, to confirm the validity of the design code system in order to start the real machine design. The summarized results are given in Section 3.1.

The practical purpose of the series experiments includes (1) demonstrating the epi-thermal neutron production as a p-Li neutron source target, (2) confirming that the neutron and gamma-ray doses to the human body are kept substantially low, and (3) validating the design code system wholly for the design of the real ABNS for BNCT by estimating the design accuracy based on the experimental results.

Experimental Condition and Mock-Up Assembly

The experiments were carried out at Birmingham University, UK. A Dynamitron accelerator of Prof. S. Green’s laboratory was utilized. The proton energies were 1.95, 2.25, and 2.65 MeV, and the beam current was several hundred μA at the target. The target was a 1 mm thick lithium sheet on a copper backing cooled with heavy water. The neutron source intensity was about 2 × 10¹¹ n/s. The irradiation time was several hours for each experiment.

The design of the mock-up assembly was performed basically with the same system as the real machine design, as detailed in Section 3.2. The assembly constructed at Birmingham University is shown in Figure 1. In the moderator assembly, heavy water, in addition to aluminum fluoride, is employed to moderate and filter neutrons. Generally speaking, source intensities in ABNS-BNCT are not so strong at present compared to reactor-based BNCT. Heavy water is thus added to save space, i.e., the distance to the patient can be kept shorter in order to make the intensity large. As a result, the spectrum can be adjusted according to users’ requests by changing the combination ratio of the two materials; for example, the neutron spectrum can be changed from a soft spectrum, similar to the nuclear reactor-based BNCT at KUR [15], to a hard spectrum, as realized at the ABNS at C-BENS [5]. Graphite is used as a reflector instead of lead from the standpoint of the structure’s weight. In Birmingham, the assembly could not be set inside the shield wall of the irradiation room, unlike the real machine. The assembly itself is thus covered with boric acid water tanks to suppress neutron leakage from the side and upper walls of the assembly. Under the moderator material, another boric acid water layer and a resin sheet containing boron are set to collimate the neutron beam. In the center, a cadmium layer is placed instead of the resin sheet to cut thermal neutrons. At the bottom, a lead layer is positioned to attenuate secondary gamma-rays. Enough space is kept under the assembly so that epi-thermal neutron irradiation experiments with a human body phantom can be carried out. The phantom was composed of acrylic boxes with sides of 10 cm to 20 cm and filled with water.

3. Results and Discussion

3.1. Radiation Measurements

The main detectors in the mock-up experiments include gold foils and a radio-photoluminescence glass dosimeter (RPLGD) [16,17] to measure three-dimensional neutron flux intensity and gamma-ray dose, respectively. The obtained results were compared with the numerical prediction to discuss the validity of the design code system. Supplementarily, an ionization chamber and pocket dosimeters were used. In addition, tritium production yield in the heavy water and gamma-ray spectrum outside the assembly with a NaI scintillation detector were measured. Moreover, small sample pieces of structural materials and spectrum index foils were irradiated and measured to evaluate the final disposal possibility of the real BNCT machine. In the present paper, the results of the neutron flux measurements mainly used in the validation examination are given. Other results are found elsewhere [18].

Measurements were carried out for three energies, namely, 1.95, 2.25, and 2.65 MeV, with and without a human body phantom. Here, the brief results for the 2.65 MeV case with an average beam current of 236 μA are described. Other energies were adopted only for the source term checking. The summarized results are as follows: Based on the comparison of induced radioactivities of over 200 gold foils, with calculations in the case without a phantom, the design accuracy is estimated to be less than 20% from the frequency distribution of C/Es [18]. Just under the assembly, where the phantom was set as shown in Figure 1, the agreement of measurement and calculation is excellent, i.e., the C/E of the foil activity is almost unity (=1.007). In addition, the cadmium ratio is also very close to unity (=0.996), showing that the thermal neutron contribution is suppressed very well. At this place, the epi-thermal neutron flux is evaluated to be the largest value of 6.36 × 10⁶ n/cm²/s. This was estimated by a measured result with a gold foil and Cd filter combined with calculation, because both the C/E and Cd ratio show almost unity, meaning, even indirectly, that the agreement in the epi-thermal neutron flux intensity between the measurement and calculation had to be excellent. At the same height, the measured radial (horizontal) epi-thermal neutron flux distribution shows an excellent collimation performance, i.e., the FWHM is 15.9 cm in the case that the moderator diameter is 20 cm. This was estimated by the gold foil activity distribution. The FWHM can also be predicted by calculation; its value is 19.9 cm, which is a little larger than the experimental result. The final performance of epi-thermal flux intensity is estimated to be 2.69 × 10⁴ n/cm²/μC, that is, 8.08 × 10⁸ n/s/cm² at 30 mA, which is the beam current spec of our accelerator shown in Table 1. Consequently, from the series experiments and comparison with the calculations using the present mock-up assembly, the validity of the neutronics design code system was confirmed.

3.2. Neutronics Design of the ABNS-BNCT System

The neutronics design of the BSA of the ABNS-BNCT system at the University of Osaka was carried out based on the results obtained in the mock-up experiment. The structure is thus quite similar to that of the Birmingham University experiment. Figure 2 shows the design result of the present BSA, which has the following specific features:

• The proton beam direction is the same as that of the epi-thermal neutrons. The BSA can thus be built into the wall. As a result, wrap-around radiations can be suppressed substantially. Hence, as shown in Figure 2, none of the boric acid solution tanks described in Figure 1 are necessary.
• The neutron moderator is a mixture of aluminum fluoride and heavy water. By changing the thicknesses, the spectrum can be adjusted in the epi-thermal energy region so that the peak energy can be shifted to control (1) the number of neutrons that remain in the high-energy region (larger than epi-thermal energy) and (2) the number of neutrons that move from the epi-thermal energy region to the thermal energy region. Also, by varying the moderator radius, the neutron beam diameter can be adjusted depending on tumor size.

3.2.1. Design Details

The neutronics design of the BSA of the ABNS-BNCT system at the University of Osaka was carried out with a general-purpose three-dimensional Monte Carlo code MCNP-5 [9], which is the same as the system used for the design of the mock-up assembly constructed at Birmingham University. For the neutron source intensity and angular distribution of emitted neutrons, the measured result by the authors’ group at FNL of Tohoku University [14] was used, combined with the experimental results obtained at Birmingham University [18]. For the neutron energy spectrum, a database compiled by the IAEA, DROSG-2000 [12], was utilized.

As shown in Figure 2, protons are incident from the upper floor to the Li target. On the upper floor, an electrostatic accelerator is installed. The Li target is positioned at around the center of the assembly. Because the Li loop design is currently underway, the preliminary structure is included in this design. The current status of the lithium loop design is as follows: The loop is like a slide, and after passing through the target position, the slide turns to the left or right according to the surrounding structure of the facility. This curving structure is necessary to realize a stable lithium flow, and, in addition, it also plays an important role in suppressing the streaming phenomenon of produced neutrons.

As shown in Figure 2, the Li target is surrounded by a graphite layer. Lead is also effective, as is well known; however, lead is too heavy in practical applications because, as mentioned later, the BSA system is planned to be installed in the ceiling. Graphite has an important role as a neutron reflector and moderator. Compared to another high-performance moderator of water or heavy water, neutrons are not over-moderated in graphite. Under the target, aluminum fluoride and heavy water are placed as the main moderator. By using aluminum fluoride and water (heavy water), the neutron energy can be controlled appropriately. Generally, epi-thermal neutrons can be obtained only with fluoride moderators; however, the spectrum becomes a little harder. Normally, as a result, the contribution of high-energy neutrons with higher energies than epi-thermal neutron energy becomes large; for instance, in the case of C-BENS [5]. If aiming at a soft spectrum, the moderator must be thick, while the flux intensity deteriorates. To appropriately make the spectrum soft, lighter atoms should be used. In the present design, heavy water was thus used. However, if only using heavy water, over-moderation occurs. And as a result, neutrons escape to the thermal neutron energy region, and, thus, the intensity decreases. From the above discussion, a mixed moderator of fluoride (aluminum fluoride) and deuteride (heavy water) was finally employed in order (1) to make the peak energy at around the center of the epi-thermal neutron energy region, (2) to decrease high-energy neutrons, and (3) to suppress the loss of neutrons escaping to the thermal neutron energy region as much as possible. As mentioned earlier, using deuteron instead of hydrogen, over-moderation can effectively be suppressed, and, in addition, unnecessary capture of neutrons can be avoided because the cross section of the ²H(n,γ) reaction is much smaller than that of the ¹H(n,γ) reaction. The concept of combining two materials with different characteristics is quite valuable for designing epi-thermal neutron sources utilizing a low-energy neutron source, like a p-Li reaction.

At the bottom of the BSA, a lead shield is positioned to remove unwanted gamma-rays. The thickness was changed from 10 cm in the Birmingham design to 4 cm because a cadmium sheet, as a thermal neutron cutter just under the moderator, was changed to a boron resin sheet. This is very effective in reducing gamma-rays and keeping the epi-thermal neutron flux intensity higher. Under the lead layer, there is an irradiation space. This means that in our ABNS-BNCT, epi-thermal neutrons are incident to a patient from the ceiling of the irradiation room. As a result, a patient on a bed could be irradiated with epi-thermal neutrons from various directions by moving the bed so that the unnecessary dose to the normal tissues could be suppressed as much as possible, and, at the same time, a quasi-uniform irradiation to the tumor could be realized. Of course, the safety of a patient should be assured and is the first priority.

3.2.2. Performance of the Designed ABNS-BNCT

Table 2 shows the performance of the ABNS-BNCT system at the University of Osaka, designed under the support of the experimental validation results obtained at Birmingham University. The protocol of BNCT used for the performance estimation was given by I. Kato of the University of Osaka, Japan [19], who has carried out BNCT in more than 30 cases with nuclear reactor-based neutron sources of KUR and JRR-4 for longer than 10 years [20]. The epi-thermal neutron flux intensity is as large as 1 × 10⁹ n/s/cm², and the fast neutron and gamma-ray contributions to epi-thermal neutron flux are suppressed to an acceptably low level.

As for tritium production, using a lithium target, tritium is created by the ⁶Li(n,α)³H reaction. However, it is known that the amount of tritium released from a lithium metal is very low [21]. This means that tritium continues to be accumulated for the whole operating period. In this design, we aim to suppress the accumulated tritium production to be less than the clearance level of 100 Bq/g [22]. This is realized for one year of operation by covering the loop with a 0.5 mm thick Cd sheet using ⁷Li-enriched liquid lithium. This liquid lithium target system is a specific feature of our system.

Table 2. Performance of the ABNS-BNCT system at the University of Osaka.

Neutron flux	Thermal	5.8 × 106 n/s/cm2/30 mA
	Epi-thermal	8.1 × 108 n/s/cm2/30 mA
Dose	Tumor	20 Gy-eq
	Normal brain	4.1 Gy-eq
Contribution	Fast neutron	5.1 × 10−13 Gy·cm2
	Gamma-ray	2.7 × 10−13 Gy·cm2
Tritium produced in the target		Less than clearance level [22]
Activity in BSA material		Less than clearance level except 60Co at around the target
Whole body dose 1		0.26 Sv/irradiation

¹ The evaluation conditions given by Kato [19] are as follows: tumor: head and neck cancer; BPA concentration: 24 ppm (ave.); T/N ratio: 3.5 (ave.); mucosa dose: 12 Gy-e; irradiation time: 30 min.

Table 2. Performance of the ABNS-BNCT system at the University of Osaka.

Neutron flux	Thermal	5.8 × 106 n/s/cm2/30 mA
	Epi-thermal	8.1 × 108 n/s/cm2/30 mA
Dose	Tumor	20 Gy-eq
	Normal brain	4.1 Gy-eq
Contribution	Fast neutron	5.1 × 10−13 Gy·cm2
	Gamma-ray	2.7 × 10−13 Gy·cm2
Tritium produced in the target		Less than clearance level [22]
Activity in BSA material		Less than clearance level except 60Co at around the target
Whole body dose 1		0.26 Sv/irradiation

¹ The evaluation conditions given by Kato [19] are as follows: tumor: head and neck cancer; BPA concentration: 24 ppm (ave.); T/N ratio: 3.5 (ave.); mucosa dose: 12 Gy-e; irradiation time: 30 min.

For the activation of other materials, it was found that, basically, very few long half-lived radioactive isotopes are produced because, considering the relatively shorter operation period of 10 years compared to the nuclear reactor, we may have to account for relatively shorter half-lived radioisotopes. For that, an appropriate impurity control must be conducted. However, from the present design calculation, considering the real disposal of the system, especially the cobalt impurity level, should be carefully controlled so as to achieve less than the clearance level. Fortunately, in our ABNS, such special care should be given only to material containing iron near the target. Practically, the target value of the cobalt content is less than 1 ppm, and the current design meets this restriction, except near the target.

The most important value of the whole-body dose can successfully be suppressed to 0.26 Sv/irradiation, which can be regarded as an acceptably low dose by our medical doctors. In the estimation calculation, each organ is assigned approximately to a corresponding phantom cell of 20 cubic cm. Radiation and organ weighting factors were obtained from ICRP Publication 60 [23]. The RBE and CBE values used for the present dose estimation are described in

Table 3. RBE and CBE values used in this design.

	Mucosa	Normal Brain	Tumor
BPA	4.9	1.35	3.8
Nitrogen 1	3	3	3
Hydrogen	3	3	3
Gamma-ray	1	1	1

¹ Abundance is 2.6 wt% [24].

.

We are now carrying out the final and precise design of the ABNS-BNCT system at the University of Osaka, including an accelerator, lithium loop, BSA, and so on, with the support of United Neutron, Ltd. (Ibaraki, Osaka, Japan) [25]. The first machine is planned to be constructed in the Institute of Free Electron Laser at the University of Osaka, Osaka, Japan [26].

4. Conclusions

The authors’ group is developing an accelerator-based neutron source for the ABNS-BNCT system at the University of Osaka. Mock-up experiments demonstrating epi-thermal neutron production by the p-Li reaction were carried out, and our neutronics design code system for the new BSA of our BNCT system was checked. The series experiments were conducted using a Dynamitron accelerator at Birmingham University, UK. In the present paper, the neutronics design goal and design code system for the ABNS-BNCT system at the University of Osaka were described, and the verification results of the design code system using the mock-up experiments were briefly introduced. As a result of the comparison between the experiments and calculations, the validity of the design code system was confirmed. Then, the neutronics design of the BSA was carried out based on the mock-up experiments. Based on the results, the BSA design with an acceptable neutronics performance for BNCT was confirmed. We are currently performing the final and precise design of the BSA, including the Li loop, in parallel with designs of the accelerator and other equipment. The first machine is planned to be constructed in the Institute of Free Electron Laser at the University of Osaka, Osaka, Japan.