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The Nagoya University Accelerator-driven Neutron Source (NUANS) is an accelerator-based neutron source by 7Li(p,n)7Be reaction with a 2.8 MeV proton beam up to 15 mA. The fast neutrons are moderated and shaped to beam with a Beam Shaping Assembly (BSA). NUANS is aiming at the basic study of the Boron Neutron Capture Therapy (BNCT) such as an in vitro cell-based irradiation experiment using a water phantom. Moreover, the BSA is developed as a prototype of one for human treatment. We have evaluated the radiation field of NUANS by a Monte Carlo code PHITS. It is confirmed that the radiation characteristics at the BNCT outlet meet the requirement of IAEA TECDOC-1223. Additionally, the radiation field in the water phantom located just in front of the BSA outlet is calculated. In the in vitro irradiation experiment, the boron dose of 30 Gy-eq, which is the dose to kill tumor cells, is expected for 20 min of irradiation at the beam current of 15 mA.

The D(d, n)3He reaction is one of the common monoenergetic neutron sources. We compile the Frag Data table of D(d, n)3He reaction from the literature as an external cross-section data for PHITS. We confirm the validity of the Frag Data table by the calculation of the total and angular neutron yield calculations for an ideal deuterium thin target. Finally, PHITS with the Frag Data table is applied to the angular neutron yield and spectrum calculations of the gas target and the deuterium-loaded titanium target of the Tohoku University Fast Neutron Laboratory.

Boron neutron capture therapy (BNCT) is an innovative cancer therapy that selectively destroys only cancer cells by utilizing the reaction between boron agents, which selectively accumulate in cancer cells, and neutron beams. However, the irradiation field in BNCT is a mixture of multiple radiation types, making accurate dose evaluation difficult. For dose evaluation in cancer tissue, which is particularly important, it is essential to discriminate between neutrons and γ-rays mixed in the BNCT irradiation field and to measure them correctly. In this paper, we introduce a new method for selective measurement of neutrons and γ-rays using a thermoluminescent phosphor, which is expected to be a promising method for improving the accuracy of dose evaluations.

In the Large Helical Device (LHD) of the National Institute for Fusion Science, the experiments using deuterium plasma were performed from March 2017. Neutrons with the energy of 2.45 MeV are generated by d (d, n) 3He reactions. The evaluation of the neutron flux in the torus hall as well as the LHD vessel is very important for designing the decommissioning of the LHD in the future, and the radiation shielding for workers and components around LHD. In this study, we performed the activation experiments using multiple activation foils to obtain a neutron spectrum in the vacuum vessel of the LHD. The reaction rate of each activation foil was generally consistent with the calculation result by MCNP6. As a result of the unfolding with SAND-II code, we experimentally obtained the neutron spectrum in the vacuum vessel of the LHD for the first time.

A simple dose calculation tool, SiDE, was developed for dose evaluation in a water phantom for boron neutron capture therapy, which makes the calculation time much shorter compared with the conventional particle transportation Monte Carlo codes and is applicable to any type of incident neutron spectra to the phantom. As the SiDE can not only calculate quantitatively the dose distribution in the phantom but also output dose indexes such as advantage depth and peak tumor dose, a comparison between different boron neutron capture therapy neutron sources can be easily performed. Consistency with a Monte Carlo transportation code was verified through comparison with the conventional dose calculation with the Particle and Heavy Ion Transport Code System, and the calculation time was nearly 1/90 in the SiDE. The dose distributions for a reactor and accelerator-based neutron sources were compared, and the differences were found to be small although large differences existed between the incident spectra.

Thermal neutron flux evaluation using a single crystal diamond detector (SDD) was carried out in the core region of the UTR-KINKI reactor where a mixed radiation field by thermal and fast neutrons and gamma-ray exists. The pulse shape discrimination method to extract pulses with a rectangular shape as well as a wide pulse-width was established to exclude pulses induced by gamma-rays. The SDD, using a 6LiF thermal neutron converter, is able to detect pulse events caused not only by fast neutrons but also by thermal neutrons through energy depositions into the diamond by energetic alpha and triton particles induced by thermal neutrons. Additionally, the SDD without the thermal neutron converter was used for the measurement of the energy deposition events only by fast neutrons. A comparison of the pulse counts of the SDD with or without the thermal neutron convertor deduced the energy deposition spectra by thermal neutrons. The thermal neutron flux in the core region of the UTR-KINKI reactor was evaluated to be 7.6 × 106 n cm−2 s−1 W−1 up to a reactor power of 1 W.

An optical fiber-based neutron detector can be used as a real-time neutron monitor for an intense neutron field. In this study, optical fiber-based neutron detectors were fabricated using Li-glass scintillator. The thermal neutron sensitivity, upper limit of the output linearity, and response to gamma rays from 60Co and Cd were evaluated. The thermal neutron sensitivity was proportional to the mass of the Li-glass scintillator, and the calibration factor was 2.06 × 10^-6 and 3.18 × 10^-6 cps/(n/cm 2/s)/g for the lower-level discrimination of the peak and valley channel, respectively. The detector output linearity was confirmed to be up to nearly 2 Mcps. While evaluating the response to gamma rays, for both 60Co and Cd, the gamma-ray counting rate was found to be smaller than the uncertainty associated with counting statistics in most expected applications where the neutron counting rate was >1 kcps.

For the estimation of the neutron field generated by deuterium plasma operation in the Large Helical Device (LHD), the first measurement of the thermal neutron distribution on the floor level of the LHD torus hall was carried out. For the thermal neutron detection, indium was used as activation foils. The radioactivity of these foils were evaluated by a high-purity germanium detector (HPGe) and an imaging plate (IP). The major components of radioactive isotope of indium was 116mIn. The mapping of thermal neutron distribution in the torus hall was performed. The interactions between neutron and components around LHD were observed in the thermal neutron distribution. Also, the borated polyethylene blocks effectively absorbed the thermal neutron. The thermal neutron distribution evaluated in this work can be helpful to predict the amount of radioactive waste in the torus hall proceeding with deuterium experiment in LHD.

Prior to the deuterium experiments of the Large Helical Device (LHD), an in situ neutron calibration experiment was carried out by using an 800 MBq 252Cf neutron source in November 2016, where a neutron flux monitor, a neutron activation system and a vertical neutron camera are calibrated. This paper concentrates on the in situ calibration for the neutron flux monitor. The neutron flux monitor based on 235U fission chambers and 10B and 3He proportional counters are positioned at three locations outside the LHD cryostat. The counts are measured during continuous rotation of the 252Cf neutron source on the typical magnetic axis of the plasma by a toy train, where the neutron source can be regarded as a ring neutron source. Detection efficiencies for the 252Cf ring source are derived by (total counts)/(total emitted neutron number) in the continuous rotation of the neutron source. Detection efficiencies for the real plasma neutron source are obtained from the measured detection efficiencies with the correction factor taking account of the deference between the 252Cf ring source and the real plasma neutron source evaluated by MCNP calculations with the three-dimensional model of the LHD. Finally, the uncertainty of the total neutron emission rate measurement is discussed.

The deuterium operation of the large helical device (LHD) began in March 7, 2017, after long-term preparation and commissioning of apparatuses necessary for execution of the deuterium experiment. A comprehensive set of neutron diagnostics was developed and installed onto LHD through numerous efforts in preparation. Neutron diagnostics play an essential role in both neutron yield management for the radiation safety and extension of energetic-particle physics study in LHD. Neutron flux monitor (NFM) characterized by fast-response and wide dynamic range capabilities is successfully working. Total neutron emission rate reached $3.3 \times 10^{15}$ (n/s) in the first deuterium campaign of LHD. The highest neutron emission rate was recorded in inward shifted configuration. Neutron yield evaluated by neutron activation system agrees with neutron yield measured with NFM. Performance of vertical neutron camera was demonstrated. Neutron emission profile was inwardly shifted in the inwardly shifted configuration, whereas it was outwardly shifted in the outwardly configuration. Secondary deuterium-tritium neutrons produced by triton burnup in LHD deuterium plasmas were detected for the first time in stellarator/heliotron devices in the world. Similar to total neutron emission rate, the inward shifted configuration provided highest triton burnup ratio.

An accelerator-based neutron source using d-Li reactions is one of the most promising neutron sources for fusion material irradiation facilities such as IFMIF, where 40 MeV deuterons bombard a liquid lithium target. The neutron yield estimation including angular neutron spectra is one of the most important issues in the design of such irradiation facilities. Recently, JAEA released deuteron nuclear data of JENDL/DEU-2020 in ACE format file for Monte Carlo codes such as MCNP, and in Frag-Data format for the PHITS code. We carry out the benchmark calculations of d-Li neutron yield by using PHITS with Frag-Data, MCNP with JENDL/DEU-2020, and MCNP/PHITS with built-in nuclear reaction models. Those calculation results are compared with experimental data. It is confirmed that PHITS with Frag Data and MCNP with JENDL/DEU-2020 reproduce well the experimental data. Those are useful for the neutron yield estimation and also the irradiation field characterization of IFMIF and similar facilities.

International Fusion Material Irradiation Facility (IFMIF) is the facility generating the high flux and high energy neutrons to develop fusion reactor materials. In IFMIF, high-speed liquid lithium (Li) jet is used as the target irradiated by deuteron beams. Since the Li jet must flow with high velocity for heat removal, it is important to research on the Li flow characteristics. These researches have been aimed toward the steady-state Li flow characteristics. On the other hand, in the actual IFMIF, it is also necessary to clarify transient flow characteristics at start and stop of the system for the operation. In this study, water experiment to obtain them at start and stop is thus conducted. Water can be substituted for liquid Li as the target, because the kinematic viscosity of the Li at operation temperature in IFMIF is nearly equal to that of water at normal temperature and pressure. Flow patterns of the water jet at test section which consists of a two-staged contraction nozzle, a vertical concave flow channel are observed by laser probe method and high-speed video camera. As a result, the flow pattern at start and stop was successfully observed. The flow at stop became stable by venting gas.

The single crystal CVD diamond detector (SDD) was installed in the torus hall of the Large Helical Device (LHD) to measure neutrons with high time resolution and neutron energy resolution. The LiF foil with 95.62 % of 6Li isotope enrichment pasted on the detector was used as the thermal neutron convertor as the energetic ions of 2.0 MeV alpha and 2.7 MeV triton particles generated in LiF foil and deposited the energy into SDD. SDD were exposed to the neutron field in the torus hall of the LHD during the 2nd campaign of the deuterium experiment. The total pulse height in SDD was linearly propotional to the neutron yield in a plasma operation in LHD over 4 orders of magnitude. The energetic alpha and triton were separately measured by SDD with LiF with the thickness of 1.9 μm, although SDD with LiF with the thickness of 350 μm showed a broadened peak due to the large energy loss of energetic particles generated in the bulk of LiF. The modeling with MCNP and PHITS codes well interpreted the pulse height spectra for SDD with LiF with different thicknesses. The results above demonstrated the sufficient time resolution and energy discrimination of SDD used in this work.

Accelerator-driven neutron sources have been developed worldwide for boron neutron capture therapy (BNCT) instead of nuclear reactors. Nagoya University is currently developing a compact DC accelerator driven BNCT system using a sealed lithium (Li) target. The target receives a high heat load by a proton beam, and the heat then requires efficient removal. This study developed a high-efficiency heat removal technique that uses V-shaped staggered ribs on the heat removal side of the flow channels. We performed experiments using an electron beam to confirm the technique performance. The results showed double the heat removal performance compared to the smooth surface channel. The value expected to sufficiently remove 6.6 MW/m2 of heat load provided by the proton beam at the Nagoya University neutron source for BNCT.

As the commissioning phase of the Nagoya University Accelerator-driven Neutron Source for boron neutron capture therapy, in-phantom thermal neutron flux measurements were conducted using a small ⁶LiF/Eu:CaF₂ scintillator detector and activation foils. The spatial distribution of the measured thermal neutron flux agreed with the Monte Carlo simulation results. Based on this comparison, the free-in-air neutron spectrum, calculated at the exit aperture, was verified and the epithermal neutron flux, at a 0.25 mA proton current, was evaluated as (1.49 ± 0.10) × 10⁷ n/(cm2 s).

The International Fusion Materials Irradiation Facility (IFMIF), presently in the Engineering Validation and Engineering Design Activities (EVEDA) phase was started from 2007 under the frame of the Broader Approach (BA) agreement. In the activities, a prototype Li loop with the world's highest flow rate of 3000 L/min was constructed in 2010, and it succeeded in generating a 100 mm wide and 25 mm thick with a free-surface lithium flow along a concave back plate steadily at a high-speed of 15 m/s at 250 °C for 1300 h. In the demonstration operation it was needed to develop the Li flowing measurement system with precious resolution less than 0.1 mm, and a new wave height measuring method which is laser-probe method was developed for measurements of the 3D geometry of the liquid Li target surface. Using the device, the stability of the variation in the Li flowing thickness which is required in the IFMIF specification was ± 1 mm or less as the liquid Li target, and the result was satisfied with it and the feasibility of the long-term stable liquid Li flow was also verified. The results of the other engineering validation tests such as lithium purification tests of lithium target facility have also been evaluated and summarized.

In the framework of the Engineering Validation and Engineering Design Activities (EVEDA) project of the International Fusion Materials Irradiation Facility (IFMIF), we measured surface velocity fields of a lithium (Li) target at the EVEDA Li test loop under specifically-designated IFMIF conditions (target speeds of 10, 15, and 20 m/s, vacuum pressure of 10−3 Pa, and Li temperature of 250 °C). In the current design of the IFMIF, the free surface of the Li target is under a most severe heat load condition with respect to Li boiling. The objective of this study is to measure the actual free-surface velocity under these IFMIF conditions to evaluate the heat removal performance of the Li target. The measured results (using the surface-wave tracking method that our team developed) showed two-dimensional time-averaged velocity distributions around the IFMIF beam footprint being virtually uniform, and close to the cross-sectional average velocity. The uniformity of the velocity distributions was less than 1 m/s. The comparison between the measured and analyzed surface velocity at the beam center showed that the analysis accurately predicts the measurement results within a margin of 3%. Finally, it was confirmed that the Li target delivers adequate heat removal performance in the IFMIF as designed.

In the international fusion materials irradiation facility (IFMIF), 14 MeV neutrons are generated by 40 MeV deuteron beam injection into a high-speed free surface liquid lithium (Li) plane jet, which flows along a vertical concave wall in vacuum. Measurement of the free surface flow or of the fluctuation in thickness is a key issue for stable neutron generation and for safety of Li target system. In the present study, a laser reflection method for the measurement of surface wave of the Li jet is proposed and tested experimentally. The method enables us to obtain time variation of wave shapes and surface fluctuations in two dimensions. As a result, it was found that slope angles of waves in transverse direction are steeper than those in flow direction, and that the surface fluctuation consists of two components of long waves and very short waves. It was confirmed that with using the present laser method, very similar wave characteristics obtained by the probe method. Although the laser method need to be improved the data processing, and that proof of principle experiment of the method was performed successfully.

The Pulse shape discrimination (PSD) technique for single crystal CVD diamond detector (SDD) was applied to the real-time thermal neutron measurement during the 3rd campaign of Deuterium-Deuterium (D-D) plasma experiment in the Large Helical Device (LHD). The PSD method is based upon the different shape of electrical pulses produced in diamond by gamma-ray (triangular-shaped pulse) and energetic ions (rectangular-shaped pulse), respectively. An improved PSD was developed which makes use of the full width at half maximum (FWHM) and the full width at 2/3 of the peak height (FW2/3 PH). These two parameters reflect the difference in the pulse shape produced by gamma-rays and energetic ions. The performance of this improved PSD technique was first tested irradiating the SDD by gamma-rays from 60Co and alphas from 241Am radiation sources, respectively. Then, using a 6LiF thermal neutron converter, a successful discrimination between gamma-rays and alpha particles and tritons generated by the 6Li(n,α)3H reaction induced by neutrons from 252Cf spontaneous fission neutron source was achieved. Thanks to this PSD processing technique, the thermal neutron detection efficiency increased about 1.7 times compared to previous measurements performed without the PSD technique. The real-time thermal neutron measurement by SDD with this enhanced PSD processing was also used during D-D plasma discharges in the Large Helical Device (LHD). It was found that both the time integrated counts and the time evolution count rate measured by the SDD reproduced well that of the fission chamber used as official LHD neutron monitor. The capability of the SDD detector using the PSD technique of precise real-time thermal neutron measurement in LHD was demonstrated.

Boron neutron capture therapy (BNCT) requires pharmaceutical innovations and molecular-based evidence of effectiveness to become a standard cancer therapeutic in the future. Recently, in Japan, 4-borono-L-phenylalanine (BPA) was approved as a boron agent for BNCT against head and neck (H&amp, N) cancers. H&amp, N cancer appears to be a suitable target for BPA-BNCT, because the expression levels of L-type amino acid transporter 1 (LAT1), one of the amino acid transporters responsible for BPA uptake, are elevated in most cases of H&amp, N cancer. However, in other types of cancer including malignant brain tumors, LAT1 is not always highly expressed. To expand the possibility of BNCT for these cases, we previously developed poly-arginine peptide (polyR)-conjugated mercaptoundecahydrododecaborate (BSH). PolyR confers the cell membrane permeability and tumor selectivity of BSH. However, the molecular determinants for the properties are not fully understood. In this present study, we have identified the cluster of differentiation 44 (CD44) protein and translational machinery proteins as a major cell surface target and intracellular targets of BSH-polyR, respectively. CD44, also known as a stem cell-associated maker in various types of cancer, is required for the cellular uptake of polyR-conjugated molecules. We showed that BSH-polyR was predominantly delivered to a CD44High cell population of cancer cells. Once delivered, BSH-polyR interacted with the translational machinery components, including the initiation factors, termination factors, and poly(A)-biding protein (PABP). As a proof of principle, we performed BSH-polyR-based BNCT against glioma stem-like cells and revealed that BSH-polyR successfully induced BNCT-dependent cell death specifically in CD44High cells. Bioinformatics analysis indicated that BSH-polyR would be suitable for certain types of malignant tumors. Our results shed light on the biochemical properties of BSH-polyR, which may further contribute to the therapeutic optimization of BSH-BNCT in the future.

The designs of neutron spectrum-shaping assembly (NSSA) composed with various shielding materials with natural isotopic abundance were evaluated to construct the neutron field dominated with thermal neutron or epi-thermal neutron for the application of neutron filed in the torus hall of LHD toward the BNCT research. According to the neutron transport calculation by MCNP6, the fast neutron moderation efficiency was higher in polyethylene (PE) compared to lithium fluoride (LiF) and magnesium fluoride (MF), although LiF showed relatively large epi-thermal and thermal neutron absorption. This comparison showed that the thermal neutron field can be effectively achieved with using PE. For constructing the NSSA which can provide the neutron field dominated with epi-thermal neutron, several NSSA designs were evaluated with respect to fast neutron flux, epi-thermal neutron flux, and gamma-ray dose. The combination of MF, lead (Pb) and cadmium (Cd) can provide the good epi-thermal neutron field with the mitigations of fast neutron flux and gamma-ray dose, which is also suitable for BNCT research in the torus hall of LHD.

BNCT is a radiation therapy using neutrons. For dosimetry in BNCT, it is important to evaluate the thermal neutron distribution in a water phantom irradiated with neutrons used in the therapy. We proposed application of a small 6Li-based scintillator neutron detector with an optical fiber into a comparison study between simulation and experimental results in BNCT. In this paper, we compared simulation and experimental results for depth profile of 6Li(n,α) reaction rate in a water phantom of BNCT using a developed small 6Li-based scintillator neutron detector. The experimental results well agreed with the simulation ones. However, we found that the self-shielding effect in the detector will be significant even for a small LiF/Eu:CaF2 eutectics scintillator having the mass of 100 μg. We concluded that we have to consider the self-shielding effect in the small 6Li-based scintillator neutron detector when comparing simulation and experimental results of the 6Li(n,α) reaction rate in a water phantom.

We developed a mixed powder neutron scintillator as a novel neutron detection material. Since LiF-Ag:ZnS mixed powder, which is a widely used mixed powder neutron scintillator, is opaque owing to the extremely high refractive index of Ag:ZnS, the scintillator thickness is limited. We proposed the use of LiF-Eu:CaF 2 mixed powder, where both LiF and Eu:CaF2 have almost the same refractive index, as a roughly transparent mixed powder neutron scintillator. Although the just mixed LiF-Eu:CaF2 powder is opaque, it can become translucent by filling the gap in the powder with a liquid glass resin as a refractive index buffer. As a result, we observe a peak shape in the pulse height spectrum obtained from the LiF-Eu:CaF2 mixed powder scintillator when irradiating thermal neutrons .

We developed optical fiber-based neutron detectors that use small pieces of 6Li-based scintillators, namely, Eu:LiCaAlF6 crystals and LiF/Eu:CaF2 eutectics. Considering that these elements have high light yields (more than 10,000 photons/neutron), they can produce pulse height spectra with clear peaks even when using optical fiber light guides. To adjust the neutron sensitivity, we controlled the amount of 6Li in the scintillator. We fabricated three different detectors using Eu:LiCaAlF6 and LiF/Eu:CaF2 with various amounts of 6Li and then evaluated the neutron sensitivity of each detector. This approach confirmed that there was a linear relationship between the neutron sensitivity and amount of 6Li, thus implying that the 6Li mass measured by a high-precision balance could be a good metric of neutron sensitivity. We also checked the linearity of the output of the detector at relatively high neutron fluxes by using an accelerator-based neutron source. Although the detectors showed saturation behavior at high count rate, we were able to cover a relatively wide dynamic range by combining multiple detectors using different amounts of 6Li. After correcting for the pulse height distortion and count loss due to pile-up events, we were able to confirm the linearity of the detectors up to a count rates of 100 kcps.

The Nagoya University Accelerator driven Neutron Source (NUANS) is constructed at the main campus of the Nagoya University. The electrostatic accelerator is used with the maximum proton energy and intensity of 2.8MeV, 15mA(42kW) respectively. Two neutron beamlines are designed at NUANS. The BL1 is dedicated to BNCT development. The BL2 is designed for research and development for neutron devices and neutron imaging. The neutrons used for the BL2 are generated by using the (p, n) reaction from a thin beryllium target. We constructed a compact target station for the BL2 and measured the neutron transmission image.

A high-speed (nominal: 15 m/s, range: 10–16 m/s) liquid lithium wall jet is planned to serve as the target for two 40 MeV and 125 mA deuteron beams in the International Fusion Materials Irradiation Facility (IFMIF). The design requirement of target thickness stability is 25 ± 1 mm under a vacuum of 10 −3 Pa. This paper presents the results of the target thickness measurement conducted in the EVEDA Li Test Loop under a wide range of conditions including the IFMIF condition (target speed of 10, 15, and 20 m/s; vacuum pressure of 10 −3 Pa; and Li temperature of 250 °C). For measurement, we use a laser probe method that we developed in advance; this method generates statistical measurements method using a laser distance meter. The measurement results obtained under the IFMIF nominal condition (15 m/s, 10 −3 Pa, 250 °C) at the IFMIF beam center are as follows: average target thickness = 26.08 ± 0.09 mm (2 σ ), mean wave amplitude = 0.26 ± 0.01 mm (2 σ ), and maximum wave amplitude = 1.46 ± 0.25 mm (2 σ ). Of the total wave components, 99.7% are within the design requirement. The analytically predicted target thickness is in excellent agreement with the experimental data, resulting in successful characterization of the Li target thickness.

A liquid metal lithium (Li) jet will be employed as a target irradiated by an ion beam to generate a neutron field in an accelerator neutron source. In these applications, Li vapor, which is produced by a decrease in boiling temperature under vacuum condition and heat load due to beam irradiation, may cause degradation of the vacuum and contamination of the equipment. The present study is intended to develop a novel measurement method of Li vapor using the optical absorption characteristics. In this experiment, we were successful to obtain the molar absorbance coefficient of Li vapor.

In situ calibration of the neutron activation system on the Large Helical Device (LHD) was performed by using an intense 252Cf neutron source. To simulate a ring-shaped neutron source, we installed a railway inside the LHD vacuum vessel and made a train loaded with the 252Cf source run along a typical magnetic axis position. Three activation capsules loaded with thirty pieces of indium foils stacked with total mass of approximately 18 g were prepared. Each capsule was irradiated over 15 h while the train was circulating. The activation response coefficient (9.4 ± 1.2) × 10−8 of 115In(n, n′)115mIn reaction obtained from the experiment is in good agreement with results from three-dimensional neutron transport calculations using the Monte Carlo neutron transport simulation code 6. The activation response coefficients of 2.45 MeV birth neutron and secondary 14.1 MeV neutron from deuterium plasma were evaluated from the activation response coefficient obtained in this calibration experiment with results from three-dimensional neutron calculations using the Monte Carlo neutron transport simulation code 6.

A technique was needed for measuring the thickness of the liquid lithium target of the International Fusion Materials Irradiation Facility (IFMIF 1 ). Thus, the applicability of a laser-based distance meter was evaluated within the framework of Engineering Validation and Engineering Design Activities (EVEDA) of IFMIF. This device is usually used against diffuse objects. However, in the present case, it is used against specular, finely deformed liquid metal surfaces. Both the measurement resolution and precision are required to be 0.1 mm or less to measure the thickness variation of ±1 mm. To ensure sufficient power of the reflected laser from the liquid metal surface, the laser spot diameter was set to 0.1 mm, considering previous experimental data regarding the dominant wavelength of surface waves. An experiment was conducted at the Osaka University Li loop in the velocity range 10–15 m/s and Li temperature of 300 °C under an argon atmosphere of 0.12 MPa. The experimental results showed that the laser-based distance meter can be used to measure the Li-jet thickness at a jet velocity of up to 15 m/s and with superior precision at 16 μm, which is a much more precise value than required.

In the Large Helical Device (LHD) of the National Institute for Fusion Science, a deuterium plasma experiment was performed in March 2017. Neutrons with an energy of 2.45 MeV are generated by d (d, n) 3He reactions. The evaluation of this neutron flux in the torus hall, as well as within the LHD vessel, is very important for the decommissioning of the LHD in the future and the radiation shielding for workers and components around the LHD. In this study, we evaluate the spatial distribution of thermal and epithermal neutron flux on the floor level of the LHD torus hall by measuring the neutron activation of gold foils. The flux information obtained in the experiment can be used for the evaluation of radio-activation and the shielding design.

The global flux distributions for thermal, epithermal and fast neutrons in the torus hall of large fusion devices were experimentally evaluated for the first time in the Large Helical Device (LHD) using the activation foil method measured by the imaging plate and high-purity germanium detector. It turned out that the thermal neutrons were effectively absorbed by borated polyethylene blocks placed beneath the LHD. This should reduce the radioactivity of the floor and would be beneficial in maintaining a good environment for radiation workers. Uniform distributions of epithermal and fast neutrons were observed near the LHD. In particular, the significant decrease in fast neutron flux with increasing distance from the LHD, due to the fast energy loss of fast neutrons, was observed.The neutron flux distribution measurement, with rough energy discrimination based on the threshold energy of the neutron activation foil, allows us to estimate the spatial radiation dose rate, as well as the radioactivity in components in the torus hall. The prediction of the radioactivity in the concrete floor indicated that radioactive isotope Fe will be a dominant source of radioactivity in the concrete after the nine-year deuterium experiment campaign finishes.

We developed a thermal neutron source for performance tests of boron agents for Boron Neutron Capture Therapy (BNCT) at Neutron exposure Accelerator System for Biological Effect Experiments (NASBEE) in National Institute of Radiological Sciences, Japan. The thermal neutron flux at the experimental position was measured to be 2.18× 108 n/cm2/s, the cadmium ratio was 9.21, and gamma ray dose rate was 2.11 Gy/h. The neutron dose rate was estimated to be 2.96 Gy/h. These values indicate that the neutron source can be applicable for tests of boron agents although application range is limited. An example of irradiation experiment is described.