Etsuo IshitsukaProfessor for Academic Advancement

■Researcher basic information

Organization

  • College of Engineering Department of Mechanical Systems Engineering
  • Faculty of Applied Science and Engineering Domain of Mechanical Systems Engineering

Research Areas

  • Energy, Quantum beam science, 原子力

Degree

  • 1999年7月　工学博士（東京大学）
  • 1986年3月　工学修士（日本大学）

Educational Background

  • Mar. 1986, 日本大学大学院, 生産工学研究科電気工学専攻 修士課程修了

External link

Message from Researchers

  • (Message from Researchers)

    原子力関係、研究用原子炉、核融合炉、高温ガス炉

■Research activity information

Paper

  • Study on T production using high-temperature gas-cooled reactor for DEMO fusion reactor − H absorption properties of Zr sphere with Ni coating
    Hideaki Matsuura; Hiromi Kawai; Aoi Furuya; Kazunari Katayama; Teppei Otsuka; Etsuo Ishitsuka; Shigeaki Nakagawa; Kenji Tobita; Satoshi Konishi; Youji Someya; Ryoji Hiwatari; Yoshiteru Sakamoto, Elsevier BV
    Fusion Engineering and Design, Jun. 2025
  • Feasibility of using BeO rods as secondary neutron sources in the long-life fuel cycle high-temperature gas-cooled reactor
    Ho H. Q.; Ishii Toshiaki; Nagasumi Satoru; Ono Masato; Shimazaki Yosuke; Ishitsuka Etsuo; Sawahata Hiroaki; Goto Minoru; Simanullang I. L.*; Fujimoto Nozomu*; Iigaki Kazuhiko, External sources of neutron provide stable and sufficient neutron for initial startup of a nuclear reactor. They also provide signals for neutron detectors to monitor the safety of reactor during shutdown. In the high temperature engineering test reactor, $^{252}$Cf is used as the external neutron source. However, the $^{252}$Cf sources must be renewed every approximately 7 years because of its relatively short half-life of 2.6 years. The renewal of $^{252}$Cf sources requires a high cost and a very complicated procedure. This study investigated the feasibility of using BeO rods as the secondary neutron sources to avoid renewing the $^{252}$Cf neutron sources periodically. The BeO rods could exist in the reactor for a long time so that if the reactor operates long enough, the neutron flux at the wide-range monitoring detectors remains significant even if the reactor is shutdown for as long as 5 years. The results of this study indicated that using BeO rods as the secondary neutron sources would be an attractive option for the future HTGR design with a long-life fuel cycle.
    Nuclear Engineering and Design, Feb. 2024, [Reviewed]
  • T production using a high-temperature gas-cooled reactor for the DEMO fusion reactor: Li rod structure for the initial irradiation test
    Hideaki Matsuura; Taisei Abe; Kanta Kitagawa; Motomasa Naoi; Hiromi Kawai; Kazunari Katayama; Teppei Otsuka; Minoru Goto; Shigeaki Nakagawa; Etsuo Ishitsuka; Shimpei Hamamoto; Kenji Tobita; Satoshi Konishi; Yuki Koga; Ryoji Hiwatari; Youji Someya; Yoshiteru Sakamoto, Elsevier BV
    Fusion Engineering and Design, Dec. 2023
  • Loading method of Li rods for tritium production using High-Temperature Gas-Cooled reactor for fusion reactors
    Yuki Koga; Hideaki Matsuura; Kazunori Katayama; Teppei Otsuka; Minoru Goto; Shimpei Hamamoto; Etsuo Ishitsuka; Shigeaki Nakagawa; Kenji Tobita; Youji Someya; Yoshiteru Sakamoto
    Nuclear Engineering and Design, Dec. 2023, [Reviewed]
  • Evaluation of power distribution calculation of the very high temperature reactor critical assembly (VHTRC) with Monte Carlo MVP3 code
    Irwan Liapto Simanullang; Naoki Nakagawa; Hai Quan Ho; Satoru Nagasumi; Etsuo Ishitsuka; Kazuhiko Iigaki; Nozomu Fujimoto, Power distribution plays a significant role in preventing the fuel temperature exceeds the safety limit of 1600$^{\circ}$C in high-temperature gas-cooled reactors. The experiment to measure the power distribution in the graphite-moderated system was carried out with the Very High Temperature Reactor Critical Assembly facility. In the previous study, the power distribution in the VHTRC was calculated using a nuclear design code system based on diffusion calculation. The results showed a maximum discrepancy of up to 20 \ between the experiment and calculated values in the axial direction. The large discrepancy occurred near the boundary of fuel and reflector regions. This study describes the evaluation results of pin-wise power distribution of the VHTRC with the Monte Carlo MVP3 code. The calculation results were in good agreement with the measured results. In the axial direction, the discrepancy was less than 1 \ around the boundary of fuel and reflector regions., Elsevier BV
    Annals of Nuclear Energy, Nov. 2022, [Reviewed]
  • MCNP6 calculation of neutron flux map in the HTTR during normal operation
    Ho H. Q.; Ishitsuka Etsuo; Iigaki Kazuhiko, Detailed neutron flux distribution is important to understand the neutronic behavior during operation as well as to precise the core optimization and safety analysis of a reactor. In the literature, no calculations have been performed to show the detailed neutron flux map for the high temperature engineering test reactor (HTTR) because of the limitation of the old neutronic codes and the low performance of the computing system. The present work deals with MCNP6 Monte-Carlo calculation to determine the detailed neutron flux map in the HTTR during normal operation. At first, the calculation of neutron flux at several positions in the reactor was validated by comparing the corresponding reaction rate between the calculation and measurement. After that detailed neutron flux with the small cells of 1cm $\times$ 1cm $\times$ 10cm was obtained for the entire reactor core using the fmesh tally of MCNP6 code.
    Recent Contributions to Physics, Sep. 2022, [Reviewed]
  • Calculation of shutdown gamma distribution in the high temperature engineering test reactor
    Hai Quan Ho; Toshiaki Ishii; Satoru Nagasumi; Masato Ono; Yosuke Shimazaki; Etsuo Ishitsuka; Minoru Goto; Irwan Liapto Simanullang; Nozomu Fujimoto; Kazuhiko Iigaki, Estimation of decay gamma distribution in a reactor core is essential for safely conducting various works after reactor shutdown such as periodic maintenance, shuffling fuel, removing spent fuel at the end of cycle, etc. Because of the dependency on the complex operating history of the reactor, attempting to calculate the decay gamma rays distribution in the core remains a challenge. This study showed a method to calculate the shutdown gamma distribution in the HTTR core by coupling a Monte-Carlo transport calculation code MCNP6 and an activation code ORIGEN2 to take advantage of spatial dependence and transportation abilities of MCNP6 and the detailed fission products tracking during burnup and cooling of ORIGEN2. As result, the three-dimensional shutdown gamma distribution in the HTTR core for different cooling times and spatial locations could be obtained accurately., Elsevier BV
    Nuclear Engineering and Design, Sep. 2022, [Reviewed]
  • Effect of nuclear heat caused by the 6Li(n,α)T reaction on tritium containment performance of tritium production module in High-Temperature Gas-Cooled reactor for fusion reactors
    Yuki Koga; Hideaki Matsuura; Kazunari Katayama; Teppei Otsuka; Minoru Goto; Shimpei Hamamoto; Etsuo Ishitsuka; Shigeaki Nakagawa; Kenji Tobita; Satoshi Konishi; Ryoji Hiwatari; Youji Someya; Yoshiteru Sakamoto, Tritium is required for research and development activities for the deuterium–tritium (DT) fusion reactor and fueling the DEMOnstration Power Station (DEMO). However, tritium is a very rare nuclide and must be produced artificially. Tritium production by loading Li compounds (Li rods) into burnable poison holes of a high-temperature gas-cooled reactor (HTGR) has been proposed (H. Matsuura, et al., Nucl. Eng. Des. 243 (2012) 95–101.). Al2O3 and Zr are used to prevent tritium leaks. Nuclear reaction heat caused by the nuclear reaction (e.g., 6Li(n,α)T reaction) can cause a spatial temperature profile in the Li rods and may change its tritium containment performance, because Al2O3 and Zr performance strongly depend on these temperatures. The effect of nuclear reaction heat by the 6Li(n,α)T reaction on the tritium containment performance of the Li rods was evaluated by simulation. The temperatures of the Li rods for the high-temperature engineering test reactor (HTTR) and gas turbine high-temperature reactor 300 (GTHTR300) increased by 36 K and 46 K, and the leaked tritium decreased by 32% and 37% via nuclear reaction heat, respectively.
    Nuclear Engineering and Design, Jan. 2022
  • Design of a portable backup shutdown system for the high temperature gas cooled reactor
    Shimpei Hamamoto; Hai Quan Ho; Kazuhiko Iigaki; Minoru Goto; Yosuke Shimazaki; Hiroaki Sawahata; Etsuo Ishitsuka, The experience of Fukushima Daiichi nuclear power plant accident caused by the great earthquake that occurred in eastern Japan in 2011 showed the importance of preparing for the loss of function of the engineered safety features. Increasing the strength of equipment to prevent loss of function in an accident is effective, but the possibility of loss of function remains. Therefore, it is important to have an alternative to lost functions in order to put the accident under control early. Thus, this study designed an alternative shutdown system, namely a portable backup shutdown system (PBSS), to make countermeasures in the event of a loss of shutdown function more robust without impairing economic efficiency of the High Temperature Gas-cooled Reactor (HTGR). The PBSS is portable and capable of being installed manually so that it can operate in a total loss of off-site electricity. Various neutron absorber materials for the PBSS were also considered from the viewpoints of technical and cost-effective properties. As results of optimization, the boron nitride (BN) was selected as it shows a good neutronic property as well as a reasonable cost in comparison with other materials.
    Nuclear Engineering and Design, Jan. 2022
  • Proposal of evaluation method of graphite incombustibility
    Hamamoto Shimpei; Ohashi Hirofumi; Iigaki Kazuhiko; Shimazaki Yosuke; Ono Masato; Shimizu Atsushi; Ishitsuka Etsuo, Since the HTGR has a large amount of graphite material in the core, it is necessary to assume an accident in which the reactor pressure boundary is damaged and air flows into the core. It is important to state that at the time of this accident, graphite does not burn and the accident does not develop due to the heat of oxidation reaction. Therefore, in this study, in order to evaluate the combustibility of graphite materials, we propose a method to compare the calorific value and heat removal amount of the material. When calculating the calorific value, the structural material of HTTR, a high-temperature gas reactor in Japan, was used as a reference. The amount of air in contact with the structural material is a value determined from the chimney effect. The amount of heat release is the sum of convection and radiation. As a result of comparing the heat generation amount with the heat removal amount, it was shown that the heat release amount was always larger than the heat generation amount. This result shows that the graphite material does not depend on the state at the time of the air inflow accident, the temperature decreases and does not burn. It is important to clearly explain the non-flammability of graphite materials when deciding how to deal with severe accidents in HTGRs. This quantitative evaluation method based on a simple theory is considered useful.
    Proceedings of 2021 International Congress on Advances in Nuclear Power Plants (ICAPP 2021) (USB Flash Drive), Oct. 2021, [Reviewed]
  • Study on chemical form of tritium in coolant helium of high temperature gas-cooled reactor with tritium production device
    Hamamoto Shimpei; Ishitsuka Etsuo; Nakagawa Shigeaki; Goto Minoru; Matsuura Hideaki*; Katayama Kazunari*; Otsuka Teppei*; Tobita Kenji*, Impurity concentrations of hydrogen and hydride in the coolant were investigated in detail for the HTTR, a block type high-temperature gas reactor owned by Japan. As a result, it was found that CH$_{4}$ was 1/10 of H$_{2}$ concentration, which was under the conventional detection limit. If the ratio of H$_{2}$ to CH$_{4}$ in the coolant is the same as the ratio of HT to CH$_{3}$T, the CH$_{3}$T has a larger dose conversion factor, and this compositional ratio is an important finding for the optimal dose evaluation. Further investigation of the origin of CH$_{4}$ suggested that CH$_{4}$ was produced as a result of a thermal equilibrium reaction rather than being released as an impurity from the core.
    Proceedings of 2021 International Congress on Advances in Nuclear Power Plants (ICAPP 2021) (USB Flash Drive), Oct. 2021, [Reviewed]
  • The T-containment properties of a Zr-containing Li rod in a high-temperature gas-cooled reactor as a T production device for fusion reactors
    Hideaki Matsuura; Takuro Suganuma; Yuki Koga; Motomasa Naoi; Kazunari Katayama; Teppei Otsuka; Minoru Goto; Shigeaki Nakagawa; Shinpei Hamamoto; Etsuo Ishitsuka; Kenji Tobita; Satoshi Konishi; Ryoji Hiwatari; Youji Someya; Yoshiteru Sakamoto, The production of tritium (T) using high-temperature gas-cooled reactors (HTGRs) has been studied for a prior engineering research with T handling and initial T possession in demonstration fusion reactors. Stable containment of T in Li-loading rods during HTGR operation is a critical issue. This study investigates this for an irradiation test to examine T-containment performance in Li-loading rods and develops an analytical model of evaluating the amount of T outflow to a He coolant. The hydrogen absorption characteristics, including the deterioration of the hydrogen absorption speed after Zr has sufficiently absorbed the hydrogen, is experimentally measured assuming an HTGR setting. We present an analytical model of evaluating the T outflow from a Li rod and, on the basis of this model, estimate the total T outflow, assuming the presence of a gas-turbine high-temperature reactor of 300 MWe with a nominal capacity and a high-temperature engineering test reactor. It is demonstrated that, by loading a sufficient amount of Zr into the Li rod, the T outflow can be suppressed to less than a small percent of the total T produced during 360 days of reactor operation.
    Fusion Engineering and Design, Aug. 2021, [Reviewed]
  • Nuclear data processing code FRENDY: A verification with HTTR criticality benchmark experiments
    Nozomu Fujimoto; Kenichi Tada; Hai Quan Ho; Shimpei Hamamoto; Satoru Nagasumi; Etsuo Ishitsuka, Japan Atomic Energy Agency has developed a new nuclear data processing code, named FRENDY, to generate ACE-formatted files from various evaluated nuclear libraries. A code-to-experiment verification of FRENDY processing was carried out in this study with criticality benchmark assessments of the high temperature engineering test reactor. The ACE files of the JENDL-4.0 and ENDF/B-VII.1 was generated successfully by FRENDY. These ACE files have been used in MCNP6 transport for various benchmark problems of the high temperature engineering test reactor. As a result, the k and reaction rate obtained by MCNP6 calculation presented a good agreement compared to the experimental data. The proper ACE files generation by FRENDY was confirmed for the HTTR criticality calculations. eff
    Annals of Nuclear Energy, Aug. 2021
  • Preparation for restarting the high temperature engineering test reactor: Development of utility tool for auto seeking critical control rod position
    Hai Quan Ho; Nozomu Fujimoto; Shimpei Hamamoto; Satoru Nagasumi; Minoru Goto; Etsuo Ishitsuka, At high power operation of the HTTR, the control rod should be kept at the top of the active core for maintaining the optimized power distribution so as to minimize the maximum fuel temperature. It is important to calculate the control rod position each time the operating conditions change in order to ensure the safe operation of the reactor. Since the Monte-Carlo code cannot change the core geometry such as control rod position during criticality and burnup calculation, the critical control rod position was determined by adjusting the control rods manually at each burnup step. This complicates the calculation procedures as well as prolongs the total time including calculation time, handling time, and waiting time. Therefore, this study develops a new utility tool that seeks the control rod position automatically without any further handling procedures and waiting time. As a result, the determination of critical control rod position becomes simpler and the total time was also reduced significantly from about 5 days to less than 2 days. The calculated critical control rod position using the new tool also gives a good agreement with the experiment data.
    Nuclear Engineering and Design, Jun. 2021
  • Feasibility study on tritium recoil barrier for neutron reflectors of research and test reactors
    Inesh Kenzhina; Etsuo Ishitsuka; Hai Quan Ho; Naoki Sakamoto; Keisuke Okumura; Noriyuki Takemoto; Yevgeni Chikhray, © 2020 Elsevier B.V. Tritium release into the primary coolant of the JMTR and the JRR-3 M had been studied, and it is found that tritium recoil release from the chain reaction of beryllium neutron reflectors is dominant. To prevent the tritium recoil release, Al, Ti, V, Ni and Zr are selected as the candidate tritium recoil barrier materials in this feasibility study. It is clear that 20∼40 μm thickness is required depending on the material to reduce by 3 orders, and that an impact on the effective multiplication factor is about 0.2 % at most. Total evaluation including the activities, fabrication and usage experiences, suggests the selection of Al as the first candidate may have the least development risk as the tritium recoil barrier.
    Fusion Engineering and Design, Mar. 2021
  • Modeling the processes of hydrogen isotopes interactions with solid surfaces
    Chikhray Y.*; Askerbekov S.*; Kenzhin Y.*; Gordienko Y.*; Ishitsuka Etsuo, The investigation of the mechanisms and dynamics of hydrogen isotopic interaction with solid surfaces (metals, ceramics, graphites, eutectics) in temperature and pressure ranges is important not only for the correct prediction of each isotope's evolution but also for substantiation of the safe operation of hydrogen-facing structural materials. The interaction of the hydrogen isotopes mix with the surface of solid metal or liquid eutectics is a complicated multistage H-D-T-O-solid interacting process depending on material property, environment, and the solid's surface parameters. To better understand the mechanisms of hydrogen isotopes interchange at a solid surface and to identify the limiting stages in the sorption-desorption processes, a reactor experiment of neutron irradiation was conducted with lithium-containing eutectics as tritium-generating media under the external flow of hydrogen. This work presents the model and results of its application to fitting the experimental results of tritium yield from the lithium-lead eutectics Pb$_{83}$Li$_{17}$under thermal neutrons irradiation at the IVG.1M reactor in Kazakhstan. The elaborated model and the approach used were also applied to the simulation of high temperature gas cooled reactor graphite corrosion in water vapors.
    Fusion Science and Technology, May 2020, [Reviewed]
  • Promising neutron irradiation applications at the high temperature engineering test reactor
    Hai Quan Ho; Yuki Honda; Shimpei Hamamoto; Toshiaki Ishii; Shoji Takada; Nozomu Fujimoto; Etsuo Ishitsuka, High temperature engineering test reactor (HTTR), a prismatic type of the HTGR, has been constructed to establish and upgrade the basic technologies for the HTGRs. Many irradiation regions are reserved in the HTTR to be served as a potential tool for an irradiation test reactor in order to promote innovative basic researches such as materials, fusion reactor technology, and radiation chemistry and so on. This study shows the overview of some possible irradiation applications at the HTTRs including neutron transmutation doping silicon (NTD-Si) and iodine-125 ($^{125}$I) productions. The HTTR has possibility to produce about 40 tons of doped Si-particles per year for fabrication of spherical silicon solar cell. Besides, the HTTR could also produce about 1.8$\times$10$^{5}$ GBq/year of $^{125}$I isotope, comparing to 3.0$\times$10$^{3}$ GBq of total $^{125}$I supplied in Japan in 2016.
    Journal of Nuclear Engineering and Radiation Science, Apr. 2020, [Reviewed]
  • Evaluation of tritium release into primary coolant for research and testing reactors
    Inesh Kenzhina; Etsuo Ishitsuka; Keisuke Okumura; Hai Quan Ho; Noriyuki Takemoto; Yevgeni Chikhray, © 2020, © 2020 Atomic Energy Society of Japan. All rights reserved. The sources and mechanisms for the tritium release into the primary coolant in the JMTR and the JRR-3M containing beryllium reflectors are evaluated. It is found that the recoil release from chain reaction of 9Be is dominant and its calculation results agree well with trends derived from the measured variation of tritium concentration in the primary coolant. It also indicates that the simple calculation method used in this study for the tritium recoil release from the beryllium reflectors can be utilized for an estimation of the tritium release into the primary coolant for a water-cooled research and testing reactors containing beryllium reflectors.
    Journal of Nuclear Science and Technology, 2020
  • Conceptual design of direct 99mTc production facility at the high temperature engineering test reactor
    Hai Quan Ho; Hiroki Ishida; Shimpei Hamamoto; Toshiaki Ishii; Nozomu Fujimoto; Naoyuki Takaki; Etsuo Ishitsuka, This study proposed a conceptual design of direct $^{\rm 99m}$Tc production facility from a natural MoO$_{3}$ target at the high temperature engineering test reactor (HTTR). $^{\rm 99m}$Tc is produced by a beta decay of $^{99}$Mo, which is formed via the $^{98}$Mo(n,$\gamma$)$^{99}$Mo reaction. $^{\rm 99m}$Tc is then extracted from the MoO$_{3}$ target by sublimation method to take advantage of the high temperature of the HTTR core. The foremost advantage of this concept is that the MoO$_{3}$ target is heated up inside the reactor without pulling out for external electric heating, and as a result, $^{\rm 99m}$Tc could be extracted directly during irradiation. With 1 kg of MoO$_{3}$ target, the HTTR could produce about 6.8$\times$10$^{8}$ MBq of $^{\rm 99m}$Tc activity in comparison with 3.0$\times$10$^{8}$ MBq of total $^{\rm 99m}$Tc supplied in Japan in 2017.
    Nuclear Engineering and Design, Oct. 2019, [Reviewed]
  • Li-rod structure in high-temperature gas-cooled reactor as a tritium production device for fusion reactors
    Matsuura, Hideaki; Okamoto, Ryo; Koga, Yuki; Suganuma, Takuro; Katayama, Kazunari; Otsuka, Teppei; Goto, Minoru; Nakagawa, Shigeaki; Ishitsuka, Etsuo; Tobita, Kenji, Production of tritium using a high-temperature gas-cooled reactor (HTGR) has been studied for a prior engineering test with tritium handling and for the startup operation of a demonstration fusion reactor. For this purpose, the hydrogen absorption speed of Zr in a Li-loading rod for the reactor operation is experimentally measured, and an analysis model is presented to evaluate the tritium outflow from the Li rod in a high-temperature engineering test reactor (HTTR). On the basis of the presented model, the structure of the Li-loading rod for the demonstration test using the HTTR is proposed., ELSEVIER SCIENCE SA
    FUSION ENGINEERING AND DESIGN, Sep. 2019, [Reviewed]
  • Study on lithium rod test module and irradiation method for tritium production using high temperature gas-cooled reactor
    Yuki Koga; Hideaki Matsuura; Yuma Ida; Ryo Okamoto; Kazunari Katayama; Teppei Otsuka; Minoru Goto; Shigeaki Nakagawa; Satoru Nagasumi; Etsuo Ishitsuka; Yosuke Shimazaki, Large quantities of tritium are required to start up fusion reactors and conducts engineering tests using tritium for a fusion blanket system. However, tritium is very rare and kg orders of tritium must be produced artificially. Tritium production, by Li(n,α)T reaction using a high temperature gas-cooled reactor (HTGR) has been proposed. This method considers the loading of Li rods into burnable poison holes in the HTGR. In this paper, the Li rod suited for use in the High Temperature engineering Test Reactor (HTTR) was designed, and tritium production and leakage from Li-rod capsules were evaluated by adjusting the thicknesses of LiAlO , alumina, and Zr layers. An irradiation test scenario to be conducted in the HTTR for demonstration of the Li rod's tritium production and containment performance was presented. 6 2
    Fusion Engineering and Design, Nov. 2018
  • Study on source of radioactive material in primary coolant of HTTR
    Ishii Toshiaki; Shimazaki Yosuke; Ono Masato; Fujiwara Yusuke; Ishitsuka Etsuo; Hamamoto Shimpei, In the primary cooling system of the High-Temperature Engineering Test Reactor, the highest dose rate was observed at the Helium Gas Circulator filter areas. To find the origin of the radioactive material, the radiation dose rates, and $\gamma$ ray spectrums were measured. From these results, it is clear that the main radioactive nuclide at the filter is $^{60}$Co after 6 years reactor shutdown, and the in-core materials are a low possibility as the candidates of radioactive materials. It is also clear that the mixtures of materials, the contained low-level impurities and other new candidate materials must be considered.
    Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), Oct. 2018, [Reviewed]
  • Feasibility study of new applications at the high-temperature gas-cooled reactor
    Ho H. Q.; Honda Yuki*; Hamamoto Shimpei; Ishii Toshiaki; Takada Shoji; Fujimoto Nozomu*; Ishitsuka Etsuo, Besides the electricity generation and hydrogen production, HTGRs have many advantages for thermal neutron irradiation applications such as stable operation in longterm, large space available for irradiation target, and high thermal neutron economy. This study summarized the feasibility of new irradiation applications at the HTGRs including neutron transmutation doping silicon and I-125 productions. The HTTR located in Japan was used as a reference HTGR in this study. Calculation results show that HTTR could irradiate about 40 tons of doped Si particles per year for fabrication of spherical silicon solar cell. Besides, the HTTR could also produce about 1.8x105 GBq in a year of I-125, comparing to 3.0x103 GBq of total I-125 supplied in Japan in 2016.
    Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), Oct. 2018, [Reviewed]
  • Feasibility study of large-scale production of iodine-125 at the high temperature engineering test reactor
    Ho H. Q.; Honda Yuki*; Hamamoto Shimpei; Ishii Toshiaki; Fujimoto Nozomu*; Ishitsuka Etsuo, The feasibility of a large-scale iodine-125 production from natural xenon gas at high-temperature gas-cooled reactors was investigated. A high-temperature engineering test reactor, which is located in Japan, was used as a reference HTGR reactor in this study. First, a computer code based on a Runge-Kutta method was developed to calculate the quantities of isotopes arising from the neutron irradiation of natural xenon gas target. This code was verified with a good agreement with a reference result. Next, optimization of irradiation planning was carried out. As results, with 4 days of irradiation and 8 days of decay, the $^{125}$I production could be maximized and the $^{126}$I contamination was within an acceptable level. The preliminary design of irradiation channels at the HTTR was also optimized. The case with 3 irradiation channels and 20-cm diameter was determined as the optimal design, which could produce approximately 180,000 GBq per year of $^{125}$I production.
    Applied Radiation and Isotopes, Oct. 2018, [Reviewed]
  • Proposal of a neutron transmutation doping facility for n-type spherical silicon solar cell at high-temperature engineering test reactor
    Ho H. Q.; Honda Yuki; Motoyama Mizuki*; Hamamoto Shimpei; Ishii Toshiaki; Ishitsuka Etsuo, The p-type spherical silicon solar cell is a candidate for future solar energy with low fabrication cost, however, its conversion efficiency is only about 10\%. The conversion efficiency of a silicon solar cell can be increased by using n-type silicon semiconductor as a substrate. This study proposed a new method of neutron transmutation doping silicon (NTD-Si) for producing the n-type spherical solar cell, in which the Si-particles are irradiated directly instead of the cylinder Si-ingot as in the conventional NTD-Si. By using a screw, an identical resistivity could be achieved for the Si-particles without a complicated procedure as in the NTD with Si-ingot. Also, the reactivity and neutron flux swing could be kept to a minimum because of the continuous irradiation of the Si-particles. A high temperature engineering test reactor (HTTR), which is located in Japan, was used as a reference reactor in this study. Neutronic calculations showed that the HTTR has a capability to produce about 40 ton of 10 $\Omega$ cm resistivity Si-particles for fabrication of the n-type spherical solar cell.
    Applied Radiation and Isotopes, May 2018, [Reviewed]
  • Improvement of exchanging method of neutron startup source of high temperature engineering test reactor
    Sawahata Hiroaki; Shimazaki Yosuke; Ishitsuka Etsuo; Yamazaki Kazunori; Yanagida Yoshinori; Fujiwara Yusuke; Takada Shoji; Shinozaki Masayuki; Hamamoto Shimpei; Tochio Daisuke, In the HTTR, $^{252}$Cf is loaded in the reactor core as a neutron startup source and changed at frequency. In this exchange work, there were two technical issues; slightly higher radiation exposure of workers by neutron leakage and reliability of neutron source transportation container in handling. To reduce the radiation dose by neutron leakage, detail numerical evaluations using PHITS code were carried out, the effective shielding method for fuel handling machine was proposed. Easily removable polyethylene blocks and particles were used as the neutron shielding, and installed in the cooling paths of the fuel handling machine. As a result, the collective effective dose by neutron was reduced from about 700 man-microSv to about 300 man-microSv. As to the neutron source transportation container, the handling performance was improved and the handling work was safety accomplished by downsizing.
    Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), Jun. 2016, [Reviewed]
  • Status of development on $^{99}$Mo production technologies in JMTR
    Inaba Yoshitomo; Iimura Koichi; Hosokawa Jinsaku; Izumo Hironobu; Hori Naohiko; Ishitsuka Etsuo, The Japan Materials Testing Reactor (JMTR) is now under refurbishment, and the operation of the new JMTR will start in FY 2011. The new JMTR has a plan to produce $^{99}$Mo, which is the parent nuclide of $^{99m}$Tc, and two $^{99}$Mo production technologies have been developed: one is a solid irradiation method, and the other is a solution irradiation method. In this paper, the present status of the development on the $^{99}$Mo production technologies with the solid and solution irradiation methods was described. In the solid irradiation method, it was found that JMTR can provide about 20\% of the $^{99}$Mo imported into Japan. In the solution irradiation method, the fundamental characteristics of the aqueous molybdate solutions selected as candidates for the irradiation target were cleared by the $\gamma$-ray irradiation test.
    IEEE Transactions on Nuclear Science, Jun. 2011, [Reviewed]
  • Application of pade approximation for calculation of epithermal neutron self-shielding factors of some materials dealing with Doppler broadening effects
    Phuong H. T.*; Nhon M. V.*; Trang V. T. T.*; Ishitsuka Etsuo, The pade approximation method has been applied to calculate the epithermal neutron self-shielding factors for some materials which used as a comparator in $k$$_{0}$-standardization method. It includes a Doppler broadening effect correctly. With an assumption of an isotropic neutron field, the epithermal neutron self-shielding factors in foils and wires of Co, Mo, Zr and Au were calculated. The present method has an advantage in accuracy as well as less time consuming for calculation. The extensions of this method to obtaining the factors for different materials with many resonances could be performed with easy. The obtained values are compared with previous values and their experimental values.
    Applied Radiation and Isotopes, Jun. 2010, [Reviewed]
  • Status of development on $^{99}$Mo production technologies in JMTR
    Inaba Yoshitomo; Iimura Koichi; Hosokawa Jinsaku; Izumo Hironobu; Hori Naohiko; Ishitsuka Etsuo, The Japan Materials Testing Reactor (JMTR) is now under refurbishment, and the operation of the new JMTR will be started in FY 2011. The new JMTR has a plan to produce $^{99}$Mo, which is the parent nuclide of $^{99m}$Tc, and two $^{99}$Mo production technologies have been developed; the one is a solid irradiation method, and the other is a solution irradiation method. $^{99}$Mo production in the JMTR will be started by the solid irradiation method, and it was found that the JMTR can produce about 20\% of the demand for $^{99}$Mo in Japan. In the solution irradiation method, the fundamental characteristics of the aqueous molybdate solutions selected as the candidates for the irradiation target were cleared.
    Proceedings of 1st International Conference on Advancements in Nuclear Instrumentation, Measurement Methods and their Applications (ANIMMA 2009) (USB Flash Drive), Jun. 2009, [Reviewed]
  • Development on $^{99}$Mo production technique by solution irradiation method; Characterization of aqueous molybdate solutions
    Inaba Yoshitomo; Ishikawa Koji*; Tatenuma Katsuyoshi*; Ishitsuka Etsuo, A solution irradiation method is proposed as a new production technique of $^{99}$Mo, which is the parent nuclide of $^{99m}$Tc used as a radiopharmaceutical. In this new method, an aqueous molybdenum solution is irradiated with neutron in a nuclear reactor, and efficient and low cost $^{99}$Mo production comparing with conventional $^{99}$Mo production methods can be realized by using the $^{98}$Mo (n,$\gamma$) $^{99}$Mo reaction and a high performance adsorbent for (n,$\gamma$) $^{99}$Mo. In the present research, some tests under unirradiation and $\gamma$-ray irradiation were carried out using two kinds of aqueous molybdate solutions to be candidates for the irradiation target of the new method, and the corrosivity for structural materials, chemical stability, radiolysis and $\gamma$ heating of the solutions were investigated. As a result, it was found that the solutions are promising as the target and that stainless steel can be used as the structural material of capsules and pipes.
    Nippon Genshiryoku Gakkai Wabun Rombunshi, Jun. 2009, [Reviewed]
  • Irradiation test of component for radiation-resistant small-sized motor
    Nakamichi Masaru; Ishitsuka Etsuo; Shimakawa Satoshi; Kan Satoshi*, 材料試験炉(JMTR)を用いた核融合炉ブランケット炉内機能試験に使用するために耐放射線性の小型モータを開発し、JMTRで照射試験を行った。本開発研究の結果、$\gamma$線量,高速及び熱中性子照射量が市販品の約700倍まで耐える小型モータの開発に成功した。本研究では開発した耐放射線性小型モータの主要構成部品を中性子照射し、それぞれの部品に対する中性子照射の影響を調べた。この結果、Nd-Fe-B磁石をSm-Co系磁石に変更することによって、さらに1桁程度の耐放射線性の向上が期待できることを明らかにした。
    Fusion Engineering and Design, Jun. 2009, [Reviewed]
  • Irradiation tests of a small-sized motor with radiation resistance
    Nakamichi Masaru; Ishitsuka Etsuo; Shimakawa Satoshi; Kan Satoshi*, 国際熱核融合実験炉(ITER)では、原型炉用ブランケット開発のため、テストブランケットモジュール(TBM)を取付けトリチウム生成・回収特性などを評価する。このTBM開発のため、材料試験炉(JMTR)を用いて、ITERパルス運転を模擬した照射試験(ブランケット照射試験)が計画されており、その照射試験体開発のため、耐放射線性を有する小型モータの開発を実施した。本モータ開発においては、構成部材を耐放射線性の高い材料に変更することに加えて、有機系潤滑剤を使用しない構造にすることによって、耐放射線性を格段に向上させることに成功した。照射試験の結果、本モータは、市販モータの仕様限度の約700倍の$\gamma$線照射量まで照射しても健全であることが明らかになった。
    Fusion Engineering and Design, Dec. 2008, [Reviewed]
  • Fuels and materials irradiation test plan at JMTR: (3) Irradiation test plan of materials
    Nishiyama Yutaka; Chimi Yasuhiro; Ise Hideo; Nakamura Takehiko; Ishitsuka Etsuo; Tsukada Takashi, 原子力機構では、材料試験炉(JMTR)に材料照射試験装置を整備し、高経年化に対応する原子炉圧力容器鋼及び炉心シュラウド材料等の中性子照射試験を実施する計画である。その概要について報告する。, Atomic Energy Society of Japan
    Proceedings of Annual / Fall Meetings of Atomic Energy Society of Japan, 2008
  • In-situ tritium recovery behavior from Li$_{2}$TiO$_{3}$ pebble bed under neutron pulse operation
    Tsuchiya Kunihiko; Kikukawa Akihiro*; Hoshino Tsuyoshi; Nakamichi Masaru; Yamada Hirokazu*; Yamaki Daiju; Enoeda Mikio; Ishitsuka Etsuo; Kawamura Hiroshi; Ito Haruhiko; Hayashi Kimio, チタン酸リチウム(Li$_{2}$TiO$_{3}$)は、核融合炉ブランケットで用いるトリチウム増殖材料の有望な候補材の1つである。大小2種類(直径2mm及び0.3mm)のLi$_{2}$TiO$_{3}$微小球を混合充填した充填体を中性子パルス運転が模擬できる照射試験体に装荷し、中性子吸収体を回転させた後一定出力とした時と、中性子吸収体を一定間隔でパルス運転した時のトリチウム生成回収特性を調べるための照射試験をJMTRを用いて行った。その結果、R/G(トリチウム回収率との生成率の比)はパルス運転に伴って、波を描きながら増加したが、マクロ的なトリチウム生成回収挙動は一定出力運転させたものと時定数がほとんど変わらないことがわかった。この原因は、トリチウム回収速度の時定数がパルスの周期より十分長いためで、パルス運転の影響はほとんどないことに起因しているものと考えられる。
    Journal of Nuclear Materials, Aug. 2004, [Reviewed]
  • Radiation-induced thermoelectric sensitivity in the mineral-insulated cable of magnetic diagnostic coils for ITER
    Nishitani Takeo; Vayakis G.*; Yamauchi Michinori*; Sugie Tatsuo; Kondoh Takashi; Shikama Tatsuo*; Ishitsuka Etsuo; Kawashima Hisato, ITERではプラズマ位置制御用磁気センサーとして無機絶縁(MI)ケーブルを用いた磁気コイルを用いるが、これまでの照射試験においてMIケーブルの中心導体と外皮導体間に数Vの起電力(RIEMF)が発生することが観測されており、磁気計測に与える影響が懸念されていた。そこで磁気コイルをJMTRで照射し、照射中に磁気コイルの中心導体両端に発生する起電力を高感度電圧計で測定した。これまで懸念されていたRIEMFによる中心導体両端間の起電力は十分小さいことを確認したが、中性子フルエンスの増加とおもに熱起電力が発生する現象を観測した。この結果をITERの使用条件に外挿すると、1000秒以上の長時間運転では問題となる可能性があることを指摘した。
    Journal of Nuclear Materials, Aug. 2004, [Reviewed]
  • Development of supercritical pressure water cooled solid breeder blanket in JAERI
    Akiba Masato; Ishitsuka Etsuo; Enoeda Mikio; Nishitani Takeo; Konishi Satoshi, 原研における超臨界圧水を冷却水に用いた核融合発電プラント用ブランケットの設計,開発の現状に関するレビュー論文である。原研では超臨界圧水を用いた核融合発電プラントの概念設計を進めた結果、システムの発電効率として40\%以上が得られる見通しを得た。この成果に基づき、発電プラント用ブランケットのより詳細な構造検討を実施した。まず2次元コードを用いてブランケット内の固体増殖・増倍材の温度分布を評価し、各々の充填層の厚さを決定した。これに基づいて2次元輸送コードを用いてトリチウム増殖比の評価した結果、局所で1.4以上、全体で1以上のTBRを得られる見通しを得た。さらに複雑な構造の製作手法として高温等方加圧法を採用して第一壁の模擬試験体を製作し、5000回以上の熱サイクルに耐えることを実証した。
    Purazuma, Kaku Yugo Gakkai-Shi, Sep. 2003, [Reviewed]
  • Thermal conductivity of neutron irradiated Be$_{12}$Ti
    Uchida Munenori*; Ishitsuka Etsuo; Kawamura Hiroshi, 原型炉用中性子増倍材として期待されているBe$_{12}$Tiについて、ブランケット内での熱的特性を評価するために、未照射及び中性子照射したBe$_{12}$Tiの熱伝導率を測定した。ベリリウム及びチタンの粉末からHIP法で製作したBe$_{12}$Tiサンプル($\phi$8 mm$\times^{t}$2mm) をJMTRで高速中性子フルエンス(E$>$1MeV) 4$\times$10$^{20}$ n/cm$^{2}$の条件で330,400 and 500$^{\circ}$Cにおいて照射した。熱拡散率と比熱をレーザーフラッシュ法にて1000$^{\circ}$Cまで測定し、熱伝導率を計算した。中性子照射したBe$_{12}$Tiは、照射による熱伝導率の低下が見られたが、充填層の有効熱伝導率を計算モデルにより推測したところ、十分に設計可能な範囲内であった。
    Fusion Engineering and Design, Sep. 2003, [Reviewed]
  • Development of EC H\&CD launcher components for fusion device
    Takahashi Koji; Ishitsuka Etsuo; Moeller C. P.*; Hayashida Kazunori*; Kasugai Atsushi; Sakamoto Keishi; Hayashi Kenichi*; Imai Tsuyoshi, For the purpose to confirm the reliability of the front steering concept, tests of the front steering launcher mock-up and neutron irradiation tests of bearings for a movable mirror were carried out under the ITER conditions. The max. stress of the flexible pipes for the movable mirror was measuerd to be 60MPa, much less than the allowable stress and agree with the calculation. No degradation on the performance of bearings was obtained for the assumed fluence of ITER. The mock-up of remote steering(RS) launcher was tested and RF radiation of 0.5MW-3sec and 0.2MW-10sec over 0-10$^{\circ}$ were performed. No degradation was obseved, either. The new launcher which has the capability of wider range rf beam steering and the corrugated square waveguide with four miter bends that yield the dog-legged structure has been designed, based on the test results. It has been fabricating for high power experiments.
    Fusion Engineering and Design, Sep. 2003, [Reviewed]
  • Development of advanced blanket materials for a solid breeder blanket of a fusion reactor
    Kawamura Hiroshi; Ishitsuka Etsuo; Tsuchiya Kunihiko; Nakamichi Masaru; Uchida Munenori*; Yamada Hirokazu*; Nakamura Kazuyuki; Ito Haruhiko; Nakazawa Tetsuya; Takahashi Heishichiro*; Tanaka Satoru*; Yoshida Naoaki*; Kato Shigeru*; Ito Yoshio*, 核融合原型炉を実現するために、先進ブランケットの設計研究が行われている。これらの設計では、より高い発電効率を目指して冷却材温度を500$^{\circ}C$以上としており、高温に耐え、また高中性子照射量まで使用できるブランケット材料(トリチウム増殖材料及び中性子増倍材料)の開発が求められている。本論文では、原研及び国内の大学、産業界が共同で実施してきたこれら先進ブランケット材料の開発の現状について報告する。トリチウム増殖材料に関しては、トリチウム放出特性に悪影響を及す高温での結晶粒径成長を抑制できる材料の開発として、TiO$_{2}$を添加したLi$_{2}$TiO$_{3}$に注目し、湿式造粒法による微小球の製造技術開発を実施した。この結果、固体ブランケットに用いる微小球製造に見通しが得られた。中性子増倍材料に関しては、融点が高く化学的に安定な材料としてベリリウム金属間化合物であるBe$_{12}$Ti等に注目し、回転電極法による微小球の製造技術開発及び特性評価を実施した。この結果、ベリリウムの含有量を化学量論値より多くすることにより、延性を増すことによって、微小球の製造に見通しが得られた。また、Be$_{12}$Tiはベリリウムより中性子照射によるスエリングが小さいことなど、優れた特性を有していることが明らかとなった。
    Nuclear Fusion, Aug. 2003, [Reviewed]
  • Tritium release properties of neutron-irradiated Be$_{12}$Ti
    Uchida Munenori*; Ishitsuka Etsuo; Kawamura Hiroshi, Be$_{12}$Ti has high melting point and good chemical stability and is expected as the advanced material for the neutron multiplier of DEMO-Reactor that requires higher temperature than 600$^{\circ}$C in a blanket. To evaluate the tritium inventory in the breeding blanket, tritium release experiment of the neutron irradiated Be$_{12}$Ti irradiated with a total fast neutron fluence of about 4 x 10$^{22}$ n/cm$^{2}$ (E$>$1MeV) at 330, 400 and 500$^{\circ}$C, was carried out. It was clear that tritium could be released easier than beryllium, and the apparent diffusion coefficient of Be$_{12}$Ti was about two orders larger than that of beryllium at 600$\sim$1100$^{\circ}$C. In addition to good tritium release property, the swelling calculated from the density change of the specimens up to 1100$^{\circ}$C in this test was smaller than that of beryllium.
    Journal of Nuclear Materials, Dec. 2002, [Reviewed]
  • Heat load rest of Be/Cu joint for ITER first wall mock-ups
    Uchida Munenori*; Ishitsuka Etsuo; Hatano Toshihisa; Barabash V.*; Kawamura Hiroshi, ITER第1壁の開発を目的として、Al,Ti,Cuから成る中間層及びCu中間層を用いて製作したベリリウム/銅合金接合体の熱負荷試験を実施し、試験体の健全性を調べた。除熱性能を確認した後に、接合部温度が約200$^{\circ}$Cとなる熱負荷条件(5MW/m$^{2}$)で15秒加熱,15秒冷却で1000回の熱サイクル試験を実施した。Al/Ti/Cu中間層の試験体は、1000回まで良好な除熱性能を維持したが、Cu中間層を用いた試験体は除熱性能の低下が見られた。試験後、接合部の断面を調べた結果、接合体のコーナー部の接合界面において剥離が確認され、これが熱伝導の低下を招いたものと推定された。
    Journal of Nuclear Materials, Dec. 2002, [Reviewed]
  • In-situ irradiation test of mica substrate bolometer at the JMTR reactor for the ITER diagnostics
    Nishitani Takeo; Shikama Tatsuo*; Reichle R.*; Hodgson E. R.*; Ishitsuka Etsuo; Kasai Satoshi; Yamamoto Shin, ITER-EDAの工学R\&Dの一環として行ったボロメーターの照射試験の結果について報告する。ボロメーターは赤外$\sim$軟X線領域の輻射を測定する素子であり、プラズマのパワーバランスを評価する重要な計測器である。ITERのボロメーターの候補であるマイカ薄膜ボロメーターの実時間照射試験をJMTRを用いて行った。原子炉出力50MWで25日間を1照射サイクルとして3サイクル照射し、全高速中性子フルエンスは0.1dpa(ITER用ボロメータの目標値)であった。照射中、マイカ薄膜に蒸着した金の抵抗体の抵抗値の著しい増加が観測され、照射後試験により分析したところ、金から水銀への核変換で水銀が45\%生成していることを確認した。感度及び応答時間はほとんど一定であったが、0.03dpaのフルエンスで断線が発生した。照射後試験時に観察したところ、マイカ薄膜自体は健全であったが、金の抵抗体に断線が発生していることを確認した。このため、蒸着抵抗体を白金等の核変換断面積が小さい物質に代える必要があることを指摘した。
    Fusion Engineering and Design, Dec. 2002, [Reviewed]
  • Application of beryllium intermetallic compounds to neutron multiplier of fusion blanket
    Kawamura Hiroshi; Takahashi Heishichiro*; Yoshida Naoaki*; Shestakov V.*; Ito Yoshio*; Uchida Munenori*; Yamada Hirokazu*; Nakamichi Masaru; Ishitsuka Etsuo, 高温ブランケット用の中性子増倍材として期待されているベリリウム金属間化合物に関し、日本国内での開発現状について報告する。ベリリウム金属間化合物の開発は、原研,大学,企業が協力して実施している。ベリリウム金属間化合物の一つであるBe$_{12}$Tiに関し、従来のベリリウム金属より、構造材との両立性が良いこと,スエリングが小さいこと,機械強度が高いこと,トリチウムインベントリが小さいことなどの優れた特性を有することが明らかとなった。また、ベリリウム金属間化合物は機械的に脆く、熱応力が生じる回転電極法で微小球を製造することができなかったが、組織制御によって延性を持たせることによって、微小球を製造できる見通しが得られた。
    Fusion Engineering and Design, Nov. 2002, [Reviewed]
  • Development of fusion nuclear technologies at Japan Atomic Energy Research Institute
    Seki Masahiro; Yamanishi Toshihiko; Shu Wataru; Nishi Masataka; Hatano Toshihisa; Akiba Masato; Takeuchi Hiroshi; Nakamura Kazuyuki; Sugimoto Masayoshi; Shiba Kiyoyuki; Jitsukawa Shiro; Ishitsuka Etsuo; Tsuji Hiroshi, Latest status on development of long-term fusion nuclear technologies at JAERI is overviewed. A tritium processing system for the ITER and DEMO reactors was designed and basic technologies for each component of this system was demonstrated successfully by an operation of the integrated system for one month. An ultra-violet laser with a wave length of 193 nm was found quite effective for removing tritium from in-vessel components of D-T fusion reactors. Blanket technologies have been developed for the Test Blanket Module of the ITER and for advanced blankets for DEMO reactors. This blanket is composed of Li$_{2}$TiO$_{3}$ breeder pebbles and neutron multiplier Be pebbles, contained in a box structures made of a reduced activation ferritic steel F82H. Mechanical properties of F82H under neutron irradiation up to 50 dpa were obtained in a temperature range from 200 to 500$^{\circ}$C. Design of the International Fusion Materials Irradiation Facility (IFMIF) has been developed so as to obtain engineering data for candidate materials for DEMO reactors, under neutron irradiation up to 100-200 dpa.
    Fusion Science and Technology, Jul. 2002, [Reviewed]
  • Irradiation Effects on Diagnostic Components for ITER
    NISHITANI Takeo; SHIKAMA Tatsuo; REICHEL Roger; SUGIE Tatsuo; KAKUTA Tsunemi; KASAI Satoshi; ISHITSUKA Etsuo; YAMAMOTO Shin, Irradiation tests on many diagnostic components have been carried out as a part of the R&D program of the ITER Engineering Design Activities (EDA). Five kinds of ITER round robin fibers were irradiated in JMTR and60Co γ-rays. Induced transmission loss of those fibers are much smaller than that of pure SiO2 core fiber. Especially, KU-H2G from Russian Federation and F-doped fibers from Japan have rather good radiation hardness even in the visible range. Those fibers might be available in the diagnostic port of ITER. The transmissivity of KU-1 quartz, which is a candidate of the window material for UV and visible spectroscopy in ITER, was measured under the irradiation of 14 MeV neutrons and γ-rays in the UV range (200 - 400 nm). Significant transmission loss was observed in the wavelength range of 200-300 nm. The Mica substrate bolometer was irradiated in JMTR during three irradiation cycles. Total neutron fluence was about 1×1024 n/m2 which was equivalent to 0.1 dpa. Significant increase in the meander resistance was observed, which was causedby the nuclear transmutation of gold into mercury. The use of gold meanders might be problematic in ITER., The Japan Society of Plasma Science and Nuclear Fusion Research
    Kakuyūgō kenkyū, 25 May 2002, [Reviewed]
  • In situ characterization of a small sized motor under neutron irradiation
    Ishitsuka Etsuo; Kan S.*; Kawamura Hiroshi; Onozawa H.*, ポリイミド巻線を使用した耐放射線小型モータを開発し、JMTRを用いて照射試験を実施した。耐放射線小型モータには、フィールドコイルとしてポリイミド巻線、マグネットとしてNd-Fe、ベアリング等の潤滑剤としてポリフェニルエーテルを用い、フィールドコイルはMgO,Al$_{2}$O$_{3}$を充填したシリコン樹脂で固定した。耐放射線小型モータは約50$^{\circ}C$で照射し、$\gamma$線量率と高速中性子束はそれぞれ7.4$\times$10$^{1}$Gy/sと6.6$\times$10$^{14}$n/m$^{2}$/sであった。モータの回転試験を実施した結果、$\gamma$線量及び高速中性子照射量が3.1$\times$10$^{7}$Gy/sと2.8$\times$10$^{20}$n/m$^{2}$まで正常に回転した。また、フィールドコイルの絶縁抵抗及び導体抵抗を測定した結果、$\gamma$線量及び高速中性子照射量が3.1$\times$10$^{8}$Gy及び2.8$\times$10$^{21}$n/m$^{2}$においても導体抵抗及び絶縁抵抗が1$\times$10$^{8}\Omega$及び12$\Omega$であり、照射開始時より顕著な劣化は観察されなかった。
    Fusion Engineering and Design, Nov. 2001, [Reviewed]
  • High heat load tests of neutron-irradiated divertor mockups
    Ishitsuka Etsuo; Uchida Munenori*; Sato Kazuyoshi; Akiba Masato; Kawamura Hiroshi, 炭素繊維強化炭素複合材とアルミナ分散強化銅からなるダイバータモックアップを中性子照射し、高熱負荷試験を実施した。試料の照射条件は、照射温度が約300$^{\circ}C$、照射損傷量が0.3及び0.4dpaであった。高熱負荷試験は、熱流束を5MW/m$^{2}$、加熱及び冷却時間を10秒として実施した。この際、冷却水の流速及び圧力は、各々11m/s及び1.5MPaであった。試験の結果、0.3dpaまで照射した試料の表面温度は約800$^{\circ}C$となり、未照射試料により約400$^{\circ}C$高くなり、0.4dpaの試料では1100$^{\circ}C$となることが明らかとなった。この原因は、中性子照射によって、炭素繊維強化炭素複合材の熱伝導率が低下したためと考えられる。さらに、同じ高熱負荷試験条件で1000回の熱サイクル試験を実施した結果、炭素繊維強化複合材とアルミナ分散強化銅の剥離はなく、冷却性能が低下しないことを確認した。
    Fusion Engineering and Design, Oct. 2001, [Reviewed]
  • Thermal cycle experiments of neutron-irradiated CFC/Cu mock-ups
    Sato Kazuyoshi; Ishitsuka Etsuo; Uchida Munenori*; Kawamura Hiroshi; Ezato Koichiro; Taniguchi Masaki; Akiba Masato, 2種類のアーマ材からなるダイバータ模擬試験体を中性子照射して高熱負荷試験を実施し、アーマ材の影響を調べた。試験体は、1次元及び2次元の炭素繊維強化炭素複合(CFC)アーマ材とアルミナ分散強化銅製冷却構造体からなり、無酸素銅の中間層を介して銀ろうで接合した構造である。試験体の照射温度は280~320$^{\circ}C$、照射損傷量0.3~0.5dpaである。本試験体をITER定常熱負荷条件を模擬した5MW/m$^{2}$で10s間の加熱を実施した結果、照射量0.43dpaの1次元材及び2次元CFC材の表面温度は、それぞれ650$^{\circ}C$及び1200$^{\circ}C$に達し、未照射材より高くなった。これ、CFC材の熱伝導率が中性子照射によって低下したためであるが、その低下する割合は1次元及び2次元とも同程度であった。また、1000回の熱サイクル試験を実施した結果、接合部の剥離等は認められなかった。
    Physica Scripta, Jul. 2001, [Reviewed]
  • Erosion characteristics of neutron-irradiated carbon-based materials under simulated disruption heat loads
    Sato Kazuyoshi; Ishitsuka Etsuo; Uda Minoru*; Kawamura Hiroshi; Suzuki Satoshi; Taniguchi Masaki; Ezato Koichiro; Akiba Masato, 中性子照射後炭素系材料の熱衝撃による損耗特性を調べるため、JMTRホットセル内に設置した電子ビーム加熱装置(OHBIS)を使用し、熱衝撃試験を実施した。その結果、試料の損耗量は中性子照射量が増えるに従って増加し、特に、中性子照射量0.46dpaの損耗量は、未照射材の約2倍に達することが明らかとなった。さらに未照射材と照射材の損耗形状を比較した結果、最大損耗深さの変化は認められず、損耗重量の差は、損耗形状がブロードになったため生じることがわかった。しかしながら、中性子照射後材料の熱衝撃試験では、試験中にビーム電流の減少が認められた。これは、中性子照射による熱伝導率の低下により損耗量が増大したため、試料への実質的な熱負荷が減少したためと思われる。このため、実負荷の減少を考慮に入れ熱解析を実施した。その結果、実験結果と同様に最大損耗深さは変化しないことが明らかとなった。
    Journal of Nuclear Materials, Dec. 2000, [Reviewed]
  • Effects of helium production and irradiation damage on tritium release behavior of neutron-irradiated beryllium pebbles
    Ishitsuka Etsuo; Kawamura Hiroshi; Terai Takayuki*; Tanaka Satoru*, ベリリウムからのトリチウム放出挙動に関しては、これまでに表面酸化膜及び結晶粒径の効果について報告されているが、ヘリウム生成量及び照射損傷量の効果については報告されていない。このため、ヘリウム生成量及び照射損傷量が異なる条件で照射したベリリウム微小球からのトリチウム放出特性を調べた。照射条件は3種類で照射温度が445,383,616($^{\circ}C$)、各照射温度に対応するヘリウム生成量及びdpaが7,5.1,10($\times$10$^{2}$appmHe)及び4,8.6,6である。これらの試料を用いてトリチウム放出率測定試験を行った結果、照射損傷が大きい試料の見かけのトリチウム拡散係数が大きくなることが明らかになった。
    Journal of Nuclear Materials, Dec. 2000, [Reviewed]
  • Verification of tritium production evaluation procedure using Monte Carlo code MCNP for in-pile test of fusion blanket with JTMR
    Nagao Yoshiharu; Nakamichi Masaru; Tsuchiya Kunihiko; Ishitsuka Etsuo; Kawamura Hiroshi, 核融合炉ブランケット炉内試験において、照射試験体内に装荷したトリチウム増殖材領域のトリチウム生成量評価のため、モンテカルロ(MCNP)コードを用いた評価手法の検証を行った。本検証のため、予備照射試験として、リチウム-アルミニウム合金を用いたトリチウムモニタ及び中性子フルエンスモニタを3次元的に複数個装荷した照射試験体を製作し、JMTRにおいて照射し、各々のモニタの測定結果とMCNPによる計算結果との比較評価を行った。中性子フルエンスモニタによる高速中性子束の測定値とMCNP計算値を比較した結果、誤差は$\pm$10\%以内と比較的良く一致した結果が得られ、トリチウム生成量評価の技術的な見通しを得ることができた。本国際会議では、トリチウムモニタによる測定値とMCNP計算値との比較結果についても報告する。
    Fusion Engineering and Design, Nov. 2000, [Reviewed]
  • Compression properties of neutron irradiated beryllium pebbles
    Ishitsuka Etsuo; Kawamura Hiroshi; Terai Takayuki*; Tanaka Satoru*, ベリリウム微小球は、核融合炉ブランケットの中性子増倍材として検討されているが、これまでに中性子照射データがほとんど取得されていない。このため、回転電極法及びMg還元法で製造した2種類のベリリウム微小球を中性子照射し、機械的特性を調べた。照射条件は、ヘリウム生成量が約500appm、dpaが約8、照射温度が400,500,600$^{\circ}$Cである。この結果、2種類のベリリウム微小球の強度はほとんど変わらないことが明らかになった。また、回転電極法で製造したベリリウム微小球に関して、これまでのデータと比較したところ、ヘリウム生成量が約500appmの場合、dpaが4から8になると強度が約7割に低下することが明らかになった。
    Fusion Engineering and Design, Nov. 2000, [Reviewed]
  • Irradiation tests on diagnostics components for ITER in 1995
    Nishitani, Takeo; Ikeda, Yujiro; Kakuta, Tsunemi; Kasai, Satoshi; Kawamura, Hiroshi; Maekawa, Fujio; Morita, Yosuke; Nagashima, Akira; Noda, Kenji; Oyama, Yukio; Sagawa, Hisashi; Sugie, Tatsuo; Yamaki, Daiju, ITER用計測装置の開発において最も重要な課題は計測機器要素の放射線照射損傷である。ITER工学設計活動の一環として、ボロメータ等の真空容器内計測センサー及びセラミックス、窓材、光ファイバー等の光/信号伝送用の基本要素の照射試験を実施した。FNSにおいて14MeV中性子に対するセラミックスの照射誘起伝導及び窓材の照射誘起発光の測定を行った。またJMTRでは窓材、光ファイバーの透過損失測定及び反射鏡のオフライン照射試験を行った。Co$^{60}$$gamma$線照射下においてボロメータの特性測定を行った。
    JAERI-Tech 96-040, Oct. 1996
  • Present Status of Investigation for Fusion Reactor with JMTR
    Kawamura Hiroshi; Sagawa Hisashi; Ishitsuka Etsuo, Components such as a blanket or divertor and instruments are exposed to a severe neutron irradiation field. As regards engineering, it is needed to grasp the characteristics of the components and instruments under neutron irradiation. Especially, a fusion blanket is the center of components, because it breeds tritium, generating heat energy and shielding neutron. Therefore, it is necessary to evaluate the in-pile functions of the blanket under neutron irradiation so as to construct ITER. JMTR of the Oarai Establishment at JAERI is conducting in-pile functional tests of a blanket mock-up, the development of blanket materials, the development of instruments, a re-weldability test for vacuum vessel materials, an electron beam irradiation test of the divertor and armor materials after neutron irradiation and so on. This report briefly describes the present status of each study., Toyama University
    Annual report of Hydrogen Isotope Research Center, Toyama University, 1995
  • Surface Characterization of Hot-Pressed Beryllium with X-ray Photctron Spectroscopy
    Ishitsuka Etsuo; Kawamura Hiroshi; Ashida Kan, In the case of using hot-pressed beryllium in a fusion reactor, the surface state of hot-pressed beryllium is one of the items most necessary to investigate the behavior of hydrogen in beryllium. Therefore, the surface characterization of the hot-pressed beryllium was examined by X-rays photoelectron spectroscopy (XPS) after vacuum heating and deuterium ion bombardment. On the surface of the as-received sample, carbon, fluorine, oxygen and beryllium were observed. However, carbon that is sure to be adsorbed and fluorine that was mixed as an impurity during fabrication process decreased to a great extent by heating, and both elements (carbon and fluorine) were barely observed after heating at 800℃for 20 min. From the element-abundance-ratio of Be to O and ratio of the two peaks of Be 1s, it was obvious that surface of hot-pressed beryllium was covered by two parts of Be and one of BeO due to heating at 800℃ for 20 min. Additionally, it was observed that the oxidation of the surface of hot-pressed beryllium is induced by deuterium implantation using an ion gun., Toyama University
    Annual report of Tritium Research Center, Toyama University, Japan, 1988

MISC

Lectures, oral presentations, etc.

Affiliated academic society

  • ATOMIC ENERGY SOCIETY OF JAPAN

Research Themes

Industrial Property Rights

  • 5403605, 2011-007733, 2009-153749, 放射線照射装置
    石塚 悦男, 稲葉 良知, 蓼沼 克嘉*
  • 5441096, 2010-127825, 2008-304522, ラジオアイソトープシートの製造方法
    石塚 悦男, 稲葉 良知, not registered
  • 2009-036379, 2008-247754, 転がり軸受
    高橋 幸司, 石塚 悦男, not registered, not registered
  • 4618732, 2008-102078, 2006-286159, 放射線モリブデンの製造方法と装置
    石塚 悦男, 蓼沼 克嘉*
  • 2104113, 07830597.6, 放射性モリブデンの製造方法及び製造装置並びにその方法によって製造されたモリブデン
    石塚 悦男, not registered
  • 2666570, 2666570, MANUFACTURING METHOD OF RADIOACTIVE MOLYBDENUM,MANUFACTURING APPARATUS AND RADIOACTIVE MOLYBDENUM MANUFACTURED THEREBY
    石塚 悦男, not registered
  • 4220754, H16-144131, H14-307212, アンテナミラー
    高橋 幸司, 石塚 悦男, 今井 剛, not registered
  • H13-286083, H12-092772, 耐放射線性コイルと耐放射線性モータ
    河村 弘, 石塚 悦男, 中道 勝, not registered
  • 3076067, H10-811556, H10-513498, 核融合炉用の金属ベリリウムペブル
    河村 弘, 石塚 悦男, not registered
  • 0872851, 0872851, 979403, Process for preparing metallic beryllium pebbles
    河村 弘, 石塚 悦男, not registered
  • H10-253526, H09-058165, 分光分析装置
    河村 弘, 石塚 悦男, not registered
  • 5958105, 09/068, Process for preparing metallic beryllium pebbles
    河村 弘, 石塚 悦男, not registered
  • 3076068, H10-513499, 金属ベリリウムペブルの製造方法
    石塚 悦男, 河村 弘, not registered