Nihon Genshiryoku Gakkai-Shi ATOMO, 64(11), p.617 - 621, 2022/11
In December 2016 Decisions were made by the Government on the Fast Breeder Prototype Reactor "Monju", which were decommissioning of the reactor and installation of a new research reactor at the Monju site. After the decisions, MEXT started research to list reactor candidates suitable for the site. Among the candidates, medium power reactor type of which thermal output less than 10,000 kW was chosen to utilize neutron beams. Then, from 2020, MEXT launched an entrusted business and adopted JAEA, Kyoto University and University of Fukui as the core institutions of the business to carry out the conceptual design. This paper describes the system to proceed the conceptual design and to examine the utilization management of the new research reactor and also shows their status.
Lu, K.; Takamizawa, Hisashi; Katsuyama, Jinya; Li, Y.
International Journal of Pressure Vessels and Piping, 199, p.104706_1 - 104706_13, 2022/10
Naoe, Takashi; Kinoshita, Hidetaka; Wakui, Takashi; Kogawa, Hiroyuki; Haga, Katsuhiro
JAEA-Technology 2022-018, 43 Pages, 2022/08
In the liquid mercury target system for the pulsed spallation neutron source of Materials and Life science experimental Facility (MLF) at the Japan in the Japan Proton Accelerator Research Complex (J-PARC), cavitation that is generated by the high-energy proton beam-induced pressure waves, resulting severe erosion damage on the interior surface of the mercury target vessel. The erosion damage is increased with increasing the proton beam power, and has the possibility to cause the leakage of mercury by the penetrated damage and/or the fatigue failure originated from erosion pits during operation. To achieve the long term stable operation under high-power proton beam, the mitigation technologies for cavitation erosion consisting of surface modification on the vessel interior surface, helium gas microbubble injection, double-walled beam window structure has been applied. The damage on interior surface of the vessel is never observed during the beam operation. Therefore, after the target operation term ends, we have cut out specimen from the target nose of the target vessel to inspect damaged surface in detail for verification of the cavitation damage mitigation technologies and lifetime estimation. We have developed the techniques of specimen cutting out by remote handling under high-radiation environment. Cutting method was gradually updated based on experience in actual cutting for the used target vessel. In this report, techniques of specimen cutting out for the beam entrance portion of the target vessel in high-radiation environment and overview of the results of specimen cutting from actual target vessels are described.
Lu, K.; Katsuyama, Jinya; Li, Y.; Yoshimura, Shinobu*
Journal of Pressure Vessel Technology, 143(2), p.021505_1 - 021505_8, 2021/04
Kenzhina, I.*; Ishitsuka, Etsuo; Ho, H. Q.; Sakamoto, Naoki*; Okumura, Keisuke; Takemoto, Noriyuki; Chikhray, Y.*
Fusion Engineering and Design, 164, p.112181_1 - 112181_5, 2021/03
Tritium release into the primary coolant during operation of the JMTR (Japan Materials Testing Reactor) and the JRR-3M (Japan Research Reactor-3M) had been studied. It is found that the recoil release by Li(n,)H reaction, which comes from a chain reaction of beryllium neutron reflectors, is dominant. To prevent tritium recoil release, the surface area of beryllium neutron reflectors needs to be minimum in the core design and/or be shielded with other material. In this paper, as the feasibility study of the tritium recoil barrier for the beryllium neutron reflectors, various materials such as Al, Ti, V, Ni, and Zr were evaluated from the viewpoint of the thickness of barriers, activities after long-term operations, and effects on the reactivities. From the results of evaluations, Al would be a suitable candidate as the tritium recoil barrier for the beryllium neutron reflectors.
Kenzhina, I.*; Ishitsuka, Etsuo; Okumura, Keisuke; Ho, H. Q.; Takemoto, Noriyuki; Chikhray, Y.*
Journal of Nuclear Science and Technology, 58(1), p.1 - 8, 2021/01
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 Be 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 research and testing reactors containing beryllium reflectors.
Chikhray, Y.*; Askerbekov, S.*; Kenzhin, Y.*; Gordienko, Y.*; Ishitsuka, Etsuo
Fusion Science and Technology, 76(4), p.494 - 502, 2020/05
Ho, H. Q.; Honda, Yuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Takada, Shoji; Fujimoto, Nozomu*; Ishitsuka, Etsuo
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021902_1 - 021902_6, 2020/04
Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09
Tobita, Toru; Nishiyama, Yutaka; Onizawa, Kunio
JAEA-Data/Code 2018-013, 60 Pages, 2018/11
Mechanical properties of materials including fracture toughness are extremely important for evaluating the structural integrity of reactor pressure vessels (RPVs). In this report, the published data of mechanical properties of nuclear RPVs steels, including neutron irradiated materials, acquired by the Japan Atomic Energy Agency (JAEA), specifically tensile test data, Charpy impact test data, drop-weight test data, and fracture toughness test data, are summarized. There are five types of RPVs steels with different toughness levels equivalent to JIS SQV2A (ASTM A533B Class 1) containing impurities in the range corresponding to the early plant to the latest plant. In addition to the base material of RPVs, the mechanical property data of the two types of stainless overlay cladding materials used as the lining of the RPV are summarized as well. These mechanical property data are organized graphically for each material and listed in tabular form to facilitate easy utilization of data.
Rodriguez, D.; Rossi, F.; Seya, Michio; Koizumi, Mitsuo
Proceedings of 2017 IEEE Nuclear Science Symposium and Medical Imaging Conference (NSS/MIC 2017) (Internet), 3 Pages, 2018/11
Proceedings of 13th International Symposium on Advanced Science and Technology in Experimental Mechanics (13th ISEM'18) (USB Flash Drive), 6 Pages, 2018/10
Issues on the engineering technologies relating to high-power spallation neutron sources with liquid metals are introduced. The present status on research activities and results was reviewed.
Ho, H. Q.; Honda, Yuki*; Hamamoto, Shimpei; Ishii, Toshiaki; Takada, Shoji; Fujimoto, Nozomu*; Ishitsuka, Etsuo
Proceedings of 9th International Topical Meeting on High Temperature Reactor Technology (HTR 2018) (USB Flash Drive), 6 Pages, 2018/10
Chimi, Yasuhiro; Kasahara, Shigeki; Seto, Hitoshi*; Kitsunai, Yuji*; Koshiishi, Masato*; Nishiyama, Yutaka
Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Vol.2, p.1039 - 1054, 2018/00
In order to understand irradiation-assisted stress corrosion cracking (IASCC) growth behavior, crack growth rate (CGR) tests have been performed in simulated Boiling Water Reactor water conditions at 288C on neutron-irradiated 316L stainless steels (SSs) at 12-14 dpa. After the tests, the microstructures near the crack tip of the specimens are examined with scanning transmission electron microscope (FE-STEM). In comparison with a previous study at 2 dpa, this result shows a less benefit of low electrochemical corrosion potential (ECP) conditions on CGR. A crack tip immersed over 1000 hours was filled with oxides, while almost no oxide film was observed near the crack front in the low-ECP conditions. In addition, a high density of deformation twins and dislocations were found near the fracture surface of the crack front. It is considered that both localized deformation and oxidation are possible dominant factors for the SCC growth in highly irradiated SSs.
Sekio, Yoshihiro; Yamagata, Ichiro; Akasaka, Naoaki; Sakaguchi, Norihito*
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Next Generation Nuclear Systems for Sustainable Development (FR-17) (USB Flash Drive), 8 Pages, 2017/06
The widths of void denuded zones (VDZs) which were formed near random grain boundaries by neutron irradiation were analyzed in order to perform quantitative evaluations for the irradiation-induced point defect behavior in the modified 316 stainless steel (PNC316) having been developed by JAEA. Namely, the temperature dependence of VDZ width was investigated and vacancy migration energy of the PNC316 steel was estimated from the VDZ width analysis for the neutron-irradiated specimens. The obtained value of vacancy migration energy was estimated as 1.46 eV, which was consistent with that from the exiting method using electron in-situ examination. This indicates that VDZ analysis could be effective method to evaluate especially vacancy migration energy during irradiation, and this would be realized from not in-situ observation but post-irradiation examination in the case of neutron irradiation.
Kasahara, Shigeki; Kitsunai, Yuji*; Chimi, Yasuhiro; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka
Journal of Nuclear Materials, 480, p.386 - 392, 2016/11
This paper addresses influence of two different temperature profiles during startup periods in the Japan Materials Testing Reactor and a boiling water reactor upon microstructural evolution and mechanical properties of austenitic stainless steel irradiated with neutrons to about 1 dpa and 3 dpa. Tensile tests at 290C and Vickers hardness tests at room temperature were carried out, and their microstructures were observed by FEG-TEM. Influence of difference in the temperature profiles was observed obviously in interstitial cluster formation, in particular, growth of Frank loops. The influence was also found certainly in loss of strain hardening capacity and ductility, although the influence on the yield strength and the Vickers hardness was not clearly observed. As a result, Frank loops, which were observed in austenitic stainless steel irradiated at doses of 1 dpa or more, were considered to contribute to deformation of the austenitic stainless steel.
Chimi, Yasuhiro; Takamizawa, Hisashi; Kasahara, Shigeki*; Iwata, Keiko; Nishiyama, Yutaka
Nuclear Engineering and Design, 307, p.411 - 417, 2016/10
To investigate influential parameters for irradiation-assisted stress corrosion cracking (IASCC) growth behavior, we attempt to analyze statistically existing data on the crack growth rate (CGR) in irradiated austenitic stainless steels (SSs) in boiling water reactor (BWR) environments using the Bayesian nonparametric (BNP) method. From the probability distribution of CGR and some input parameters, such as yield stress of irradiated material (), stress intensity factor (), electrochemical corrosion potential (ECP), and fast neutron fluence, the mean CGR is estimated and compared with the measured CGR. The analytical results show good reproducibility of the measured CGR. The results also indicate the possible neutron fluence effects on CGR in high CGR region (i.e., high neutron fluence condition) by radiation-induced segregation (RIS), localized deformation, and/or other mechanisms than radiation hardening.
Chimi, Yasuhiro; Kitsunai, Yuji*; Kasahara, Shigeki; Chatani, Kazuhiro*; Koshiishi, Masato*; Nishiyama, Yutaka
Journal of Nuclear Materials, 475, p.71 - 80, 2016/07
To elucidate the mechanism of irradiation-assisted stress corrosion cracking (IASCC) in high-temperature water for neutron-irradiated austenitic stainless steels (SSs), the locally deformed structures, the oxide films formed on the deformed areas, and their correlation were investigated. Tensile specimens made of irradiated 316L SSs were strained 0.1%-2% at room temperature or at 563 K, and the surface structures and crystal misorientation among grains were evaluated. The strained specimens were immersed in high-temperature water, and the microstructures of the oxide films on the locally deformed areas were observed. The appearance of visible step structures on the specimens' surface depended on the neutron dose and the applied strain. The surface oxides were observed to be prone to increase in thickness around grain boundaries (GBs) with increasing neutron dose and increasing local strain at the GBs. No penetrative oxidation was observed along GBs or along surface steps.
Takemoto, Noriyuki; Romanova, N.*; Kimura, Nobuaki; Gizatulin, S.*; Saito, Takashi; Martyushov, A.*; Nakipov, D.*; Tsuchiya, Kunihiko; Chakrov, P.*
JAEA-Technology 2015-021, 32 Pages, 2015/08
Silicon semiconductor production by neutron transmutation doping (NTD) method using the JMTR has been investigated in Neutron Irradiation and Testing Reactor Center, Japan Atomic Energy Agency in order to expand the industry use. As a part of investigations, irradiation test with a silicon ingot was planned using WWR-K in Institute of Nuclear Physics, Republic of Kazakhstan. A device rotating the ingot made with the silicon was fabricated and was installed in the WWR-K for the irradiation test. And that, a preliminary irradiation test was carried out using neutron fluence monitors to evaluate the neutronic irradiation field. Based on the result, two silicon ingots were irradiated as scheduled, and the resistivity of each irradiated silicon ingot was measured to confirm the applicability of high-quality silicon semiconductor by the NTD method (NTD-Si) to its commercial production.
Nishimura, Arata*; Muroga, Takeo*; Takeuchi, Takao*; Nishitani, Takeo; Morioka, Atsuhiko
Fusion Engineering and Design, 81(8-14), p.1675 - 1681, 2006/02
In a fusion reactor plant, a neutral beam injector (NBI) will be operated for a long time, and it will allow neutron streaming from NBI ports to outside of the plasma vacuum vessel. It requires the superconducting magnet to develop nuclear technology to produce stable magnetic field and to reduce activation of the magnet components. In this report, the back ground of the necessity and the contents of the nuclear technology of the superconducting magnets for fusion application are discussed and some typical investigation results are presented, which are the neutron irradiation effect on NbSn wire, the development of low activation superconducting wire, and the design concept to reduce nuclear heating and nuclear transformation by streaming. In addition, recent activities in high energy particle physics are introduced and potential ripple effect of the technology of the superconducting magnets is described briefly.