Tsai, T.-H.; Sasaki, Shinji; Maeda, Koji
Journal of Nuclear Science and Technology, 60(6), p.715 - 723, 2023/06
Takabe, Yugo; Otsuka, Noriaki; Fuyushima, Takumi; Sayato, Natsuki; Inoue, Shuichi; Morita, Hisashi; Jaroszewicz, J.*; Migdal, M.*; Onuma, Yuichi; Tobita, Masahiro*; et al.
JAEA-Technology 2022-040, 45 Pages, 2023/03
Because of the decommission of the Japan Materials Testing Reactor (JMTR), the domestic neutron irradiation facility, which had played a central role in the development of innovative nuclear reactors and the development of technologies to further improve the safety, reliability, and efficiency of light water reactors, was lost. Therefore, it has become difficult to pass on the operation techniques of the irradiation test reactors and irradiation technologies, and to train human resources. In order to cope with these issues, we conducted a study on the implementation of irradiation tests using overseas reactors as neutron irradiation sites as an alternative method. Based on the "Arrangement between the National Centre for Nuclear Research and the Japan Atomic Energy Agency for Cooperation in Research and Development on Testing Reactor," the feasibility of conducting an irradiation test at the MARIA reactor (30 MW) owned by the National Centre for Nuclear Research (NCBJ) using the temperature control system, which is one of the JMTR irradiation technologies, was examined. As a result, it was found that the irradiation test was possible by modifying the ready-made capsule manufactured in accordance with the design and manufacturing standards of the JMTR. After the modification, a penetration test, an insulation continuity test, and an operation test in the range of room temperature to 300C, which is the operating temperature of the capsule, were conducted and favorable results were obtained. We have completed the preparations prior to transport to the MARIA reactor.
Miyazawa, Takeshi; Kikuchi, Yuta*; Ando, Masami*; Yu, J.-H.*; Yabuuchi, Kiyohiro*; Nozawa, Takashi*; Tanigawa, Hiroyasu*; Nogami, Shuhei*; Hasegawa, Akira*
Journal of Nuclear Materials, 575, p.154239_1 - 154239_11, 2023/03
Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka
Journal of Nuclear Materials, 573, p.154110_1 - 154110_7, 2023/01
JAEA-Conf 2022-001, p.7 - 13, 2022/11
The partitioning and transmutation (P-T) technology has promising potential for volume reduction and mitigation of degree of harmfulness of high-level radioactive waste. JAEA is developing the P-T technology combined with accelerator driven systems (ADS). One of critical issues affecting the feasibility of ADS is the proton beam window (PBW) which functions as a boundary between the accelerator and the sub-critical reactor core. The PBW is damaged by a high-intensity proton beam and spallation neutrons produced in the target, and also by flowing high-temperature liquid lead bismuth eutectic alloy which is corrosive to steel materials. To study the materials damage under the ADS environment, J-PARC is proposing a plan of proton irradiation facility which equips with a liquid lead-bismuth spallation target bombarded by a 400 MeV - 250 kW proton beam. The facility is also open for versatile purposes such as soft error testing of semi-conductor devises, RI production, materials irradiation for fission and fusion reactors, and so on. Application to nuclear data research with using the proton beam and spallation neutrons is also one of such versatile purposes, and we welcome unique ideas from the nuclear data community.
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.
Shaimerdenov, A.*; Gizatulin, S.*; Dyussambayev, D.*; Askerbekov, S.*; Ueta, Shohei; Aihara, Jun; Shibata, Taiju; Sakaba, Nariaki
Nuclear Engineering and Technology, 54(8), p.2792 - 2800, 2022/08
Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka
Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06
Maekawa, Fujio; Takei, Hayanori
Purazuma, Kaku Yugo Gakkai-Shi, 98(5), p.206 - 210, 2022/05
In developing an accelerator-driven nuclear transmutation system (ADS), it is necessary to solve technical issues related to proton beams, such as the development of materials that can withstand high-intensity proton beams and the characterization of subcritical cores driven by proton beams. Therefore, at the high-intensity proton accelerator facility J-PARC, a transmutation experimental facility that actually conducts various tests using a high-intensity proton beam is being planned. This paper introduces the outline and future direction of the transmutation experimental facility.
Zhao, Q.*; Saito, Takeshi*; Miyakawa, Kazuya; Sasamoto, Hiroshi; Kobayashi, Taishi*; Sasaki, Takayuki*
Journal of Hazardous Materials, 428, p.128211_1 - 128211_10, 2022/04
The influence of humic acid and its radiological degradation on the sorption of Cs and Eu by sedimentary rock was investigated to understand the sorption process of metal ions and humic substances. Aldrich humic acid (HA) solution was irradiated with different doses of gamma irradiation using a Co-60 gamma-ray source prior to the contact between the metal ions and the solid sorbent. The HA molecule decomposed to smaller molecules with a lower complexation affinity. Batch sorption experiments were performed to evaluate the effect of gamma-irradiated HA on the sorption of Cs and Eu ions. The addition of non-irradiated HA weakened the sorption of Eu because of the lower sorption of the neutral or negatively charged Eu-HA complexes compared with free Eu ions. The sorption of monovalent Cs ions was barely affected by the presence of HA and its gamma irradiation. The concentration ratio of HA complexed species and non-complexed species in the solid and liquid phases was evaluated by sequential filtration and chemical equilibrium calculations. The ratios supported the minimal contribution of HA to Cs sorption. However, the concentration ratio for Eu in the liquid phase was high, indicating that the complexing ability of HA to Eu was higher than that of HA to Cs ions. Therefore, the sorption of free Eu would predominate with the gamma irradiation dose applied to the HA solution under a radiation field near the HLW package.
Jensen, C. B.*; Wachs, D. M.*; Woolstenhulme, N. E.*; Ozawa, Takayuki; Hirooka, Shun; Kato, Masato
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
Kato, Chiaki; Yamagishi, Isao; Sato, Tomonori; Yamamoto, Masahiro*
Zairyo To Kankyo, 70(12), p.441 - 447, 2021/12
Zeolite particles have been used in a Cs adsorption vessel for purification of contaminated water in Fukushima Dai-ich Nuclear Power Station (1F). The used Cs adsorption vessels were kept in storage space on 1F site. The risk of localized corrosion of stainless steel used in the vessel was worried. To evaluate the risk of localized corrosion, using specially designed electrochemical testing apparatus was used under gamma-ray irradiation test. And, real size mock-up test conducted. The results showed the potential change caused by creation of HO by water radiolysis decreased by zeolite particles and the enrichment of chloride ion concentration in the vessel do not propagate during dry up procedure of Cs adsorption vessel. These data indicate the risk of localized corrosion of Cs adsorption vessel may stay at considerably low level.
Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi
Nuclear Technology, 207(8), p.1280 - 1289, 2021/08
Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.
Iwata, Keiko; Hata, Kuniki; Tobita, Toru; Hirota, Takatoshi*; Takamizawa, Hisashi; Chimi, Yasuhiro; Nishiyama, Yutaka
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07
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.
JPS Conference Proceedings (Internet), 33, p.011042_1 - 011042_6, 2021/03
Development of beam window (BW) materials is one of crucial issues in development of accelerator driven nuclear transmutation systems (ADS). The BW is exposed to high energy protons and spallation neutrons, and also to corrosive lead-bismuth eutectic (LBE) alloy at high temperature of about 500C. Recently, not only high-power accelerators but also high-power targets are the rate-limiting factor for increasing the power of accelerator facilities in terms of radiation damage and heat removal. To study radiation damage on BW and target materials for high-power accelerator facilities including ADS, we are planning a materials irradiation facility by utilizing the proton beam of 400 MeV and 250 kW provided by the J-PARC's Linac. The target is flowing LBE alloy which is a candidate target and coolant material of ADS. When a steel sample is irradiated in the target for one year, the sample receives radiation damage of about 10 dpa at maximum which is equivalent to the yearly radiation damage of ADS's BW. In the current facility concept, the facility is equipped with a hot-laboratory for efficient post-irradiation examination. The facility will be outlined in this presentation.
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.