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Saito, Shigeru; Meigo, Shinichiro; Makimura, Shunsuke*; Hirano, Yukinori*; Tsutsumi, Kazuyoshi*; Maekawa, Fujio
JAEA-Technology 2023-025, 48 Pages, 2024/03
JAEA-Technology-2023-025.pdf:3.11MB
JAEA has been developing Accelerator-Driven Systems (ADS) for research and development of nuclear transmutation using accelerators in order to reduce the volume and hazardousness of high-level radioactive waste generated by nuclear power plants. In order to prepare the material irradiation database necessary for the design of ADS and to study the irradiation effects in Lead-Bismuth Eutectic (LBE) alloys, a proton irradiation facility is under consideration at J-PARC. In this proton irradiation facility, 250 kW proton beams will be injected into the LBE spallation target, and irradiation tests under LBE flow will be performed for candidate structural materials for ADS. Furthermore, semiconductor soft-error tests, medical RI production, and proton beam applications will be performed. Among these, Post Irradiation Examination (PIE) of irradiated samples and RI separation and purification will be carried out in the PIE facility to be constructed near the proton irradiation facility. In this PIE facility, PIE of the equipment and samples irradiated in other facilities in J-PARC will also be performed. This report describes the conceptual study of the PIE facility, including the items to be tested, the test flow, the facilities, the test equipment, etc., and the proposed layout of the facility.
HTGR Design Group, HTGR Project Management Office
JAEA-Technology 2023-019, 39 Pages, 2024/01
JAEA-Technology-2023-019.pdf:1.34MB
In order to realize the development of the demonstration reactor of High Temperature Gas-cooled Reactor (HTGR) with a target of starting operation in the 2030s, as indicated in the "Basic Policy for GX Realization" (Cabinet Decision on February 10, 2023) and the Working Group on Innovative Reactors of METI, Japan Atomic Energy Agency (JAEA) has been working on the development of a standard for the development of a HTGR under the Atomic Energy Society of Japan and the Japan Society of Mechanical Engineers. In addition, JAEA has been commissioned by the Agency for Natural Resources and Energy of the Ministry of Economy, Trade and Industry (METI) to conduct the "Demonstration Project for Mass Hydrogen Production Technology Using Ultra-High Temperatures" and has been promoting a hydrogen production project using the HTTR (High Temperature Engineering Test Reactor). Furthermore, in collaboration with the National Nuclear Laboratory (NNL) of the United Kingdom and the National Centre for Nuclear Research (NCBJ) of Poland, JAEA are aiming to strengthen the international competitiveness of HTGR technology by further upgrading the HTGR technology developed in Japan through the construction and operation of the HTTR. In response to the growing interest in HTGR development in Japan and abroad, we have developed FAQs on HTGR related technologies in order to provide accurate technical information on HTGRs.
Nakata, Hisakazu
Genkan Senta Topikkusu, (147), p.4 - 11, 2023/10
no abstracts in English
Akiyama, Yoichi; Shibanuma, So; Yanagisawa, Kenichi*; Yamada, Taichi; Suzuki, Kenta; Yoshida, Moeka; Ono, Takahiro; Kawabata, Kuniaki; Watanabe, Kaho; Morimoto, Kyoichi; et al.
JAEA-Review 2023-015, 60 Pages, 2023/09
JAEA-Review-2023-015.pdf:4.78MB
Naraha Center for Remote Control Technology Development (NARREC) was established in Japan Atomic Energy Agency to promote a decommissioning work of Fukushima Daiichi Nuclear Power Station (Fukushima Daiichi NPS). NARREC consists of a Full-scale Mock-up Test Building and Research Management Building. Various test facilities are installed in these buildings for the decommissioning work of Fukushima Daiichi NPS. These test facilities are intended to be used for various users, such as companies engaged in the decommissioning work, research and development institutions, educational institutions and so on. The number of NARREC facility uses was 84 in FY2021. We participated booth exhibitions and presentations on the decommissioning related events. Moreover, we also contributed to the development of human resources by supporting the 6th Creative Robot Contest for Decommissioning. As a new project, "Narahakko Children's Classroom" was implemented for elementary school students in Naraha Town. This report summarizes the activities of NARREC in FY2021, such as the utilization of facilities and equipment of NARREC, the development of remote-control technologies for supporting the decommissioning work, arrangement of the remote-control machines for emergency response, and training for operators by using the machines.
Nishino, Saki; Okada, Jumpei; Watanabe, Kazuki; Furuuchi, Yuta; Yokota, Satoru; Yada, Yuji; Kusaka, Shota; Morokado, Shiori; Nakamura, Yoshinobu
JAEA-Technology 2023-011, 39 Pages, 2023/06
JAEA-Technology-2023-011.pdf:2.51MB
Tokai Reprocessing Plant (TRP) which shifted to decommissioning phase in 2014 had nuclear fuel materials such as the spent fuel sheared powder, the diluted plutonium solution and the uranium solution in a part of the reprocessing main equipment because TRP intended to resume reprocessing operations when it suspended the operations in 2007. Therefore, we have planned to remove these nuclear materials in sequence as Flush-out before beginning the decommissioning, and conducted removal of the spent fuel sheared powder as the first stage. The spent fuel sheared powder that had accumulated in the cell of the Main Plant (MP) as a result of the spent fuel shearing process was recovered from the cell floor, the shearing machine and the distributor between April 2016 and April 2017 as part of maintenance. Removing the recovered spent fuel sheared powder was conducted between June 2022 and September 2022. In this work, the recovered powder was dissolved in nitric acid at the dissolver in a small amount in order to remove it safely and early, and the dissolved solution was sent to the highly radioactive waste storage tanks without separating uranium and plutonium. Then, the dissolved solution transfer route was rinsed with nitric acid and water. Although about 15 years had passed since previous process operations, the removing work was successfully completed without any equipment failure because of the organization of a system that combines veterans experienced the operation with young workers, careful equipment inspections, and worker education and training. Removing this powder was conducted after revising the decommissioning project and obtaining approval from the Nuclear Regulation Authority owing to operating a part of process equipment.
Watanabe, Kazuki; Kimura, Norimichi*; Okada, Jumpei; Furuuchi, Yuta; Kuwana, Hideharu*; Otani, Takehisa; Yokota, Satoru; Nakamura, Yoshinobu
JAEA-Technology 2023-010, 29 Pages, 2023/06
JAEA-Technology-2023-010.pdf:3.12MB
The Krypton Recovery Development Facility reached an intended technical target (krypton purity of over 90% and recovery rate of over 90%) by separation and rectification of krypton gas from receiving off-gas produced by the shearing and the dissolution process in the spent fuel reprocessing at the Tokai Reprocessing Plant (TRP) between 1988 and 2001. In addition, the feasibility of the technology was confirmed through immobilization test with ion-implantation in a small test vessel from 2000 to 2002, using a part of recovered krypton gas. As there were no intentions to use the remaining radioactive krypton gas in the krypton storage cylinders, we planned to release this gas by controlling the release amount from the main stack, and conducted it from February 14 to April 26, 2022. In this work, all the radioactive krypton gas in the cylinders (about 7.1
10
GBq) was released at the rate of 50 GBq/min or less lower than the maximum release rate from the main stuck stipulated in safety regulations (3.710
GBq/min). Then, the equipment used in the controlled release of radioactive krypton gas and the main process (all systems, including branch pipes connected to the main process) were cleaned with nitrogen gas. Although there were delays due to weather, we were able to complete the controlled release of radioactive krypton gas by the end of April 2022, as originally targeted without any problems such as equipment failure.
Yamamoto, Masahiko; Nishida, Naoki; Kobayashi, Daisuke; Nemoto, Ryo*; Hayashi, Hiroyuki*; Kitao, Takahiko; Kuno, Takehiko
JAEA-Technology 2023-004, 30 Pages, 2023/06
JAEA-Technology-2023-004.pdf:1.94MB
Glove-box gloves, that are used for handling nuclear fuel materials at the Tokai Reprocessing Plant (TRP) of the Japan Atomic Energy Agency, have an expiration date by internal rules. All gloves are replaced at a maximum of every 4-year. However, degrees of glove deterioration varies depending on its usage environment such as frequency, chemicals, and radiation dose. Therefore, physical properties such as tensile strength, elongation, hardness of gloves are measured and technical evaluation method for the glove life-time is established. It was found that gloves without any defects in its appearance have enough physical properties and satisfies the acceptance criteria values of new gloves. Thus, it was considered that the expired gloves could be used for total of 8-year, by adding 4-year of new glove life-time. In addition, the results of extrapolation by plotting the glove's physical properties versus the used years showed that the physical properties at 8-year is on the safer side than the reported physical properties of broken glove. Also, the data are not significantly different from the physical properties of the long-term storage glove (8 and 23 years). Based on these results, life-time of gloves at TRP is set to be 8-year. The frequency of glove inspections are not changed, and if any defects is found, the glove is promptly replaced. Thus, the risk related to glove usage is not increased. The cost of purchasing gloves, labor for glove replacement, and the amount of generated waste can be reduced by approximately 40%, respectively, resulting in more efficient and rationalized glove management.
Yoshitomi, Hiroshi
Isotope News, (786), p.26 - 29, 2023/04
no abstracts in English
Kageyama, Tomio; Denuma, Akio; Koizumi, Jin*; Odakura, Manabu*; Haginoya, Masahiro*; Isaka, Shinichi*; Kadowaki, Hiroyuki*; Kobayashi, Shingo*; Morimoto, Taisei*; Kato, Yoshiaki*; et al.
JAEA-Technology 2022-033, 130 Pages, 2023/03
JAEA-Technology-2022-033.pdf:9.87MB
Uranium handling facility for development of nuclear fuel manufacturing equipment (Mockup room) was constructed in 1972. The Mockup room has a weak seismic resistance and is deteriorating with age. Also, the original purpose with this facility have been achieved and there are no new development plans using this facility. Therefore, interior equipment installed in this facility had been dismantled and removed since March 2019. After that, the Mockup room was inspected for contamination, and then controlled area in the Mockup room was cancelled on March 29th 2022. A total of 6,549 workers (not including security witnesses) were required for this work. The amount of non-radioactive waste generated by this work was 31,300 kg. The amount of radioactive waste generated by this work was 3,734 kg of combustible waste (103 drums), 4,393 kg of flame resistance waste (61 drums), 37,790 kg of non-combustible waste (124 drums, 19 containers). This report describes about the dismantling and removing the interior equipment in the Mockup room, the amount of waste generated by this work, and procedure for cancellation the controlled area in the facility.
Kusakabe, Kazuaki*; Watanabe, Masanori; Nishiuchi, Masashi*; Yamasaki, Takuhei*; Inoue, Hiromi*
Kankyo Hoshano Josen Gakkai-Shi, 11(1), p.15 - 23, 2023/03
The spread of radioactive materials caused by the Fukushima Daiichi Nuclear Power Plant accident that occurred in March 2011 contaminated a wide area that includes Fukushima Prefecture. Although air dose rates in Fukushima Prefecture have been steadily decreasing because of decontamination and the physical decay of radioactive materials, it is important to confirm the sustainability of decontamination effects in living areas and to predict future trends in air dose rates to reassure residents who are concerned regarding radiation exposure. This report aims to confirm the sustainability of the decontamination effects in public facilities after decontamination on a continuous and detailed basis, and to verify whether the future transition in air dose rates can be predicted using existing model. The air dose rates in public facilities after decontamination were measured via fixed-point and walking surveys, and the changes in air dose rates were clarified quantitatively for each facility. The measured values were compared with values obtained using existing model, and prediction accuracy was considered. The results showed that there was no evident recontamination after decontamination at any of the surveyed facilities, indicating that the decontamination effects were sustained. It was also confirmed that future trends in air dose rates at the facilities after decontamination could be accurately predicted by existing model. Key words: air dose rate, decontamination, future prediction, public facilities.
Kawasaki, Kohei; Ono, Takanori; Shibanuma, Kimikazu; Goto, Kenta; Aita, Takahiro; Okamoto, Naritoshi; Shinada, Kenta; Ichige, Hidekazu; Takase, Tatsuya; Osaka, Yuki; et al.
JAEA-Technology 2022-031, 91 Pages, 2023/02
JAEA-Technology-2022-031.pdf:6.57MB
The document for back-end policy opened to the public in 2018 by Japan Atomic Energy Agency (hereafter, JAEA) states the decommissioning of facilities of Nuclear Fuel Cycle Engineering Laboratories and JAEA have started gathering up nuclear fuel material of the facilities into Plutonium Fuel Production Facilities (hereafter, PFPF) in order to put it long-term, stable and safe storage. Because we planned to manufacture scrap assemblies almost same with Monju fuel assembly using unsealed plutonium-uranium mixed-oxide (hereafter, MOX) powder held in PFPF and transfer them to storage facilities as part of this "concentration" task of nuclear fuel material, we obtained permission to change the use of nuclear fuel material in response to the new regulatory Requirements in Japan for that. The amount of plutonium (which is neither sintered pellets nor in a lidded powder-transport container) that could be handled in the pellet-manufacturing process was limited to 50 kg Pu or less in order to decrease the facility risk in this manufacture. Therefore, we developed and installed the "MOX weighing and blending equipment" corresponding with small batch sizes that functioned in a starting process and the equipment would decrease handling amounts of plutonium on its downstream processes. The failure data based on our operation and maintenance experiences of MOX fuel production facilities was reflected in the design of the equipment to further improve reliability and maintainability in this development. The completed equipment started its operation using MOX powder in February 2022 and the design has been validated through this half-a-year operation. This report organizes the knowledge obtained through the development of the equipment, the evaluation of the design based on the half-a-year operation results and the issues in future equipment development.
Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 181, p.109534_1 - 109534_10, 2023/02
Times Cited Count：0 Percentile：0.01(Nuclear Science & Technology)Feasibility of reprocessing of High Temperature Gas-cooled Reactor (HTGR) spent fuel by existing Plutonium Uranium Redox EXtraction (PUREX) plant and technology has been investigated. The spent fuel dissolved solution includes approximately 3 times amount of uranium-235 and 1.5 times amount of protonium because of the 3 times higher burnup compared with that of Light Water Reactor (LWR). Then, the heavy metal of the spent fuel is planned to be diluted to 3.1 times by depleted uranium to satisfy the limitation of Rokkasho Reprocessing Plant (RRP) plant. In the present study, recoverability of uranium and plutonium with the dilution is confirmed by a simulation with a reprocessing process calculation code. Moreover, the case without the dilution from the economic perspective is investigated. As a result, the feasibility is confirmed without the dilution, and it is expected that the reprocessed amount is reduced to 1/3 compared with a diluted case even though the facility should be optimized from the perspective of mass flow and criticality.
Sakai, Akihiro; Kamei, Gento; Sakamoto, Yoshiaki
Nihon Genshiryoku Gakkai-Shi ATOMO
, 65(1), p.25 - 29, 2023/01
Currently, radioactive waste generated from research institutes, etc. is keeping in storage facilities without being disposed of. In order to solve this problem, the Japan Atomic Energy Agency (JAEA) is proceeding with the project for concrete-pit disposal and trench disposal of these waste. This paper introduces the characteristics of the waste and disposal facilities planned by the JAEA, as well as the status of development of the siting criteria for the disposal facility.
Yamamoto, Masahiko; Horigome, Kazushi; Kuno, Takehiko
Applied Radiation and Isotopes, 190, p.110460_1 - 110460_7, 2022/12
Times Cited Count：1 Percentile：29.26(Chemistry, Inorganic & Nuclear)Gravimetric measurement of U content in UO
with ignition in the air has been investigated. The ignition temperature, ignition time and aliquot sample mass are optimized as 900
C, 60 minutes, and 1 g, respectively. The method is validated by IDMS with uncertainty estimation. The obtained result by gravimetry is 0.78236
0.00051 g/g (k=2) and agreed with IDMS value within its uncertainty. It has been found that U in UO can be measured accurately and precisely by gravimetry.
Geological Disposal Research and Development Department
JAEA-Evaluation 2022-007, 81 Pages, 2022/11
JAEA-Evaluation-2022-007.pdf:2.06MB
JAEA-Evaluation-2022-007-appendix(CD-ROM).zip:37.06MB
Japan Atomic Energy Agency (JAEA) consulted the advisory committee, "Evaluation Committee on Research and Development (R&D) Activities for Geological Disposal of High-Level Radioactive Waste", for post- and pre-review assessment of R&D activities on high-level radioactive waste disposal in accordance with "General Guideline for the Evaluation of Government Research and Development (R&D) Activities" by the Cabinet Office, Government of Japan, "Guideline for Evaluation of R&D in Ministry of Education, Culture, Sports, Science and Technology" and JAEA's "Regulation on Conduct for Evaluation of R&D Activities". In response to JAEA's request, the Committee reviewed mainly the progress of the R&D project on geological disposal, the relevance of the project outcome and the efficiency of the project implementation during the period of the current and next plan. This report summarizes the results of the assessment by the Committee with the Committee report attached.
Nakatani, Takayoshi; Shimizu, Ryo; Tazaki, Makiko; Kimura, Takashi; Hori, Masato
Dai-43-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2022/11
Denuclearization is the verification of nuclear development and the freezing, disabling, decommissioning, and verification of nuclear weapons, nuclear materials that can be used for nuclear weapons, and their manufacturing facilities, equipment, equipment, and materials. In this study, the technical process of measures and verification methods to effectively and efficiently achieve denuclearization at nuclear installations is examined and considered from a technical point of view.
Ogawa, Rina; Nakata, Hisakazu; Sugaya, Toshikatsu; Sakai, Akihiro
JAEA-Technology 2022-010, 54 Pages, 2022/07
JAEA-Technology-2022-010.pdf:11.07MB
Japan Atomic Energy Agency has considered trench disposal as one of the disposal methods for radioactive wastes generated from research facilities and other facilities. The trench disposal facility is regulated by "Act on the Regulation of Nuclear Source Material, Nuclear Fuel Material and Reactors". In particular, the design of the trench facility is regulated by a rule under the law. When the rule was amended in 2019, the design of the trench disposal facility required equipment to reduce ingress of rain water and groundwater. In the report, studies on the design of a trench disposal facility to adapt to the amended rule were performed. The trench disposal facility has considered being established in a place lower than groundwater level. Therefore, it was decided to study covering soil at the upper part of the trench facility, because the ingress water in the facility is mainly derived from rain water. In this study, it was decided to evaluate the design of covering soil of the radioactive waste categorized into chemically stable materials. Therefore, as the examination method, a parameter study on varying the permeability coefficient and thickness of the layers composing cover soil. In the parameter study, the velocity of the water infiltrating into the trench facility was evaluated. Based on the results, more efficient design of the layers composing the covering soil was considered. The result showed that the impermeable efficiency of the covering soil was different depending on the thickness and the permeability conductivity of each layer. As a result, it was possible to understand the impermeable performance of covering soil by the permeability coefficient and thickness of each layer. We will plan to decide the specification of the cover soil while examination of future tasks and cost in the basic design.
Omori, Kazuki; Yamauchi, Sho; Yanagibashi, Futoshi; Sasaki, Shunichi; Wada, Takuya; Suzuki, Hisanori; Domura, Kazuyuki; Takeuchi, Kenji
Nihon Hozen Gakkai Dai-18-Kai Gakujutsu Koenkai Yoshishu, p.245 - 248, 2022/07
Tokai Reprocessing Plant (TRP), which is shifted to decommissioning stage, stores large amount of high-level radioactive liquid waste (HLLW). Although TRP is implementing vitrification of HLLW to reduce the risks related to HLLW storage, additional 20 years are required to complete vitrification of HLLW. Therefore, TRP is implementing safety countermeasure related to seismic resistance of HLLW storage facility as one of the top priorities. The results of the seismic evaluation indicate that although the facility itself is seismically resistant, there is a risk of insufficient binding force acting between the facility and the surrounding ground. Thus, replacement of the surrounding ground with concrete is performed. Since the countermeasures, to protect existing buries structure and coordinate with the other construction projects around the site, are required, the dedicated team was setup to handle the process and safety management of the concrete replacement construction.
Aomori Research and Development Center
JAEA-Review 2021-065, 54 Pages, 2022/06
JAEA-Review-2021-065.pdf:4.0MB
Aomori Research and Development Center consists of Nuclear Facilities Management Section, General Affairs and Purchase Section, Facility Maintenance and Engineering Section, AMS Management Section and Nuclear Fuel Cycle Cooperation Office. Each sections are carrying out management of facility operation, decommissioning of reactor facility, etc. to achieve the Medium to long-term plan. In this report, the activities of Aomori Research and Development Center are described to contribute to future facility management and business promotion.
Sakai, Akihiro
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 29(1), p.48 - 54, 2022/06
no abstracts in English
