Sakamoto, Naoki; Fujishima, Tadatsune; Mizukoshi, Yasutaka
Hozengaku, 19(2), p.125 - 126, 2020/07
The five post-irradiation examination facilities in JAEA's Oarai research and development institute have been operated for over 40 years in order to investigate the irradiation performance of fast reactor fuel materials. The equipment associated with these facilities has been managed to maintain secure from the problems occurred in the process of aging. Therefore, we established a safety assessment method for aging facilities in 2002, and we have been conducting maintenance management of facilities since then. In this study, improvement plans of the safety assessment method are considered in order to solve the issues detected as a result of analysis of past maintenance information.
Sato, Hiroyuki; Ohashi, Hirofumi
Mechanical Engineering Journal (Internet), 7(3), p.19-00332_1 - 19-00332_11, 2020/06
An uncertainty analysis method for control room habitability under toxic gas leakage accidents in cogeneration HTGR is proposed to support risk-informed design of the plant. The method is applied to representative toxic gas leakage accidents in a IS process hydrogen production plant coupled to the HTTR gas turbine test plant. Epistemic and aleatory uncertainties for each variable parameter are identified and are propagated using Latin hypercube sampling. The analyses show that the suggested method can successfully characterize and quantify uncertainties in the toxic gas concentration in control room. The results lead us to the conclusion that toxic gas dispersion behavior analysis should combine two evaluation methods: dense gas dispersion model and computational fluid dynamics simulation.
Kawaguchi, Munemichi; Miyahara, Shinya*; Uno, Masayoshi*
Journal of Nuclear Engineering and Radiation Science, 6(2), p.021305_1 - 021305_9, 2020/04
Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater were performed to investigate the chemical reaction beneath the internal heater (800C), which was used to simulate the obstacle and heating effect on SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. The internal heater on the concrete hindered the transport of Na into the concrete. Therefore, Na could start to react with the concrete at the periphery of the internal heater, and the concrete ablation depth at the periphery was larger than under the internal heater. The high Na pool temperature of 800C increased largely the Na aerosol release rate, which was explained by Na evaporation and hydrogen bubbling, and formed the porous reaction product layer, whose porosity was 0.54-0.59 from the mass balance of Si and image analyzing EPMA mapping. They had good agreement with each other. The porous reaction products decreased the amount of Na transport into the reaction front. The Na concentration around the reaction front became about 30wt.% despite the position of the internal heater. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.
Sato, Hiroyuki; Aoki, Takeshi; Ohashi, Hirofumi; Yan, X. L.
Nuclear Engineering and Design, 360, p.110493_1 - 110493_8, 2020/04
JAEA has been conducting research and development with a central focus on the utilization of HTTR, the first HTGR in Japan, towards the realization of industrial use of nuclear heat. On the basis of licensing experience through the HTTR construction, JAEA initiated an activity to establish an international safety standard for licensing of commercial HTGR cogeneration systems fully taking into account safety features of HTGRs. We have developed a roadmap towards licensing of commercial HTGR cogeneration systems. A test plan using the HTTR to support the establishment of safety standards and safety analysis methods are also presented. In addition, we confirmed that a vessel cooling system, a passive air-cooled decay heat removal system, satisfies the safety requirement.
Kondo, Satoru; Tobita, Yoshiharu
JAEA-Research 2019-009, 382 Pages, 2020/03
The SIMMER-III computer code, developed at the Japan Atomic Energy Agency (JAEA, the former Power Reactor and Nuclear Fuel Development Corporation), is a two-dimensional, multi-velocity-field, multi-component fluid-dynamics code, coupled with a space- and time-dependent neutron kinetics model. The code is being used widely for simulating complex phenomena during core-disruptive accidents (CDAs) in liquid-metal fast reactors (LMFRs). In parallel to the code development, a comprehensive assessment program was performed in two phases: Phase 1 for verifying individual fluid-dynamics models; and Phase 2 for validating its applicability to integral phenomena important to evaluating LMFR CDAs. The SIMMERIII assessment program was participated by European research and development organizations, and the achievement of Phase 1 was compiled and synthesized in 1996. This report has been edited by revising and reproducing the original 1996 informal report, which compiled the achievement of Phase 1 assessment. A total of 34 test problems were studied in the areas: fluid convection, interfacial area and momentum exchange, heat transfer, melting and freezing, and vaporization and condensation. The problems identified have been reflected to the Phase 2 assessment and later model development and improvement. Although the revisions were made in the light of knowledge base obtained later, the original individual contributions by the participants, both positive and negative, are retained except for editorial changes.
Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*
Annals of Nuclear Energy, 133, p.830 - 836, 2019/11
A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. Moreover, the authors started experiment research with using a scaled-down heat-removal test facility. Therefore, this study propose a comparative methodology between an actual RCCS and a scaled-down heat-removal test facility.
Tazaki, Makiko; Shimizu, Ryo; Kimura, Takashi; Tamai, Hiroshi; Nakatani, Takayoshi; Suda, Kazunori
Nihon Kaku Busshitsu Kanri Gakkai Dai-40-Kai Nenji Taikai Puroshidhingusu, p.93 - 96, 2019/11
Iraq, defeated by the Gulf War, had to accept denuclearization under UN Security Council Resolution (UNSCR) 687, but did not immediately and accurately report all its past nuclear activities to the IAEA. IAEA, as an internationally authorized nuclear inspection and verification organization to Iraqi denuclearization by the UNSCR 687, gradually revealed existence and the whole picture of Iraqi clandestine nuclear activities and verified destruction of weapon related facilities and equipment as well as removal of nuclear materials outside Iraq. Analyzing Iraq's characteristics of denuclearization could serve as a good reference for not only preventing nuclear weapons development but also considering their denuclearization in current and future states.
Gijutsushi, (634), p.8 - 11, 2019/10
Since the FUKUSHIMA-DAIICHI Nuclear Power Plant accident, a significant issue to develop human resources for nuclear and radiation technology has been growing up. A strong effort will be made, based on the serious experiences of troubles/accidents, to restore social confidence by developing engineers having high sense of ethics beyond logic of organization.
Gunji, Satoshi; Clavel, J.-B.*; Tonoike, Kotaro; Duhamel, I.*
Proceedings of 11th International Conference on Nuclear Criticality Safety (ICNC 2019) (Internet), 11 Pages, 2019/09
The new criticality experiments facility STACY will be able to contribute to the validation of criticality calculations related to the fuel debris. The experimental core design is in progress in the frame of JAEA/IRSN collaboration. This paper presents the method applied to optimize the design of core configurations of the new STACY to measure the criticality characteristics of pseudo fuel debris focused on Molten Core Concrete Interaction (MCCI) debris. To ensure that a core configuration is relevant for code validation, it is important to evaluate the reactivity worth of the main isotopes and the keff sensitivity to their cross sections. To obtain maximum sensitivity of Si capture reaction, some parameters of the core configuration, as for example the lattice pitch or the core dimensions, were adjusted using optimization algorithm to research efficiently the optimal core configurations.
Ota, Hirokazu*; Ohgama, Kazuya; Yamano, Hidemasa
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.30 - 39, 2019/09
Liu, X.*; Morita, Koji*; Yamano, Hidemasa
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.47 - 51, 2019/09
Investigation of the eutectic reaction in a core disruptive accident of sodium cooled reactor is of importance since reactor criticality will be affected by the change in reactivity after eutectic reaction. In this study, we performed 1st step of validation analysis using a fast reactor safety analysis code, SIMMER-III, with the developed model based on a new series of experiments, where a BC pellet was immersed into a molten stainless steel (SS) pool. The simulation results showed the general behavior of eutectic material formation measured in the experiments reasonably. The eutectic reaction consumes solid BC and liquid SS, and then the liquid eutectic composition is produced at the early stage of reaction due to the high temperature of molten SS. Movement of the eutectic material in the molten pool leads to the redistribution of boron element. Molten SS pool then freezes to solid SS and movement of eutectic material is stopped by surrounding solid SS. Boron concentration in the pool was measured after molten SS freezes into a solid. Simulation results indicate that boron tends to accumulate in the upper part of the molten pool. This is attributed to the buoyancy force acting on lighter boron in the molten SS pool. A parametric study was also conducted by changing the initial temperature of BC pellet and SS to investigate the temperature sensitivity on the eutectic reaction behavior.
Tsuruga Comprehensive Research and Development Center
JAEA-Technology 2019-007, 159 Pages, 2019/07
This report summarizes the history and achievements of the prototype fast breeder reactor Monju. The development of Monju started in 1968 as a prototype reactor following the experimental fast reactor Joyo. The development covers all the activity related to the fast reactor; plant design, mockup tests, construction, operation, and plant management. This report summarizes the history and achievements for 11 technical areas: history and principal achievements, design and construction, operation test, plant safety, core physics, fuel, plant system, sodium technology, materials and mechanical design, plant management, and trouble management.
Onoe, Hironori; Kosaka, Hiroshi*; Matsuoka, Toshiyuki; Komatsu, Tetsuya; Takeuchi, Ryuji; Iwatsuki, Teruki; Yasue, Kenichi
Genshiryoku Bakkuendo Kenkyu (CD-ROM), 26(1), p.3 - 14, 2019/06
In this study, it is focused on topographic changes due to uplift and denudation, also climate perturbations, a method which is able to assess the long-term variability of groundwater flow conditions using the coefficient variation based on some steady-state groundwater flow simulation results was developed. Spatial distribution of long residence time area which is not much influenced due to long-term topographic change and recharge rate change during the past one million years was able to estimate through the case study of the Tono area, Central Japan. By applying this evaluation method, it is possible to identify the local area that has low variability of groundwater flow conditions due to topographic changes and climate perturbations from the regional area quantitatively and spatially.
Kato, Atsushi; Mukaida, Kyoko
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 10 Pages, 2019/05
Improvement of economic competitiveness is a part of key requirement in the project. By adopting innovative technologies to reduce plant commodities, JSFR could achieve economic competitiveness compared with LWR. After the Fukushima-Dai-ichi Nuclear Power Plants accident, safety enhancement measures were added on LWR in Japan mainly against external hazards. In parallel, Safety Design Criteria and Guidelines (SDC/SDG) for SFR were constructed in the framework of Generation IV international forum. Design studies of JSFR were carried out responding to GIF SDC/SDG and lessons learn from the Fukushima accident. This reports an impact of recent safety design enhancements on JSFR construction cost. Safety design enhancement adopted in JSFR.
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 4 Pages, 2019/05
The decommissioning of Fukushima Daiichi Nuclear Power Plant accident is one of the most important issues in Japan. The criticality safety of fuel debris is imperative to prevent exposure of workers. The investigating criticality monitoring system cannot detect the criticality of fuel debris quickly. The estimation of criticality of fuel debris is required for the fuel debris retrieval. Though the expert knowledge of reactor physics is necessary to estimate the criticality of fuel debris, many people who make a plan of fuel debris retrieval may not know well about criticality analysis. We developed a handy criticality analysis tool HAND to quickly estimate the criticality of fuel debris without expert knowledge of reactor physics. Since the input data of HAND is so simple and users can intuitively understand the calculation results, this tool is expected to be the effective tool to estimate the criticality of fuel debris.
Yamano, Hidemasa; Vasile, A.*; Kang, S.-H.*; Summer, T.*; Tsige-Tamirat, H.*; Wang, J.*; Ashurko, I.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 7 Pages, 2019/05
The Generation IV (GEN-IV) international forum is a framework for international co-operation in research and development for the next generation of nuclear energy systems. Within the GEN-IV sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aims of the SO project include (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 "Methods, Models and Codes", WP-SO-2 "Experimental Programs and Operational Experience", and WP-SO-3 "Studies of Innovative Design and Safety Systems". This paper reports recent activities within the SO project.
Matsunaga, Shoko*; Matsubara, Shinichiro*; Kato, Atsushi; Yamano, Hidemasa; Dderlein, C.*; Guillemin, E.*; Hirn, J.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 8 Pages, 2019/05
This paper presents a design of Curie Point Electro-Magnet (CPEM) which will be installed as a passive shutdown system for a French Sodium-cooled Fast Reactor (ASTRID) development program which is conducted in collaboration between France and Japan. To confirm CPEM design validity, a qualification program for CPEM is developed on the basis of past comprehensive test series of Self-Actuated Shutdown System (SASS) in Japan. The main outcome of this paper is results of holding force tests in hot gas, which satisfy design requirements. Moreover, the result of a numerical magnetic field analysis showed the same tendency as that of the holding force test.
Someya, Takayuki*; Chitose, Hiromasa*; Watanabe, Satoshi*; Nemoto, Yoshiyuki; Kaji, Yoshiyuki
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05
In this study, CFD analysis has been conducted for the assessment of spent fuel integrity in large LOCA event and the maximum temperature of spent fuel assemblies has been evaluated. Then, it has been compared with the result of the simple assessment method. As a case study, additional CFD analysis has been conducted, where water level in SFP decreases to the Bottom of Active Fuel (BAF) due to boil-off. Since this scenario might be more severe than large LOCA scenario, the number of spent fuel assemblies, their decay heat and loading pattern to maintain spent fuel integrity are investigated.
Nippon Genshiryoku Gakkai-Shi, 61(4), p.270 - 272, 2019/04
no abstracts in English
Genshiryoku No Ima To Ashita, p.264 - 265, 2019/03
no abstracts in English