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Brear, D. J.*; Kondo, Satoru; Sogabe, Joji; Tobita, Yoshiharu*; Kamiyama, Kenji
JAEA-Research 2024-009, 134 Pages, 2024/10
JAEA-Research-2024-009.pdf:2.45MB
The SIMMER-III/SIMMER-IV computer codes are being used for liquid-metal fast reactor (LMFR) core disruptive accident (CDA) analysis. The sequence of events predicted in a CDA is often influenced by the heat exchanges between LMFR materials, which are controlled by heat transfer coefficients (HTCs) in the respective materials. The mass transfer processes of melting and freezing, and vaporization and condensation are also controlled by HTCs. The complexities in determining HTCs in a multi-component and multi-phase system are the number of HTCs to be defined at binary contact areas of a fluid with other fluids and structure surfaces, and the modes of heat transfer taking into account different flow topologies representing flow regimes with and without structure. As a result, dozens of HTCs are evaluated in each mesh cell for the heat and mass transfer calculations. This report describes the role of HTCs in SIMMER-III/SIMMER-IV, the heat transfer correlations implemented and the calculation of HTCs in all topologies in multi-component, multi-phase flows. A complete description of the physical basis of HTCs and available experimental correlations is contained in Appendices to this report. The major achievement of the code assessment program conducted in parallel with code development is summarized with respect to HTC modeling to demonstrate that the coding is reliable and that the model is applicable to various multi-phase problems with and without reactor materials.
Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*
JAEA-Review 2024-013, 48 Pages, 2024/07
JAEA-Review-2024-013.pdf:1.99MB
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2022. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2021, this report summarizes the research results of the "Fuel debris criticality analysis technology using non-contact measurement method" conducted in FY2022. The purpose of research was to improve the fuel debris criticality analysis technology using non-contact measurement method by the development of the fuel debris criticality characteristics measurement system and the multi-region integral kinetic analysis code. It was performed by Tokyo Institute of Technology, National Institute of Advanced Industrial Science and Technology, and Nagaoka University of Technology as the second year of three years research project.
Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05
Times Cited Count：1 Percentile：35.82(Nuclear Science & Technology)Collaborative Laboratories for Advanced Decommissioning Science; Tokyo Institute of Technology*
JAEA-Review 2022-043, 52 Pages, 2023/01
JAEA-Review-2022-043.pdf:3.48MB
The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2021. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2021, this report summarizes the research results of the "Fuel debris criticality analysis technology using non-contact measurement method" conducted in FY2021. The purpose of research was to improve the fuel debris criticality analysis technology using non-contact measurement method by the development of the fuel debris criticality characteristics measurement system and the multi-region integral kinetic analysis code. It was performed by Tokyo Institute of Technology (Tokyo Tech), National Institute of Advanced Industrial Science and Technology (AIST), and National Research Nuclear University (MEPhI) as the first year of four years research project. For the criticality characteristic measurement systems to be developed by the Japanese and Russian sides, …
Onoda, Yuichi; Uchita, Masato*; Tokizaki, Minako*; Okazaki, Hitoshi*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 6 Pages, 2022/08
The safety analyses were carried out to confirm the sufficiency of the function of the plant protection system against the pump/diagrid link rupture. The target plant is a pool-type SFR of about 600 MWe class equipped with an axially homogeneous core currently under development in Japan. In the pool-type SFR, the primary system piping connects primary pump and the high-pressure sodium plenum located at the inlet of fuel sub-assemblies and called "pump/diagrid link". Because this piping is submerged in the reactor vessel, it is difficult to detect small scale sodium leakage in this piping, and thus a certain large pipe break like guillotine should be assumed and evaluated as a design basis event. In order to confirm the detectability of pump/diagrid link rupture by safety protection system signals, a series of analyses of the guillotine break for a pump/diagrid link were carried out. Sensitivity study had also been performed to consider the uncertainty of the reactivity coefficient in the analyses. The sufficiency of the function of the plant protection system against the pump/diagrid link rupture was confirmed by the analysis results that at least two signals are transmitted for the detection of the event, which is the development target of the plant protection system in pool-type SFR.
Yonomoto, Taisuke; Mineo, Hideaki; Murayama, Yoji; Hohara, Shinya*; Nakajima, Ken*; Nakatsuka, Toru; Uesaka, Mitsuru*
Nihon Genshiryoku Gakkai-Shi ATOMO
, 63(1), p.73 - 77, 2021/01
no abstracts in English
Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr
, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo
Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09
Times Cited Count：4 Percentile：18.87(Nuclear Science & Technology)Ono, Masato; Fujiwara, Yusuke; Matsumoto, Tetsuro*; Iigaki, Kazuhiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(2), p.110 - 120, 2020/06
Integrity confirmation for buildings against collisions of projectiles has been conducted to evaluate collisions between a projectile with simple shape and a wall using empirical formulas. It is a matter of fact, there is a possibility that structures with complex shape such as stack may collide with a reactor building. However, there were not so many studies of collisions between structures with complex shape and buildings in the literature. Impact evaluation was carried out using reactor building and stack with real shape and adequate physical property. It was found that ceiling of reactor building was not damaged by the collision, confirming that there was no effect inside of reactor building.
Nuclear Safety Research Center, Sector of Nuclear Safety Research and Emergency Preparedness
JAEA-Review 2018-022, 201 Pages, 2019/01
JAEA-Review-2018-022.pdf:20.61MB
Nuclear Safety Research Center (NSRC), Sector of Nuclear Safety Research and Emergency Preparedness, Japan Atomic Energy Agency (JAEA) is conducting technical support to nuclear safety regulation and safety research based on the Mid-Long Term Target determined by Japanese government. This report summarizes the research structure of NSRC and the cooperative research activities with domestic and international organizations as well as the nuclear safety research activities and results in the period from JFY 2015 to 2017 on the nine research fields in NSRC; (1) severe accident analysis, (2) radiation risk analysis, (3) safety of nuclear fuels in light water reactors (LWRs), (4) thermohydraulic behavior under severe accident in LWRs, (5) materials degradation and structural integrity, (6) safety of nuclear fuel cycle facilities, (7) safety management on criticality, (8) safety of radioactive waste management, and (9) nuclear safeguards.
Nakamura, Hideo
Proceedings of ANS International Conference on Best Estimate Plus Uncertainties Methods (BEPU 2018) (USB Flash Drive), 8 Pages, 2018/00
no abstracts in English
Nakamura, Hideo
Proceedings of Seminar on the Transfer of Competence, Knowledge and Experience gained through CSNI Activities in the Field of Thermal-Hydraulics (THICKET 2016) (CD-ROM), 29 Pages, 2016/06
no abstracts in English
Yamamoto, Kento*; Akie, Hiroshi; Suyama, Kenya; Hosoyamada, Ryuji*
JAEA-Technology 2015-019, 110 Pages, 2015/10
JAEA-Technology-2015-019.pdf:3.67MB
In the direct disposal of used nuclear fuel (UNF), criticality safety evaluation is one of the important issues since UNF contains some amount of fissile material. The recent development of higher-enrichment fuel has enhanced the benefit of the application of Burnup Credit. In the present study, the effects of the several parameters on the reactivity of disposal canister model were evaluated for used PWR fuel. The parameters are relevant to the uncertainties of depletion calculation code, irradiation history, and axial and horizontal burnup distribution, which are known to be important effect in the criticality safety evaluation adopting burnup credit. The latest data or methodology was adopted in this evaluation, based on the various latest studies. The appropriate margin of neutron multiplication factor in the criticality safety evaluation for UNF can be determined by adopting the methodology described in the present study.
Watanabe, Osamu*; Oyama, Kazuhiro*; Endo, Junji*; Doda, Norihiro; Ono, Ayako; Kamide, Hideki; Murakami, Takahiro*; Eguchi, Yuzuru*
Journal of Nuclear Science and Technology, 52(9), p.1102 - 1121, 2015/09
Times Cited Count：14 Percentile：70.76(Nuclear Science & Technology)A natural circulation (NC) evaluation methodology has been developed to ensure the safety of a sodium-cooled fast reactor (SFR) of 1500MW adopting the NC decay heat removal system (DHRS). The methodology consists of a 1D safety analysis which can evaluate the core hot spot temperature taking into account the temperature flattening effect in the core, a 3D fluid flow analysis which can evaluate the thermal-hydraulics for local convections and thermal stratifications in the primary system and DHRS, and a statistical safety evaluation method. The safety analysis method and the 3D analysis method have been validated using results of a 1/10 scaled water test simulating the primary system of the SFR and a 1/7 scaled sodium test simulating the primary system and the DHRS, and the applicability of the safety analysis for the SFR has been confirmed by comparing with the 3D analysis. Finally, a statistical safety evaluation has been performed for the SFR using the safety analysis method.
Mascari, F.*; Nakamura, Hideo; Umminger, K.*; De Rosa, F.*; D'Auria, F.*
Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.4921 - 4934, 2015/08
Ono, Tomio*; Subekti, M.*; Kudo, Kazuhiko*; Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Nabeshima, Kunihiko
Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.115 - 126, 2005/06
Control-rod withdrawal tests simulating reactivity insertion are carried out in the HTTR to verify the inherent safety features of HTGRs. This paper describes pre-test analysis method using artificial neural networks to predict the changes of reactor power and reactivity. The network model applied in this study is based on recurrent neural networks. The inputs of the network are the changes of the central control rods position and other significant core parameters, and the outputs are the changes of reactor power and reactivity. Furthermore, Time Synchronizing Signal(TSS) is added to input to improve the modeling of time series data. The actual tests data, which were previously carried out in the HTTR, were used for learning the model of the plant dynamics. After the learning, the network can predict the changes of reactor power and reactivity in the following tests.
Editorial Committee on Nuclear Safety Research Results
JAERI-Review 2005-009, 151 Pages, 2005/03
JAERI-Review-2005-009.pdf:31.04MB
no abstracts in English
Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka
JAERI-Tech 2004-068, 27 Pages, 2004/12
JAERI-Tech-2004-068.pdf:7.68MB
ITER vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed. This independent concept has two advantages: (1) thermal load due to the temperature deference between VV and the lower temperature components such as TF coil becomes lower and (2) the other components such as TF coil is categorized as a non-safety component because of its independence from VV. Stress analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support.
Takeda, Nobukazu; Omori, Junji*; Nakahira, Masataka; Shibanuma, Kiyoshi
Journal of Nuclear Science and Technology, 41(12), p.1280 - 1286, 2004/12
Times Cited Count：4 Percentile：28.88(Nuclear Science & Technology)ITER vacuum vessel (VV) is a safety component confining radioactive materials. An independent VV support structure located at the bottom of VV lower port is proposed as an alternative concept, which is deferent from the current reference, i.e., the VV support is directly connected to the toroidal coil (TF coil). This independent concept has two advantages comparing to the reference one: (1) thermal load becomes lower and (2) the TF coil is categorized as a non-safety component. Stress Analyses have been performed to assess the integrity of the VV support structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coil is found to be 15 mm, much less than the current design value of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as an alternative VV support.
Takamatsu, Kuniyoshi; Furusawa, Takayuki; Hamamoto, Shimpei; Nakagawa, Shigeaki
Proceedings of 6th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operations and Safety (NUTHOS-6) (CD-ROM), 11 Pages, 2004/10
Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are in progress to verify the inherent safety features for High Temperature Gas-cooled Reactors (HTGRs). The coolant flow reduction test by tripping one or two out of three gas circulators is one of the safety demonstration tests. The reactor power safely brings to a stable level without a reactor scram and the temperature transient of the reactor-core is very slow. Through the safety demonstration test, the two dimensional temperature analysis code (TAC-NC code) was improved. This paper describes the validation of the TAC-NC code using the measured value of the test by tripping of one and two out of three gas circulators at 30%(9MW). The TAC-NC code could evaluate accurately the temperature transient within 10% during the test. Also, it was confirmed that the temperature transient by tripping all gas circulators is very slow.
Editorial Committee on Nuclear Safety Research Results
JAERI-Review 2004-010, 155 Pages, 2004/03
JAERI-Review-2004-010.pdf:16.43MB
JAERl is conducting nuclear safety research primarily at the Nuclear Safety Research Center in close cooperation with the related departments in accordance with the Long Term Plan for Development and Utilization of Nuclear Energy and Annual Plan for Safety Research issued by the Japanese government. The fields of conducting safety research at JAERl are the engineering safety of nuclear power plants and nuclear fuel cycle facilities, and radioactive waste management as well as advanced technology for safety improvement or assessment. Also, JAERI has conducted international collaboration to share the information on common global issues of nuclear safety and to supplement own research. This report summarizes the nuclear safety research activities of JAERI from April 2001 through March 2003 and utilized facilities.
