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Wen, J.*; Kamada, Yuto*; Yokoyama, Kosei*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Imaizumi, Yuya; Tagami, Hirotaka; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 8 Pages, 2024/11
Wen, J.*; Kamada, Yuto*; Yokoyama, Kosei*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Imaizumi, Yuya; Tagami, Hirotaka; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 8 Pages, 2024/11
Kondo, Satoru; Tobita, Yoshiharu*; Morita, Koji*; Kamiyama, Kenji; Yamano, Hidemasa; Suzuki, Toru*; Tagami, Hirotaka; Sogabe, Joji; Ishida, Shinya
JAEA-Research 2024-008, 235 Pages, 2024/10
JAEA-Research-2024-008.pdf:4.77MB
The SIMMER-III and SIMMER-IV computer codes, developed at the Japan Atomic Energy Agency are the codes with two- and three-dimensional, multi-field, multi-component fluid-dynamics models, coupled with a space- and time-dependent neutron kinetics model. The codes have been used widely for simulating complex phenomena during core-disruptive accidents in liquid-metal fast reactors. Advanced features of the codes in comparison with the former codes include: stable and robust fluid-dynamics algorithm with up to 8 velocity fields, improved representation of structures and multi-phase flow topology, comprehensive treatment of complex heat and mass transfer processes, accurate analytic equations of state, a stable and efficient neutron flux shape solution method and decay heat model. This report describes the models and methods of SIMMER-III and SIMMER-IV. For those individual models, the details of which have been reported elsewhere, only the outlines of the models are presented. The reports of code verification and validation have been already published.
Zhang, T.*; Yao, Y.*; Morita, Koji*; Liu, X.*; Liu, W.*; Imaizumi, Yuya; Kamiyama, Kenji
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 9 Pages, 2023/05
Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji
Annals of Nuclear Energy, 179, p.109389_1 - 109389_10, 2022/12
Times Cited Count：3 Percentile：47.03(Nuclear Science & Technology)Takatsuka, Daichi*; Morita, Koji*; Liu, W.*; Zhang, T.*; Nakamura, Takeshi*; Kamiyama, Kenji
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 10 Pages, 2022/10
Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.
Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07
This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.
Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji
Extended abstracts of the 2nd Asian Conference on Thermal Sciences (Internet), 2 Pages, 2021/10
For the Japanese sodium cooled fast reactor, a fuel subassembly with an inner duct structure (FAIDUS) was designed to avoid the re-criticality by preventing the large-scale pool formation. In the present study, using the finite volume particle method, the EAGLE ID1 test which was an in-pile test performed to demonstrate the effectiveness of FAIDUS was numerically simulated and the thermal-hydraulic mechanisms underlying the heat transfer process were analyzed.
Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 11 Pages, 2021/08
One of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors is eutectic reactions between boron carbide (B
C) and stainless steel (SS) as well as its relocation. Such behaviors have never been simulated in CDA numerical analyses in the past, therefore it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study focuses on BC-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in a range from solid to liquid state. The physical model is developed for a CDA computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies conducted until 2019. Specific results in this paper are the validation of physical model describing BC-SS eutectic reaction in the CDA analysis code, SIMMER-III, through the numerical analysis of the BC-SS eutectic melting experiments in which a BC block was placed in a SS pool.
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Song, T.*; Gubler, P.; Hong, J.*; Lee, S. H.*; Morita, Kenji*
Physics Letters B, 813, p.136065_1 - 136065_5, 2021/02
Times Cited Count：3 Percentile：37.54(Astronomy & Astrophysics)Zhang, T.*; Funakoshi, Kanji*; Liu, X.*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji
Annals of Nuclear Energy, 150, p.107856_1 - 107856_10, 2021/01
Times Cited Count：6 Percentile：59.46(Nuclear Science & Technology)Sheikh, M. A. R.*; Liu, X.*; Matsumoto, Tatsuya*; Morita, Koji*; Guo, L.*; Suzuki, Toru*; Kamiyama, Kenji
Energies (Internet), 13(19), p.5018_1 - 5018_15, 2020/10
Times Cited Count：10 Percentile：27.17(Energy & Fuels)Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 10 Pages, 2020/08
One of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors is eutectic reactions between boron carbide (BC) and stainless steel (SS) as well as its relocation. Such behaviors have never been simulated in CDA numerical analyses in the past, therefore it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study focuses on BC-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in a range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies conducted until 2018. Specific results in this paper are boron concentration distributions of solidified BC-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.
Fujihara, Masayoshi*; Morita, Katsuhiro*; Mole, R.*; Mitsuda, Setsuo*; Toyama, Takami*; Yano, Shinichiro*; Yu, D.*; Sota, Shigetoshi*; Kuwai, Tomohiko*; Koda, Akihiro*; et al.
Nature Communications (Internet), 11, p.3429_1 - 3429_7, 2020/07
Times Cited Count：50 Percentile：92.72(Multidisciplinary Sciences)Yamano, Hidemasa; Takai, Toshihide; Furukawa, Tomohiro; Kikuchi, Shin; Emura, Yuki; Kamiyama, Kenji; Fukuyama, Hiroyuki*; Higashi, Hideo*; Nishi, Tsuyoshi*; Ota, Hiromichi*; et al.
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.418 - 427, 2019/09
Eutectic reactions between boron carbide (BC) and stainless steel (SS) as well as its relocation are one of the key issues in a core disruptive accident (CDA) evaluation in sodium-cooled fast reactors. Since such behaviors have never been simulated in CDA numerical analyses, it is necessary to develop a physical model and incorporate the model into the CDA analysis code. This study is focusing on BC-SS eutectic melting experiments, thermophysical property measurement of the eutectic melt, and physical model development for the eutectic melting reaction. The eutectic experiments involve the visualization experiments, eutectic reaction rate experiments and material analyses. The thermophysical properties are measured in the range from solid to liquid state. The physical model is developed for a severe accident computer code based on the measured data of the eutectic reaction rate and the physical properties. This paper describes the project overview and progress of experimental and analytical studies by 2017. Specific results in this paper is boron concentration distributions of solidified BC-SS eutectic sample in the eutectic melting experiments, which would be used for the validation of the eutectic physical model implemented into the computer code.
Morita, Yasuji; Nishihara, Kenji; Tsubata, Yasuhiro
JAEA-Data/Code 2018-017, 32 Pages, 2019/02
JAEA-Data-Code-2018-017.pdf:2.35MB
Potential radiotoxicity defined as a summation of intake dose was estimated for each actinide element to suppose target of recovery ratio of minor actinide (MA). Importance of each element from the viewpoint of the radiotoxicity was evaluated from the evolution of the radiotoxicity and ratio to the total radiotoxicity. In all the 4 types of spent fuels examined, Am is the most important element. For instance, the potential radiotoxicity of Am accounts for 93% of the total radiotoxicity of actinide elements in HLW produced by reprocessing of spent fuel from pressurized water reactor (PWR). Residual Pu after the recovery of 99.5% in reprocessing still gives contribution that cannot be ignored in radiotoxicity. When the burn-up of the UO
fuel in PWR increased, the potential radiotoxicity of actinide elements increased almost in proportion to the burn-up, but in case of MOX fuel in PWR and minor-actinide-recycled MOX fuel in fast reactor, the radiotoxicity of actinide elements increased further. Much consideration is required for the recovery of actinide elements in HLW from different types of fuel.
Kato, Masatsugu*; Funakoshi, Kanji*; Liu, X.*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11
Phan, L. H. S.*; Ohara, Yohei*; Kawata, Ryo*; Liu, X.*; Liu, W.*; Morita, Koji*; Guo, L.*; Kamiyama, Kenji; Tagami, Hirotaka
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 12 Pages, 2018/10
Self-leveling behavior of core fuel debris beds is one of the key phenomena for the safety assessment of core disruptive accidents (CDAs) in sodium-cooled fast reactors (SFRs). The SIMMER code has been developed for CDA analysis of SFRs, and the code has been successfully applied to numerical simulations for key thermal-hydraulic phenomena involved in CDAs as well as reactor safety assessment. However, in SIMMER's fluid-dynamics model, it is always difficult to represent the strong interactions between solid particles as well as the discrete particle characteristics. To solve this problem, a new method has been developed by combining the multi-fluid model of the SIMMER code with the discrete element method (DEM) for the solid phase to reasonably simulate the particle behaviors as well as the fluid-particle interactions in multi-phase flows. In this study, in order to validate the multi-fluid model of the SIMMER code coupled with DEM, numerical simulations were performed on a series of self-leveling experiments using a gas injection method in cylindrical particle beds. The effects of friction coefficient on the simulation results were investigated by sensitivity analysis. Though more extensive validations are needed, the reasonable agreement between simulation results and corresponding experimental data preliminarily demonstrates the potential ability of the present method in simulating the self-leveling behaviors of debris bed. It is expected that the SIMMER code coupled with DEM is a prospective computational tool for analysis of safety issues related to solid particle debris bed in SFRs.
Morita, Koji*; Ogawa, Ryusei*; Tokioka, Hiromi*; Liu, X.*; Liu, W.*; Kamiyama, Kenji
Proceedings of 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12) (USB Flash Drive), 11 Pages, 2018/10
The EAGLE in-pile ID1 test has been performed by Japan Atomic Energy Agency to demonstrate early fuel discharge from a fuel subassembly with an inner duct structure, which is named FAIDUS. It was deduced that early duct wall failure observed in the test was initiated by high heat flux from the molten pool of fuel and steel mixture. The posttest analyses suggest that molten pool-to-duct wall heat transfer might be enhanced effectively by the molten steel with large thermal conductivity in the pool without the presence of fuel crust on the duct wall. In this study, mechanisms of heat transfer from the molten pool to the duct wall was analyzed using a fully Lagrangian approach based on the finite volume particle method for multi-component, multi-phase flows. A series of pin disruption, molten pool formation and duct wall failure behaviors was simulated to investigate mixing and separation behavior of molten steel and fuel in the pool, and their effect on molten pool-to-duct wall heat transfer. The present 2D particle-based simulations demonstrated that large thermal load beyond 10 MW/m
on the duct wall was caused by effective heat transfer due to direct contact of liquid fuel with nuclear heat to the duct wall.
Sheikh, Md. A. R.*; Son, E.*; Kamiyama, Motoki*; Morioka, Toru*; Matsumoto, Tatsuya*; Morita, Koji*; Matsuba, Kenichi; Kamiyama, Kenji; Suzuki, Toru*
Journal of Nuclear Science and Technology, 55(6), p.623 - 633, 2018/06
Times Cited Count：18 Percentile：82.83(Nuclear Science & Technology)During core-disruptive accidents in sodium-cooled fast reactors, the sedimentation behavior of fragmented debris is crucial for in-vessel retention. The height of the beds formed may influence both the cooling of the bed and the neutronic characteristics. To develop an experimental database of bed formation behavior, a series of experiments using simulant materials, namely, AlO
, ZrO, and stainless steel, were performed under gravity-driven discharge of solid particles from a nozzle into a quiescent cylindrical water pool. The bed height was measured for particles of different size, density, and sphericity, and an injection nozzle with varying diameter, injection velocity, and injection height. From these experiments, an empirical correlation was established to predict the bed height for both homogeneous and mixed particles for the different properties. This correlation reproduces reasonably well the experimental trend in bed height.
