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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
Johnson, M.*; Delacroix, J.*; Journeau, C.*; Brayer, C.*; Clavier, R.*; Montazel, A.*; Pluyette, E.*; Matsuba, Kenichi; Emura, Yuki; Kamiyama, Kenji
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04
Fuel-coolant interactions in the event of molten fuel discharge to the lower plenum of a sodium cooled fast reactor is under investigation as part of a French-Japanese experimental collaboration on severe accidents. The MELT facility enables the X-ray visualisation of the quenching of molten core material jets in sodium at kilogram-scale. The SERUA facility, currently under preparation, is presented for the investigation of boiling heat transfer at elevated melt-coolant interface temperatures. In this article, the status of the collaboration using these facilities is presented.
Bachrata, A.*; Gentet, D.*; Bertrand, F.*; Marie, N.*; Kubota, Ryuzaburo*; Sogabe, Joji; Sasaki, Keisuke; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 9 Pages, 2022/04
In the frame of France-Japan collaboration, one of the objectives is to define and assess the calculation methodologies, and to investigate the phenomenology and the consequences of severe accident scenarios in sodium fast reactors (SFRs). A methodology whose purpose is to assess the loadings of the structures induced by a Fuel Coolant Interaction (FCI) taking place in the sodium plenum of SFR has been defined in the frame of the collaboration between France and Japan during 2014-2019. The work progress will be spread over the period 2020-2024 and the main objectives and milestones will be introduced in the paper. The objective of studies is to comprehensively address the margin between the limit of integrity of the main vessel structures and the loadings resulting from severe accidents. For this purpose, the SIMMER mechanistic calculation code simulates core disruptive accident sequences in SFRs. A fluid structure dynamics tool evaluates this interaction i.e. EUROPLEXUS is used in CEA studies and AUTODYN tool is used in JAEA studies. In the paper, a benchmark study is described in order to illustrate the evaluation of vapour expansion phase in the hot plenum. To do that, joint input data are used on the basis of an ASTRID 1500 MWth core degraded state after the power excursion which leads to vapour expansion. The most penalizing case was evidenced in this study by suppressing the action of transfer tube in-core mitigation devices in SIMMER input deck and thus privileging the upward molten core ejection. Even if the most penalizing case was evidenced in this paper, no significant RV deformation was observed in both EUROPLEXUS and AUTODYN calculation results. The assumed mechanical energy was small for the core expansion phase.
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.
Matsuba, Kenichi; Kato, Shinya; Kamiyama, Kenji; Akayev, A. S.*; Baklanov, V. V.*
Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 4 Pages, 2021/08
In order to obtain experimental knowledge on fragmentation and cooling behavior of molten core material discharged into regions where the depth and volume of sodium are limited, a series of out-of-pile experiments using molten alumina as a simulant for molten core material was conducted. It was found that following mechanisms might be involved in the fragmentation and cooling behavior in a shallow sodium pool: (1) FCI which occurs at location of impingement of the molten jet on the bottom plate promotes fragmentation. (2) If there is a sufficient amount of sodium as a heat sink outside the region, heat exchange by sodium flow in and out due to vapor expansion and condensation suppresses the sodium temperature rise. (3) This temperature suppression contributes to effective cooling of molten core material. In the future study, in order to confirm the mechanisms which was clarified in this study, analytical evaluation of the experimental result will be carried out using a simulation tool.
Mori, Kazuhiro*; Okumura, Ryo*; Yoshino, Hirofumi*; Kanayama, Masaya*; Sato, Setsuo*; Oba, Yojiro; Iwase, Kenji*; Hiraka, Haruhiro*; Hino, Masahiro*; Sano, Tadafumi*; et al.
JPS Conference Proceedings (Internet), 33, p.011093_1 - 011093_6, 2021/03
no abstracts in English
Johnson, M.*; Journeau, C.*; Matsuba, Kenichi; Emura, Yuki; Kamiyama, Kenji
Annals of Nuclear Energy, 151, p.107881_1 - 107881_13, 2021/02
Times Cited Count：5 Percentile：86.54(Nuclear Science & Technology)High-resolution X-ray imaging was employed at the JAEA MELT facility to visualize a kilogram-scale interaction between a jet of high temperature molten stainless steel and sodium. A novel software, SPECTRA, has been developed for the quantitative characterization of jet quenching and fragmentation. Tracking and 3D reconstruction of the melt phase traversing the imaging window enabled the detection of 72% of the debris mass recovered post-experiment. The rebounding of melt fragments confirmed a solid outer crust at the melt-coolant interface, while a thermal fragmentation event induced rapid vapor expansion. Jet fragmentation is best explained by the vaporization of coolant entrained within the melt jet generating an internal over-pressure sufficient for fragmentation of the crust. Thermal fragmentation produced a bimodal debris size distribution of coarse jet shells and finer fragments.
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：4 Percentile：80.46(Nuclear Science & Technology)Kamiyama, Kenji
Human Energy Atom, 2021(2), p.30 - 35, 2021/00
Japan Atomic Energy Agency (JAEA) has agreed to the research cooperation on the core safety of sodium-cooled fast reactors (SFRs) with the National Nuclear Center of the Republic of Kazakhstan (NNC-RK), and it has been going on for over 20 years. This research cooperation is called the EAGLE project, which is an advanced and challenging research program utilizing the facilities of NNC-RK. The background and outline of this EAGLE program, as well as the implementation status and major achievements so far, are introduced here.
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：2 Percentile：5.38(Energy & Fuels)Wu, P.*; Fan, F.-R.*; Hagihara, Masato*; Kofu, Maiko; Peng, K.*; Ishikawa, Yoshihisa*; Lee, S.*; Honda, Takashi*; Yonemura, Masao*; Ikeda, Kazutaka*; et al.
New Journal of Physics (Internet), 22(8), p.083083_1 - 083083_9, 2020/08
Times Cited Count：4 Percentile：49.11(Physics, Multidisciplinary)Thermoelectric material SnSe has aroused world-wide interests in the past years, and its inherent strong lattice anharmonicity is regarded as a crucial factor for its outstanding thermoelectric performance. However, the understanding of lattice anharmonicity in SnSe system remains inadequate, especially regarding how phonon dynamics are affected by this behavior. In this work, we present a comprehensive study of lattice dynamics on Na
Sn
Se
S
by means of neutron total scattering, inelastic neutron scattering, Raman spectroscopy as well as frozen-phonon calculations. Lattice anharmonicity is evidenced by pair distribution function, inelastic neutron scattering and Raman measurements. By separating the effects of thermal expansion and multi-phonon scattering, we found that the latter is very significant in high-energy optical phonon modes. The strong temperature-dependence of these phonon modes indicate the anharmonicity in this system. Moreover, our data reveals that the linewidths of high-energy optical phonons become broadened with mild doping of sulfur. Our studies suggest that the thermoelectric performance of SnSe could be further enhanced by reducing the contributions of high-energy optical phonon modes to the lattice thermal conductivity via phonon engineering.
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.
Igarashi, Kai*; Onuki, Ryoji*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08
Sr
NiO studied by neutron diffractionKajimoto, Ryoichi; Nakajima, Kenji; Fujita, Masaki*; Ishikado, Motoyuki*; Torii, Shuki*; Ishikawa, Yoshihisa*; Miao, P.*; Kamiyama, Takashi*
Journal of the Physical Society of Japan, 88(11), p.114602_1 - 114602_6, 2019/11
Times Cited Count：0 Percentile：0(Physics, Multidisciplinary)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.
Ito, Daisuke*; Kurisaki, Tatsuya*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.6430 - 6439, 2019/08
In core disruptive accident of sodium-cooled fast reactor, cooling of residual fuel debris formed in the reactor core is one of important factors to achieve in-vessel retention of the fuel. To clarify the feasibility of the cooling which is called "in-place cooling", characteristics of gas-liquid two-phase flow in the debris bed must be well understood. Since the debris bed can be formed in a confined flow channel in the core, effect of the channel wall cannot be neglected. Thus, this study aims to clarify the effect of the wall on two-phase flow characteristics in the debris bed, which was simulated as a particle bed packed in a pipe. The pressure drop was measured and compared with results by previous models, and porosity and void fraction distributions were measured by X-ray radiography. Then, the pressure drop evaluation model was modified considering the wall effect, and the applicability of the models was discussed.
Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05
In order to ensure In-Vessel Retention (IVR) of molten-core in Core Disruptive Accident (CDA), we are investigating the possibility of the molten-core discharge through the control rod guide tube (CRGT) to prevent energetics due to exceeding the prompt criticality. Internal structures of the CRGT, such as a sodium-flow regulator when the CRGT is connected to the high-pressure plenum, may disturb the discharge of molten-core from the core region. Based on above background, an experimental program to clarify characteristics of molten-core discharge through the CRGT has been commenced as one of subjects under a joint study with National Nuclear Center of the Republic of Kazakhstan (NNC-RK) named EAGLE-3 project. An experiment using molten-alumina as fuel simulant and sodium was conducted at the out-of-pile test facility owned by NNC-RK to investigate sodium cooling effect around the sodium flow regulator on its destruction. The experimental result represented that void development at the initiation of molten-alumina discharge eliminated liquid-phase sodium from the discharge path and this also eliminated sodium cooling effect around the sodium flow regulator. As a result, early destruction of the sodium flow regulator and massive discharge of molten alumina occurred in turn.
Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 11 Pages, 2019/05
The cooling of the residual core materials after the fuel discharge from the core in the accident of SFRs can significantly affect the distribution fraction of the core materials, which is an important factor for the in-vessel retention (IVR). For the evaluation of the cooling of the residual core materials which is called "in-place cooling", behavior in a SFR core was analyzed preliminary by SIMMER-III. Based on the analysis result, method of phenomena identification and ranking table (PIRT) was applied. Fundamental experiment focusing on three thermal-hydraulic phenomena those were extracted by PIRT was considered in order to investigate them and utilize it for validation of the SIMMER-III. To achieve continuous oscillation of sodium level which can occur in the phase of in-place cooling of SFRs, analytical survey was conducted by SIMMER-III. As a result of that, the effects of experimental conditions on the oscillation amplitude and the duration time were clarified quantitatively, which are necessary to determine the specific experimental conditions.
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
