Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
Initialising ...
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：1 Percentile：83.53(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：1 Percentile：0.02(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：0 Percentile：0(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：2 Percentile：54.33(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 (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 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
Kurisaki, Tatsuya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11
In the evaluation of the in-place cooling which is for the residual core materials in the severe accident of sodium-cooled fast reactors, pressure loss of two-phase flow in debris bed is one of the important factors. Although Lipinski model is already proposed for the pressure loss evaluation, the accuracy would decrease when the porosity is not homogeneous. Thus, experiment to measure the pressure loss in a packed bed of non-homogeneous porosity distribution was conducted, and the Lipinski model was modified dividing the cross section to evaluate the pressure loss in it. As a result, it was confirmed that agreement of the experimental values with the values by modified Lipinski model was better than that with the original Lipinski model.
Matsuba, Kenichi; Kamiyama, Kenji; Ganovichev, D. A.*; Baklanov, V. V.*
Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 4 Pages, 2018/11
In Core Disruptive Accidents of Sodium-cooled Fast Reactors, molten core material would be discharged through control rod guide tubes into the inlet coolant plenums beneath the rector cores. The inlet coolant plenums have quite limited heights and sodium inventories. Therefore, in the inlet plenums, molten core material with a jet-like shape would impinge on the bottom of the plenum before it breaks up into fragments. In this study, to clarify fragmentation behavior in a shallow sodium pool whose height and volume are so limited that jet impingement on the bottom is expected, an out-of-pile experiment discharging molten alumina into a sodium pool was conducted. Although a small amount of alumina agglomeration was found on the bottom plate (steel disk) installed in the sodium pool, most of the molten alumina was fragmented into debris particles. Results obtained in the present experiment suggest that molten core material is fragmented and quenched even in a shallow sodium pool.
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.
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.
Wu, P.*; Zhang, B.*; Peng, K. L.*; Hagiwara, Masayuki*; Ishikawa, Yoshihisa*; Kofu, Maiko; Lee, S. H.*; Kumigashira, Hiroshi*; Hu, C. S.*; Qi, Z. M.*; et al.
Physical Review B, 98(9), p.094305_1 - 094305_7, 2018/09
Times Cited Count：5 Percentile：40.47(Materials Science, Multidisciplinary)Using angle-resolved photoemission spectroscopy and inelastic neutron scattering, we have studied how electronic structures and lattice dynamics evolve with temperature in Na-doped SnSe.
Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Aoyagi, Mitsuhiro; Matsuba, Kenichi; Kamiyama, Kenji
Nuclear Engineering and Design, 334, p.90 - 95, 2018/08
Times Cited Count：5 Percentile：65.87(Nuclear Science & Technology)Two-phase flow through porous media must be well understood to develop a severe accident analysis code not only for light water reactor but also sodium-cooled fast reactor. When a core disruptive accident occurs in sodium-cooled fast reactor, the fuel inside the core become melted and interacts with the coolant. As a result, gas-liquid two-phase flow will be formed in the debris bed, which may have porous nature depending on the cooling process. In such condition, the local porosity and its distribution are very important to characterize two-phase flow field in the porous media. In this study, X-ray radiography was applied to measure the local porosity in the packed bed of spheres. The radial profiles were estimated from the chordal profiles measured by the X-ray method and compared with the previous porosity model. In addition, the void fraction radial profiles were also obtained in air-water two-phase flow.
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：5 Percentile：65.87(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, Al
O
, 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.
