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Ahmed, Z.*; Wu, S.*; Sharma, A.*; Kumar, R.*; Yamano, Hidemasa; Pellegrini, M.*; Yokoyama, Ryo*; Okamoto, Koji*
International Journal of Heat and Mass Transfer, 250, p.127343_1 - 127343_17, 2025/11
Nishi, Tsuyoshi*; Matsumoto, Saori*; Yamano, Hidemasa; Hayashi, Kiichiro*; Endo, Rie*; Bell, M. R.*; Neubert, L.*; Volkova, O.*
Steel Research International, 96(5), p.2300766_1 - 2300766_6, 2025/05
Times Cited Count:4 Percentile:75.40(Metallurgy & Metallurgical Engineering)The density of Ni-based superalloys is measured using the maximum bubble pressure (MBP) method. The viscosity is evaluated using the oscillating crucible method. The surface tension is simultaneously measured using the MBP method.
Neubert, L.*; Bell, M. R.*; Yamamoto, Taisei*; Nishi, Tsuyoshi*; Yamano, Hidemasa; Ahrenhold, F.*; Volkova, O.*
Steel Research International, 96(5), p.202400237_1 - 202400237_8, 2025/05
Times Cited Count:2 Percentile:0.00(Metallurgy & Metallurgical Engineering)Ishikawa, Takehiko*; Oda, Hirohisa*; Koyama, Chihiro*; Shimonishi, Rina*; Ikeuchi, Rumiko*; Paradis, P.-F.*; Okada, Jumpei*; Fukuyama, Hiroyuki*; Yamano, Hidemasa
International Journal of Microgravity Science and Application, 42(2), p.420202_1 - 420202_10, 2025/04
Ahmed, Z.*; Sharma, A. K.*; Pellegrini, M.*; Yamano, Hidemasa; Okamoto, Koji*
Arabian Journal for Science and Engineering, 50(5), p.3361 - 3371, 2025/03
Times Cited Count:0 Percentile:0.00(Multidisciplinary Sciences)Bell, M. R.*; Neubert, L.*; Sherstneva, A.*; Yamamoto, Taisei*; Nishi, Tsuyoshi*; Yamano, Hidemasa; Weinberg, M.*; Volkova, O.*
Steel Research International, p.2400252_1 - 2400252_10, 2025/00
Times Cited Count:1 Percentile:0.00(Metallurgy & Metallurgical Engineering)In this study, the thermophysical properties of low-sulfur manganese-boron steel with varying boron and sulfur contents at different temperatures are investigated.
Yamano, Hidemasa; Toyooka, Junichi; Sato, Hiroyuki; Sakaba, Nariaki
Nihon Genshiryoku Gakkai-Shi ATOMO, 66(12), p.607 - 611, 2024/12
This report mainly introduces trends in fast reactor development in Japan in addition to introducing overseas development trends for major developing countries.
Yamano, Hidemasa; Morita, Koji*
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (Internet), 9 Pages, 2024/11
Emura, Yuki; Matsuba, Kenichi; Kikuchi, Shin; Yamano, Hidemasa
Proceedings of 13th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS13) (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
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.
Kasahara, Naoto*; Yamano, Hidemasa; Nakamura, Izumi*; Demachi, Kazuyuki*; Sato, Takuya*; Ichimiya, Masakazu*
International Journal of Pressure Vessels and Piping, 211, p.105298_1 - 105298_6, 2024/10
Sogabe, Joji; Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Kamiyama, Kenji; Onoda, Yuichi; Matsuba, Kenichi; Yamano, Hidemasa; Kubo, Shigenobu; Kubota, Ryuzaburo*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
In the frame of France-Japan collaboration, the calculational methodologies were defined and assessed, and the phenomenology and the severe accident consequences were investigated in a pool-type sodium-cooled fast reactor.
Wozniak, N.*; Shemon, E.*; Feng, B.*; Ohgama, Kazuya; Doda, Norihiro; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ota, Hirokazu*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
To enhance the accuracy of the safety evaluations in sodium-cooled fast reactors, it is necessary to develop a method to realistically evaluate the reactivity caused by core deformation. In this regard, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts of Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and test results revealed that the core bowing analysis codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies.
Yamano, Hidemasa; Emura, Yuki; Takai, Toshihide; Kubo, Shigenobu; Quaini, A.*; Fossati, P.*; Delacroix, J.*; Journeau, C.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
This report mainly introduces trends in fast reactor development in Japan in addition to introducing overseas development trends for major developing countries. The paper describes major severe accident study results focusing on kinetics of interaction in core material mixtures, physical properties of core material mixtures, high temperature thermodynamic data for the uranium oxide (UO)-iron (Fe)-boron carbide (B
C) system, experimental studies on B
C-stainless steel (SS) kinetics and B
C-SS eutectic material relocation (freezing), and B
C-SS eutectic and kinetics models for severe accident code systems,
Onoda, Yuichi; Ishida, Shinya; Fukano, Yoshitaka; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Shibata, Akihiro*; Bertrand, F.*; Seiler, N.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
Hayashi, Masaaki*; Nakahara, Hirotaka*; Shirakura, Shota*; Yamano, Hidemasa
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
As part of the development of risk assessment technologies for sodium-cooled fast reactor coupled to thermal energy storage (TES) system with sodium-molten salt heat exchanger (HX), simple evaluation of heat transfer performance using heat transfer coefficient formula is performed. And Computational Fluid Dynamics (CFD) thermal analyses by STAR-CCM+ with partial HX model are performed to develop evaluation technology. The performance evaluation technology of a HX between sodium and molten salt and the confirmation of heat transfer improvement measures effects is developed.
Ohgama, Kazuya; Doda, Norihiro; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ota, Hirokazu*; Ogata, Takanari*; Wozniak, N.*; Shemon, E.*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
To enhance the accuracy of the safety evaluations in sodium-cooled fast reactors, it is necessary to develop a method to realistically evaluate the reactivity caused by core deformation. In this regard, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments of a single duct of Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and test results revealed that the core bowing analysis codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement of a single duct due to thermal bowing and the contact load on the duct pad.
Futagami, Satoshi; Kondo, Yuki; Yamano, Hidemasa; Kurisaka, Kenichi
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 9 Pages, 2024/10
Onoda, Yuichi; Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
We developed the measures for improving resilience of the sodium-cooled fast reactor structure using the failure mitigation technology and evaluated the effectiveness of the measures. To prevent core damage in the event of an accident progressing to an ultra-high temperature state, both measures to prevent overpressure in the reactor vessel and measures to cool the reactor core are required. As a core cooling measure, we developed a core cooling concept that promotes radiant heat transfer from the reactor vessel and cools the containment vessel outer surface by natural convection named Containment Vessel Auxiliary Cooling System (CVACS). We developed a method to use the reduction rate of core damage frequency as an indicator for effectiveness of the measures for improving resilience. The core damage frequency was evaluated by calculating the core cooling performance using CVACS, reflecting the results of structural analysis and human reliability analysis. By implementing measures for improving resilience in addition to existing measures, the core damage frequency of Japan loop-type sodium-cooled fast reactor caused by LOHRS has been reduced to about one-hundredth of the previous level.
Kurisaka, Kenichi; Nishino, Hiroyuki; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
The objective of this study is to implement an effectiveness evaluation of the measures for improving resilience of nuclear structures against excessive earthquake. In this study, those measures for improving resilience have an effect to enlarge their seismic safety margin. To evaluate effectiveness of those measures, seismic core damage frequency (CDF) is selected as an index. Reduction of CDF as an effectiveness index is quantified by applying seismic PRA technology. Target system is a loop-type next-generation sodium-cooled fast reactor, which adopts the building isolated from horizontal seismic ground motion. Even if the reactor vessel (RV) is buckled due to seismic shaking, it is expected that the RV maintains stable state without unstable failure such as rupture, collapse. Realistic consideration of the post-buckling behavior is regarded as a measure for improving resilience in this study. We set two cases for improving the resilience in the accident sequences analysis. The first case assumes low-cycle fatigue failure after buckling as the realistic failure mode of the RV, and we applied the fragility evaluated in our study. After the RV fatigue failure, the behavior of failure propagation is very uncertain. As the second case, the median seismic capacity to loss of reactor level is assumed to be slightly larger than that of fatigue failure of the RV. Analyses for both cases were performed, and the results were compared to the base case indicating significant reduction of CDF. Within the assumption, the measures for improving the resilience were significantly effective for decreasing CDF in excessive earthquake up to several times of a design basis ground motion. The seismic PRA technology could serve to the effectiveness evaluation of the measures for improving resilience of nuclear structures against excessive earthquake.