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Yamano, Hidemasa; Futagami, Satoshi; Sasa, Kyohei*; Nakamura, Hironori*; Tokizaki, Minako*; Kubota, Ryuzaburo*
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 12 Pages, 2025/09
This study examined the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on the passive reactor shutdown capability to sodium-cooled fast reactors (SFRs) recently designed in Japan.
Wozniak, N.*; Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Shemon, E.*; Feng, B.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.165 - 170, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies in multiple duct configuration.
Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Wozniak, N.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ogata, Takanari*; Shemon, E.*; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.160 - 164, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using a single duct from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement caused by thermal bowing as well as the contact load on the duct pad.
Watakabe, Tomoyoshi; Yamano, Hidemasa; Futagami, Satoshi
Transactions of the 28th International Conference on Structural Mechanics in Reactor Technology (SMiRT28) (Internet), 10 Pages, 2025/08
The fragility evaluation method of the RV in the SFR has been developed for both failure mode buckling and fatigue. The past studies showed the fragility using the RV of a loop-type SFR: radius R / thickness t is about 100. Recently, Japan Atomic Energy Agency together with Japanese industrial partners has been conducting a pool-type SFR design. The fragility of the pool-type RV has never been evaluated by the proposed method, which is based on the fatigue failure mode instead of the buckling. To investigate the basic behaviour of the pool-type RV under excessive seismic loading, this report investigated the fragility by using a simple cylindrical model with a wall thickness close to the design of the pool-type RV: R/t is 200, as the first trial.
Futagami, Satoshi
Nihon Genshiryoku Gakkai-Shi ATOMO
, 66(11), p.555 - 559, 2024/11
no abstracts in English
Kinoshita, Takahiro*; Okamura, Shigeki; Nishino, Hiroyuki; Yamano, Hidemasa; Kurisaka, Kenichi; Futagami, Satoshi; Fukasawa, Tsuyoshi*
International Journal of Pressure Vessels and Piping, 211, p.105299_1 - 105299_5, 2024/10
Times Cited Count：1 Percentile：16.65(Engineering, Multidisciplinary)The seismic evaluation of key components such as reactor vessel is important for the Seismic Probabilistic Risk Assessment (S-PRA) in a Sodium-Cooled Fast Reactor (SFR). Many components were fractured by integrated damage like fatigue damage during seismic ground motion. In this paper, failure probability evaluation method with integrated energy was developed by comparing the energy with vibration tests and fatigue tests. Vibration tests were performed to evaluate integrated vibration energy at failure by energy balance equation, and fatigue tests were performed to evaluate integrated vibration energy at failure based on experimental results. As results, it is shown that integrated energy at failure time by vibration tests were estimated and its values were in range the energy based on results of fatigue tests.
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.
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.
Nishino, Hiroyuki; Kurisaka, Kenichi; Futagami, Satoshi; Watakabe, Tomoyoshi; Yamano, Hidemasa
Proceedings of Probabilistic Safety Assessment and Management & Asian Symposium on Risk Assessment and Management (PSAM17 & ASRAM2024) (Internet), 10 Pages, 2024/10
The reactor vessel (RV) buckling was a dominant contributor to core damage. However, even if the 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 the resilience in this study. The purpose of this study is to understand the post-buckling deformation behavior of the RV and to evaluate the RV fragility based on fatigue failure. This study performed structural analysis using a finite element method to quantify time histories of displacement, strain, etc. As the result of the analysis, wrinkles of the buckling appeared at the elevation higher than the liquid level in the RV. The largest strain value was also indicated around this elevation. The cumulative fatigue damage fraction was evaluated in this analysis to evaluate the fragility of fatigue failure in addition to the buckling fragility. The result showed that the seismic intensity for the median fragility of the fatigue failure was about six times larger than the design-basis ground motion. This is 1.2 times larger than the buckling-based result, which suggests that realistic evaluation of the post-buckling behavior could contribute to improving the resilience of the nuclear structure.
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
Yamano, Hidemasa; Futagami, Satoshi; Doda, Norihiro; Tagami, Hirotaka; Uchibori, Akihiro; Ogata, Takanari*; Ota, Hirokazu*
Nihon Kikai Gakkai 2024-Nendo Nenji Taikai Koen Rombunshu (Internet), 5 Pages, 2024/09
Yamano, Hidemasa; Futagami, Satoshi; Higurashi, Koichi*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-14) (Internet), 12 Pages, 2024/08
This paper describes the application of safety design criteria (SDC) and safety design guidelines (SDG) developed in the Generation-IV International Forum on the natural circulation of sodium to sodium-cooled fast reactors (SFRs) recently designed in Japan.
Yamano, Hidemasa; Futagami, Satoshi; Shibata, Akihiro*
Proceedings of Advanced Reactor Safety (ARS 2024), p.151 - 160, 2024/08
This study examined the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on the active reactor shutdown system (RSS) to sodium-cooled fast reactors (SFRs) recently designed in Japan.
Yamano, Hidemasa; Futagami, Satoshi; Higurashi, Koichi*
Proceedings of Advanced Reactor Safety (ARS 2024), p.121 - 130, 2024/08
This paper describes the application of safety design criteria (SDC) and safety design guideline (SDG) developed in the Generation-IV international forum on decay heat removal system (DHRS) enhancing reliability to sodium-cooled fast reactors (SFRs) recently designed in Japan.
Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori; Kurisaka, Kenichi
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 11 Pages, 2024/03
In this study, the dynamic structural analysis of the reactor vessel for excessive earthquake using the FINAS/STAR code has shown the elephant foot buckling deformation and calculated the cumulative fatigue failure fraction. Using the calculation results, this paper describes the fragility curve using the safety factor method, indicating the significantly improved curve compared the previous one.
Futagami, Satoshi; Ando, Masanori; Yamano, Hidemasa
Transactions of the 27th International Conference on Structural Mechanics in Reactor Technology (SMiRT 27) (Internet), 10 Pages, 2024/03
Futagami, Satoshi; Yamano, Hidemasa; Kurisaka, Kenichi; Ujita, Hiroshi*
Proceedings of PSAM 2023 Topical Conference AI & Risk Analysis for Probabilistic Safety/Security Assessment & Management, 8 Pages, 2023/10
To create an innovation for efficient and effective social implementation of nuclear power plant PRA, automatic construction tool for fault tree architecture and automatic failure judgment tool to construct reliability database are developed by using AI and digitization technology. This paper describes overall development plan of PRA methodology using the AI technology and the progress of automatic FT creation tools development.
Ujita, Hiroshi*; Morimoto, Tatsuya*; Futagami, Satoshi; Yamano, Hidemasa; Kurisaka, Kenichi
Proceedings of PSAM 2023 Topical Conference AI & Risk Analysis for Probabilistic Safety/Security Assessment & Management, 8 Pages, 2023/10
This study is intended to develop PRA methodology using the AI technology. For this purpose, as a first step, the authors have been conducting a three-year program including the development of AI tools for automatic fault tree (FT) creation and automatic fault detection methodology for building reliability database. These AI tools are intended to enable any users to easily perform PRA with the same quality without user effect. For the automatic fault detection method, The AI tool is developed for extracting failure occurrence locations (system/equipment), failure modes, and causes from Japanese reliability databases of NUCIA(for light water reactors) and CORDS(for sodium-cooled fast reactors), and transforming them into a database using AI technologies.
Yamano, Hidemasa; Futagami, Satoshi; Ando, Masanori
Mechanical Engineering Journal (Internet), 10(4), p.23-00043_1 - 23-00043_12, 2023/08
This study has conducted a detailed structural analysis of a reactor vessel (RV) in a loop-type sodium-cooled fast reactor using a general-purpose finite element analysis code, FINAS/STAR, to understand its deformation behavior under extremely high temperature conditions and to identify the areas which should be focused to mitigate impacts of failure. The RV was heated from the normal operation condition to the sodium boiling temperature in the upper sodium plenum during 20 hours assuming depressurization. The analysis has revealed less significant stress and strain which were sufficiently lower than failure criteria. The upper body of RV was identified as the important area in terms of mitigation of structural failure. The RV was eventually deformed downward about 16 cm, resulting in no failure. This effect contributes to maintaining RV sodium level in a long term, thereby enhancing the RV resilience.
Ota, Hirokazu*; Ogata, Takanari*; Yamano, Hidemasa; Futagami, Satoshi; Shimada, Sadae*; Yamada, Yumi*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
