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Zou, Z.*; 守田 幸路*; 石田 真也; 山野 秀将
International Communications in Heat and Mass Transfer, 180, Part 2, p.112532_1 - 112532_22, 2026/11
被引用回数：0During severe accidents in sodium-cooled fast reactors, a large amount of molten fuel may form within the core and accumulate at its center, leading to central sloshing. This behavior can be triggered by pressure surges caused by fuel-coolant interactions between molten fuel and liquid sodium, which drive the fuel outward, followed by an inward motion that leads to central accumulation. However, unlike ordinary engineering fluids, molten fuel may undergo prompt criticality when compressed during sloshing, resulting in rapid energy release that can threaten the structural integrity of the core and reactor vessel. Therefore, from the viewpoint of nuclear safety, it is essential to investigate the dynamic core response during central sloshing of molten fuel. To address this issue, this study proposes a novel theoretical model that uses the density-normalized pressure increase rate at the onset of prompt criticality to evaluate the potential energy release.
今泉 悠也; 石田 真也; 松場 賢一; 山野 秀将
Proceedings of the 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-15) (Internet), 9 Pages, 2026/09
In severe accidents of sodium-cooled fast reactors (SFRs), blockage formation in a large-diameter relocation duct by the solidification of molten core materials is one of the most important factors determining the reactivity transient in an in-vessel retention strategy. The severe accident analysis code, SIMMER, has several physical models which can simulate blockage formation by solid particles and/or crust formation in a pin-bundle geometry. In this study, those models were improved for the blockage formation in a large-diameter duct for fuel relocation. The improved model was validated by SIMMER analyses of two experiments. One investigated narrow channel flow and blockage formation by solidification. The other investigated molten material flow in a large-diameter duct simulating fuel relocation in SFRs. The improved model well simulated these experiments. In particular, no blockage formation observed in the experiment was simulated by the improved model. Effect by the model improvement was investigated in a severe accident analysis of an SFR, and significant decrease of core fuel amount and reactivity decrease occurred in the analysis.
山野 秀将; 高井 俊秀; 江村 優軌; 福山 博之*; 東 英生*; 西 剛史*; 守田 幸路*; 中村 勤也*; Ahmed, Z.*
Proceedings of the 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NUTHOS-15) (Internet), 12 Pages, 2026/09
This paper summarizes the research project conducted until 2024. As part of state-of-the-art studies, this paper describes validation of the developed physical model using two experiments: eutectic reaction experiment with a B
C-SS pellet into liquid SS pool and freezing experiment using the BC-SS eutectic reaction product.
石田 真也; 今泉 悠也; 松場 賢一; 山野 秀将
第30回動力・エネルギー技術シンポジウム講演論文集(インターネット), 5 Pages, 2026/06
酸化物燃料を用いたナトリウム冷却高速炉実証炉におけるシビアアクシデント時の再臨界回避に係る評価を適切に行うためには、解析手法を整備し、当該手法の妥当性を確認する必要がある。ここでは、高速炉のシビアアクシデントの代表的な事象である流量喪失時炉停止失敗事象に対しPIRT手法を適用し、実証炉への導入が検討されている溶融燃料流出管に関連する構造壁の溶融・破損や燃料流出といった現象を重要現象として同定するとともに、各重要現象に対応するSIMMER-III/-IVコードの解析モデルの妥当性確認の方法を整理した。
山野 秀将; 松場 賢一; 佐々 京平*; 木下 円機*; 守田 幸路*; 齊藤 泰司*; 伊藤 啓*; 鈴木 徹*
第30回動力・エネルギー技術シンポジウム講演論文集(インターネット), 5 Pages, 2026/06
A research project on severe accidents of oxide-fueled core has been started since 2024 as research and development in a conceptual design stage of demonstration sodium-cooled fast reactors in Japan. It is necessary to develop analytical methodologies and technical bases to confirm the effectiveness of various design measures, which are introduced to the demonstration reactor in order to achieve recriticality elimination and in-vessel retention. The project implements element testing, integral testing, analytical methodology development and reactor application analyses, in which involved are important phenomena of fuel discharge behavior, core residual fuel coolability, and debris-bed formation behavior and coolability. This paper describes the project overview and progress until 2025.
C-stainless steel composites using the improved MPS methodAhmed, Z.*; Wu, S.*; Sharma, A.*; Kumar, R.*; 山野 秀将; Pellegrini, M.*; 横山 諒*; 岡本 孝司*
International Journal of Heat and Mass Transfer, 250, p.127343_1 - 127343_17, 2025/11
被引用回数：5 パーセンタイル：72.82(Thermodynamics)The study aims to measure boron concentration through the unidirectional diffusion of boron within the stainless steel (SS) layers while evaluating the updated model ability to replicate melt relocation behavior and geometry. In the current MPS simulations, one scenario employed dummy walls as heat sources, while another scenario used SS surface particles as heat sources to avoid interference with the melt flow as it reached the bottom of the specimen.
石田 真也; 田上 浩孝*; 山野 秀将; 久保 重信
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
An improved and detailed fuel pin model has been developed at JAEA as part of a new SIMMER-V code under a joint research program being conducted in the frame of the French-Japanese collaboration. The model allows a continuous analysis from the initiating phase to the transition phase in future sodium-cooled fast reactors (SFRs). In addition, the model can accommodate specific core geometries such as annular fuel pellets and axially heterogeneous cores. This paper describes the validation study of the new pin model based on the analysis of a CABRI test. CABRI E7 test analyses are conducted by SIMMER-V and SAS4A codes. The validation results show that the new pin model has a capability to express the physical phenomena related to mechanical fuel pin failure for severe accidents of SFRs. Through the new pin model validation work, the model improvement enables more realistic evaluation of the mechanical fuel pin failure and the reliability of the new pin model has been improved.
小野田 雄一; 久保田 龍三朗*; 山野 秀将
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 11 Pages, 2025/09
Calculations of core degradation behavior under Unprotected Loss of Flow (ULOF) accident of ASTRID CFV-type core equipped Sodium-cooled Fast Reactor (SFR) had been conducted using severe accident analysis code SIMMER-III. The main objective of this study is to understand event sequences accompanied by re-criticality due to core material motion. Fuel melting and significant core degradation leading to fuel discharge into the upper and lower direction of the core occurred in the reference case calculation. Two types of sensitivity calculations have been done. One aims to investigate the effect of Fuel-Coolant Interaction (FCI) in the failed control rod guide tube. The other one aims to investigate the effect of pressure generation due to FCI inside a special device which discharges molten core material into the lower plenum. This study is a part of France-Japan collaboration on R&Ds of SFRs conducted from 2020 to 2024.
今泉 悠也; 曽我部 丞司*; 小野田 雄一; 山野 秀将
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 9 Pages, 2025/09
Cooling capability of core remaining fuel after core degradation in ULOF accident in a SFR was evaluated. Two analysis cases with and without coolant re-entry from lower side of transfer tubes were defined in this study, because the cooling capability depends on it. For the no coolant re-entry case where stratified layers of fuel and steel are formed and cooled down in a long time, new analysis code with one-dimensional heat and mass transfer model, called PAHRCR, was developed and applied. As a result of the analyses, the melting of lower core structure stopped in the 60 % fuel inventory case, but it was melted and almost lost in the 70 % fuel inventory case. This result showed the coolable inventory of the core fuel. For the analysis of the coolant re-entry case with assumption of molten fuel discharge, the SIMMER-III code was used. As a result, the remaining core fuel was cooled down in relatively short time, and all fuel particles were finally immersed in the coolant.
山野 秀将; 二神 敏; 佐々 京平*; 中村 博紀*; 時崎 美奈子*; 久保田 龍三朗*
Proceedings of 2025 International Congress on Advances in Nuclear Power Plants (ICAPP 2025) (Internet), 12 Pages, 2025/09
本研究では、受動的炉停止能力に関して、第4世代国際フォーラムで開発された安全設計クライテリアとガイドラインを我が国で最近に設計されたナトリウム冷却高速炉へ適用した。
高野 和也; 栗坂 健一; 山野 秀将
Progress in Nuclear Science and Technology (Internet), 8, p.82 - 85, 2025/09
ナトリウム-溶融塩熱交換器を有する蓄熱式高速炉のリスク評価技術開発の一環として、溶融塩を活用した既存の太陽熱蓄熱発電システムにおける事故トラブル事例結果に基づき、熱交換器における伝熱管破損件数と溶融塩暴露時間を整理するとともに、ベイズ推定手法に基づき伝熱管破損発生率を評価する方法を検討した。
大釜 和也; 堂田 哲広; 太田 宏一*; Wozniak, N.*; 上羽 智之; 二神 敏; 田中 正暁; 山野 秀将; 尾形 孝成*; Shemon, E.*; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.160 - 164, 2025/09
ナトリウム冷却高速炉の安全評価の精度向上のため、炉心変形による反応度を現実的に評価する手法の開発が必要である。炉心を構成する集合体の熱湾曲解析モデルの妥当性を確認することを目的に、「常陽」燃料集合体の単体ダクトの熱湾曲試験を対象としたベンチマーク解析を日米共同で実施した。解析結果と試験結果の比較から、両国の炉心湾曲コードは単体ダクトの水平変位の軸方向分布とダクトパッド部の接触荷重を合理的に予測できることがわかった。
Wozniak, N.*; 大釜 和也; 堂田 哲広; 太田 宏一*; Shemon, E.*; Feng, B.*; 上羽 智之; 二神 敏; 田中 正暁; 山野 秀将; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.165 - 170, 2025/09
ナトリウム冷却高速炉の安全評価の精度向上のため、炉心変形による反応度を現実的に評価する手法の開発が必要である。炉心を構成する集合体の熱湾曲解析モデルの妥当性を確認することを目的に、「常陽」燃料集合体の複数ダクトの熱湾曲試験を対象としたベンチマーク解析を日米共同で実施した。解析結果と試験結果の比較から、両国の炉心湾曲コードは、複数のダクトで構成されるダクト列の熱湾曲を合理的に予測できることがわかった。
佐藤 理花; 近藤 俊樹; 梅田 良太; 菊地 晋; 山野 秀将
Progress in Nuclear Science and Technology (Internet), 8, p.137 - 142, 2025/09
ナトリウム-溶融塩熱交換器を有する蓄熱式高速炉では、ナトリウム(Na)と硝酸系溶融塩との熱交換器伝熱管破損に至るような仮想的な事故条件下でNaと硝酸系溶融塩との化学反応が発生する可能性がある。そのため、Naと硝酸系溶融塩の反応挙動は、当該システムの安全評価上、重要現象の一つとなっている。本研究では、NaNO
-KNOの混合物であるソーラーソルトとNaとの反応試験を実施し、得られた試験結果について検討を行った。その結果、ソーラーソルトの融解が開始した後にNaとの反応が起こることが分かった。試験で得られた反応温度から、速度論的パラメータおよび反応速度を求め、Na-水反応と比較した。その結果、Na-溶融塩熱交換器を有する蓄熱式高速炉の伝熱管破損時の事象進展で勘案すべき時間スケール内にソーラーソルト反応が生じ得ることが分かった。
田上 浩孝*; 岡野 靖; 山野 秀将
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 12 Pages, 2025/08
The metal-fuel specific physical models for uranium-iron eutectic reaction and metal fuel pin behavior have been developed and incorporated into the SIMMER-III/IV code for safety analyses of metal-fuel sodium-cooled fast reactors. The TREAT M6 experimental analysis was performed to validate the metal-fuel pin model.
C-stainless steel composite under severe reactor conditionsAhmed, Z.*; Kumar, R.*; 横山 諒*; 岡本 孝司*; Pellegrini, M.*; 山野 秀将
Proceedings of 21st International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-21) (Internet), 15 Pages, 2025/08
Scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM-EDS) and X-ray photoelectron spectroscopy (XPS) confirmed the formation of metal carbides and the presence of boron in the eutectic melt. Characterizations results tell that not only boron, but carbon too diffuses in the eutectic melt which is a concern of recriticality. The current findings confirm the relocation of the BC-SS eutectic mixture with diffusion and precipitation of both boron and carbon, conditions likely to occur under extreme reactor conditions.
渡壁 智祥; 山野 秀将; 二神 敏
Transactions of the 28th International Conference on Structural Mechanics in Reactor Technology (SMiRT28) (Internet), 10 Pages, 2025/08
座屈と疲労の両破損モードを対象にした、高速炉の原子炉容器のフラジリティ評価法がこれまでに開発された。既往報告において、ループ型高速炉の原子炉容器を対象にしたフラジリティを提示した。近年、JAEAは日本の民間協力会社と共同でタンク型高速炉の設計を検討している。タンク型容器のフラジリティは、座屈の代わりに疲労を破損と想定した提案法によってまだ評価されていない。本報告では、最初の試みとして、過大地震荷重下でのタンク型容器の基本特性を調査するため、タンク型容器の設計形状に近い薄肉単純円筒モデルを用いてフラジリティを評価した。
山野 秀将; 豊岡 淳一
日本原子力学会誌ATOMO
, 67(7), p.389 - 393, 2025/07
再生可能エネルギー導入拡大を見据えた新型炉に期待される新たな技術開発として、IAEAとGIFの取組を中心に、再エネ協調技術開発の国内外動向を紹介する。
Neubert, L.*; Bell
, M. R.*; 山本 泰生*; 西 剛史*; 山野 秀将; Ahrenhold, F.*; Volkova, O.*
Steel Research International, 96(5), p.2400237_1 - 2400237_8, 2025/05
被引用回数：6 パーセンタイル：48.92(Metallurgy & Metallurgical Engineering)Density, surface tension, and viscosity of various liquid electrical steels are measured at different temperatures, varying in their silicon content between 3 and 6mass%.
Bell, M. R.*; Neubert, L.*; Sherstneva, A.*; 山本 泰生*; 西 剛史*; 山野 秀将; Weinberg, M.*; Volkova, O.*
Steel Research International, 96(5), p.2400252_1 - 2400252_10, 2025/05
被引用回数：11 パーセンタイル：71.90(Metallurgy & Metallurgical Engineering)低硫黄マンガンボロン鋼について、異なる温度域でボロンと硫黄の含有量を変えて、物性値を測定した。
