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The OECD/NEA Working Group on the Analysis and Management of Accidents (WGAMA); Advances in codes and analyses to support safety demonstration of nuclear technology innovations

中村 秀夫; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*

Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10

The WGAMA activity achievements have been published as technical reports, becoming reference materials to discuss innovative methods, materials and technologies in the fields of thermal-hydraulics, computational fluid dynamics (CFD) and severe accidents (SAs). The International Standard Problems (ISPs) and Benchmarks of computer codes have been supported by a huge amount of the databases for the code validation necessary for the reactor safety assessment with accuracy. The paper aims to review and summarize the recent WGAMA outcomes with focus on new advanced reactor applications including small modular reactors (SMRs). Particularly, discussed are applicability of major outcomes in the relevant subjects of passive system, modelling innovation in CFD, severe accident management (SAM) countermeasures, advanced measurement methods and instrumentation, and modelling robustness of safety analysis codes. Although large portions of the outcomes are considered applicable, design-specific subjects may need careful considerations when applied. The WGAMA efforts, experiences and achievements for the safety assessment of operating nuclear power plants including SA will be of great help for the continuous safety improvements required for the advanced reactors including SMRs.


Development of an integrated computer code system for analyzing irradiation behaviors of a fast reactor fuel

上羽 智之; 根本 潤一*; 伊藤 昌弘*; 石谷 行生*; 堂田 哲広; 田中 正暁; 大塚 智史

Nuclear Technology, 207(8), p.1280 - 1289, 2021/08

 被引用回数:3 パーセンタイル:38.8(Nuclear Science & Technology)



熱流動とリスク評価,1; リスク評価における熱流動解析の役割

丸山 結; 吉田 一雄

日本原子力学会誌ATOMO$$Sigma$$, 63(7), p.517 - 522, 2021/07



The Working group on the analysis and management of accidents (WGAMA); A Historical review of major contributions

Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr$'e$, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; 中村 秀夫

Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09

 被引用回数:2 パーセンタイル:12.91(Nuclear Science & Technology)

WGAMA started on Dec. 31st 1999 to assess and strengthen the technical basis needed for the prevention, mitigation and management of potential accidents in NPP and to facilitate international convergence on safety issues and AM analyses and strategies. WGAMA addresses reactor thermal-hydraulics (Thys), in-vessel behavior of degraded cores, containment behavior and protection, and FP release, transport, deposition and retention, for both current and advanced reactors. This paper summarizes such WGAMA contributions in Thys, CFD and severe accidents, which include the Fukushima-Daiichi accident impacts on the WGAMA activities and their substantial outcomes. Around 50 technical reports have become reference in the related fields, which appear in References. Recommendations in these reports include further research, some of which have given rise to the joint projects conducted or underway within the OECD framework. Ongoing WGAMA activities are numerous and a number of them are to be launched in the near future, which are shortly mentioned too.


Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

小野 綾子; 田中 正暁; 三宅 康洋*; 浜瀬 枝里菜; 江連 俊樹

Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06



Effect of coolant water temperature of ECCS on failure probability of RPV

勝山 仁哉; 眞崎 浩一; Lu, K.; 渡辺 正*; Li, Y.

Proceedings of 2019 ASME Pressure Vessels and Piping Conference (PVP 2019) (Internet), 7 Pages, 2019/07



熱水力安全評価基盤技術高度化戦略マップ2017; 軽水炉の継続的な安全性向上に向けたアプローチ

糸井 達哉*; 岩城 智香子*; 大貫 晃*; 木藤 和明*; 中村 秀夫; 西田 明美; 西 義久*

日本原子力学会誌ATOMO$$Sigma$$, 60(4), p.221 - 225, 2018/04



Thermal-hydraulic analyses of the High-Temperature engineering Test Reactor for loss of forced cooling at 30% reactor power

高松 邦吉

Annals of Nuclear Energy, 106, p.71 - 83, 2017/08



Evaluation of sodium pool fire and thermal consequence in two-cell configuration

高田 孝; 大野 修司; 田嶋 雄次*

Mechanical Engineering Journal (Internet), 4(3), p.16-00577_1 - 16-00577_11, 2017/06



Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

上羽 智之; 大島 宏之; 伊藤 昌弘*

Nuclear Engineering and Design, 317, p.133 - 145, 2017/06

 被引用回数:9 パーセンタイル:67.26(Nuclear Science & Technology)



Development of sodium-water coupled thermal-hydraulics simulation code for sodium-heated straight tube steam generator of fast reactors

吉川 龍志; 田中 正暁; 大島 宏之; 今井 康友*

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10



Outcome of first containment cooling experiments using CIGMA

柴本 泰照; 与能本 泰介; 石垣 将宏; 安部 諭

Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 10 Pages, 2016/10

The Japan Atomic Energy Agency (JAEA) initiated the ROSA-SA project in 2013 for the purpose of studying thermal hydraulics relevant to over-temperature containment damage, hydrogen risk, and fission product transport. For this purpose, the JAEA newly constructed the Containment InteGral Measurement Apparatus (CIGMA) in 2015 for the experiments addressing containment responses, separate effects, and accident managements. Recently, we successfully conducted first experiments using CIGMA to characterize the facility under typical experimental conditions. Among these experiments, the present paper focuses on the results of containment cooling tests, for which an upper part of the vessel outer surface was cooled by spray water. Several distinctive phenomena were observed in the tests, including inverse temperature stratification in the vessel due to the cooling in the upper region. The RELAP5 analysis result was also presented to roughly indicate the prediction capability of the best-estimate two-phase flow code in predicting the containment thermal hydraulics.


鉛ビスマス冷却加速器駆動システムの熱設計,1; 定格運転条件に対する熱流動解析

秋本 肇; 菅原 隆徳

JAEA-Data/Code 2016-008, 87 Pages, 2016/09




An Experimental study on natural circulation decay heat removal system for a loop type fast reactor

小野 綾子; 上出 英樹; 小林 順; 堂田 哲広; 渡辺 収*

Journal of Nuclear Science and Technology, 53(9), p.1385 - 1396, 2016/09

 被引用回数:11 パーセンタイル:72.83(Nuclear Science & Technology)



大型装置CIGMAを用いた格納容器熱水力安全研究; 重大事故の評価手法と安全対策の高度化を目指して

柴本 泰照; 与能本 泰介; 堀田 亮年*

日本原子力学会誌ATOMO$$Sigma$$, 58(9), p.553 - 557, 2016/09

日本原子力研究開発機構安全研究センターでは、シビアアクシデント対策の強化を特徴とする新しい安全規制を支援するため、2013年にROSA-SA計画を開始し、今般、本計画の中核となる大型格納容器実験装置CIGMA(Containment InteGral Measurement Apparatus)を完成させた。CIGMAは、設計温度や計測点密度において世界有数の性能を有しており、シビアアクシデント時の格納容器内の事故進展挙動や事故拡大防止に係る熱水力実験を実施することができる。本稿では、本計画と既往研究の概要を述べるとともに、CIGMA装置の特徴、及びこれまで実施した一連の実験結果を紹介する。


First experiments at the CIGMA facility for investigations of LWR containment thermal hydraulics

柴本 泰照; 安部 諭; 石垣 将宏; 与能本 泰介

Proceedings of 24th International Conference on Nuclear Engineering (ICONE-24) (DVD-ROM), 9 Pages, 2016/06

There has been an extensive reorientation of the light water reactor research in Japan since the Fukushima Dai-ichi Nuclear Power Station accident, which focuses on severe accidents and accident managements. The Japan Atomic Energy Agency (JAEA) initiated the ROSA-SA project in 2013 for the purpose of studying thermal hydraulics relevant to over-temperature containment damage, hydrogen risk, and fission product transport. For this purpose, the JAEA newly constructed the Containment InteGral Measurement Apparatus (CIGMA) in 2015 for the experiments addressing containment responses, separate effects, and accident managements. Recently, we successfully conducted first experiments using CIGMA to characterize the facility under typical experimental conditions investigating basic phenomena such as buildup of pressure by steam injection, containment cooling and depressurization by internal or external cooling, and density stratified layer mixing by impinging jet. This paper provides an overview of the research programs, the brief description of the facility specification and the outcomes obtained from the first experiments.


A Study on the thermal-hydraulics in the damaged subassemblies under the operation of decay heat removal system

小野 綾子; 小野島 貴光; 堂田 哲広; 三宅 康洋*; 上出 英樹

Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.2183 - 2192, 2016/04

ナトリウム高速冷却炉において崩壊熱を除去するいくつかの補助冷却系が検討されている。そのうちの二つがPRACSとDRACSである。本研究では、炉心溶融を引き起こすようなシビアアクシデントを仮定した状況下においてPRACSとDRACSの適用性を確かめるために、模擬炉心やPRACS, DRACSが備え付けられているプラント過渡応答試験装置を用いてナトリウム試験を実施した。炉心溶融は集合体の入口をバルブで閉止することで模擬した。実験結果は、部分破損および全体破損をした炉心においても長期的に安定した冷却がPRACSやDRACSにより可能であることを示した。


熱水力安全評価基盤技術高度化戦略マップの改訂; 軽水炉の継続的な安全性向上に向けて

新井 健司*; 梅澤 成光*; 及川 弘秀*; 大貫 晃*; 中村 秀夫; 西 義久*; 藤井 正*

日本原子力学会誌ATOMO$$Sigma$$, 58(3), p.161 - 166, 2016/03



Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 2; Heat transfer and flow visualization experiment by using internally heated annulus

上澤 伸一郎; 永武 拓; Jiao, L.; Liu, W.; 高瀬 和之; 吉田 啓之

Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 11 Pages, 2015/11

To understand the current status of the TEPCO's Fukushima Daiichi Nuclear Power Station, the progress of the accident has been calculated by severe accident analysis codes, for example, MAAP, SAMPSON and so on. However, effects of seawater are not considered in these calculations, although the seawater was attempted to inject into the reactors to cool down the nuclear fuels. In the present study, the objective is to understand the basic physical effect of the seawater on the thermal-hydraulic behavior without boiling. We measured and compared the thermal-hydraulic behavior in pure water, NaCl solution and artificial seawater with the concentration of 3.5wt% in a heat transfer and flow visualization experiment by using an internally heated annulus. Above Re = 2300 [-], the correlations between Nusselt number and Reynolds number in the NaCl solution and the artificial seawater were the same with that in the pure water. Moreover, the correlation can be predicted by Dittus-Boelter equation. Below Re = 2300 [-], the Nusselt numbers of each fluid correlated with the Rayleigh number. Therefore, considering physical properties of the NaCl solution and the artificial seawater, the thermal-hydraulics behavior without boiling in the NaCl solution and the artificial seawater was not different from the behavior in the pure water.


Evaluation of seawater effects on thermal-hydraulic behavior for severe accident conditions, 1; Outline of the research project

吉田 啓之; 上澤 伸一郎; 永武 拓; Jiao, L.; Liu, W.; 高瀬 和之

Proceedings of International Conference on Power Engineering 2015 (ICOPE 2015) (CD-ROM), 9 Pages, 2015/11

In the Fukushima Daiichi Nuclear Power Plant accident, seawater was injected into the reactor to cool down the nuclear fuels. The injection of the seawater may change the thermal-hydraulic characteristics. Therefore, the thermal hydraulic behavior of seawater has to be evaluated to consider the current status of Fukushima Daiiichi Nuclear Power Plants. However, there is little information about the thermal-hydraulic characteristics of seawater. In order to understand the effects of the seawater on the thermal hydraulic behaviors, a research project was started in Japan Atomic Energy Agency. In this research project, we performed two-different type experiments, one is a heat transfer and visualization experiment by using an internally heated annulus, the other is a heat transfer experiment by using a degraded core simulated test section. In this paper, the outline of the research project and examples of results are reported. For single phase flow conditions, heat transfer coefficients of evaluated by the existing correlation and thermal properties of the artificial seawater almost agreed with the experimental results. For two-phase flow conditions, the results of the artificial seawater were different from that of pure water and the NaCl solution. In the artificial seawater, small solid depositions were observed, and it was considered that these solid depositions affected the thermal hydraulic behavior of the artificial seawater.

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