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Estimation of the core degradation and relocation at the Fukushima Daiichi Nuclear Power Station Unit 2 based on RELAP/SCDAPSIM analysis

間所 寛; 佐藤 一憲

Nuclear Engineering and Design, 376, p.111123_1 - 111123_15, 2021/05

 被引用回数:5 パーセンタイル:79.86(Nuclear Science & Technology)

Estimation of the final debris distribution at the Fukushima Daiichi Nuclear Power Plant (1F) is inevitable for a safe and effective decommissioning. It is necessary to clarify possible failure modes of the reactor pressure vessel (RPV), which is influenced by the thermal status of slumped debris that highly depends on the in-vessel accident progression. The accident analysis of 1F Unit 2 (1F2) was conducted using the RELAP/SCDAPSIM code. One of the unsolved issues of 1F2 is the mechanism of three pressure peaks measured through late Mar. 14 to early March 15, 2011. Comparing the results of previous boiling water reactor (BWR) core degradation experiments and that of 1F2 numerical analysis, it can be estimated that most relocated metallic materials had solidified at the core bottom at the onset of first pressure peak. It is likely that the pressure increase occurred due to the evaporation of injected water reaching the heated core plate structures. Between the first and second pressure peaks, the water is assumed to have been injected continuously and the water level was likely to have recovered to BAF at the initiation of the second pressure peak. Probable slumping of a certain amount of molten materials initiated the second pressure peak and the subsequent gradual pressure increase continued possibly due to massive reaction between coolant and remaining Zircaloy in the core. Assuming the closure of the safety relief valve (SRV) at 0:00 on Mar. 15, the third pressure peak was well reproduced in the analysis.


Validation study of SAS4A code for the unprotected loss-of-flow accident in an SFR

石田 真也; 川田 賢一; 深野 義隆

Mechanical Engineering Journal (Internet), 7(3), p.19-00523_1 - 19-00523_17, 2020/06

ナトリウム冷却高速炉の炉心損傷事故(CDA)の起因過程を評価する安全解析コードSAS4Aの客観的な検証の十分性を示すためにSAS4Aの検証にPIRT(Phenomena Identification and Ranking Table)手法を導入した。当該手法に基づいて、課題と検証の目的の明確化、対象施設とシナリオの選定、FOMと重要現象の選定を行い、解析モデルと試験の検討結果を併せて検証マトリクスを作成した。作成した検証マトリクスと試験解析の結果によって、起因過程評価に必要な解析モデルが不足なく検証されていることを示した。加えて、今回の検証マトリクスは各物理現象の関連性も含んだ総合的な検証となっているため、この検証マトリクスを用いた検証は高い信頼性を有する検証であると言える。すなわち、本研究によって、SAS4Aコードの信頼性を大きく向上させることができた。


Development of a fast reactor and related thermal hydraulics studies in Japan

大島 宏之; 上出 英樹

Proceedings of 18th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-18) (USB Flash Drive), p.2095 - 2107, 2019/08

日本ではシビアアクシデント(CDA)対策を重要な視点としてナトリウム冷却高速炉の開発を行ってきた。安全性強化とCDA対策に関連した研究の近年の進捗として、外部事象の一つである火山噴火のPRA, CDA時の崩壊熱除去に関する模擬試験、CDA時の溶融炉心燃料の挙動を評価する上で重要な、ナトリウムプール中の溶融燃料の分散特性にかかる基礎試験などを実施してきた。本論文では、これらの成果について述べる。


Study on the discharge behavior of molten-core through the control rod guide tube in the core disruptive accident of SFR

加藤 慎也; 松場 賢一; 神山 健司; Ganovichev, D. A.*; Baklanov, V. V.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 9 Pages, 2019/05



Validation study of initiating phase evaluation method for the core disruptive accident in an SFR

石田 真也; 川田 賢一; 深野 義隆

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 10 Pages, 2019/05

高速炉の安全研究の分野では炉心損傷事故(CDA)が評価上重要な課題であるとして、当該事故に関する評価手法の研究開発が進められて来ている。その中でSAS4AはCDAの起因過程(IP)の事象進展を評価するために開発が進められている解析コードである。本研究ではSAS4Aの信頼性向上のため、PIRT手法を適用したSAS4Aの検証を行った。SAS4Aの検証は、(1) CDAの代表的な事象であるULOFに対する評価指標(FOM)の選択、(2) ULOFに関連する物理現象の抽出、(3)物理現象のランク付け、(4)評価マトリクスの構築、(5)評価マトリクスに基づく試験解析、という流れで実施し、これによりSAS4Aの信頼性向上を図ることができた。


Coolability evaluation of debris bed on core catcher in a sodium-cooled fast reactor

松尾 英治*; 佐々 京平*; 小山 和也*; 山野 秀将; 久保 重信; Hourcade, E.*; Bertrand, F.*; Marie, N.*; Bachrata, A.*; Dirat, J. F.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 5 Pages, 2019/05



Analysis of fuel subassembly innerduct configurational effects on the core characteristics and power distribution of a sodium-cooled fast breeder reactor

大釜 和也; 中野 佳洋; 大木 繁夫

Journal of Nuclear Science and Technology, 53(8), p.1155 - 1163, 2016/08

 被引用回数:1 パーセンタイル:11.66(Nuclear Science & Technology)

JSFR(Japan Sodium-cooled Fast Reactor)では、炉心崩壊事故(CDA)対策として、内部ダクト付燃料集合体を採用している。炉心核計算において、この内部ダクト構造を直接取扱い、全内部ダクトが炉心中心に対して外側を向くように集合体を配列した場合(外向)、全内部ダクトが内側を向くように集合体を配列した場合(内向)に比較して、炉心中心付近の出力分布が高くなることが報告されている。この要因を分析するため、本研究では、モンテカルロ法に基づく輸送計算および燃焼計算コードを使用し、種々の内部ダクト配列において炉心の出力分布および炉心特性を評価した。この結果、外向および内向配置における炉心中心の出力分布の違いの主要因は、内部ダクト配列の違いに起因する核物質の空間分布の違いであることがわかった。同じメカニズムで、炉心中心以外においても内部ダクト配置の違いにより出力分布に影響が生じることがわかった。また、内部ダクト配置の違いによる制御棒価値への影響を確認した。


A Preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

鈴木 徹; 飛田 吉春; 川田 賢一; 田上 浩孝; 曽我部 丞司; 松場 賢一; 伊藤 啓; 大島 宏之

Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04

 被引用回数:23 パーセンタイル:90.07(Nuclear Science & Technology)

In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss-of-flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of In-Vessel Retention (IVR) for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of IVR against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.


An Experimental study on heat transfer from a mixture of solid-fuel and liquid-steel during core disruptive accidents in Sodium-Cooled Fast Reactors

神山 健司; 小西 賢介; 佐藤 一憲; 豊岡 淳一; 松場 賢一; 鈴木 徹; 飛田 吉春; Pakhnits, A. V.*; Vityuk, V. A.*; Vurim, A. D.*; et al.

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 8 Pages, 2014/12

The relocation of degraded core material through the Control Rod Guide Tubes (CRGTs) is one of essential subjects to achieve the in-vessel retention (IVR) in the case of postulated core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). The CRGT is available as the discharge path by its failure in the core region and heat-transfer from the core-material to the CRGT is one of dominant factors in its failure. In case of a core design into which a fuel subassembly with an inner duct structure (FAIDUS) is introduced, a mixture of solid-fuel and liquid-steel is supposed to remain in the core region since the FAIDUS could effectively eliminate fuel in liquid-state from the core region. Therefore, the objective of the present study is to obtain experimental knowledge for the evaluation of heat-transfer from the mixture of solid-fuel and liquid-steel to the CRGT. In the present study, an experiment was conducted using Impulse Graphite Reactor which is an experimental facility in National Nuclear Center of the Republic of Kazakhstan. In the experiment, the mixture of solid-fuel and liquid-steel was generated by a low-power nuclear heating of fuel and transferring its heat to steel, and then, data to consider the heat-transfer characteristics from the mixture of solid-fuel and liquid-steel to the CRGT were obtained. The heat-transfer characteristic was revealed by evaluating thermocouple responses observed in the experiment. Through the present study, knowledge was obtained to evaluate heat-transfer from the remaining core-materials to the CRGT.


Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

日高 昭秀; 浅香 英明; 上野 信吾*; 吉野 丈人*; 杉本 純

JAERI-Research 99-067, p.55 - 0, 1999/12




Analyses of ALPHA in-vessel debris coolability experiments with SCDAPSIM code

日高 昭秀; 丸山 結; 上野 信吾*; 杉本 純

JAERI-Conf 99-005, p.49 - 55, 1999/07



国際核融合材料照射施設(IFMIF)の設計活動; 現状と今後の展望

勝田 博司; 野田 健治; 加藤 義夫; 杉本 昌義; 前川 洋; 小西 哲之; 中村 秀夫; 井田 瑞穂*; 大山 幸夫; 實川 資朗; et al.

日本原子力学会誌, 40(3), p.162 - 191, 1998/00



Conceptual design study of IFMIF target system

加藤 義夫; 中村 秀夫; 井田 瑞穂*; 前川 洋; 勝田 博司; T.Hua*; S.Cevolani*

Eighth Int. Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-8), 3, p.1260 - 1267, 1997/00



Addendum to IFMIF-CDA interim report


JAERI-Tech 96-036, 32 Pages, 1996/08




Minutes of the Second IFMIF-CDA Design Integration Workshop; May 20-25,1996,JAERI,Tokai,Japan


JAERI-Conf 96-012, 394 Pages, 1996/08




Analytical study on depressurization during PWR station blackout

日高 昭秀; Ezzidi, A.*; 杉本 純

PSA 96: Int. Topical Meeting on Probabilistic Safety Assessment, 3, p.1548 - 1556, 1996/00



IFMIF-CDA Technical Workshop on Lithium Target System; July 18-21, 1995, JAERI, Tokai, Japan


JAERI-Conf 95-019, 257 Pages, 1995/09




SCDAP/RELAP5 analysis of station blackout with pump seal LOCA in Surry plant

日高 昭秀; 早田 邦久; 杉本 純

Journal of Nuclear Science and Technology, 32(6), p.527 - 538, 1995/06

 被引用回数:2 パーセンタイル:28.33(Nuclear Science & Technology)



High heat flux experiments of saddle type divertor module

鈴木 哲; 秋場 真人; 荒木 政則; 佐藤 和義; 横山 堅二; 大楽 正幸

Journal of Nuclear Materials, 212-215(1), p.1365 - 1369, 1994/09



DANKE: 球,棒,平板に対するダンコフ補正因子モンテカルロ計算プログラム

奥野 浩; 小室 雄一

JAERI-M 94-049, 28 Pages, 1994/03



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