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宇佐美 博士; 吉永 恭平*; 藤川 圭吾*
日本原子力学会誌ATOMO
, 67(5), p.295 - 299, 2025/05
日本原子力研究開発機構では、東京電力HD(株)福島第一原子力発電所の廃炉等を始めとした原子力分野の課題解決に資するため、国内外の英知を結集し、様々な分野の知見や経験を、従前の機関や分野の壁を超えて緊密に融合・連携させることにより、基礎的・基盤的研究や、産学が連携した人材育成の取組を推進している。令和6年度から「シビアエンジニアリングマネジメント学」という新たな学問体系を基軸としたこれまでにないユニークな研究人材育成事業を開始したため、本事業の概要や狙い、現在までの取組状況について紹介する。
寺田 敦彦; Thwe Thwe, A.; 日野 竜太郎*
JAEA-Review 2024-049, 400 Pages, 2025/03
JAEA-Review-2024-049.pdf:13.94MB
福島第一原子力発電所(1F)事故における水素爆発を鑑みて、原子力技術者が理解しておくべき水素安全技術の先端を示しつつ、原子力技術者に協力すべき燃焼、爆発専門家向けに原子力における水素安全の要点を示し、事故後廃棄物管理までを視野に入れて放射線分解水素に関する情報を加えた「原子力における水素安全対策高度化ハンドブック(第1版)」を2017年に刊行した。その後、水素安全対策の合理的な高度化や水素安全評価のさらなる信頼性の向上に向けて、原子力事故の解析に一般的に用いられている集中定数系(LP)コードを補完するうえで、原子炉格納容器(CV)内での局所的な水素濃度上昇の影響、着火後の火炎伝播加速による安全機器の健全性、水素処理対策の妥当性等をより精緻、かつ定量的に評価できる数値流体力学(CFD)解析への期待が高まっている。これは、さらなる水素挙動や爆発燃焼に対する安全性の向上を図ることが必要とされていることにもよる。そこで、加圧水型原子炉(PWR)を対象に、シビアアクシデント(SA)時の水素拡散から爆発燃焼、それによるCV及びCV内の安全機器への影響評価までを解析するCFDによる水素挙動統合解析システムを構築・整備してきた。ここで得られたSA安全対策、それによる水素安全向上、安全対策を踏まえた水素発生事故に対する安全性評価などについて、LP及びCFD解析の役割や活用例を本「原子力における水素安全対策高度化ハンドブック(第2版)」にまとめた。ハンドブックに記載した実機サイズの解析結果は解析モデル等を既存の代表的な小型、中型、大型試験による照合解析で確認した。
Brumm, S.*; Gabrielli, F.*; Sanchez Espinoza, V.*; Stakhanova, A.*; Groudev, P.*; Petrova, P.*; Vryashkova, P.*; Ou, P.*; Zhang, W.*; Malkhasyan, A.*; et al.
Annals of Nuclear Energy, 211, p.110962_1 - 110962_16, 2025/02
被引用回数：5 パーセンタイル：90.97(Nuclear Science & Technology)The completed Horizon-2020 project on "Management and Uncertainties of Severe Accidents (MUSA)" has reviewed uncertainty sources and Uncertainty Quantification methodology for the purpose of assessing Severe Accidents (SA). The key motivation of the project has been to bring the advantages of the Best Estimate Plus Uncertainty approach to the field of Severe Accident. The applications brought together a large group of participants that set out to apply uncertainty analysis (UA) within their field of SA modelling expertise, in particular reactor types, but also SA code used (ASTEC, MELCOR, etc.), uncertainty quantification tools used (DAKOTA, RAVEN, etc.), detailed accident scenarios, and in some cases SAM actions. This paper synthesizes the reactor-application work at the end of the project. Analyses of 23 partners are sorted into different categories, depending on whether their main goal is/are (i) uncertainty bands of simulation results; (ii) the understanding of dominating uncertainties in specific sub-models of the SA code; (iii) improving the understanding of specific accident scenarios, with or without the application of SAM actions; or, (iv) a demonstration of the tools used and developed, and of the capability to carry out an uncertainty analysis in the presence of the challenges faced. The partners' experiences made during the project have been evaluated and are presented as good practice recommendations. The paper ends with conclusions on the level of readiness of UA in SA modelling, on the determination of governing uncertainties, and on the analysis of SAM actions.
月森 和之; 矢田 浩基
Journal of Pressure Vessel Technology, 147, p.031901_1 - 031901_9, 2025/00
日本では、福島第一原子力発電所の事故以来、原子力プラントに対して厳しい安全対策が取られている。シビアアクシデント時において、放射性物質を内包する危機がそのバウンダリ機能を維持できるか否かが重要な関心事となる。本研究では、高速炉の1時冷却材を内包する容器等のバウンダリを構成する構造部材である鏡板とベローズに着目し、設計を超える過大な圧力を受けた場合の座屈、座屈後変形、さらにバウンダリの貫通破損(バウンダリ機能喪失)までの挙動を検討した。研究は、2013年度に始まり、段階的に進められたが、最終段階として、新たに提案する破損クライテリアの鏡板、ベローズへの適用結果を示すものである。
山田 剛司*; Li, X.; 山下 拓哉; 山路 哲史*
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 10 Pages, 2024/11
本研究では、MPS法によるMCCI対応溶融物挙動解析コードに、長期間のコンクリート浸食挙動の解析を可能とするような新たなクラストモデルを開発した。新クラストモデルでは、長時間にわたるクラスト粒子の物理的移動の累積を可能にしつつ、数値的移動の累積(数値的クリープ)を防ぐことができる。CEAで実施されたVULCANO VBS-U3実験の公開文献を参考に試解析を実施し、模擬炉心物質とコンクリート壁との境界にクラストが形成された後も継続するコンクリートの溶融浸食(アブレーション)挙動を定性的に解析できることを示した。
曽我部 丞司; 石田 真也; 田上 浩孝; 岡野 靖; 神山 健司; 小野田 雄一; 松場 賢一; 山野 秀将; 久保 重信; 久保田 龍三郎*; et al.
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
日仏協力の枠組みにおいて、タンク型ナトリウム冷却高速を対象とした過酷事故の評価手法を定義し、解析評価を実施した。
小野田 雄一; 石田 真也; 深野 義隆; 神山 健司; 山野 秀将; 久保 重信; 柴田 明裕*; Bertrand, F.*; Seiler, N.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
PIRTs have been developed and are reported for the 3 sequence event families of SFR severe accidents. For ULOF, there are 13 phenomena ranked with high importance and large uncertainty. Two PIRTs for primary phase of UTOP have been developed based on those of ULOF. Two phenomena with high importance and large uncertainty both in FRN and JPN ranking are highlighted. For USAF PIRT, they are eight phenomena ranked important and uncertain by both sides related to heat transfer coefficient, chunk relocation in the molten pool of the initiating SA and to thermomechanical loading on the hexcan of the initiating SA. These phenomena are recognized to deserve priority study. The event progression regarding FP transport focusing on phenomena of ULOF is investigated. Seven phenomenological phases were identified along with the accident sequences and of their events progression. The summary of the elementary phenomena on this PIRT, and the vote for the table are foreseen in the future study.
at temperature range 170 - 290
CLuu, V. N.; 中島 邦久
Nuclear Engineering and Design, 426, p.113402_1 - 113402_7, 2024/09
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)A field assessment at the Fukushima-Daiichi Nuclear Power Station revealed high radioactivity on the concrete shield plugs, which is estimated above 20 PBq for Cs-137 at units 2 and 3. This leads to significant interest in the retention of Cs on concrete during severe accidents (SA). However, the interaction of CsOH, as one of the main Cs forms released in SA, with concrete surfaces at elevated temperatures remains poorly researched. In this study, we have experimentally investigated the deposition behavior of CsOH on CaCO, which is the primary phase existing on the surface of concrete, under humid atmosphere. As a result, the chemical reaction enhanced deposition rate (N), and increased linearly with CsOH concentration (C
), as following expression: N(
g/cm
s) = v
C, where v is temperature-dependent deposition velocity as given by ln v (cm/s) = -3785.8/T + 3.766, for T in the range of 170 and 290 C. This empirical model can be integrated into severe accident codes to quantify the chemical trapping of cesium on concrete surfaces during ex-vessel release. Moreover, it can contribute to understanding the reasons behind the high dose rate on concrete shield plugs at the Fukushima Daiichi Nuclear power stations and aid in developing effective decommissioning practices for concrete structures.
松本 俊慶; 日引 俊*; 丸山 結
International Journal of Energy Research, 2024(1), p.9748588_1 - 9748588_18, 2024/08
被引用回数：0 パーセンタイル：0.00(Energy & Fuels)ウェットキャビティ戦略(格納容器内への事前注水方策)の有効性を評価するために、圧力容器から放出される溶融物条件の不確かさを考慮した確率論的な評価手法を構築した。第1段階では、MELCORコードにより溶融物条件を求めた。炉心溶融進展に関係する5つの不確かさパラメータが選択された。インプットパラメータのセットはラテン超方格サンプリングにより発生された。第2段階ではJASMINEコードにより溶融物挙動が解析された。JASMINE解析のパラメータの確率分布はMELCOR解析の結果から決定された。初期水位は0.5、1.0、2.0mに設定された。デブリの高さが冷却性判定のために基準と比較された。一連の計算の結果としてデブリの冷却確率が取得された。さらにMELCOR-JASMINEを組み合わせた解析手法の成立性と技術的な課題について議論された。
中村 勇気*; 小島 良洋*; 山下 拓哉; 下村 健太; 溝上 伸也
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
At the Fukushima Daiichi Nuclear Power Plant accident, it has been reported that several units of containment vessel had failed, and large quantity of radionuclides had been released into the environment. However, the detailed accident progression of such a containment failure, which includes core melt, reactor vessel failure and following containment vessel behavior, has still large uncertainties. Especially for the unit 2 and 3, they had succeeded in the initial core cooling, but at last lost their cooling system and fell into severe accident to release the fission product into the environment. Nowadays, several information has been obtained by the internal inspection into the containment of the Fukushima Daiichi Nuclear Power Plants. To clarify the uncertainties in the accident scenario, considering the information and several insights already accustomed by previous research, the latest accident scenario in unit 2 and unit 3 of the Fukushima Daiichi Nuclear Power Plants accident are suggested and tested by the severe accident analysis code, MAAP in this study. It is shown that unit 2 and 3 both accident scenario would have resulted in the thermal stratification in suppression pool which encouraged the containment pressure response in the early phase of the accident. In addition, containment vessel leakage would have occurred and affected the containment depressurization.
石田 真也; 内堀 昭寛; 岡野 靖
第28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), 4 Pages, 2024/06
本研究では、炉心損傷事故の起因過程から遷移過程までの一貫解析も可能な炉心損傷挙動評価モジュールの開発を行い、ナトリウム冷却高速炉のシビアアクシデント時の原子炉全体の挙動を一貫して評価する解析コードSPECTRAに導入した。本モジュールを含むSPECTRAの統合的な妥当性確認の一環として、混合酸化物(MOX)燃料炉心における炉心流量喪失時原子炉停止機能喪失事象(ULOF)を対象とした解析を実施し、冷却材の沸騰から燃料ピンの破損、損傷領域の拡大に至るまでの高速炉の炉心損傷事故を評価するための機能がSPECTRAに備わっていることを確認した。
石田 真也; 深野 義隆; 飛田 吉春; 岡野 靖
Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05
被引用回数：1 パーセンタイル：27.70(Nuclear Science & Technology)To improve the safety of future SFRs, the development of SFRs with low void reactivity has been promoted. Small SFRs can have a negative void coefficient of reactivity, so the analysis of the CDA event sequence in small SFRs is valuable for the investigation of the reactor characteristics for the future research and development of SFRs. In this study, the typical initiating events of a CDA in small SFRs were evaluated with the computational code, SAS4A. The event progression of ULOF and UTOP in the low void reactivity reactor is found to be slow due to the effective operation of the negative reactivity feedback and the absence of significant positive reactivity insertion. No power excursion occurs in the initiating phase. In ULOF, the cladding melt and relocation behavior becomes more important for the evaluation of the event progression due to its positive reactivity.
Li, X.; 山路 哲史*; 佐藤 一憲*; 山下 拓哉; 永江 勇二
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 12 Pages, 2024/05
For Fukushima Daiichi Nuclear Power Station (1F) Unit-2, the muon radiography investigation results indicate that the fuel debris are largely retained inside the RPV. The current study focuses on the analysis of metallic melt penetration behavior in the CRD housing with Moving Particle Semi-implicit (MPS) method. A three-dimensional CRD housing model with simplified inner structures was established. The injection of SS-Zircaloy eutectic melt into the CRD housing was simulated and its downstream penetration and freezing behavior under vertically varying temperature boundary conditions was analyzed. It is found that the melt would start to freeze and form channel blockages soon after it enters the region with a relatively cold boundary in the downstream.
Li, N.*; Sun, Y.*; 中島 邦久; 黒崎 健*
Journal of Nuclear Science and Technology, 61(3), p.343 - 353, 2024/03
被引用回数：1 パーセンタイル：27.70(Nuclear Science & Technology)福島原子力発電所(1F)事故では、表面積の大きなステンレス鋼(SS304)製の気水分離器や蒸気乾燥器にセシウムが大量に残っている可能性がある。そして、1F廃止措置においてこのようなCsは、放射性粉塵を生成する可能性があるため、安全上問題になることが予想される。しかし、水酸化セシウム(CsOH)の化学吸着により生成した酸化被膜の付着強度については、まだ、明らかになっていない。本研究では、CsOHによる化学吸着がどの程度酸化被膜の付着強度に影響するかスクラッチ試験機を用いて調査した。その結果、CsOHの化学吸着により酸化被膜の付着強度は低下したが、剥離させることはできなかった。
佐藤 一憲; 吉川 信治; 山下 拓哉; 下村 健太; Cibula, M.*; 溝上 伸也*
Nuclear Engineering and Design, 414, p.112574_1 - 112574_20, 2023/12
Based on the updated knowledge from plant-internal investigations, experiments and computer-model simulations until now, the in-vessel phase of Fukushima-Daiichi Nuclear Power Station Unit 3 was analyzed using the MAAP code. In Unit 3, it is considered that ca. 40 percent of UO
fuel was molten when core materials relocated to the lower plenum of the reactor pressure vessel. Initially relocated molten materials would have been fragmented by mixing with liquid water, while solid materials would have relocated later on. With this two-step relocation, debris in the lower plenum seems to have been permeable for coolant, thus debris seems to have been once cooled down effectively. Although the present MAAP analysis seems to slightly underestimate core-material oxidation during the relocation period, this probable underestimation was compensated for by an existing study that was considered more reliable, so that more realistic debris conditions in the lower plenum could be obtained. Probable debris reheat-up behavior was evaluated based on interpretation of the pressure data. This evaluation predicted that the fuel debris in the lower plenum was basically in solid-phase at the time when it relocated to the pedestal. With this study, basic validity of the former prediction of the Unit 3 accident progression behavior was confirmed, and detailed boundary conditions for future studies addressing the later phases were provided.
勝村 庸介*; 高木 純一*; 細見 憲治*; 宮原 直哉*; 駒 義和; 井元 純平; 唐澤 英年; 三輪 周平; 塩津 弘之; 日高 昭秀*; et al.
日本原子力学会誌ATOMO, 65(11), p.674 - 679, 2023/11
本委員会では、東京電力ホールディングス株式会社(東電)福島第一原子力発電所(1F)事故後の 核分裂生成物(FP)挙動を予測可能な技術に高めて廃炉作業に貢献することと、1F事故進展事象の把握で得られた情報をソースターム(ST)の予測技術の向上に反映させ、原子炉安全の一層の向上に繋げることを目標とした活動を実施している。この2年間では、これまでの12年間の1F実機調査や1F関連研究で得られた情報を調査し、1F廃炉における燃料デブリやFP挙動の予測、及びST予測精度向上に必要な課題として「FPの量・物質収支と化学形態」「サンプリング目的とデータ活用」「環境への移行経路」を摘出した。今後、これらの課題の解決に向けた道筋の議論を進める。
丸山 結; 杉山 智之*; 島田 亜佐子; Lind, T.*; Bentaib, A.*; Sogalla, M.*; Pellegrini, M.*; Albright, L.*; Clayton, D.*
Proceedings of 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20) (Internet), p.4782 - 4795, 2023/08
The Analysis of Information from Reactor Buildings and Containment Vessels of Fukushima Daiichi Nuclear Power Station (FDNPS) (ARC-F) project was initiated in January 2019 for three years with 22 signatories from 12 countries. Three main tasks were implemented in the ARC-F project, which were relevant to 1) refinement of analysis for accident scenarios and associated fission product (FP) transport and dispersion, 2) compilation and management of data and information, and 3) discussion for the next-phase project. Various activities were performed in Task 1, covering improvement of analysis for accident scenarios, and in-depth analyses for specific phenomena such as in-vessel melt progression, molten core/concrete interaction, FP transport and source term, hydrogen combustion and atmospheric dispersion of FPs. Through these studies, analyses for accident scenarios with severe accident codes were refined and important phenomena with large uncertainties were clarified. In order to share well selected and organized information from the FDNPS with the project partners, two databases, information source database and sample database, were built under Task 2. The analysis techniques including the separation of iodine species were developed also in Task 2 and applied to the analysis of FPs in several samples taken from the FDNPS. The next-phase project was discussed in Task 3, resulting in launching the Fukushima Daiichi Nuclear Power Station Information Collection and Evaluation (FACE) project. The FACE project officially started in July 2022 with the participation of 23 organizations from 12 countries and the European Commission.
南上 光太郎; 塩津 弘之; 丸山 結; 杉山 智之; 岡本 孝司*
Journal of Nuclear Science and Technology, 60(7), p.816 - 823, 2023/07
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)For proper source term evaluation, we constructed the theoretical model to estimate the mass transfer coefficient of gaseous iodine species under two-phase flow conditions, which complicates the direct experimental measurements. The mass transfer speed is determined by the product of the overall mass transfer coefficient and the interfacial area. By using the ratio of two gas components, the interfacial area, which is an important parameter that is difficult to measure, can be canceled out and the ratio of their overall mass transfer coefficients can be obtained. This ratio is expected to be equal to the ratio of their diffusion coefficients. Therefore, the unknown mass transfer coefficient such as iodine species can be estimated using the diffusion coefficients of two gas components and the reference mass transfer coefficient such as O. We carried out the experiments using the bubble column to confirm this relationship. From the results in this study, we confirmed that the ratio of the overall mass transfer coefficient was in good agreement with the ratio of diffusion coefficient under the bubbly flow conditions. Using this relationship confirmed in this study, we estimated the mass transfer coefficient of I, one of the iodine species.
山下 拓哉; 本多 剛*; 溝上 暢人*; 野崎 謙一朗*; 鈴木 博之*; Pellegrini, M.*; 酒井 健*; 佐藤 一憲; 溝上 伸也*
Nuclear Technology, 209(6), p.902 - 927, 2023/06
被引用回数：5 パーセンタイル：82.11(Nuclear Science & Technology)The estimation and understanding of the state of fuel debris and fission products inside the plant is an essential step in the decommissioning of the TEPCO Fukushima Daiichi Nuclear Power Station (1F). However, the direct observation of the plant interior, which is under a high radiation environment, is difficult and limited. Therefore, in order to understand the plant interior conditions, a comprehensive analysis and evaluation is necessary, based on various measurement data from the plant, analysis of plant data during the accident progression phase and information obtained from computer simulations for this phase. These evaluations can be used to estimate the conditions of the interior of the reactor pressure vessel (RPV) and the primary containment vessel (PCV). Herein, 1F Unit 3 was addressed as the subject to produce an estimated diagram of the fuel debris distribution from data obtained about the RPV and PCV based on the comprehensive evaluation of various measurement data and information obtained from the accident progression analysis, which were released to the public in November 2022.
Li, C.-Y.; 渡部 晃*; 内堀 昭寛; 岡野 靖
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Quantitative assessment of the effect of accident management on the various external hazards is essential in the nuclear safety analysis. This study aims to establish the dynamic probabilistic risk assessment methodology for sodium-cooled fast reactors that can consider the transient plant status under continuous external hazards with corresponding countermeasures operating stochastically. Specifically, the Continuous Markov chain Monte Carlo (CMMC) and Deterministic and Stochastic Petri Nets (DSPN) methods are newly applied to the severe accident analysis code, SPECTRA, which can conduct dynamic plant evaluation in the different severe accident conditions of nuclear reactors, to develop an evaluation methodology for typical external hazards. In the DSPN-CMMC-SPECTRA coupled frame, the latest safety functions of the plant components/systems can be stochastically determined by the DSPN-CMMC grounded on the current plant states under continuous hazard and the interaction between the multi-state components/systems; then, SPECTRA can evaluate the following plant state determined by the latest safety function of the components/systems. Therefore, the advantage of this newly developed DSPN-CMMC-SPECTRA frame is having the capability to quantitatively and stochastically evaluate the transient accident progressions that potentially lead to the core damage under the continuous external hazard scenario. As for the preliminary exam on the DSPN-CMMC-SPECTRA frame, one of the typical external hazards of continuous volcanic ashfall is selected in this research. In addition, the numerical investigation of alternative accident management' effects has also been carried out and quantitatively confirmed in this research.
