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Li, C.-Y.; Wang, K.*; 内堀 昭寛; 岡野 靖; Pellegrini, M.*; Erkan, N.*; 高田 孝*; 岡本 孝司*
Applied Sciences (Internet), 13(13), p.7705_1 - 7705_29, 2023/07
被引用回数：1 パーセンタイル：56.82(Chemistry, Multidisciplinary)For a sodium-cooled fast reactor, the capability for stable cooling and avoiding re-criticality on the debris bed is essential for achieving in-vessel retention when severe accidents occur. However, an unexploited uncertainty still existed regarding the compound effect of the heterogeneous configuration and dynamic particle redistribution for the debris bed's criticality and cooling safety assessment. Therefore, this research aims to develop a numerical tool for investigating the effects of the different transformations of the heterogeneous configurations on the debris bed's criticality/cooling assessment. Based on the newly proposed methodology in this research, via integrating the Discrete Element Method (DEM) with Computational Fluid Dynamics (CFD) and Monte-Carlo-based Neutronics (MCN), the coupled CFD-DEM-MCN solver was constructed with the originally created interface to integrate two existing codes. The effects of the different bed configurations' transformations on the bed safety assessments were also quantitively confirmed, indicating that the effect of the particle-centralized fissile material had the dominant negative effect on the safety margin of avoiding re-criticality and particle re-melting accidents and had a more evident impact than the net bed-centralized effect. This coupled solver can serve to further assess the debris bed's safety via a multi-physics simulation approach, leading to safer SFR design concepts.
上澤 伸一郎; 山下 晋; 柴田 光彦; 吉田 啓之
Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 8 Pages, 2022/10
Various types of cooling of fuel debris such as reduction in the amount of water injection, intermittent injection water and air cooling are considered for contaminated water management in the decommissioning of Fukushima Daiichi Nuclear Power Station (NPS). As a method for estimating the thermal behavior of fuel debris in the preliminary containment vessel (PCV) of Fukushima Daiichi NPS, JAEA develops a method for estimating the thermal behavior in the air cooling, including the influence of the position, heat generation, and the porosity of fuel debris by JUPITER added a porous model. In order to validate the natural convective heat transfer analysis with porous bodies by the JUPITER added the porous model, the heat transfer and the flow visualization experiments of air natural convection with a porous body was performed in this study. The temperature distributions of the heat transfer surface, the natural convection area and the porous body were measured with thermocouples. In addition, the flow visualization was performed by a particle image velocimetry (PIV). We compared the temperature and flow results between the experiment and the numerical analysis. The temperature distributions at the central axis were generally agreed on the condition without the porous body. As for with the porous body, the tendency to increase the temperature inside the porous body was reproduced as in the analysis, but quantitative differences were observed. For the flow distribution of the test vessel, flow along the experimental vessel wall to form a vortex was observed in the experiment. On the other hand, an upward flow occurred from the center of the heating surface, and the flow was separated by the ceiling to form two vortices in the numerical analysis. The difference between the analysis and the experiment indicates that discussions for various thermal conductivity models to porous bodies is necessary.
佐藤 一憲
Nuclear Engineering and Design, 383, p.111426_1 - 111426_19, 2021/11
被引用回数：5 パーセンタイル：64.12(Nuclear Science & Technology)The D/W (Drywell) and S/C (Suppression Chamber) pressure data of Fukushima-Daiichi Nuclear Power Plant Unit 3 was analyzed in depth. This analysis provided valuable information related to the accident progression behavior on one hand, and gave a hint for understanding of the debris-to-coolant heat transfer when fuel debris relocated to the pedestal on the other hand. In this unit, the D/W and S/C pressure increased and decreased cyclically with a relationship, which seems to have been dependent on the composition of vapor and non-condensable gases in the S/C cover gas region. Based on this characteristic, the vapor pressure in the S/C cover gas region was evaluated for two pressure decrease cycles during and after the expected debris relocation to the pedestal respectively. This evaluation allowed an understanding that the S/C vapor pressure increased due to the heat transfer from the debris relocated to the pedestal.
佐藤 一憲; 荒井 雄太*; 吉川 信治
Journal of Nuclear Science and Technology, 58(4), p.434 - 460, 2021/04
被引用回数：6 パーセンタイル：70.8(Nuclear Science & Technology)The vapor formation within the reactor pressure vessel (RPV) is regarded to represent heat removal from core materials to the coolant, while the hydrogen generation within the RPV is regarded to represent heat generation by metal oxidation. Based on this understanding, the history of the vapor/hydrogen generation in the in-vessel phase of Fukushima Daiichi Nuclear Power Station Unit 3 was evaluated based on the comparison of the observed pressure data and the GOTHIC code analysis results. The resultant vapor/hydrogen generation histories were then converted to heat removal by coolant and heat generation by oxidation. The effects of the decay power and the heat transfer to the structures on the core material energy were also evaluated. The core materials are suggested to be significantly cooled by water within the RPV, especially when the core materials are relocated to the lower plenum.
松本 俊慶; 岩澤 譲; 安島 航平*; 杉山 智之
Proceedings of Asian Symposium on Risk Assessment and Management 2020 (ASRAM 2020) (Internet), 10 Pages, 2020/11
本研究では、事前注水した格納容器内デブリの冷却確率を評価した。まず、落下溶融物条件を求めるため、シビアアクシデント解析コードMELCORによる不確かさ解析を行った。この解析では炉心の溶融・移行過程に関連する5つの不確かさパラメータを選択し、仮定された確率分布を用いて、ラテン超方格法(LHS)により入力パラメータセットを生成した。これを用いたMELCORによる多ケース解析の結果から落下溶融物条件を抽出した。次に、MELCOR解析結果をもとに、パラメータの確率分布を決定し、LHSにより生成した59個のパラメータセットを用いてJASMINEコードによる水中の溶融物挙動の解析を行った。水位の条件は0.5m, 1.0m及び2.0mとした。広がり半径とデブリ質量の解析結果からデブリの堆積高さを求め、判定基準と比較することで冷却の成否判定を行った。以上の一連の解析の結果、デブリ冷却の成功確率を求めた。また、MELCOR及びJASMINEを組み合わせた冷却性解析の課題について論じた。
堀田 亮年*; 秋葉 美幸*; 森田 彰伸*; Konovalenko, A.*; Vilanueva, W.*; Bechta, S.*; Komlev, A.*; Thakre, S.*; Hoseyni, S. M.*; Sk
ld, P.*; et al.
Journal of Nuclear Science and Technology, 57(4), p.353 - 369, 2020/04
被引用回数：14 パーセンタイル：71.27(Nuclear Science & Technology)Key phenomena in the cooling states of debris beds under wet cavity conditions were classified into several groups based on the complicated geometry, nonhomogeneous porosity and volumetric heat of debris beds. These configurations may change due to the molten jet breakup, droplet agglomeration, anisotropic melt spreading, two-phase flow in a debris bed, particle self-leveling and penetration of molten metals into a particle bed. The modular code system THERMOS was designed for evaluating the cooling states of underwater debris beds. Three additional tests, DEFOR-A, PULiMS and REMCOD were employed to validate implemented models. This paper summarizes the entire test plan and representative data trends prior to starting individual data analyses and validations of specific models that are planned to be performed in the later phases. It also tries to report research questions to be answered in future works, such as various scales of melt-coolant interactions observed in the PULiMS tests.
丸山 結*; 森山 清史; 中村 秀夫; 平野 雅司; 中島 研吾*
Journal of Nuclear Science and Technology, 40(1), p.12 - 21, 2003/01
被引用回数：6 パーセンタイル：41.7(Nuclear Science & Technology)デブリと下部ヘッドとの境界におけるギャップを満たす水のデブリ冷却効果を、実炉規模の定常計算に基づいて概略的に評価した。本計算では、狭隘流路における限界熱流束及び内部発熱流体の自然対流熱伝達に関する相関式を用い、ギャップ幅に依存する下部プレナム内冷却可能最大デブリ堆積深さを算出した。計算結果は、TMI-2事故の条件下では1mmから2mmの幅を有するギャップによりデブリを冷却可能なこと、より多量のデブリを冷却するためにはギャップ幅の大幅な増大が必要であることを示唆した。定常計算と併せて、過渡挙動の重要性を明確にするため、ギャップの成長及びギャップ内浸水に関するモデルを構築し、低レイノルズ数型二方程式モデルに分類されるYinらの乱流自然対流モデルとともにCAMPコードに組み込んだ。乱流モデルはUCLAで実施された内部発熱流体自然対流実験の解析により検証し、容器壁面における局所熱伝達係数の分布を精度良く予測することを確認した。ギャップ冷却に関するモデルの検証は、原研で実施した、水で満たした半球容器に溶融アルミナを注いだ圧力容器内デブリ冷却性実験の解析を通して行った。構築したギャップ冷却モデルを用いることにより、CAMPコードが容器壁の温度履歴を再現可能であることを明らかにした。
上澤 伸一郎; 小野 綾子; 山下 晋; 吉田 啓之
no journal, ,
多孔質体と考えられている、東京電力福島第一原子力発電所のPCV内燃料デブリの熱挙動を推定するため、JUPITERを用いた空冷時における燃料デブリ熱挙動解析手法の開発を進めている。前報までは、多孔質体モデルを追加したJUPITERの妥当性確認や東京電力福島第一原子力発電所2号機PCV内の予備解析結果について報告した。本報では、多孔質体モデルにおける有効熱伝導率モデルの違いがPCV内熱挙動へ与える影響について報告する。本解析結果から、モデルの違いにより燃料デブリの温度や流速分布が異なり、燃料デブリの除熱量が大きく異なることが確認された。このことから、PCV内の熱挙動を推定するためには、燃料デブリ等の多孔質体としての内部構造の把握と、それに応じたモデルの適切な選択が必要である。
