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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:0The 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.
福田 航大
Annals of Nuclear Energy, 208(1), p.110748_1 - 110748_10, 2024/12
被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)The Windscale Works criticality accident in 1970 resulted from mixing an aqueous solution with an organic solvent with different plutonium concentrations and densities. Although this accident has been studied using improved computer capabilities in recent years, a precise criticality scenario has not yet been identified. This study aims to clarify a possible criticality scenario of the accident-the time variation of reactivity and its mechanism. The accident was simulated by combining the multiphase computational fluid dynamics solver of OpenFOAM and the delta-tracking-based Monte Carlo neutron transport code Serpent2. Consequently, the periodic uneven arrangement of fluids might have caused oscillations in neutron leakage and absorption, resulting in periodic wavy reactivity changes. Furthermore, the emulsion, which was thought to be the primary cause, might not be the dominant mechanism for reactivity change, although it contributed to the criticality of the accident.
郡司 智; 荒木 祥平; 井澤 一彦; 須山 賢也
Annals of Nuclear Energy, 209, p.110783_1 - 110783_7, 2024/12
被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)燃料デブリの組成や性状は不確実であるため、臨界安全性評価に使用される計算コードや核データを検証するには臨界実験が必要である。このため、原子力機構は、臨界集合体STACYの改造を行っている。STACY更新炉の初臨界は2024年春に予定されている。本稿では、STACY更新炉の初臨界時の基本炉心構成仕様の特性について事前解析の結果を報告する。初臨界時には中性子減速条件の異なる2種類の格子板(格子間隔は1.50cmと1.27cm)を用意される。一方で、利用可能なUO燃料棒の数には制限がある。これらの実験的制約を満たす最初の臨界のための炉心構成は、計算解析によって設計された。最適減速条件に近い1.50cmピッチの格子板を備えた円柱形の炉心構成では、臨界に達するには253本の燃料棒が必要となる。1.27cmピッチの格子板については、ピッチを2倍にして2.54cmピッチの炉心構成を検討した。この場合、臨界に達するには213本の燃料棒が必要となる。さらに、燃料デブリの様態をシミュレートするために、鉄またはコンクリート模擬棒を使用した実験炉心構成についても検討した。本稿では、これらの炉心構成と炉心特性の解析結果を示す。
尾崎 裕介; 石井 英一
Geoenergy (Internet), 2(1), p.geoenergy2023-056_1 - geoenergy2023-056_11, 2024/12
本研究では、幌延深地層研究センターにおける約10年間の研究坑道の掘削時の坑内への湧水データおよびHDB-6孔で観測された水圧変化を再現解析することで、幌延深地層研究センター周辺における稚内層内部における有効透水係数を推定した。求めた有効透水係数をLandau-Lifshitz-Matheronの式により断層の透水量係数や断層における流れの次元と定量的に関連付けた。これらの結果、稚内層における有効透水係数は、透水量係数のダクタリティインデックスに対する依存性と透水量係数の次元への依存性の双方を考慮した場合に推定される透水量係数と整合的であることが示された。
大野 宏和; 石井 英一; 武田 匡樹
Geoenergy (Internet), 2(1), p.geoenergy2023-047_1 - geoenergy2023-047_10, 2024/12
Faults in some deep mudstones have poor hydraulic connectivity owing to high normal stress on the fault planes. Designing a method for modeling solute transport pathways in such faults/fractures using available data is a critical issue vis-a-vis the safety assessment of radioactive waste disposal. In this study, faults in deep siliceous mudstones with low swelling capacity are investigated using cross-hole hydraulic and tracer tests between two boreholes. These results indicate that transport pathways in faults with low hydraulic connectivity can be modeled using a highly tortuous 1D pipe flow path.
土井 大輔
International Journal of Hydrogen Energy, 91, p.1245 - 1252, 2024/11
被引用回数:0Hydrogen is a major nonmetallic impurity in the coolant of sodium-cooled fast reactors (SFRs) during normal operation. A higher hydrogen concentration than the gas-liquid equilibrium has been transiently detected in the gas space of actual SFR plants. The presence of several sodium compounds can increase hydrogen generation; however, a thorough understanding of the thermal behavior of candidate reactions is lacking. Herein, thermal analysis reveals the hydrogen release behavior of sodium hydride. Mass spectrometry indicates hydrogen generation with decreasing sample mass, indicating thermal decomposition. Detailed kinetic analysis based on master plot methods indicates that the hydrogen release reaction occurred through a mechanism involving random nucleation and growth of nuclei. Furthermore, the reaction rate was newly formulated based on a kinetic model function representing the above mechanism and the Arrhenius-type reaction rate constant comprising an activation energy of 119.0 0.8 kJ mol and a frequency factor of 1.8 10 s. These findings will enable the numerical simulation of the hydrogen release behavior in SFRs.
今泉 悠也; 神山 健司; 松場 賢一
Annals of Nuclear Energy, 206, p.110658_1 - 110658_10, 2024/10
被引用回数:1 パーセンタイル:0.00(Nuclear Science & Technology)In severe accidents of SFRs, molten core materials can discharge from the core, and the jet can impinge on the lower structure plate. After the jet impingement, fragmented discharged materials can form ring-shape solidification. A fundamental experiment was conducted to simulate the behavior. In order to simulate the behavior of solid body creation and motion, a new solid body formation model by inter-particle attraction force in particle method was developed. The advantage of the new model is that it can simulate creation, formation, and motion of solid bodies without any artificial treatment as solid bodies. The movable solid bodies by the new model have any size, shape, and number, and they are created and grown by solidification, and diminish and disappear by melting. The mechanism based on the inter-particle attraction force is common with that in real world where interatomic attraction force is the cause of solid body formation.
丸山 修平; 山本 章夫*; 遠藤 知弘*
Annals of Nuclear Energy, 205, p.110591_1 - 110591_13, 2024/09
被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)This study developed a new method for evaluating the uncertainty in reactor core/shielding characteristics attributable to the scattering angle distribution, employing a random sampling (RS) technique integrated with continuous energy Monte Carlo (CEMC) calculations. The impact of neutron scattering angle is not negligible in the analysis of fast reactor cores and shielding. Recent advancements have enabled the high-accuracy assessment of nuclear data-induced uncertainty by merging CEMC calculations and the RS technique. Nonetheless, a method to quantify uncertainty due to scattering angle distribution remains unestablished. This study introduces a new approach for uncertainty quantification related to scattering angle distribution in CEMC-RS, utilizing the maximum entropy method. The effectiveness of this method was verified through comparison with results from the classical deterministic uncertainty quantification approach based on generalized perturbation theory. Overall, this method offers a more accurate tool for nuclear engineers and researchers in evaluating and managing uncertainties in reactor design and safety analysis.
植木 太郎
Progress in Nuclear Energy, 173, p.105236_1 - 105236_10, 2024/08
被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)本論文は、不完全乱雑化ワイエルシュトラス関数(IRWF)によってモデル化された燃料デブリの乱雑化レプリカに対して、中性子実効増倍率(keff)計算の極値評価を効率化したことを報告する。対象とした効率化手法は、有界増幅(BA)として特徴づけられるものである。数値計算結果により、BAのIRWFへの適用が、必要とされる乱雑化レプリカ数を、少なくとも1桁減少させることが分かった。この効率性向上検証のため、BAを適用せずに計算されたkeffのデータセットに一般化極値(GEV)統計を適用し、極値分布がワイブル分布に従うことを明らかにした。GEV理論はワイブル分布の上限値の存在を保証しており、実際に計算された上限は、BAが適用されて乱雑化レプリカ数を1桁以上減らした場合に得られる上位2つのkeff値よりも小さいことが判明した。これは、BAによる効率性向上がGEV解析によって確認されたことを意味する。
安部 諭; 柴本 泰照
Annals of Nuclear Energy, 202, p.110461_1 - 110461_16, 2024/07
被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)During a severe accident in a nuclear containment vessel, jets released from the primary system exhibit complex thermohydraulic behavior due to buoyancy effects and impingement on internal obstacles such as inner walls and floors. Thus, the obstacle-influenced jets are of interest in recent research activities. This paper describes an experimental investigation of the behavior of jets passing through a grid-type obstacle. The flow field was acquired by a particle image velocimetry system. The experiment captured the jet fragmentation by the grid-type obstacle and their recoupling. The mean velocity field obtained by postprocessing indicates a "Rectifying effect," with the axial velocity increasing at the center and the magnitude of the radial velocity decreasing. The meandering flow was suppressed due to this effect. In the near grid-obstacle region, the axial turbulence intensity was relatively large at the edge of each fragmented region due to shear stress. Moreover, the spatial distribution of the radial turbulence fluctuation became more complex. Further investigation is required to clarify the budget of the transport equation for turbulence fluctuation. The experimental data shown in this paper is useful for computational fluid dynamics validation.
中原 将海; 渡部 創; 木村 修也; 佐々木 美咲*; 稲垣 博光*; 森口 哲次*
Progress in Nuclear Energy, 172, p.105195_1 - 105195_8, 2024/07
被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)原子力施設の廃止措置に伴い発生する二次廃棄物の低減を目指したウルトラファインバブルを用いた新しい固体汚染物質の除去技術を提案している。非放射性物質であるCo酸化物の模擬汚染物を用いた除去試験により除去技術の性能評価を実施した。また、放射性物質で汚染された燃料ピン端栓部を使用した除染試験をホットセルで実施し、異なる化学形の違いによる影響を調べた。
Nguyen, B. V. C.*; 村上 健太*; Chena, L.*; Phongsakorn, P. T.*; Chen, X.*; 橋本 貴司; Hwang, T.*; 古澤 彰憲; 鈴木 達也*
Nuclear Materials and Energy (Internet), 39, p.101639_1 - 101639_9, 2024/06
被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)In reactor pressure vessel materials, the formation of Mn- and Ni-rich nanoclusters is a major cause of neutron irradiation embrittlement. The segregation of these solute atoms into dislocation loops has attracted attention as a mechanism to accelerate solute clustering. In this study, the behaviors of solute Mn and Ni atoms in Fe-0.6wt.%Ni, Fe-1.4wt.%Mn, and Fe-1.4wt.%Mn-0.6wt.%Ni alloys irradiated at 400 C up to 3 dpa were analyzed using three-dimensional atom probe tomography. Solute atom clusters were observed in all materials, and their shapes were spherical, flat, and torus in FeNi, FeMn, and FeMnNi, respectively. In ternary alloy FeMnNi, Mn and Ni atoms were concentrated in the sample in the form of arcs, and the orientation of the plane containing the arcs was estimated by comparing field desorption images. The size, number density, and orientation of this structure were found to be in good agreement with those of both types of dislocation loops, namely, b = 1/2 111 and b = 100, identified in a previous study using the same material. The positions of Ni and Mn enrichment did not fully overlap. Ni atoms tended to be concentrated more in the inner part of the loop than the Mn atoms. Mn atoms were enriched only in the vicinity of the dislocation loops, whereas Ni atoms showed a higher concentration inside the dislocation loops than in the bulk.
Hong, Z.*; Ahmed, Z.*; Pellegrini, M.*; 山野 秀将; Erkan, N.*; Sharma, A. K.*; 岡本 孝司*
Progress in Nuclear Energy, 171, p.105160_1 - 105160_13, 2024/06
被引用回数:2 パーセンタイル:94.57(Nuclear Science & Technology)本研究では、BC粉末とステンレス鋼(SS)間との共晶反応はBC粉ペレットとSS間とのそれより相当に速いことが分かった。粉末及びペレットに対して導出された反応速度定数は参考文献値によく一致している。また、粉末とペレットの場合の詳細微細構造をSEM/EDSを用いて組成分析を行った。粉末の場合、(Fe,Cr)Bからなる反応層として厚肉層が見られた。一方、ペレットの場合、2つの反応層が見られた。
岡垣 百合亜; 日引 俊詞*; 柴本 泰照
International Journal of Energy Research, 2024, p.5114542_1 - 5114542_37, 2024/04
被引用回数:0 パーセンタイル:0.00(Energy & Fuels)In pressurized water reactor accident scenarios, the injection of water from the ECCS (ECC injection) might induce a PTS, affecting the RPV integrity. Therefore, PTS is a vital research issue in reactor safety, and its analysis is essential for evaluating the integrity of RPVs, which determines the reactor life. The PTS analysis comprises a coupled analysis between thermal-hydraulic and structural analysis. The thermal-hydraulic approach is particularly crucial, and reliable Computational Fluid Dynamics (CFD) simulations should play a vital role in the future because predicting the temperature gradient of the RPV wall requires data on the transient temperature distribution of the downcomer. Since one-dimensional codes cannot predict the complex three-dimensional flow features during ECC injection, PTS is one reactor safety issue where CFD can benefit from complement evaluations with thermal-hydraulic system analysis codes. This study reviewed the code validation efforts for turbulence models most affecting PTS analysis based on papers published since 2010 on single- and two-phase flow CFD analysis for the experiment on PTS performed in the ROCOM, TOPFLOW, UPTF, and LSTF. The results revealed that in single-phase flow CFD analysis, where knowledge and experience are sufficient, various turbulence models have been considered, and many analyses using LES have been reported. For two-phase flow analysis of air-water conditions, interface capturing/tracking methods were used in addition to two-fluid models. The standard k- and SST k- models were still in the validated phase, and various turbulence models have yet to be fully validated. In the two-phase flow analysis of steam-water conditions, many studies have used two-fluid models and RANS, and NEPTUNE_CFD, in particular, has been reported to show excellent prediction performance based on years of accumulated validation.
廣瀬 意育; 安部 諭; 石垣 将宏*; 柴本 泰照; 日引 俊*
Progress in Nuclear Energy, 169, p.105085_1 - 105085_13, 2024/04
被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)Immersed boundary methods (IBMs) have been developed as complementary methods for computational fluid dynamics (CFD). They allow a flow simulation in a mock-up model that includes complex-shaped inner structures and/or boundaries with a non-body conformal mesh. Such a model might force us to create a complicated body-fitted mesh with a high cost in the conventional CFD (CCFD) approach. We focus on the Brinkman penalization (BP) method and its extended version, which we call here the extended Brinkman penalization method (EBP), among the different types of IBMs, aiming to apply them to the phenomena that occur during severe accidents in a nuclear reactor containment vessel and explore the possibility that the methods can partially replace the CCFD. In this paper, as a preliminary step to validate the applicability of these methods, we measure the jet flow rectified by a grating-type structure used for the validation of numerical techniques and apply them to simulate the behavior of an upward jet rectified by a horizontally placed grating-type structure modeled as an immersed body. This type of structure is generally used in reactor buildings, and it is crucial to evaluate their influence on gaseous flows because the behaviors of hydrogen produced during severe accidents may be influenced by them. The structure is selected as our subject because it has moderate complexity, enabling us to examine the effects of the IBMs and compare them with CCFD. We investigate whether these methods can reproduce a result of corresponding CCFD in which the grating is modeled as body-conformal mesh and show that the former can produce the latter with equivalent accuracy. All these results are also compared with the experimental data on the flow velocity distributions downstream of the grating measured using particle image velocimetry.
嶋田 和真; 櫻原 達也*; Farshadmanesh, P.*; Reihani, S.*; Mohagehgh, Z.*
Annals of Nuclear Energy, 197, p.110243_1 - 110243_12, 2024/03
被引用回数:0 パーセンタイル:0.00(Nuclear Science & Technology)本研究は、原子力発電所に対するレベル3確率論的リスク評価(PRA)において住民の避難行動を設定する際の主観的な専門家判断を回避したレベル3PRA手法を開発する。そのために、交通シミュレーションコードMATSimで出力した避難速度をレベル3PRAコードMACCSに入力した。さらに、道路封鎖を検討する箇所の優先順位を設定するために、自然災害リスク評価コードHAZUSを用いて地震による道路封鎖リスクを評価する手法を開発した。そして、米国原子力規制委員会が実施した最先端の原子炉事故影響研究において採用されたSequoyah原子力発電所のケーススタディに対して、住民の避難経路と放射線被ばく線量の関係を評価した。その結果、地震封鎖リスクは小さいが、封鎖されると住民の被ばく線量が増加する避難経路を見出した。この結果は、提案するレベル3PRA手法が避難経路を強化する意思決定を支援することを示した。
廣瀬 意育; 柴本 泰照; 日引 俊*
Progress in Nuclear Energy, 168, p.105027_1 - 105027_17, 2024/03
被引用回数:1 パーセンタイル:41.04(Nuclear Science & Technology)This study reviewed the literature that measured critical heat flux (CHF) for downward flow in round pipes and arranged the proposed correlations. Each correlation shows relatively good prediction accuracy for experimental data from their literature, but the accuracies sometimes decrease for experimental data from other literature. No correlation accurately predicts all the experimental data of the literature, indicating an issue in extrapolating existing correlations. Therefore, we developed a correlation that can accurately predict the experimental data of the collected literature. First, we used a neural network to select the essential dimensionless quantities that comprise the correlation. Then, we regarded the prediction accuracy when all candidate dimensionless quantities extracted from the literature were used for the input variables of the network as the achievable limit prediction accuracy and searched for the minimum combination of dimensionless quantities required to achieve it. The results showed that only the dimensionless mass flux and the ratio of the heating length to the channel diameter are the essential parameters to achieve it. We developed a correlation equation using these two dimensionless quantities and achieved 17.6% of the average prediction accuracy. This result considerably improved existing correlation equations with 25%-40% average prediction accuracy for the same experimental data.
吉本 将隆*; 田村 和久; 渡邊 健太*; 清水 啓佑*; 堀澤 侑平*; Kobayashi, Takeshi*; Tsurita, Hanae*; 鈴木 耕太*; 菅野 了次*; 平山 雅章*
Sustainable Energy & Fuels (Internet), 8(6), p.1236 - 1244, 2024/03
被引用回数:0 パーセンタイル:0.00(Chemistry, Physical)1つのデバイスで、効率的に太陽光エネルギーを化学エネルギーに変換可能である光再充電システムは、太陽光を効率的に使用するのに重要である。光-(デ)インターカレーションは、光再充電システムの機能において重要な役割を果たしている。しかしながら、光-(デ)インターカレーション過程は、電解液の分解や電極材料の溶出などの副反応などのため、いまだ十分に理解されていない。本研究では、エピタキシャル成長させて作成したNbドープのアナターゼTiO薄膜から構成される薄膜全固体電池を用いて光応答Liデインターカレーションに成功したことを示す。光照射下、放電時は、Liのデインターカレーションが起き、引き続いて、可逆的にTiOにインターカレーションする。さらに、光照射下では、充電容量の一部はキャパシタと同様、インターカレーションに寄らない電子移動に基づいていることがわかった。
Sahboun, N. F.; 松本 俊慶; 岩澤 譲; Wang, Z.; 杉山 智之
Annals of Nuclear Energy, 195, p.110145_1 - 110145_12, 2024/01
被引用回数:1 パーセンタイル:41.04(Nuclear Science & Technology)Relocated corium into the Primary Containment Vessel needs to be properly cooled to avoid or mitigate molten core concrete interactions in the PCV in order to maintain its supporting capability for the reactor pressure vessel and to suppress combustible or non-condensable gas releases. To know how effective the cooling is, it became important to know the geometry of the relocated corium. The present study chooses to focus on the "Wet Cavity" strategy and to build a reliable tool to evaluate the corium coolability in such a case. To achieve this goal, a previously developed formulation built to predict the corium geometry under the "Dry Cavity" strategy was extended to the conditions used in the "Wet Cavity" strategy. This extension includes the effect of solidification and cooling from the water by using a newly developed expression for the dimensionless thickness s, the water subcooling, and the melts super heat. After the validation of the extended formulation was confirmed, potential restrictions and limitations were investigated.
浜瀬 枝里菜; 大釜 和也; 河村 拓己*; 堂田 哲広; 田中 正暁; 山野 秀将
Annals of Nuclear Energy, 195, p.110157_1 - 110157_14, 2024/01
被引用回数:1 パーセンタイル:41.04(Nuclear Science & Technology)高速炉プラント動特性解析コードSuper-COPDのスクラム不作動流量喪失事象に対する妥当性確認のため、FFTFの受動的安全性試験LOFWOS No.13試験を対象としたIAEAベンチマークに参加した。ブラインドフェーズで課題として抽出された燃料集合体出口温度及び全反応度の評価精度向上のため、集合体間熱移行及び集合体間ギャップ部流れを考慮した全炉心モデル及び炉心湾曲反応度簡易評価モデルを導入した。最終フェーズ解析の結果、2次ピーク時の集合体出口温度を良好に再現するとともに、全反応度の実測値の挙動を概ね評価できたことから、LOFWOSに対するSuper-COPDの妥当性を確認した。