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福田 航大
Annals of Nuclear Energy, 208(1), p.110748_1 - 110748_10, 2024/12
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.
今泉 悠也; 神山 健司; 松場 賢一
Annals of Nuclear Energy, 206, p.110658_1 - 110658_10, 2024/10
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.
Vauchy, R.; 廣岡 瞬; 堀井 雄太; 小笠原 誠洋*; 砂押 剛雄*; 山田 忠久*; 田村 哲也*; 村上 龍敏
Journal of Nuclear Materials, 599, p.155233_1 - 155233_11, 2024/10
U
Pu
O
(y=0.30および0.45)およびPuOにおける蛍石の溶出/再結合は、示差走査熱量測定を使用して調査された。結果は、プルトニアを除いて、文献データと比較的よく一致している。我々の値は、Pu-Oの混和ギャップの臨界温度が以前に報告されたものより30
50K低いことを示している。最後に、体系的な実験手順により、低化学量論的U
0Pu
O、U
Pu
O、およびPuO二酸化物に存在するソルバスの軌跡を精密化することができた。
外山 健*; 丹野 敬嗣; 矢野 康英; 井上 耕治*; 永井 康介*; 大塚 智史; 宮澤 健; 光原 昌寿*; 中島 英治*; 大沼 正人*; et al.
Journal of Nuclear Materials, 599, p.155252_1 - 155252_14, 2024/10
高速実験炉「常陽」で中性子照射した14Cr-ODS鋼(MA957)中の酸化物の安定性について3D-APとTEMを用いて評価を行った。中性子照射は、(502
C, 130dpa)、(589C, 154dpa)及び(709C, 158dpa)の3条件で実施した。709C照射では僅かな数密度の減少が認められたが、酸化物は高い数密度を有しており、相対的に照射前後で顕著な変化は確認されず安定に存在していた。これらのことから、ODS鋼は、700C照射で約160dpaまで照射されたとしても強度は維持されることが示唆された。本研究成果の一部は、文部科学省の原子力システム研究開発事業による委託業務(JPMXD0219214482)として実施した。
Am
O
at 1473, 1573, and 1673 K渡部 雅; 横山 佳祐; Vauchy, R.; 加藤 正人; 菅田 博正*; 関 崇行*; 日野 哲士*
Journal of Nuclear Materials, 599, p.155232_1 - 155232_5, 2024/10
本研究では熱重量法を用いて1473、1573及び1673 KにおけるUAmOの新たな酸素ポテンシャルデータを取得した。同じy及びO/M比で比較した場合、UAmOの酸素ポテンシャルは、UPuOよりも高いことがわかった。また、hypostoichiometric領域のカチオン価数はNd含有UOと類似しており、定比組成では、Am
, U
, and U
となると推定された。実験データを欠陥化学モデルを用いて解析し、O/M比を温度と酸素分圧の関数として表すことができた。
丸山 修平; 山本 章夫*; 遠藤 知弘*
Annals of Nuclear Energy, 205, p.110591_1 - 110591_13, 2024/09
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.
Pu
)O (x = 0, 0.18, 0.45, and 1) and analysis of heat capacity廣岡 瞬; 森本 恭一; 松本 卓; 小笠原 誠洋*; 加藤 正人; 村上 龍敏
Journal of Nuclear Materials, 598, p.155188_1 - 155188_9, 2024/09
酸化物燃料の温度解析において重要な役割を持つ比熱は、特に高温領域において文献間でばらつきが大きい。さらに、UOのデータと比べてPuOやMOXのデータは報告例が少ないため、比熱においてPu含有率の依存性の評価が困難である。本研究では、UO、PuO、MOX (Pu=0.18, 0.45)を対象に、ドロップカロリメータを用いて最高2200Kのエンタルピーのデータを取得した。取得したエンタルピーの温度依存性を評価することで比熱を算出した。エンタルピー、比熱ともに、2000Kまでは温度とともにほぼ線形に上昇し、2000Kを超えると急激に上昇する結果が得られた。2000K以下のデータは文献値とよく一致し、2000K以上のデータは文献値と大きく異なる結果となった。この結果について、酸素及び電子正孔対の欠陥の観点で考察を行った。
浪江 将成; 斉藤 淳一; 池田 明日香; 岡 涼太郎*; Kim, J.-H.*
Surfaces (Internet), 7(3), p.550 - 559, 2024/09
The iron (Fe) specimens selected as the substrate metal for this study were surface-treated using fluorine gas, and their wettability with liquid sodium (Na) was evaluated using the sliding angle. Additionally, the surface morphology and binding state were analyzed, and the applicability of wettability control with liquid sodium by fluorination was discussed using the analysis results. Fluorination formed a fluoride layer comprising FeF and FeF
bonds on the iron surface. The composition of the fluoride layer varied, depending on the treatment conditions. The surface of the specimen that contains a lot of FeF bonds had a small sliding angle for the liquid sodium droplet and was harder to wet than the untreated specimen. In contrast, the surface of the specimen that contains a lot of FeF bonds had a large sliding angle for the liquid sodium droplet and was easier to wet than the untreated specimen. These results indicate that fluorination is an effective surface modification technique that can be applied to control the wettability of iron with liquid sodium.
丸山 修平
JAEA-Data/Code 2024-009, 16 Pages, 2024/08
JAEA-Data-Code-2024-009.pdf:1.08MB
国産の核データ処理コードFRENDY にはランダムサンプリング法に基づくACE ファイル摂動ツールが実装されており、これを利用して核データ起因不確かさを定量化することが可能である。しかしながら、高速炉の炉心解析や遮蔽解析における不確かさ評価で有意となる散乱角度分布起因の不確かさを評価する機能はこれまで開発されていなかった。近年、平均散乱角余弦の共分散データの情報から最大エントロピー法に基づき、この不確かさを定量化する手法が著者らによって提案された。本報告では、この提案手法に基づく弾性散乱角度分布の不確かさに対する摂動機能をFRENDY/ACE ファイル摂動ツールに追加する。
吉田 圭佑; 加藤 慎吾; 奥山 慎一; 石森 有; 井上 睦夫*
Journal of Nuclear and Radiochemical Sciences (Internet), 24, p.1 - 12, 2024/08
The factors causing the temporal variation of 
Be deposition in the Hokuriku region (the Sea of Japan side of central Honshu, the main island of Japan) during winter (November to February) were examined using monthly samples of Be deposition conducted over 30 years, spanning from 1991 to 2021. The predominant factors on Be deposition at a Hokuriku region site were as follows: 1) the amount of Be generated by cosmic rays, 2) the volume of air transported from the Arctic, and 3) the amount of precipitation at the observation site. The contribution of each of these factors fluctuated depending on the sampling period. The temporal variations in Be deposition during the first half of the sampling period (1991-2005) were primarily driven by cosmic rays. In contrast, during the latter half of the period (2006-2021), meteorological factors, particularly snowfall, emerged as significant contributors. This shift in influence was attributed to the effects of climate change in the Hokuriku region.
酒瀬川 英雄; 中島 基樹*; 加藤 太一朗*; 野澤 貴史*; 安堂 正己*
Materials Today Communications (Internet), 40, p.109659_1 - 109659_8, 2024/08
酸化物分散強化型鋼鋼のナノメートルサイズの酸化物粒子はクリープ強度の向上に対して重要な役割を持つ。以前の研究では旧粉末境界という焼結前に機械的合金粉末の表面であった組織因子に注目した。その結果、より小さなサイズの粉末で製作され微細な旧粉末境界を持つ酸化物分散強化型鋼は、より大きなサイズの粉末で製作され粗大な旧粉末境界を持つ酸化物分散強化型鋼よりも、短いクリープ寿命を示すことを確かめた。これより、機械的合金粉末の大きさがクリープ強度特性に影響を及ぼすことを明らかとなった。本研究では非球状である機械的合金粉末の形状がクリープ強度特性に及ぼす影響に注目した。このような形状がクリープ強度特性に異方性を生じさせる可能性が考えられたからである。ここでは異なった切り出し方位を持つ試験片に対してスモールパンチクリープ試験を実施することで異方性に注目した。これより、クリープ寿命は試験片の切り出し方位によって変化することを確かめて、形状がクリープ強度特性に及ぼす影響を明らかとした。
大釜 和也; 羽様 平; 片桐 寛樹*; 竹越 淳*; 毛利 哲也
Nuclear Technology, 210(8), p.1336 - 1353, 2024/08
もんじゅ性能試験では、Pu-239などの核分裂率反応率分布やU-238の捕獲反応率分布が箔放射化法により測定されている。炉心と径ブランケット領域の反応率分布測定値の最確値を算出し、再現解析値との比較により高速炉設計手法の検証データとしての適用性を評価した。集合体中心での歪を適切に補正する解析モデルを導入するなどの工夫の結果、対象としたデータはすべて信頼でき、数%の精度で検証データとして使用できることを確認した。U-238の捕獲反応率については、箔の実効断面積の計算に箔自身と箔周辺の燃料の共鳴断面積を考慮する必要がある。
浜瀬 枝里菜; 堂田 哲広; 小野 綾子; 田中 正暁; 三宅 康洋*; 今井 康友*
Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation, and Safety (NTHOS-14) (Internet), 12 Pages, 2024/08
ナトリウム冷却高速炉の設計において、浸漬型直接炉内冷却器を用いた自然循環崩壊熱除去時に生じる炉心-プレナム相互作用現象を評価するため、炉心部の熱流動解析の計算負荷を合理的に低減した炉容器内熱流動解析評価手法を整備している。本研究では、集合体間ギャップ部(IWG)に着目し、計算負荷を軽減した実用的なIWGモデル整備を目的として、IWGのメッシュ分割と圧力損失相関式を用いたモデルの組合せが炉心内温度分布の再現性に与える影響についてナトリウム試験解析により確認した。
Wen, J.*; 鎌田 悠斗*; 横山 貢成*; 松元 達也*; Liu, W.*; 守田 幸路*; 今泉 悠也; 田上 浩孝; 松場 賢一; 神山 健司
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 8 Pages, 2024/08
To investigate the coolability of fuel debris bed immersed in molten steel, a rectangular experimental system was built in which the particle bed was volumetrically heated via direct current heating. The experimental apparatus consists of a particle bed immersed in water and a water pool above it, which simulate disrupted solid fuel and molten steel, respectively. Computer code simulations with reactor safety analysis code SIMMER-IV were performed to help understanding the heat transfer characteristics and to validate the applicability of the newly embedded momentum exchange function (MXF) models. Under the current experimental conditions, some key parameters like the particle bed average temperature, water pool average temperature, and temperature difference between the bed and the pool were evaluated to compare with the simulation results. The comparison results showed the most applicable MXF model under the current experimental conditions, and the analysis with it well reproduced the phenomena which was observed in the experiments.
矢田 浩基; 高屋 茂; 町田 秀夫*
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 7 Pages, 2024/08
ASME Code Case N-875 provides a rational in-service inspection (ISI) approach for liquid-metal cooled reactors. This is a risk informed approach and ISI requirements including acceptance criteria are established by considering features of respective plants in terms of the effects of flaws on the plant safety. In this approach, fracture mechanics is essential. Not only part-through-wall cracks but also through-wall cracks need to be evaluated, for example, to determine the maximum allowable size of flaws in reactor internal components and investigate the applicability of continuous leakage monitoring to flaws in sodium-retaining components. The basic procedure of the Code Case has also been incorporated in Fitness-for-service code (Section XI, Div. 2) in ASME, which provides requirements for Reliability and Integrity Management programs for nuclear power plants, including advanced reactors. The demonstration sodium-cooled fast reactor currently under development in Japan is expected to be designed with thin wall and large diameter. For some components, the ratio of radius to thickness (R/t) is expected to exceed 100. There are currently no generalized Stress Intensity Factors (SIFs), which are required for fracture mechanics, applicable to components with such a large R/t ratio. In this study, as a part of the development of a flaw evaluation method applicable to components with large R/t ratio, the conservatism of applying the SIF solutions for a through-wall crack in a plate to a circumferential through-wall crack in a cylinder with large R/t ratio was discussed. As a result, it was clarified that the SIF solution for a plate should not be used for circumferential through-wall cracks. Then, a new SIF solution of a circumferential through-wall crack in a cylinder was developed.
進藤 愛美*; 上奥 あや*; 岡村 和奏*; 菊地 晋; 山崎 淳司*; 古賀 信吉*
Thermochimica Acta, 738, p.179801_1 - 179801_12, 2024/08
This study investigated the kinetics of the multistep thermal dehydration/decomposition of the metakaolin-based geopolymer paste. The component two reaction steps were characterized by the evolution of water vapor and the simultaneous evolution of water vapor and CO, respectively. In a stream of dry N, the kinetics of the first and second reaction steps were characterized by the apparent activation energy (Ea) values of 92 and 166 kJ mol
, respectively. Both reaction steps exhibited a diffusion-controlled rate behavior. In a stream of wet N, the mass loss curves systematically shifted to higher temperatures with an increase in the water vapor pressure (p(HO)). The first reaction step was significantly influenced by p(HO), and the apparent Ea increased to 175 kJ mol at p(HO) = 11.4 kPa. The second reaction step was less sensitive to the atmospheric water vapor, as characterized by its Ea of 165 kJ mol , irrespective of the p(HO).
Pu
O powder using master sintering curve theory中道 晋哉; 砂押 剛雄*; 廣岡 瞬; Vauchy, R.; 村上 龍敏
Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07
Using dry recycled powders for uranium and plutonium mixed oxide (MOX) fuel production can reduce unnecessary storage and accountability of nuclear material in facilities. The shrinkage behavior of green compacts of dry recycled powders differs from that of conventional raw powders because the dry recycled MOX powder is obtained from the fabrication scrap of sintered pellets. The shrinkage behavior of dry recycled MOX powder has been investigated by dilatometry. Based on the shrinkage curves, sintering apparent activation energies were evaluated using the master sintering curve (MSC) and the constant rate of heating methods. The obtained values were higher than the energy evaluated for raw powder experiments. The sigmoid sintering prediction equation using the MSC function was constructed. The accumulation of data on the activation energy for various sintering conditions will lead to the wide application of this prediction formula in the future.
江村 優軌; 高井 俊秀; 菊地 晋; 神山 健司; 山野 秀将; 横山 博紀*; 坂本 寛*
Journal of Nuclear Science and Technology, 61(7), p.911 - 920, 2024/07
被引用回数：0 パーセンタイル：0.01(Nuclear Science & Technology)Boron carbide (B
C)- stainless steel (SS) eutectic reaction behavior is one of the most important issues in the core disruptive accidents (CDAs) of sodium-cooled fast reactors (SFRs). In this study, the immersion experiments using BC pellets with molten SS were conducted to evaluate the CDA sequences such as contact event of solid BC with degraded core materials including SS at very high temperature. The immersion experiment aims at understanding the kinetic behavior of solid BC-liquid SS reaction based on the reduced thickness of BC pellet after the experiment in the temperature ranges from 1763 to 1943 K, which is higher than the temperature of solid BC-solid SS reaction. Based on the kinetic consideration of the reaction rate constants for solid BC-liquid SS reaction, it was found that similar temperature dependency was identified between solid BC-liquid SS and solid BC-solid SS. Besides, the reaction rate constants of solid BC-liquid SS were smaller than those of solid BC-solid SS extrapolated in higher temperature region by two or more orders of magnitude due to two different evaluation method for BC side/SS side. It was confirmed that this difference was reasonable through the consideration of previous reaction tests in solid-solid contact for BC side/SS side.
Li, C.-Y.; 渡部 晃*; 内堀 昭寛; 岡野 靖
Journal of Nuclear Science and Technology, 61(7), p.935 - 957, 2024/07
被引用回数：1 パーセンタイル：63.33(Nuclear Science & Technology)For all the nuclear reactor systems, quantitative assessment of the accident management (AM) effects against long-term external hazards became one of the essential issues after the lesson learned from the Fukushima Daiichi Nuclear Power Plant accident. However, the influence from the safety systems' stochastic and dynamic shifting between multiple working states, which is related to the interaction with the adjacent components/systems in general, has not been accounted for yet. Therefore, this research aims to develop a dynamic probability risk assessment tool considering repairable multi-component interdependency for investigating the AM influences on the multi-state safety systems under long-term external hazards. Based on the newly proposed methodology in this research via integrating the Petri Net (PN) model with the continuous Markov chain Monte Carlo (CMMC) method, a framework applying PN-CMMC methodology to a severe accident analysis code, SPECTRA, had been originally constructed. Different AM influences on the multi-state decay heat removal systems against long-term volcanic ashfall were also quantitatively confirmed, indicating that halving the repairing time is more influential in suppressing the core damage frequency than doubling the number of adjacent electricity support systems. Therefore, the PN-CMMC-SPECTRA framework can further assess the uncharted dynamic multi-state concerns, leading to a safer AM strategy.
compoundsFrazer, D.*; Saleh, T. A.*; 松本 卓; 廣岡 瞬; 加藤 正人; McClellan, K.*; White, J. T.*
Nuclear Engineering and Design, 423, p.113136_1 - 113136_7, 2024/07
ナノインデンテーション法では、微小な試験片を用いてヤング率,硬度及びクリープ強度といった機械物性を評価することが可能である。本研究ではMOX燃料の代替物質として(U,Ce)Oを用いて、高温ナノインデンテーション試験を実施した。試料のCe含有率は0.1、0.2及び0.3mol%とし、温度は800Cまでの測定を行い、ヤング率、硬度及びクリープ強度の評価を行った。温度の上昇に伴い、ヤング率は線形的に低下し、硬度は指数関数的に低下する結果が得られた。また、800Cにおいては、応力指数n=4.76.9のクリープ変形が得られた。
