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Qin, T. Y.*; Hu, F. F.*; 徐 平光; Zhang, R.*; Su, Y. H.; Ao, N.*; Li, Z. W.*; 篠原 武尚; 菖蒲 敬久; Wu, S. C.*
International Journal of Fatigue, 202, p.109233_1 - 109233_16, 2026/01
The surface induction-hardened S38C medium carbon steel shows a good balance of strength and toughness, but complicates the evaluation of fatigue resistance, mainly because of gradient residual stress (RS) and grains. An integrated fatigue resistance assessment (AIFA) framework was proposed to consider the residual stress relief under stochastic loads. To this end, quasi-in situ neutron diffraction and Bragg-edge imaging were combined to probe the evolution of residual stress during crack propagation. Firstly, a rigid-flexible coupled vehicle dynamics model was adopted to obtain the time-domain variable amplitude loading spectrum. Then, Fortran subroutines were developed to assign these data into full-scale S38C axle model, and the remaining life was predicted using the damage tolerance approach. The results demonstrate that crack propagation would accelerate when residual stress is considered in the case of the crack depth exceeding 3.0 mm. It is, for the first time, found that 15 mm- and 5 mm-thickness fan-shaped specimens can retain the axial and hoop residual strain in terms of diffraction angle variation, respectively, for full-scale structural S38C steel axles. In the absence of RS, the remaining life of the axle decreases sharply from 624,800 to 51,300 km as the crack depth increases from 3.0 to 16 mm. Compared with the standard method under constant amplitude loading without residual stress relief, the present AIFA method provides the more accurate but conservative fatigue life prediction.
入澤 恵理子; 加藤 千明
Corrosion Science, 256, p.113173_1 - 113173_16, 2025/11
This study investigates the corrosion behavior of extra-high-purity Type 316 austenitic (316EHP) stainless steel with reduced impurity segregation at the grain boundaries in a liquid lead-bismuth eutectic (LBE) at 530
C to evaluate (1) the resistance of the steel to intergranular oxidation in the LBE with oxygen saturation and (2) its dissolution corrosion resistance at lower oxygen concentrations than the equilibrium oxygen potential of magnetite. Under oxygen saturation conditions in the LBE, 316EHP generated protective uniform oxide layers without severe intragranular oxidation. Compared with the case of the conventional 316L stainless steel, enhanced Cr diffusion along the grain boundaries in 316EHP considerably improved the intergranular-oxidation resistance of the steel. However, in the LBE with a low oxygen concentration, 316EHP exhibited high susceptibility to dissolution corrosion, thus undergoing a rapid intergranular attack particularly for short exposure durations, and island-like ferritic particles were formed for long exposure durations. Future studies should explore the optimal oxygen concentrations for oxide scale formation and the long-term corrosion behavior of the steel in dynamic LBE systems.
萩原 大樹; 渡辺 勇輔; 小西 博巳*; 舟木 泰智; 藤原 健壮; 飯島 和毅
Applied Geochemistry, 190, p.106490_1 - 106490_10, 2025/10
Radiocesium (
Cs and 
Cs) was sorbed on minerals and transported to river systems due to the Fukushima Daiichi Nuclear Power Plant accident. Recently, the authors have reported that mafic minerals sorb Cs equally or more strongly than micaceous minerals in fine sands. We characterized mafic minerals and elucidated their weathering using electron microscopy to determine whether they can sorb Cs. The surface of hornblende particles is weathered and altered to vermiculite. The surface of micas is less weathered than that of hornblende, indicating the Cs activity concentrations of highly weathered mafic minerals are higher than those of micas in part of sampling site. The results indicate that the effects of Cs sorption for hornblende depend on the weathering product at the surface.
市原 義孝*; 中村 尚弘*; 飯島 国彦*; 崔 炳賢; 西田 明美
Nuclear Engineering and Design, 441, p.114160_1 - 114160_10, 2025/09
本論文は、振動数に依存しない複素減衰を用いた計算負荷の小さい鉄筋コンクリートモデルを対象とした等価線形解析手法の原子力発電所の原子炉建屋の耐震設計への適用性を検討することを目的とする。これを実現するため、柏崎刈羽原子力発電所7号機原子炉建屋を対象にある理想的な一様地盤条件下での非線形及び等価線形地震時挙動に着目した地盤-建物連成系の三次元有限要素法解析を実施した。その結果、等価線形解析手法は、非線形解析手法に対しせん断ひずみ度、加速度、変位、加速度応答スペクトルの評価において全体的に 概ね良好な対応関係が得られ、その手法の有効性を確認した。また、等価線形解析手法は、原子炉建屋外壁のせん断ひずみ度の評価で材料構成則による非線形解析手法の結果を全体的に上回る安全側の評価となった。このことから、本論文で示す解析条件において、本手法は非線形解析手法より建屋の剛性を低めに評価する傾向にあることを明らかにした。
寺阪 祐太; 佐藤 優樹; 一場 雄太*
Radiation Measurements, 187, p.107486_1 - 107486_8, 2025/09
We measured the distribution of beta-ray emitters inside the Fukushima Daiichi Nuclear Power Station Unit 3 reactor building using a novel optical fiber-based position-sensitive radiation sensor designed for operation in high dose rate environments. Plastic scintillation fibers (PSFs) were installed inside the Unit 3 reactor building, where scintillation light generated through interactions between radiation and the PSFs was detected by a spectrometer to obtain the wavelength spectrum. By applying an unfolding method to the wavelength spectrum, we estimated the distribution of beta ray emitters along the PSFs. To isolate the beta ray contribution in a high gamma dose rate field, we compared measurements taken with and without a stainless steel tube serving as a beta ray shield. As a result, we identified a hotspot predominantly influenced by beta rays for the first time in the high dose rate area on the southern side of the first floor of the Unit 3 reactor building.
O
and/or SiO for methanol synthesis; Deciphering the additive-induced boost in catalytic performance by XAFSIwasaki, Kosei*; Ashida, Yuya*; 松村 大樹; Kawakami, Kotaro*; Shibuya, Kenta*; Tazawa, Masaru*; 辻 卓也; Shimizu, Hajime*
Journal of CO Utilization, 97, p.103111_1 - 103111_9, 2025/07
The development of active, long-lived methanol synthesis catalysts can be expedited by thoroughly understanding the operating mechanisms of promoter additives. SiO serves as an effective support and promoter for Cu/ZnO-type methanol synthesis catalysts; however, it has not been studied as extensively as the industrially predominant AlO, despite being similarly potent. Therefore, we conducted X-ray absorption fine structure predominant AlO, despite being similarly potent. Therefore, we conducted X-ray absorption fine structure studies to probe the effects of incorporating AlO and SiO additives into Cu/ZnO-based heterogeneous catalysts. The results revealed that the additive elements primarily affected the ability of the Zn species to generate oxygen vacancies by distorting the local structure around element-doped ZnO and that the number of oxygen vacancies correlated to the catalytic activity. Additionally, the oxygen vacancies in ZnO in the Al/Si-incorporated catalysts diminished under catalytic reaction conditions, thereby providing information on the deactivation of catalytic reactions. Our findings can facilitate the development of highly active industrial-grade methanol synthesis catalysts.
Ho(n,
)
Ho reactions中村 詔司; 芝原 雄司*; 遠藤 駿典; Rovira Leveroni, G.; 木村 敦
Journal of Nuclear Science and Technology, 14 Pages, 2025/07
During the decommissioning of nuclear facilities, large amounts of radioactive wastes are generated from structural materials. When considering the disposal or reuse of such wastes, accurate neutron capture cross-sections are required to evaluate the amounts of radioactive nuclides among the wastes. The present work selected Ho among nuclides included in the list for clearance levels in decommissioning, and measured the thermal-neutron capture cross-sections for the Ho(n,)
Ho, 
Ho reactions by the neutron activation method. The thermal cross-section measurements were performed with the graphite thermal column of the Kyoto University Research Reactor under the 5-MW operation and the thermal-neutron capture cross-sections were derived on the basis of Westcott's convention. In this work, a value of 2.79
0.04 barn was obtained for the Ho(n,)Ho reaction, and 61.20.6 barn for the Ho(n,)Ho reaction. The combination of these cross-sections presented 64.00.6 barn, which supports the recent evaluated data of 64.69 barn and 64.41.2 barn within the limit of uncertainties.
Hamdani, A.; 相馬 秀; 安部 諭; 柴本 泰照
Progress in Nuclear Energy, 185, p.105771_1 - 105771_13, 2025/07
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)This study, motivated by previous TEPSYS analysis, examined how different temperatures on the 4th and 5th floors of the Fukushima Daiichi Unit 3 reactor building (R/B) influenced non-condensable gas distribution during the 2011 severe accident. Understanding this is vital for assessing risks related to gas accumulation, especially since the hydrogen explosion may have involved multiple stages. An experimental study was conducted using the CIGMA facility, designed to mimic the R/B structure, where steam and helium (as a substitute for hydrogen) were injected for 10,000 seconds to simulate leakage. Two cooling conditions were tested: 50C (Case 1) and 90C (Case 2). Results showed that the highest concentration of non-condensable gases was often found downstream rather than near the injection point. In Case 1, after 10,000 seconds, helium concentration reached 65% in the middle region (4th floor) and 45% in the top region (5th floor). Analysis indicated that the gas mixture in the middle region posed a potential detonation risk. This study offers crucial insights for enhancing safety measures and risk mitigation strategies in nuclear reactor designs.
吉田 一雄; 桧山 美奈*; 玉置 等史
JAEA-Research 2025-003, 24 Pages, 2025/06
JAEA-Research-2025-003.pdf:2.06MB
再処理施設の過酷事故の一つである高レベル放射性廃液貯槽の冷却機能喪失による蒸発乾固事故では、沸騰により廃液貯槽から発生する硝酸-水混合蒸気とともにルテニウム(RuO
)の揮発性の化学種が放出される。このためリスク評価の観点からは、Ruの定量的な放出量の評価が重要な課題である。揮発性Ruは施設内を移行する過程で床面に停留するプール水中の亜硝酸によって化学吸収が促進されることが想定され、施設内の硝酸-水混合蒸気の凝縮水量がRuの施設内での移行に重要な役割を担う。当該事故の施設内の熱流動解析では、水の熱流動を解析対象とするMELCORコードを用いている。解析では、凝縮の支配因子である蒸発潜熱が、事故時での施設内の温度帯域で同程度であることから硝酸をモル数が等しい水として扱っている。本報では、この解析モデルの妥当性を確認するために、MELCORの制御関数機能を用いて硝酸-水混合蒸気を水蒸気で近似することによって生じる誤差を補正する解析モデルを作成し解析を実施し補正効果を比較することで従来の解析モデルの妥当性を確認した。その結果、補正解析モデルの適用によって各区画のプール水量の分布は変化するものの施設内のプール水量の総和には影響しないことを確認した。
柳澤 宏司; 求 惟子
JAEA-Research 2025-001, 99 Pages, 2025/06
JAEA-Research-2025-001.pdf:1.98MB
中性子吸収棒の反応度価値に関する安全検査データのより深い理解と反応度価値の測定技術の向上のために、TRIGA-ACPR(環状炉心パルス炉)に分類されるNSRR(原子炉安全性研究炉)の初回起動炉心の臨界解析用詳細計算モデルを作成した。本モデルの形状、材料、運転データの誤差から伝播する中性子実効増倍率(keff)の不確かさを、最新の核データライブラリJENDL-5及び旧版のJENDLライブラリとMVP第3版コードを用いて詳細に評価した。その結果、本モデルにおけるk
の全体的な不確かさは、0.0027から0.0029
kの範囲と評価した。本モデルは、TRIGA-ACPRのkのベンチマークとして利用されることが期待される。さらに、全体的な不確かさは、NSRRで測定された吸収棒価値よりも十分小さいことを確認した。よって、本モデルはNSRRにおける吸収棒反応度価値に関する今後の解析にも適用できる。
青山 高士; Choudhary, S.*; Pandaleon, A.*; Burns, J. T.*; Kokaly, M.*; Restis, J.*; Ross, J.*; Kelly, R. G.*
Corrosion, 81(6), p.609 - 621, 2025/06
This study presents a new test method for inducing controlled corrosion damage within simulated fastener holes of aluminum alloys, aimed at pretreating fatigue test specimens. The method involves insulating the outer surface while exposing the fastener hole surface to electrolytes containing 0.66 M NaCl + 0.1 M AlCl with varying concentrations of K S O 
. The evolution of corrosion damage within the fastener hole was examined as a function of exposure duration, electrolyte composition, and volume, as well as the effect of galvanic coupling with a SS316 cathode. Results indicate that fissure depth increases with an increase in K S O concentration but does not progress further after 24-48 hours of exposure in the chemical, or freely-corroding, exposure test. In contrast, galvanic coupling with a SS316 plate significantly accelerates corrosion, leading to much deeper fissures in a shorter time. The importance of electrolyte replenishment has been explored using electrochemical measurements, revealing the impact of evolving electrolyte chemistry. Beyond its application in fatigue specimen pretreatment, this method provides a simple yet effective approach for studying localized corrosion and evaluating mitigation strategies for fastener holes in aerospace structures.
杉田 裕; 大野 宏和; Beese, S.*; Pan, P.*; Kim, M.*; Lee, C.*; Jove-Colon, C.*; Lopez, C. M.*; Liang, S.-Y.*
Geomechanics for Energy and the Environment, 42, p.100668_1 - 100668_21, 2025/06
被引用回数：1 パーセンタイル：77.92(Energy & Fuels)国際共同プロジェクトDECOVALEX-2023は、数値解析を使用してベントナイト系人工バリアの熱-水-応力(または熱-水)相互作用を研究するためのタスクDとして、幌延人工バリア性能確認試験を対象とした。このタスクは、モデル化のために、1つの実物大の原位置試験と、補完的な4つの室内試験が選択された。幌延人工バリア性能確認試験は、人工的な地下水注入と組み合わせた温度制御非等温の試験であり、加熱フェーズと冷却フェーズで構成されている。6つの研究チームが、さまざまなコンピューターコード、定式化、構成法則を使用して、熱-水-応力または熱-水(研究チームのアプローチによって異なる)数値解析を実行した。
Birkholzer, J. T.*; Graupner, B. J.*; Harrington, J.*; Jayne, R.*; Kolditz, O.*; Kuhlman, K. L.*; LaForce, T.*; Leone, R. C.*; Mariner, P. E.*; McDermott, C.*; et al.
Geomechanics for Energy and the Environment, 42, p.100685_1 - 100685_17, 2025/06
被引用回数：0 パーセンタイル：0.00The DECOVALEX initiative is an international research collaboration (www.decovalex.org), initiated in 1992, for advancing the understanding and modeling of coupled thermo-hydro-mechanical-chemical (THMC) processes in geological systems. DECOVALEX stands for "DEvelopment of COupled Models and VALidation against EXperiments". DECOVALEX emphasizes joint analysis and comparative modeling of the complex perturbations and coupled processes in geologic repositories and how these impact long-term performance predictions. More than fifty research teams associated with 17 international DECOVALEX partner organizations participated in the comparative evaluation of eight modeling tasks covering a wide range of spatial and temporal scales, geological formations, and coupled processes. This Virtual Special Issue on DECOVALEX-2023 provides an in-depth overview of these collaborative research efforts and how these have advanced the state-of-the-art of understanding and modeling coupled THMC processes. While primarily focused on radioactive waste, much of the work included here has wider application to many geoengineering topics.
中村 詔司; 遠藤 駿典; Rovira Leveroni, G.; 木村 敦; 芝原 雄司*
KURNS Progress Report 2024, P. 31, 2025/06
本研究は、生成放射能を評価するために、廃止措置で問題となる核種について熱中性子捕獲断面積を測定するものである。本件では、対象核種の中から
Hoを選定し、京大原子炉にて放射化法によりその熱中性子捕獲断面積を測定した。今回、Ho(n,)Ho反応について2.790.04 barnを得た。従来の報告値は3.40.5(b)に対して、誤差1.4
に向上して導出することができた。また副産物として、Ho(n,)Ho反応について61.20.6 barnを得た。HoとHoの生成の断面積を合わせて、64.00.6 barnを求めた。今回の結果は、TOF法による過去の報告値64.42.8(b)や、最近の評価値64.69 barn、64.41.2 barnを誤差の範囲で支持した。
Alzahrani, H.*; 松下 健太郎; 堺 公明*; 江連 俊樹; 田中 正暁
Nuclear Technology, 13 Pages, 2025/06
ナトリウム冷却高速炉の炉上部プレナム部自由液面において発生する渦によるカバーガス巻込み現象の評価手法の開発が求められており、CFD解析によって得られた流速分布から渦を予測する評価手法の開発が進められている。本研究は、渦の予測のためのCFD解析の効率化の観点からAdaptive Mesh Refinement(AMR)法の適用性について検討した。解析メッシュを詳細化する基準指標として、速度勾配テンソルの第二不変量Qが負となる指標(Index-1)およびIndex-1に圧力勾配の条件を追加した指標(Index-2)を選択した。垂直平板のある非定常渦体系にAMR法を適用し、得られた詳細化メッシュを用いて解析を行った結果、Index-2によって詳細化された解析メッシュにおける圧力分布や垂直平板周りの流速分布が、リファレンスとなる一様詳細メッシュのものと同等となり、Index-2によってはく離点近傍のメッシュが詳細化されることで渦の発生や成長をより正確に予測できることが確認された。
豊田 晃大; 鬼澤 高志; 若井 栄一*
Research & Development in Material Science (Internet), 21(5), p.2632 - 2637, 2025/06
316FR steel, a modification of 316 austenitic stainless steel, will be used as a structural material in the sodium cooled fast reactor (SFR), one of the initiatives being developed in Japan to achieve carbon neutrality in order to combat global warming. To withstand the high-temperature operating environment of the SFR, the alloy design of the 316FR steel has been optimized to have high creep strength for a long time with controlled precipitation by optimizing the alloy composition. In order to clarify that 316FR steel can maintain its properties under the high temperature (around 550C) irradiation environment of the SFR, the authors mainly conducted in-situ observations under electron beam irradiation at high temperatures to investigate in detail the irradiation effects on the precipitates (mainly carbides), which are characteristic of 316FR steel. As a result, it was found that the precipitates in 316FR steel are more stable than those in type 304 stainless steel under irradiation without coarsening at grain boundaries or within grains. The characteristics and attractiveness of 316FR steel, the results obtained, and the mechanism of creep behavior under irradiation are also explained.
方野 量太; 阿部 拓海; Cibert, H.*
JAEA-Research 2024-019, 22 Pages, 2025/05
JAEA-Research-2024-019.pdf:1.03MB
マイナーアクチノイドの核変換を目的とする加速器駆動システム(ADS)は未臨界状態で運転される。ADSの未臨界度管理においては、燃焼反応度の予測が重要であるが、予測精度の検証のためには、特に第一サイクル運転時では燃焼反応度を精度良く測定する必要がある。本検討では、燃焼反応度測定手法としてCurrent-To-Flux(CTF)法に着目し、連続エネルギーモンテカルロ計算コードSERPENT2を用いて固定源燃焼計算を実施し、炉内に配置する核分裂計数管を模したタリーを用いることで、CTF法によるADS通常運転時の燃焼反応度測定のシミュレーションを実施した。シミュレーション結果から測定手法起因の燃焼反応度測定不確かさの推定を行い、燃焼期間に依らず燃焼反応度に対して10%程度のバイアスが生じ、その検出器位置依存性が体系外側で小さいことを明らかにした。
Li, X.; 山路 哲史*; 佐藤 一憲*; 山下 拓哉
Annals of Nuclear Energy, 214, p.111217_1 - 111217_13, 2025/05
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)The decommissioning of Fukushima Daiichi NPP Unit-2 requires understanding of reactor damage and fuel debris distribution for effective debris retrieval. This study numerically analyzes potential Reactor Pressure Vessel (RPV) boundary failure due to eutectic melting of Control Rod Drive (CRD) housings during reheating after debris bed dryout. The Moving Particle Semi-implicit (MPS) method, with an enthalpy-based temperature algorithm and Boussinesq approximation, is applied to simulate melt/solid interactions in a 2-D model of the lower plenum. The CRD housing melting temperature is set at 1523 K based on a quasi-binary phase diagram of 304 Stainless Steel (SS) and Zirconium (Zr) and ELSA experiments. Results suggest local RPV failure at CRD housings, leading to melt release and refreezing. The estimated failure occurs 8-12 hours post-dryout (ca. 12:00-16:00 on 3/15/2011), providing insights into melt progression and boundary breach scenarios in Unit-2.
上澤 伸一郎; 小野 綾子; 永武 拓; 山下 晋; 吉田 啓之
Journal of Nuclear Science and Technology, 62(5), p.432 - 456, 2025/05
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)ワイヤメッシュセンサ(WMS)の精度を明らかにするため、単一の球形気泡と気泡流に対してWMSの静電場シミュレーションを実施した。単一気泡の静電場シミュレーションでは、様々な気泡位置における電流密度分布と、送信ワイヤから受信ワイヤまでの電流経路を示した。その結果、WMS周囲の不均一な電流密度分布に基づく系統的誤差があることを明らかにした。また、数値流体解析コードJAEA Utility Program for Interdisciplinary Thermal-hydraulics Engineering and Research (JUPITER)で得られた気泡流結果に対して静電場シミュレーションを実施したところ、線形近似やMaxwellの式などの、WMS信号からボイド率への既存の変換方法では0と1の間の瞬間ボイド率の中間値を定量的に推定できなかった。また、WMS信号に対してボイド率0.2という大きなばらつきがあり、瞬間ボイド率を定量的に計測することが困難であることがわかった。一方で、時間平均ボイド率においては、流路の中心付近のボイド率は線形近似を使用して推定でき、流路壁面近くのボイド率はMaxwellの式を使用して推定できることがわかった。
福田 航大; 小原 徹*; 須山 賢也
Nuclear Technology, 211(5), p.963 - 973, 2025/05
被引用回数：1 パーセンタイル：43.12(Nuclear Science & Technology)An application of the boiling water reactor (BWR) to an offshore floating nuclear power plant (OFNP) is discussed in Japan. The BWR-type OFNP has some challenges for practical use, although it has high economic efficiency because of downsizing and simplification. One challenge is understanding reactor kinetics under conditions specific to the marine environment. This study quantitatively clarifies the total and spatial changes in power when the BWR is inclined during regular operation. Therefore, the TRAC/RELAP Advanced Computational Engine (TRACE) and Purdue Advanced Reactor Core Simulator (PARCS) codes were used to perform a three-dimensional neutronics-thermal-hydraulics-coupled transient analysis. The calculation model is based on Peach Bottom II. This study clarifies the changing trend in total and local BWR power by inclination with simplified modeling and conditions. Reasons for such changes are discussed based on changes in several thermal-hydraulic parameters. The difference in BWR power against the inclinations is small. Thus, it was implied that the BWR-type OFNP is expected to have a stable power supply capability during natural disasters. Finally, requires further studies to support the obtained conclusions are discussed.
