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CPham V. H.; 倉田 正輝; 永江 勇二; 石橋 良*; 佐々木 政名*
Corrosion Science, 255, p.113098_1 - 113098_9, 2025/10
被引用回数：0Being expected as materials for accident tolerant fuel cladding tube, oxidation behavior and kinetics of silicon carbide (SiC) under extreme conditions like severe accidents must be elucidated. In this study, oxidation tests of SiC at 1400-1600 C for 1-5 h, at atmospheric pressure, under two different flow rates of H
O/Ar gas mixture have been conducted to investigate the influence of steam flow rate on the formation of SiO scale and its subsequent volatilization. The oxidation tests were conducted via a newly developed test facility using laser as a heat source. Oxidation kinetics of SiC was evaluated via mass change of samples before and after the oxidation tests. Parabolic oxidation rate representative for SiO formation and linear volatilization rate reflecting its volatilization were calculated, based on these mass changes. The Arrhenius dependence of the parabolic oxidation and linear volatilization rate constants were then plotted. Results of this study indicated that SiC exhibits excellent performance under the conditions investigated. Steam flow rate has a significant influence on volatilization of SiO but has minor effects over its formation. Oxidation of SiC in steam at high temperature may follow mass gain or mass loss regime, depending on the steam flow rate. Two oxidation patterns were suggested and discussed. In the first oxidation pattern, the SiO formation is dominated over its volatilization. The second oxidation pattern (steady stage) is reached when the SiO formation rate is equivalent to its volatilization rate. Time to reach this steady stage was defined, based on the parabolic oxidation rate and linear volatilization rate.
和田 裕貴; 柴本 泰照; 日引 俊詞*
International Journal of Heat and Mass Transfer, 249, p.127219_1 - 127219_16, 2025/10
被引用回数：0Two saturated boiling heat transfer correlations for downward flows in vertical circular pipes depending on wall superheat or wall heat flux as input parameters were developed based on a heat transfer experimental database. Owing to the absence of heat transfer correlations specifically developed for downward flows, existing heat transfer correlations for different flow directions were evaluated to determine their applicability to predicting the downward flow heat transfer coefficient. The results revealed that even the most accurate correlation showed a mean absolute percentage error (MAPE) of 66.5%, highlighting the need for improving predictive performance. In response, the downward flow heat transfer correlation was modeled by integrating a nucleate boiling heat transfer term and a forced convection heat transfer term. The Dong-Hibiki correlation, a two-component, two-phase heat transfer correlation for downward flows, was adopted for the forced convection heat transfer term. The Forster-Zuber correlation, developed as a wall superheat function, and the Cooper correlation, developed as a wall heat flux function, were used for the nucleate boiling term to develop the heat transfer correlations where either wall superheat or wall heat flux is known. Notably, the Dong-Hibiki correlation has been validated over a wide range of experimental conditions. A correction factor was applied to the nucleate boiling term to address errors caused by applying Foster-Zuber and Cooper correlations to downward flows. The two developed correlations achieved an MAPE value of approximately 20%, representing an improvement of roughly 40% over existing correlations of heat transfer coefficients.
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。
和田 裕貴; 柴本 泰照; 日引 俊*
International Journal of Heat and Mass Transfer, 239, p.126598_1 - 126598_18, 2025/04
被引用回数：3 パーセンタイル：21.83(Thermodynamics)This study reviewed the saturated boiling heat transfer research in downward flows. A database of downward flow heat transfer experiments was created using experimental studies. Saturated boiling heat transfer correlations in internal flows were collected, and no downward flow-specific heat transfer correlations were identified. The applicability of heat transfer correlations to downward flow heat transfer experiments was evaluated, and no correlation could predict the heat transfer coefficients accurately for all experimental databases. However, correlations that could predict heat transfer coefficients reasonably well were determined for each channel size. Cooper's correlation [Int. Chem. Eng. Symp. Ser. 86 (1984) 785-792] had a mean absolute percentage error (MAPE) of 11.7% for mini-channels and Kim and Mudawar's correlation [Int. J. Heat Mass Transf. 64 (2013) 1239-1256] had an MAPE of 66.5% for macro-channels. Furthermore, because the advection direction between the liquid-phase and the generated bubbles differed depending on the liquid-phase velocity in downward flows, we evaluated the prediction performance of the heat transfer coefficient for the liquid-phase velocity. For some experimental data, the prediction performance of the existing correlation for downward flow heat transfer worsened as the advection velocity of the bubbles decreased. This result is one of the issues to be addressed in the future development of heat transfer correlations.
櫻井 惇也*; 鳥形 啓輔*; 松永 学*; 高梨 直人*; 日比野 真也*; 木津 健一*; 森田 聡*; 井元 雅弘*; 下畠 伸朗*; 豊田 晃大; et al.
鉄と鋼, 111(5), p.246 - 262, 2025/04
Creep testing is time-consuming and costly, leading institutions to limit the number of tests conducted to the minimum necessary for their specific objectives. By pooling data from each institution, it is anticipated that predictive models can be developed for a wide range of materials, including welded joints and degraded materials exposed to service conditions. However, the data obtained by each institution is often highly confidential, making it challenging to share with others. Federated learning, a type of privacy-preserving computation technology, allows for learning while keeping data confidential. Utilizing this approach, it is possible to develop creep life prediction models by leveraging data from various institutions. In this paper, we constructed global deep neural network models for predicting creep rupture life of heat-resistant ferritic steels in collaboration with eight institutions using the federated learning system we developed for this purpose. Each institution built a local model using only its own data for comparison. While these local models demonstrated good predictive accuracy for their respective datasets, their predictive performance declined when applied to data from other institutions. In contrast, the global model constructed using federated learning showed reasonably good predictive performance across all institutions. The distance between each institution's data was defined in the space of explanatory variables, with the NIMS data, which had the largest dataset, serving as the reference point. The global model maintained high predictive accuracy regardless of the distance from the NIMS data, whereas the predictive accuracy of the NIMS local model significantly decreased as the distance increased.
田代 信介; 内山 軍蔵; 大野 卓也; 天野 祐希; 吉田 涼一朗; 渡邉 浩二*; 阿部 仁; 山根 祐一
Nuclear Technology, 211(3), p.429 - 438, 2025/03
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)火災事故の下での放射性物質を閉じ込めるHEPAフィルターと関係付けたグローブボックス(GB)における閉じ込め安全性の評価に寄与するために、工学規模の装置を用いて代表的なGBパネル樹脂として可燃性のポリマーであるポリメチルメタクリレート(PMMA)や難燃性のポリマーであるポリカーボネートの燃焼試験を行った。燃焼試験ではPMMAやPCの質量減少速度(MLR)ならびに放熱速度(HRR)のような燃焼特性を調べた。同一寸法の平板形状のPMMAやPCの燃焼では、燃焼させるセルへの給気流量条件を変えた場合のMLRやHRRはPMMAよりPCの方が大きくかつ給気流量に対して一定であり、さらに直径を変えて断面積(S)条件も変えた場合のPMMAの燃焼におけるMLRやHRRはSに対して比例する特性が得られた。これらの結果を用いて、平板形状のPMMAやPCの断面積に対するMLRならびにHRRの関係式を導出した。
相原 純; 植田 祥平; 本田 真樹*; 笠原 清司; 岡本 孝司*
JAEA-Research 2024-012, 98 Pages, 2025/02
JAEA-Research-2024-012.pdf:32.24MB
Pu燃焼高温ガス炉とは、再処理Puの量を安全に減らすための高温ガス炉である。Pu燃焼高温ガス炉では、PuO-イットリア安定化ジルコニア(PuO-YSZ)の微小球にZrC層を被覆し、更にSiC-TRISO被覆を施した核拡散抵抗性の高い被覆燃料粒子(CFP)を用いる計画である。ZrC層の役割は酸素ゲッターである。平成26-29年に行われたPu燃焼高温ガス炉研究プロジェクトでは、Puの模擬物質としてCeを用いて模擬CFPが製造され、更に、この模擬CFPがHTTR燃料と同様に黒鉛母材で焼き固められ模擬燃料コンパクトが製造された。本報告では、模擬燃料コンパクト製造までの各段階におけるCeO-YSZ核及びZrC層の微細構造観察の結果を報告する。
沖田 将一朗; 青木 健; 深谷 裕司; 橘 幸男
Proceedings of 31st International Conference on Nuclear Engineering (ICONE31) (Internet), 5 Pages, 2024/11
We have been developing a methodology for nuclide production and annihilation and decay heat evaluations for High Temperature Gas-cooled Reactors (HTGRs). We are planning to perform validation of the evaluation method with isotopic composition data obtained from HTGR type fuel irradiation tests (AGR tests) performed at the Idaho National Laboratory. As a first step of this plan, preliminary validation of a calculation code and a nuclear data library to be used in the evaluation methodology should be conducted. We made a calculation model of the Advanced Test Reactor (ATR) with a continuous-energy Monte Carlo code MVP-3 and the latest nuclear data library in Japan JENDL-5 on the basis of a calculation input for another Monte Carlo code MCNP5 documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE). We also calculated effective multiplication factors and relative power densities for the ATR calculation model. As a result of comparison with measured values reported in the IRPhE handbook, the JENDL-5 and the calculation model built with MVP-3 shows an enough calculation accuracy in the ATR. Our results will help us to perform our planned validation of our nuclide production and annihilation and decay heat evaluation methodology with the AGR test data.
Brear, D. J.*; 近藤 悟; 曽我部 丞司; 飛田 吉春*; 神山 健司
JAEA-Research 2024-009, 134 Pages, 2024/10
JAEA-Research-2024-009.pdf:2.45MB
SIMMER-III/SIMMER-IVは液体金属高速炉の炉心崩壊事故(CDA)の解析に使用する計算コードである。CDAの事象進展は炉心物質間の熱伝達係数(HTC)により大きく影響される。溶融・固化、蒸発・凝縮といった質量移行現象も熱伝達により支配される。複雑な多相・多成分系においては、一つの流体成分と他の流体又は構造材表面との間での多数の異なるHTCを計算する必要がある。また、多相流の流動様式や構造材の配位に従って異なる伝熱モードを考慮する必要もある。結果として、各計算セルごとに数十のHTCが計算される。本報告書には、SIMMER-III/SIMMER-IVのHTCモデルの役割、選定したHTC相関式とその技術的背景、流動様式の取扱いとHTCの内挿方法、検証及び妥当性確認の成果概要を記載する。
Rochman, D.*; 湊 太志; 渡邉 友章; 他52名*
EPJ Nuclear Sciences & Technologies (Internet), 10, p.9_1 - 9_83, 2024/10
This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments.
茂木 孝介; 柴本 泰照; 日引 俊詞*; 塚本 直史*; 金子 順一*
JAEA-Review 2024-039, 45 Pages, 2024/09
JAEA-Review-2024-039.pdf:2.23MB
既往研究において様々な対向複合対流の熱伝達相関式が提案されているが、それらは様々な試験装置、流路形状、試験流体、熱流動パラメータの範囲で実施された実験結果に基づいている。従って、使用に際してその適用範囲や外挿性を踏まえた上でどの相関式を選択すべきかを整理しておくことは重要である。本稿では既存の対向複合対流の熱伝達相関式についてレビューした。また、複数の既往実験データと各相関式との比較を行い、相関式の予測性能を評価した。その結果、Jackson and Fewster相関式、Churchill相関式、Swanson and Catton (IJHMT)相関式は、全ての実験データを精度良く予測可能であった。さらに、代表長さに水力等価直径を用いることにより流路形状の違いに関わらず相関式が適用可能であり、支配パラメータの無次元化により試験流体によらず相関式が適用可能であることを確認した。
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
被引用回数：0 パーセンタイル：0.00(Materials Science, Multidisciplinary)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。
永塚 健太郎; 野口 弘喜; 長住 達; 野本 恭信; 清水 厚志; 佐藤 博之; 西原 哲夫; 坂場 成昭
Nuclear Engineering and Design, 425, p.113338_1 - 113338_11, 2024/08
被引用回数：4 パーセンタイル：91.70(Nuclear Science & Technology)高温ガス炉は固有の安全性を有し、二酸化炭素を排出することなく大量の水素や高温の熱供給が可能なことから、産業分野の脱炭素化に貢献できる。本報では、原子力機構で進めるHTTR(高温工学試験研究炉)を利用した炉心強制冷却喪失(LOFC)試験等の研究開発成果に加え、現在設計を進めるHTTRを用いた水素製造実証試験(HTTR-熱利用試験)の計画を紹介する。加えて、2030年代後半の運転開始に向け、基本設計が進められている高温ガス炉実証炉計画を紹介する。
山野 秀将; 二神 敏; 日暮 浩一*
Proceedings of Advanced Reactor Safety (ARS 2024), p.121 - 130, 2024/08
本論文は、信頼性を向上させた崩壊熱除去系について、第4世代炉国際フォーラムで開発された安全設計クライテリアと安全設計ガイドラインを我が国で最近設計されたナトリウム冷却高速炉へ適用したことを記述する。
CoSi山内 宏樹; Sari, D. P.*; 安井 幸夫*; 坂倉 輝俊*; 木村 宏之*; 中尾 朗子*; 大原 高志; 本田 孝志*; 樹神 克明; 井川 直樹; et al.
Physical Review Research (Internet), 6(1), p.013144_1 - 013144_9, 2024/02
-Mn-type family alloys Mn
have three-dimensional antiferromagnetic (AFM) corner-shared triangular network. The antiferromagnet MnRhSi shows magnetic short-range order (SRO) over a wide temperature range of approximately 500 K above the N
el temperature 
= 190 K. MnCoSi has the smallest lattice parameter and the lowest in the family compounds. The quantum critical point (QCP) from AFM to the quantum paramagnetic state is expected near a cubic lattice parameter of 6.15 
of MnCoSi is only 140 K, quantum critical behavior is observed in MnCoSi as the enhancement of the electronic specific heat coefficient 
. We study how the magnetic SRO appears in MnCoSi by using neutron scattering, 
SR, and physical property measurements. The experimental results show that the neutron scattering intensity of the magnetic SRO does not change much regardless of the suppressed magnetic moment in the long-range magnetic ordered state compared to those of MnRhSi. The initial asymmetry drop ratio of SR above becomes small, and the magnetic SRO temperature 
is suppressed to 240 K. The results suggest that the MnCoSi is close to the QCP in the Mn system.
今野 力
Journal of Nuclear Science and Technology, 61(1), p.121 - 126, 2024/01
被引用回数：1 パーセンタイル：23.64(Nuclear Science & Technology)JENDL-4.0/HEの中性子と陽子ACEファイルは2017年に作られ、そのうちの22核種の中性子ACEファイルと25核種の陽子ACEファイルがPHITSコードと一緒に公開されている。最近、JENDL-4.0/HEの中性子と陽子ACEファイルに入っている以下の5つのデータに問題があることが見つかった; 
Nと
OのACEファイル、発熱数、損傷エネルギー生成断面積、2次中性子多重度、核分裂断面積。そこで、これらの問題を修正したJENDL-4.0/HEの新しい中性子と陽子ACEファイルを作成した。この論文では問題点及び新しい中性子と陽子ACEファイルをどのように作成したかについて詳述する。
鈴木 誠也; 根本 善弘*; 椎木 菜摘*; 中山 佳子*; 竹口 雅樹*
Annalen der Physik, 535(9), p.2300122_1 - 2300122_12, 2023/09
被引用回数：0 パーセンタイル：0.00(Physics, Multidisciplinary)Germanene is a two-dimensional (2D) germanium (Ge) analogous of graphene, and its unique topological properties are expected to be a material for next-generation electronics. However, no germanene electronic devices have yet been reported. One of the reasons for this is that germanene is easily oxidized in air due to its lack of chemical stability. Therefore, growing germanene at solid interfaces where it is not oxidized is one of the key ideas for realizing electronic devices based on germanene. In this study, the behavior of Ge at the solid interface at high temperatures was observed by transmission electron microscopy (TEM). To achieve such in situ heating TEM observation, we fabricated a graphene/Ge/graphene encapsulated structure. In situ heating TEM experiments revealed that Ge like droplets moved and coalesced with other Ge droplets, indicating that Ge remained as a liquid phase between graphene layers at temperatures higher than the Ge melting point.
Thwe Thwe, A.; 門脇 敏; 永石 隆二
Journal of Nuclear Science and Technology, 60(6), p.731 - 742, 2023/06
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)本研究では、詳細な化学反応を考慮した非定常反応流れの数値計算を遂行し、固有不安定性による水素-空気希薄予混合火炎の不安定挙動を調べ、未燃ガス温度と圧力の影響を明らかにした。広い空間における火炎の不安定挙動をシミュレートし、セル状火炎の燃焼速度を求めた。そして、火炎不安定挙動に及ぼす熱損失および火炎スケールの効果を精査した。平面火炎の燃焼速度は、未燃ガスの温度が上昇すると増加し、未燃ガスの圧力と熱損失が上昇すると減少する。平面火炎の燃焼速度で標準化したセル状火炎の燃焼速度は、圧力(温度)の上昇と共に増大(減少)する。熱損失が存在する場合、標準化したセル状火炎の燃焼速度は、断熱の場合より大きくなる。これは、未燃ガスの高圧力と熱損失は、火炎の不安定挙動と不安定性をプロモートするからである。
安部 諭; 柴本 泰照
Nuclear Engineering and Technology, 55(5), p.1742 - 1756, 2023/05
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high temperature flow of approximately 390C was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high temperature region. The phenomenological discussion in this paper help understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.
野本 恭信; 水田 直紀; 守田 圭介; 青木 健; 沖田 将一朗; 石井 克典; 倉林 薫; 安田 貴則; 田中 真人; 井坂 和義; et al.
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 7 Pages, 2023/05
JAEA initiated an HTTR heat application test plan to develop for coupling technology between HTGR and hydrogen production facility. The principal objective of this test plan is to establish the high safety coupling technology for coupling a hydrogen production facility to HTGR through the demonstration of a hydrogen production by the proven technology of methane steam reforming method utilizing the HTTR as a high temperature heat source. The other objective is to develop for coupling equipment such as a high temperature isolation valve, a helium gas circulator and a high temperature insulation pipe. This paper describes the overview of an HTTR heat application test plan such as a draft test schedule and test targets for the demonstration of a hydrogen production. This paper also presents basic specifications of an HTTR heat application test facility such as the HTTR modification strategy, overall system configuration and heat and mass balance at rated test operation for the demonstration of a hydrogen production. Furthermore, the operation plan during the normal start-up and shut-down processes is proposed.
