Field-induced insulator-metal transition in EuTe

竹内 徹也*; 本多 史憲*; 青木 大*; 芳賀 芳範; 木田 孝則*; 鳴海 康雄*; 萩原 政幸*; 金道 浩一*; 軽部 皓介*; 播磨 尚朝*; et al.

Journal of the Physical Society of Japan, 93(4), p.044708_1 - 044708_10, 2024/04

https://doi.org/10.7566/JPSJ.93.044708

Electrical and magnetic properties of divalent europium antiferromagnet EuTe are studied on single crystals grown from Te-flux method. Antiferromagnetism occurs at 11 K with an insulating ground state. Application of magnetic field induces a sudden resistance drop leading to a phase transition to metallic state. Energy band calculations reveal the appearance of a Fermi surface pocket with Te-5p character induced by the polarized Eu moments.

原子力防災を中心とした専門用語に関する和英対訳の調査と提案

外川 織彦; 奥野 浩

JAEA-Review 2023-043, 94 Pages, 2024/03

JAEA-Review-2023-043.pdf:1.53MB

日本語で記載された原子力防災分野の文書を英語に翻訳するために、災害対策基本法、原子力災害対策特別措置法及び原子力の安全に関する条約について、日本語と英語の対訳を調査した。調査結果を統合し、統一的な英対訳を選択した。この結果として、原子力防災分野における専門 用語の和英対訳表を作成し、提案した。

Evaluation of temporal changes in fracture transmissivity in an excavation damaged zone after backfilling a gallery excavated in mudstone

青柳 和平; 石井 英一

Environmental Earth Sciences, 83(3), p.98_1 - 98_15, 2024/02

https://doi.org/10.1007/s12665-023-11416-x

高レベル放射性廃棄物の地層処分において、坑道掘削損傷領域の透水性の経時変化の予測が、掘削損傷領域内部の割れ目を介した放射性核種の移行特性の評価の信頼性向上において重要となる。特に人工バリア定置後の緩衝材の膨潤圧発生に伴う割れ目の透水性の変化を検討することを目的として、幌延深地層研究センターにおいて実施している実規模の人工バリア性能確認試験領域付近で4年間透水試験を実施した。試験の結果、透水性は時間とともに低下し、膨潤から4年経過した時点で、試験開始時よりも透水性が41%に低下することが確認された。さらに、Barton-Bandisのモデルを適用することにより、膨潤圧の変化に伴う透水性の低下挙動を定量的に再現することができた。割れ目の閉塞の妥当性については、人工バリア性能確認試験実施領域で実施した弾性波トモグラフィ調査において、特に底盤部分の弾性波速度の増大が見られたことから確認することができた。これらの結果から、人工バリアの膨潤に伴う掘削損傷割れ目の閉塞について、実証することができた。また、地層処分事業において、廃棄体定置後の緩衝材の膨潤による掘削損傷割れ目の透水性の事前予測に際してBarton-Bandisモデルの適用可能性が示された。

健全性崩壊をもたらす微生物による視認不可腐食の分子生物・電気化学的診断及び抑制技術の開発(委託研究); 令和4年度英知を結集した原子力科学技術・人材育成推進事業

廃炉環境国際共同研究センター; 物質・材料研究機構*

JAEA-Review 2023-031, 101 Pages, 2024/01

JAEA-Review-2023-031.pdf:24.47MB

日本原子力研究開発機構(JAEA)廃炉環境国際共同研究センター(CLADS)では、令和4年度英知を結集した原子力科学技術・人材育成推進事業(以下、「本事業」という。)を実施している。本事業は、東京電力ホールディングス株式会社福島第一原子力発電所(1F)の廃炉等をはじめとした原子力分野の課題解決に貢献するため、国内外の英知を結集し、様々な分野の知見や経験を、従前の機関や分野の壁を越えて緊密に融合・連携させた基礎的・基盤的研究及び人材育成を推進することを目的としている。平成30年度の新規採択課題から実施主体を文部科学省からJAEAに移行することで、JAEAとアカデミアとの連携を強化し、廃炉に資する中長期的な研究開発・人材育成をより安定的かつ継続的に実施する体制を構築した。本研究は、令和2年度に採択された研究課題のうち、「健全性崩壊をもたらす微生物による視認不可腐食の分子生物・電気化学的診断及び抑制技術の開発」の令和2年度から令和4年度分の研究成果について取りまとめたものである。本研究では、微生物が関与して誘起・加速する金属材料の腐食(微生物鉄腐食)に関して、加速検査試験片、オンサイト遺伝子検査などの革新的な診断技術を開発、1Fで想定される微生物腐食の発生・進行条件の洗い出しを行い、水質や環境コントロールによって微生物腐食を防ぐ方法を提案することを目的とした。具体的な研究成果として、微生物腐食反応の加速技術を材料と細菌集積の2つの方向から実現し、採水によるオンサイト加速試験の要素技術開発を行なった。さらに、局所的に激しい腐食を行う細菌の単離に成功し、新しい微生物腐食の抑止技術へと繋がる成果が得られた。さらに、これらの成果の大部分は、春に博士課程に進学した2人の修士課程の学生によって達成されたもので有り、高い学際性を持つリーダー人材の育成に貢献した。

Convergence behavior of statistical uncertainty in probability table for cross section in unresolved resonance region

多田 健一; 遠藤 知弘*

Journal of Nuclear Science and Technology, 60(11), p.1397 - 1405, 2023/11

https://doi.org/10.1080/00223131.2023.2204102

確率テーブルは非分離共鳴領域の自己遮蔽効果を取り扱う手法としてよく知られている手法である。確率テーブルの生成には長い計算時間が必要となる。この確率テーブルの生成時間を減らすには、ラダー数を削減することが効果的である。本研究の目的は、確率テーブルの統計誤差を用いて、最適なラダー数を推定することである。そこで本研究では、確率テーブルの統計誤差の定量化方法を開発し、統計誤差の収束挙動を調査した。確率テーブルと平均断面積の積を、統計誤差の指標と見なした。本研究により、統計誤差の収束率は核種ごとに異なっており、統計誤差の収束率は平均レベル間隔と平均中性子幅の影響を受けることが明らかになった。また、入力パラメータをラダー数から統計誤差の許容値に変更した場合、確率テーブルの生成時間が半分以下になることが分かった。

On the velocity and frequency of disturbance waves in vertical annular flow with different surface tension and gas-liquid density ratio

Zhang, H.*; 梅原 裕太郎*; 吉田 啓之; 森 昌司*

International Journal of Heat and Mass Transfer, 211, p.124253_1 - 124253_13, 2023/09

https://doi.org/10.1016/j.ijheatmasstransfer.2023.124253

In this study, the characteristics of disturbance waves, including velocity, longitudinal size, and frequency, are comprehensively investigated by conducting gas-liquid upward annular flow experiments in a 5 mm tube through the constant electric current method (CECM). The film thickness time trace is recorded concerning time, and the information on disturbance waves is subsequently extracted. The effect of liquid and gas flow rate, density ratio, and surface tension on the disturbance wave is quantitatively analyzed. A predictive model of the wave velocity based on the shear stress balance of the liquid film is proposed. A physical model for wave frequency prediction is derived on the basis of the mass conservation equation. A simple empirical model with good prediction accuracy of wave frequency is also derived. Compared with existing correlations available in the literature, the newly derived models show better performance under a wide range of flow conditions.

Hierarchical Bayesian modeling to quantify fracture limit uncertainty of high-burnup advanced fuel cladding tubes under loss-of-coolant accident conditions

成川 隆文; 濱口 修輔*; 高田 孝*; 宇田川 豊

Nuclear Engineering and Design, 411, p.112443_1 - 112443_12, 2023/09

https://doi.org/10.1016/j.nucengdes.2023.112443

For realizing a highly reliable fracture limit evaluation of fuel cladding tubes during loss-of-coolant accidents (LOCAs) in light-water reactors, we developed a method to quantify the fracture limit uncertainty of high-burnup advanced fuel cladding tubes. This method employs a hierarchical Bayesian model that can quantify uncertainty even with limited experimental data. The fracture limit uncertainty was quantified as a probability using the amount of oxidation (Equivalent cladding reacted: ECR) and the initial hydrogen concentration (the hydrogen concentration in the fuel cladding tubes before the LOCA-simulated tests) as explanatory variables. We divided the regression coefficients of this model into a hierarchical structure with an overall average term common to all types of fuel cladding tubes and a term representing differences among various types of fuel cladding tubes. This hierarchical structure enabled us to quantify the fracture limit uncertainty through the effective use of prior knowledge and data, even for high-burnup advanced fuel cladding tubes with a small number of data points. The fracture limits representing a 5% fracture probability with 95% confidence of the high-burnup advanced fuel cladding tubes evaluated by the hierarchical Bayesian model were higher than 15% ECR for the initial hydrogen concentrations of up to 700-900 wtppm and restraint loads below 535 N. These fracture limits were comparable to the limit of the unirradiated Zircaloy-4 cladding tube, indicating that the burnup extension and use of the advanced fuel cladding tubes do not significantly lower the fracture limit of fuel cladding tubes. Further, we proposed a method to reduce the fracture limit uncertainty by using non-binary data, instead of the binary data, depending on the condition of the fuel cladding tube specimens after performing the LOCA-simulated test, thereby increasing the amount of information in the data.

Verification of probabilistic fracture mechanics analysis code PASCAL for reactor pressure vessel

Lu, K.; 高見澤 悠; Li, Y.; 眞崎 浩一*; 高越 大輝*; 永井 政貴*; 南日 卓*; 村上 健太*; 関東 康祐*; 八代醍 健志*; et al.

Mechanical Engineering Journal (Internet), 10(4), p.22-00484_1 - 22-00484_13, 2023/08

https://doi.org/10.1299/mej.22-00484

A probabilistic fracture mechanics (PFM) analysis code, PASCAL, has been developed by Japan Atomic Energy Agency for failure probability and failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. To strengthen the applicability of PASCAL, considerable efforts on verifications of the PASCAL code have been made in the past years. As a part of the verification activities, a working group consisted of different organizations from industry, universities and institutes, was established in Japan. In the early phase, the working group focused on verifying the PFM analysis functions for RPVs in pressurized water reactors (PWRs) subjected to pressurized thermal shock (PTS) events. Recently, the PASCAL code has been improved in order to run PFM analyses for both RPVs in PWRs and boiling water reactors (BWRs) subjected to a broad range of transients. Simultaneously, the working group initiated a verification plan for the improved PASCAL through independent PFM analyses by different organizations. Concretely, verification analyses for a PWR-type RPV subjected to PTS transients and a BWR-type RPV subjected to a low-temperature over pressure transient were performed using PASCAL. This paper summarizes those verification activities, including the verification plan, analysis conditions and results. Based on the verification studies, the reliability of PASCAL for probabilistic integrity assessments of Japanese RPVs was confirmed with confidence.

The Development of a Multiphysics Coupled Solver for Studying the Effect of Dynamic Heterogeneous Configuration on Particulate Debris Bed Criticality and Cooling Characteristics

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

https://doi.org/10.3390/app13137705

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.

Accident sequence precursor analysis of an incident in a Japanese nuclear power plant based on dynamic probabilistic risk assessment

久保 光太郎

Science and Technology of Nuclear Installations, 2023, p.7402217_1 - 7402217_12, 2023/06

https://doi.org/10.1155/2023/7402217

Probabilistic risk assessment (PRA) is an effective methodology that could be used to improve the safety of nuclear power plants in a reasonable manner. Dynamic PRA, as an advanced PRA allows for more realistic and detailed analyses by handling time-dependent information. However, the applications of this method to practical problems are limited because it remains in the research and development stage. This study aimed to investigate the possibility of utilizing dynamic PRA in risk-informed decision-making. Specifically, the author performed an accident sequence precursor (ASP) analysis on the failure of emergency diesel generators that occurred at Unit 1 of the Tomari Nuclear Power Plant in Japan using dynamic PRA. The results were evaluated by comparison with the results of simplified classical PRA. The findings indicated that dynamic PRA may estimate lower risks compared with those obtained from classical PRA by reasonable modeling of alternating current power recovery. The author also showed that dynamic PRA can provide detailed information that cannot be obtained with classical PRA, such as uncertainty distribution of core damage timing and importance measure considering the system failure timing.