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for brittle fracture in ductile-brittle transition temperature region北条 公伸*; 廣田 貴俊*; 名越 康人*; 深堀 拓也*; 清水 万真*; 下平 昌樹; 小川 琢矢*; 八代醍 健志*; 大畑 充*; 南 二三吉*
Proceedings of ASME 2024 Pressure Vessels & Piping Conference (PVP 2024) (Internet), 9 Pages, 2024/07
加圧熱衝撃事象における延性-脆性遷移温度域の原子炉圧力容器の破壊挙動を予測するため、日本溶接協会規格(WES)として塑性拘束補正係数を導入した評価手法の策定を目指している。WESでは当該評価手法として、簡易法と詳細法の2種類を定める予定である。簡易法による塑性拘束補正係数の算出では、材料の降伏応力、降伏比、ワイブル形状母数をパラメータとした式を用いる。また、塑性拘束補正係数は評価対象の欠陥寸法や構造物の板厚にも依存する。本研究では、様々な原子炉圧力容器を対象として簡易法による塑性拘束補正係数を求めるため、構造物の板厚や亀裂寸法、降伏比やワイブル形状母数を変化させた感度解析を実施した。また、加圧熱衝撃事象は温度変化を伴う事象であることから、ワイブル形状母数等の温度依存性に関する検討も行った。
岩田 景子; 端 邦樹; 飛田 徹; 廣田 貴俊*; 高見澤 悠; 知見 康弘; 西山 裕孝
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 7 Pages, 2021/07
The crack arrest fracture toughness, K
, values for highly-irradiated reactor pressure vessel (RPV) steels are estimated according to a linear relationship between crack arrest toughness reference temperature, T
, and the temperature corresponding to a fixed arrest load, equal to 4 kN, T
, obtained by instrumented Charpy impact test. The relationship between T derived from the instrumented Chrapy impact test and fracture toughness reference temperature, T
, was expressed as an equation proposed in a previous report. The coefficients in the equation could be fine-tuned to obtain a better fitting curve using the present experimental data and previous K data. The K curve for RPV;A533B class1 steels irradiated up to 1.3
10
n/cm
(E 
1 MeV) was compared with a K
curve defined in JEAC4206-2016. It was shown that the K curve was always lower than the 1%ile curve of K for these irradiated RPV steels. This result indicates that the conservativeness of the method defined in JEAC4206-2016 to evaluate K using K curve is confirmed for highly-irradiated RPV steels.
廣田 貴俊*; 名越 康人*; 北条 公伸*; 岡田 裕*; 高橋 昭如*; 勝山 仁哉; 上田 貴志*; 小川 琢矢*; 八代醍 健志*; 大畑 充*; et al.
Proceedings of ASME 2021 Pressure Vessels and Piping Conference (PVP 2021) (Internet), 9 Pages, 2021/07
In order to establish a guideline for fracture evaluation by considering plastic constraint in the ductile-brittle transition temperature (DBTT) region, the CAF (Constraint-Based Assessment of Fracture in Ductile-Brittle Transition Temperature Region) subcommittee has been launched in 2018 in the Japan Welding Engineering Society. In the committee, fracture tests are conducted using C(T), SE(B), and 50mm-thick flat plate with a surface flaw subjected to bending load or tensile load to verify fracture evaluation methods. Since simulation results are easily affected by analysis conditions, benchmark analysis is essential for the potential users of the guideline. Therefore, benchmark analyses are executed on brittle and ductile damages by Beremin and Gurson-Tvergaard-Needleman (GTN) models implemented in the finite element (FE) codes. The benchmark analyses are carried out in four steps; Step 0 is to confirm the output of FE codes in each member with the same input data and the same FE model. Step 1 is to confirm the result of Weibull stress analysis for C(T) specimens tested at -125
C. The Weibull parameter, m, was fixed in this step. At step 2, sensitivity analyses are conducted on Weibull stresses in different conditions. The outputs by the GTN model are also confirmed. At the final step, the fracture simulation will be run for flat plate specimens with less plastic constraint than the standard fracture toughness specimen. As the results of the benchmark analyses up to step 2, a significant difference is not observed in the Weibull stress computed by committee members with the same input data and FE model and it is confirmed that the effects of element type, nonlinear deformation theory employed in FE analysis. For the calculation of the Weibull parameter m by using the fracture toughness test results and the developed programs by committee members, the converged values of m show good agreement among them.
Li, Y.; 廣田 貴俊*; 板橋 遊*; 山本 真人*; 関東 康祐*; 鈴木 雅秀*; 宮本 裕平*
JAEA-Review 2020-011, 130 Pages, 2020/09
JAEA-Review-2020-011.pdf:9.31MB
日本原子力研究開発機構(以下「原子力機構」という。)では、原子炉圧力容器(Reactor Pressure Vessel、以下「RPV」という。)の構造健全性評価手法の高度化を目的として、加圧熱衝撃等の過渡事象が発生した場合のRPVの破損確率や破損頻度を評価する確率論的破壊力学解析コードPASCALを開発し、最新知見に基づきその機能の高度化を進めてきた。RPVの構造健全性評価において確率論的手法の活用が期待される中で、RPVの健全性評価に係る取組みを促進するためには、複数の機関によりPASCALの機能確認を実施し、その確認過程や確認結果を取りまとめておくことにより、コードの信頼性を向上させることが不可欠である。こうした背景を踏まえ、原子力機構では開発機関以外の当該分野に関する専門家の下で、本コードの信頼性を向上させることを目的として、PASCAL信頼性向上ワーキンググループを設立し、PASCALのソースコードレベルの確認を含む機能確認を行ってきた。本報は、PASCAL信頼性向上ワーキンググループの平成28及び29年度における活動内容及び活動結果についてまとめたものである。
Lu, K.; 勝山 仁哉; Li, Y.; 宮本 裕平*; 廣田 貴俊*; 板橋 遊*; 永井 政貴*; 鈴木 雅秀*; 関東 康祐*
Mechanical Engineering Journal (Internet), 7(3), p.19-00573_1 - 19-00573_14, 2020/06
Probabilistic fracture mechanics (PFM) is considered a promising methodology in assessing the integrity of structural components in nuclear power plants because it can rationally represent the influence parameters in their probabilistic distributions without over-conservativeness. In Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 (PFM Analysis of Structural Components in Aging LWRs Version 4) which enables the probabilistic integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. Several efforts have been made to verify PASCAL4 to ensure that this code can provide reliable analysis results. In particular, a Japanese working group, which consists of different participants from the industry and from universities and institutes, has been established to conduct the verification studies. This paper summarizes verification activities of the working group in the past two years. Based on those verification activities, the reliability and applicability of PASCAL4 for structural integrity assessments of Japanese RPVs have been confirmed with great confidence.
Lu, K.; 勝山 仁哉; Li, Y.; 宮本 裕平*; 廣田 貴俊*; 板橋 遊*; 永井 政貴*; 鈴木 雅秀*; 関東 康祐*
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 9 Pages, 2019/05
Probabilistic fracture mechanics (PFM) is considered as a promising methodology in the integrity assessment of structural components in a nuclear power plant since it can rationally represent the influence parameters in their inherent probabilistic distributions without over-conservativeness. In Japan, a PFM analysis code called PASCAL4 (PFM Analysis of Structural Components in Aging LWRs Version 4) has been developed by Japan Atomic Energy Agency, which can be used for structural integrity assessments of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock events. Up till now, many efforts have been made on verifying the PASCAL4 code. Among them, a Japanese working group which is consisted of seven participants from industries, universities and institutes was established to conduct the verification studies. Based on verification activities during the past two years, the reliability and applicability of PASCAL4 for structural integrity assessments of Japanese RPVs were confirmed with great confidence. This paper summarizes the verification activities in this working group including the verification plan, analysis conditions and results.
廣田 貴俊*; 平野 隆*; 鬼沢 邦雄
Proceedings of 2013 ASME Pressure Vessels and Piping Conference (PVP 2013) (DVD-ROM), 7 Pages, 2013/07
破壊靭性マスターカーブ法は、原子炉圧力容器に使用される鋼材の破壊靭性を精度よく統計的に評価するために有効な方法である。国内では、JEAC4216-2011としてマスターカーブ法に対する参照温度(To)評価の規格が制定されている。この規格は、米国ASTM E1921をもとに、国産圧力容器鋼材への適用性の検討結果等をもとに策定された。本研究では、国産鋼材の破壊靭性データベースを用いて、このJEAC4216により求めたToをもとに、平面ひずみ破壊靭性K
曲線及び参照破壊靭性K曲線を推定する代替参照温度RT
を検討した。この際、従来の関連温度RT
に基づく破壊靭性曲線と等価な安全裕度を有するように統計処理を行った。結果として、RTとして、Toに係数C
及びマージンを加える式を開発した。本式は、原子炉圧力容器の健全性評価を規定するJEAC4206の改定に向けて提案される予定である。
