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高見澤 悠; 西山 裕孝
Journal of Pressure Vessel Technology, 143(5), p.051502_1 - 051502_8, 2021/10
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)本研究では、照射脆化予測に取込むべき化学成分を特定し、原子炉圧力容器鋼の照射脆化予測の不確実性を評価した。日本の加圧水型原子炉の監視試験データに対してノンパラメトリックベイズ(BNP)法を用いた統計分析を行った。BNP法は入力データの複雑さと不確かさを考慮可能な機械学習手法である。脆化への影響が大きい入力変数の組合せを評価可能な統計的指標を導入し、中性子照射量, Cu, Ni, Si含有量の4つの変数の組合せが脆化予測に最も効果的であることを明らかにした。また、化学成分では脆化量に対してCu含有量の影響が最も大きく、Ni, Siの順番で影響が大きいことを示した。関連温度移行量(
RT
)をBNP法を用いて算出した結果、計算値と実測値の残差の標準偏差は8.4
Cであり、現行の国内脆化予測法(JEAC4201-2007(2013年追補))と同等かそれ以上の予測性有していることを確認した。BNP法によって計算されたRTの事後分布の95%確信区間(入力データの不確実性を考慮した場合にデータが存在し得る範囲)は国内脆化予測法のマージンと同等かそれよりも小さく、JEAC4201-2007(2013年追補)において、適切なマージンが設定されていることを定量的に示した。
Li, Y.; 勝又 源七郎*; 眞崎 浩一; 林 翔太郎*; 板橋 遊*; 永井 政貴*; 鈴木 雅秀*; 関東 康祐*
Journal of Pressure Vessel Technology, 143(4), p.041501_1 - 041501_8, 2021/08
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)It has been recognized that probabilistic fracture mechanics (PFM) is a promising methodology in structural integrity assessments of pressure boundary components of nuclear power plants, because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. A PFM analysis code PASCAL has been developed by the Japan Atomic Energy Agency to evaluate the through-wall cracking frequencies of domestic reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In this study, as a part of the verification activities, a working group was established with seven organizations from industry, universities and institutes. Through one year activities, the applicability of PASCAL for structural integrity assessments of domestic RPVs was confirmed with great confidence. This paper presents the details of the verification activities of the working group.
Lu, K.; 勝山 仁哉; 眞崎 浩一; 渡辺 正*; Li, Y.
Journal of Pressure Vessel Technology, 143(3), p.031704_1 - 031704_8, 2021/06
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)Structural integrity assessment of reactor pressure vessel (RPV) is important for the safe operation of nuclear power plant. For an RPV in a pressurized water reactor (PWR), pressurized thermal shock (PTS) resulted from rapid coolant water injection due to a loss-of-coolant accident is an issue of particular concern. The coolant water temperature in the emergency core cooling system (ECCS) can influence the integrity of RPV subjected to PTS events; thus, this paper is focused on investigating the effect of coolant water temperature of ECCS on failure probability of an RPV. First, thermal-hydraulic (TH) analyses were conducted for a Japanese PWR model plant by using RELAP5, and different coolant water temperatures in ECCS were considered to investigate the effect of coolant water temperature on TH behaviors during a PTS event. Using the TH analysis results, probabilistic fracture mechanics (PFM) analyses were performed for the RPV of the Japanese model plant. Based on the PFM analysis results, the effect of coolant water temperature on failure probability of the RPV was quantified.
Lu, K.; 勝山 仁哉; Li, Y.; 吉村 忍*
Journal of Pressure Vessel Technology, 143(2), p.021505_1 - 021505_8, 2021/04
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)Probabilistic fracture mechanics (PFM) is considered to be a promising methodology in structural integrity assessments of pressure-boundary components in nuclear power plants since it can rationally represent the inherent probabilistic distributions for influence parameters without over-conservativeness. To strengthen the applicability of PFM methodology in Japan, Japan Atomic Energy Agency has developed a PFM analysis code PASCAL4 which enables the failure frequency evaluation of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and thermal transients. PASCAL4 is expected to make a significant contribution to the probabilistic integrity assessment of Japanese RPVs. In this study, PFM analysis for a Japanese model RPV in a pressurized water reactor (PWR) was conducted using PASCAL4, and the effects of nondestructive examination (NDE) and neutron flux reduction on failure frequencies of the RPV were quantitatively evaluated. From the analysis results, it is concluded that PASCAL4 is useful for probabilistic integrity assessments of embrittled RPVs and can enhance the applicability of PFM methodology.
Lu, K.; 勝山 仁哉; Li, Y.
Journal of Pressure Vessel Technology, 142(5), p.051501_1 - 051501_10, 2020/10
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)Structural integrity assessment of reactor pressure vessels (RPVs) is essential for the safe operation of nuclear power plants. For RPVs in pressurized water reactors (PWRs), the assessment should be performed by considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. To assess the structural integrity of an RPV, a traditional method is usually employed by comparing fracture toughness of the RPV material with the stress intensity factor (
) of a crack postulated near the RPV inner surface. When an underclad crack (i.e., a crack beneath the cladding of an RPV) is postulated, of this crack can be increased owing to the plasticity effect of cladding. This is because the yield stress of cladding is lower than that of base metal and the cladding may yield earlier than base metal. In this paper, detailed three-dimensional (3D) finite element analyses (FEAs) were performed in consideration of the plasticity effect of cladding for underclad cracks postulated in Japanese RPVs. Based on the 3D FEA results, a plasticity correction method was proposed on calculations of underclad cracks. In addition, the effects of RPV geometries and loading conditions were investigated using the proposed plasticity correction method. Moreover, the applicability of the proposed method to the case which considers the hardening effect of materials after neutron irradiation was also investigated. All of these results indicate that the proposed plasticity correction method can be used for calculations of underclad cracks and is applicable to structural integrity assessment of Japanese RPVs containing underclad cracks.
山口 義仁; 長谷川 邦夫; Li, Y.
Journal of Pressure Vessel Technology, 142(4), p.041507_1 - 041507_6, 2020/08
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)疲労亀裂進展中における亀裂の開閉口は、亀裂進展速度の評価において重要な現象である。ASME Code Section XIのAppendix A-4300は、負の応力比におけるフェライト鋼の疲労亀裂進展速度を算出する式について、負荷の大きさに応じて二つ提示している。一つは、負荷が小さい場合に、亀裂の閉口を考慮する式である。もう一つは、負荷が大きい場合に、亀裂の閉口を考慮しない式である。本研究では、フェライト鋼に対して、負荷の大きさを徐々に変えながら疲労亀裂進展試験を実施し、負荷の大きさが亀裂閉口に及ぼす影響を調査した。その結果、Appendix A-4300における疲労亀裂進展速度算出式を切り替える負荷の大きさと比較して、より小さい負荷で亀裂が閉口することを明らかにした。
長谷川 邦夫; Li, Y.; Lacroix, V.*; Mares, V.*
Journal of Pressure Vessel Technology, 142(3), p.031506_1 - 031506_7, 2020/06
被引用回数：1 パーセンタイル：34.12(Engineering, Mechanical)周方向に亀裂を有する管の曲げによる塑性崩壊応力はASME Section XI規格のAppendix Cで用意されている極限荷重評価法で推定される。この推定式は管の内外表面の亀裂に適用される。しかし、近年著者らが開発した精度の高い式によると、管の外表面にある亀裂の管は、塑性崩壊応力は小さく、Appendix Cの式を用いると非安全側になることを報告してきた。本報はこの精度の高い式を用いてAppendix Cの式の適用限界を明らかにした。
山口 義仁; 勝山 仁哉; Li, Y.; 鬼沢 邦雄
Journal of Pressure Vessel Technology, 142(2), p.021906_1 - 021906_11, 2020/04
被引用回数：1 パーセンタイル：34.12(Engineering, Mechanical)Some Japanese nuclear power plants have experienced several large earthquakes beyond the design basis ground motion. In addition, cracks resulting from long-term operation have been detected in piping systems. Therefore, to assess the structure integrity of cracked pipes taking the occurrence of large earthquakes into account, it is very important to establish a crack growth evaluation method for cracked pipes that are subjected to large seismic cyclic response loading. In our previous study, we proposed an evaluation method for crack growth during large earthquakes through experimental study using small specimens and investigation using finite element analyses. In the present study, to confirm applicability of the proposed method, crack growth tests were conducted on both stainless and carbon steel pipe specimens with a circumferential through-wall crack, considering large seismic cyclic response loading with complex wave forms. The predicted crack growth values are in good agreement with the experimental results and the applicability of the proposed method was confirmed.
東 喜三郎*; Li, Y.; Xu, S.*
Journal of Pressure Vessel Technology, 142(2), p.021207_1 - 021207_10, 2020/04
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)Alloy 82/182/600, which is used in light-water reactors, is known to be susceptible to stress-corrosion cracking. The depth of some of these cracks may exceed the value of half-length on the surface. Although the stress intensity factor (SIF) for cracks plays an important role in predicting crack propagation and failure, Section XI of the ASME Boiler and Pressure Vessel Code does not provide SIF solutions for such deep cracks. In this study, closed-form SIF solutions for deep surface cracks in plates are discussed using an influence coefficient approach. The stress distribution at the crack location is represented by a fourth-degree-polynomial equation. Tables for influence coefficients obtained by finite element analysis in the previous studies are used for curve fitting. The closed-form solutions for the influence coefficients were developed at the surface point, the deepest point, and the maximum point of a crack with an aspect ratio a/c ranging from 1.0 to 8.0, where a is the crack depth and c is one-half of the crack length. The maximum point of a crack refers to the location on the crack front where the SIF reaches a maximum value.
Lu, K.; 勝山 仁哉; Li, Y.
Journal of Pressure Vessel Technology, 142(2), p.021208_1 - 021208_11, 2020/04
被引用回数：3 パーセンタイル：68.53(Engineering, Mechanical)In Japan, a probabilistic fracture mechanics (PFM) analysis code PASCAL was developed for structural integrity assessment of reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. By reflecting the latest knowledge and findings, the evaluation functions are continuously improved and have been incorporated into PASCAL4 which is the most recent version of the PASCAL code. In this paper, the improvements made in PASCAL4 are explained in detail, such as the evaluation model of warm prestressing (WPS) effect, evaluation function of confidence levels for PFM analysis results by considering the epistemic and aleatory uncertainties in probabilistic variables, the recent stress intensity factor (KI) solutions, and improved methods for KI calculations when considering complicated stress distributions. Moreover, using PASCAL4, PFM analysis examples considering these improvements are presented.
勝山 仁哉; 小坂部 和也*; 宇野 隼平*; Li, Y.; 吉村 忍*
Journal of Pressure Vessel Technology, 142(2), p.021205_1 - 021205_10, 2020/04
被引用回数：1 パーセンタイル：34.12(Engineering, Mechanical)確率論的破壊力学(PFM)に基づく構造健全性評価手法は、経年劣化に関連する様々な因子の確率分布を考慮して原子炉圧力容器(RPV)の破損頻度を評価できる合理的な手法である。我々は、中性子照射脆化や加圧熱衝撃事象(PTS)事象を考慮してRPVの破損頻度を評価するPFM解析コードPASCALを開発してきた。また、国内におけるPFMの適用性向上を図るため、破壊力学に関する知識を有する解析者がそれを参照することでPFM解析を行い亀裂貫通頻度を評価できるよう、標準的解析要領を整備した。本要領は、本文、解説及び付属書で構成されており、PFM解析に関する技術的根拠や最新知見が取りまとめられたものになっている。本論では、本要領の概要について述べるとともに、本要領とPTS評価に関する国内データベースに基づき得られた国内モデルRPVに対する破損頻度の評価結果について述べる。
長谷川 邦夫; Li, Y.; Kim, Y.-J.*; Lacroix, V.*; Strnadel, B.*
Journal of Pressure Vessel Technology, 141(3), p.031201_1 - 031201_5, 2019/06
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)2個の欠陥が近接して存在する場合は、これらの欠陥は1つの欠陥に合体される。この合体評価は多くの国の維持規格に採用されているが、具体的な合体クライテリアは異なっている。一方、2個の周方向欠陥を有するステンレス鋼配管の曲げ試験が行われており、塑性崩壊荷重は求められている。また、解析的な式も導かれている。本研究では、これらの実験と解析結果から得られる塑性崩壊応力を合体クライテリアから導かれる塑性崩壊応力と比較した。その結果、合体クライテリアを用いた塑性崩壊応力は、実験や解析結果と極めて異なることが分かった。
Mares, V.*; 長谷川 邦夫; Li, Y.; Lacroix, V.*
Journal of Pressure Vessel Technology, 141(2), p.021203_1 - 021203_6, 2019/04
被引用回数：2 パーセンタイル：29.38(Engineering, Mechanical)周方向に内外表面亀裂を有する管の塑性崩壊応力は、米国機械学会のボイラーと圧力容器の規格のSection XIのAppendix Cで推定式が記載されている。このAppendix Cの推定式は欠陥形状が同じであれば内外表面亀裂の塑性崩壊応力は同じである。われわれは、管の平均半径を欠陥面と欠陥以外の面の2つの平均半径を考慮し、内外表面亀裂を有する管の塑性崩壊応力を導いた。その結果、外表面欠陥の塑性崩壊応力は、管の厚さが大きく亀裂が深くて長いとき、Appendix Cの推定式は大きく、非安全側になることが分かった。
長谷川 邦夫*; Strnadel, B.*; Li, Y.; Lacroix, V.*
Journal of Pressure Vessel Technology, 140(5), p.051204_1 - 051204_7, 2018/10
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)Subsurface flaws are sometimes found as blowholes near free surfaces of structural components. It can be easily expected that the stress intensity factor at the tip of the subsurface flaw increases with decreasing the ligament distance. Fitness-for-service (FFS) codes provide flaw-to-surface proximity rules which are transformation from subsurface to surface flaw. Although the concept of the proximity rules of the FFS codes are the same, the specific criteria for the rules on transforming subsurface flaws to surface flaws are different amongst FFS codes. This study demonstrates the proximity criteria provided by the FFS codes and indicates that the increment of the stress intensity factors before and after the transformation from subsurface to surface flaws. In addition, it is shown that remaining fatigue lives for pipes with flaws are strongly affected by the location at the transformation from subsurface to surface flaws.
河 侑成; 飛田 徹; 大津 拓与; 高見澤 悠; 西山 裕孝
Journal of Pressure Vessel Technology, 140(5), p.051402_1 - 051402_6, 2018/10
被引用回数：1 パーセンタイル：11.69(Engineering, Mechanical)The applicability of miniature compact tension (Mini-C(T)) specimens to fracture toughness evaluation of neutron-irradiated reactor pressure vessel (RPV) steels was investigated. Three types of RPV steels neutron-irradiated to a high-fluence region were prepared and manufactured as Mini-C(T) specimens according to Japan Electric Association Code (JEAC) 4216-2015. Through careful selection of the test temperature by considering previously obtained mechanical properties data, valid fracture toughness, and reference temperature T
was obtained with a relatively small number of specimens. Comparing the fracture toughness and T values determined using other larger specimens with those determined using the Mini-C(T) specimens, T values of both unirradiated and irradiated Mini-C(T) specimens were found to be the acceptable margin of error. The scatter of 1T-equivalent fracture toughness values of both unirradiated and irradiated materials obtained using Mini-C(T) specimens did not differ significantly from the values obtained using larger specimens. The correlation between the Charpy 41J transition temperature (T
) and the T values agreed very well with that of the data in the literature, regardless of specimen size and fracture toughness of the materials before irradiation. Based on these findings, it was concluded that Mini-C(T) specimens can be applied to fracture toughness evaluation of neutron-irradiated materials without significant specimen size dependence.
Lu, K.; 真野 晃宏; 勝山 仁哉; Li, Y.; 岩松 史則*
Journal of Pressure Vessel Technology, 140(3), p.031201_1 - 031201_11, 2018/06
被引用回数：7 パーセンタイル：57.86(Engineering, Mechanical)The stress intensity factor (SIF) solutions for subsurface flaws near the free surfaces of components, which are known to be important in engineering applications, have not been provided yet. Thus, in this paper, SIF solutions for subsurface flaws near the free surfaces in flat plates were numerically investigated based on finite element analyses. The flaws with aspect ratios a/l = 0.0, 0.1, 0.2, 0.3, 0.4 and 0.5, the normalized ratios a/d = 0.0, 0.1, 0.2, 0.4, 0.6 and 0.8, and d/t = 0.01 and 0.10 were taken into account, where a is the half flaw depth, l is the flaw length, d is the distance from the center of the subsurface flaw to the nearest free surface and t is the wall thickness. Fourth-order polynomial stress distribution in the thickness direction was considered. In addition, the developed SIF solutions were incorporated into a Japanese probabilistic fracture mechanics (PFM) code, and PFM analyses were performed for a Japanese reactor pressure vessel containing a subsurface flaw near the inner surface. The PFM analysis results indicate that the obtained SIF solutions are effective in engineering applications.
安藤 勝訓; 矢田 浩基; 月森 和之; 一宮 正和*; 安濃田 良成*
Journal of Pressure Vessel Technology, 139(6), p.061201_1 - 061201_12, 2017/08
被引用回数：1 パーセンタイル：11.11(Engineering, Mechanical)本研究では、内圧の影響を受けたベローズ構造の到達圧力の評価方法を開発するために、ベローズ構造の破壊試験と有限要素解析(FEA)を行った。一連の試験により破壊モードを実証し、3種類の破壊モードを確認した。試験中の座屈および変形挙動をシミュレートするために、陰解法および陽解法による解析を実施し、試験結果と比較した。
長谷川 邦夫*; Mares, V.*; 山口 義仁
Journal of Pressure Vessel Technology, 139(3), p.034501_1 - 034501_5, 2017/06
被引用回数：2 パーセンタイル：19.31(Engineering, Mechanical)Reference curves of fatigue crack growth rates for ferritic steels in air environment are provided by the ASME Code Section XI Appendix A. The fatigue crack growth rates under negative R ratio are given as da/dN vs. K
. It is generally well known that the growth rates decreases with decreasing R ratios. However, the da/dN as a function of K are the same curves under R = 0, -1 and -2. In addition, the da/dN increases with decreasing R ratio for R
-2. This paper converts from da/dN vs. K to da/dN vs. K
, using crack closure U. It can be seen that the growth rate da/dN vs. K is better equation than da/dN vs. K from the view point of stress ratio R. Furthermore, extending crack closure U to R = -5, it can be explained that the da/dN decreases with decreasing R ratio in the range of -5 R 0. This tendency is consistent with the experimental data.
Li, Y.; 長谷川 邦夫*; 宇田川 誠
Journal of Pressure Vessel Technology, 139(2), p.021202_1 - 021202_13, 2017/04
被引用回数：1 パーセンタイル：11.11(Engineering, Mechanical)The stress intensity factors (SIFs) for pipes containing semi-elliptical cracks with large aspect ratios were calculated by finite element analysis. The cracks were circumferential and axial surface cracks inside the pipes. The parameters of the SIFs are crack aspect ratio, crack depth and the ratio of pipe radius to wall thickness. In comparing SIFs for plates and pipes, it can be clarified that SIFs for both plates and thin pipes with 
1/10 are almost the same, and the SIFs for plates are used as a substitute for pipes with 1/10, where 
is the pipe wall thickness and 
is the inner radius of the pipe. This means that it is not necessary to provide SIF solutions for pipes with 1/10, and it is suggested that number of tables for influence coefficients G values for pipes can significantly reduce.
Li, Y.; 長谷川 邦夫; 三浦 直樹*; 星野 克明*
Journal of Pressure Vessel Technology, 139(2), p.021203_1 - 021203_7, 2017/04
被引用回数：0 パーセンタイル：0(Engineering, Mechanical)原子力配管系の重要な荷重条件は、内圧,曲げ荷重及びねじり荷重である。しかし、これらの荷重条件を全て考慮した減肉配管の破壊評価手法は確立されていない。われわれはこれまでに有限要素法による極限荷重解析により、これらの多軸荷重条件が負荷された場合の局部減肉を有する配管の破壊評価手法を提案した。本研究では、局部減肉を有する炭素鋼小型配管試験体を対象に内圧による引張荷重,曲げ荷重及びねじり荷重を同時に負荷した破壊試験を実施した。これまでに提案した破壊評価手法により予測した破壊荷重は試験結果とよく一致したことから、提案手法の妥当性が確認できた。
