Genshichiro Katsumata

Remaining Lives of Fatigue Crack Growths for Pipes With Subsurface Flaws and Subsurface-to-Surface Flaw Proximity Rules

If a subsurface flaw is located near a component surface, the subsurface flaw is transformed into a surface flaw in accordance with a subsurface-to-surface flaw proximity rule. The recharacterization process from subsurface to surface flaw is adopted in all fitness-for-service (FFS) codes. However, the specific criteria of the recharacterizations are different among the FFS codes. Recently, the authors have proposed a new subsurface-to-surface flaw proximity rule based on experimental data and equivalent fatigue crack growth rate calculations. In this study, fatigue crack growth calculations were carried out for pipes with subsurface flaws, using the proximity rule provided in the current ASME (American Society of Mechanical Engineers) Section XI and JSME (The Japan Society of Mechanical Engineers) codes and the proposed subsurface-to-surface flaw proximity rule. Different pipe sizes, flaw aspect ratios, and ligament distances from subsurface flaws to inner surface of pipes were taken into account. The results indicate the current proximity rule gives less conservative fatigue lives, when the aspect ratios of the subsurface flaws are small.

Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL

1. 緒言 国内において確率論的破壊力学 (PFM) の適用性向上を図るためには、破損頻度の算出に用いら れる PFM 解析コードの検証が不可欠である。著者らは、原子力機構が整備を進めている原子炉圧力容器 (RPV) に対する PFM 解析コード PASCAL に対する検証の一環として、PASCAL に導入した機能の検証を 行うとともに、米国の PFM 解析コード FAVORとのベンチマーク解析を実施した。 2. PASCALの機能整備 PASCALの適用性向上を目的として、条件付亀裂進展確率 (CPI) 及び条件付亀裂 貫通確率 (CPF) に関する低損傷確率評価機能、認識論的不確実さを考慮した信頼度評価機能等を PASCAL に導入した。例えば、信頼度評価機能の整備では、これまで不確実さの分類を考慮せずにモンテカルロ法 により CPI 及び CPFを算出していたのに対し、二重ループ化する改良を行った。具体的には、内側のルー プで偶然的不確実さのみを考慮して求めた CPIと CPFを、外側のループで認識論的不確実さに基づき統計 処理することにより、前報で示したとおり破損頻度の信頼度を評価できるようにした。また、整備した全 ての機能について検証を実施し、適切に動作していることを確認した。 3.ベンチマーク解析 PASCAL の信頼性向上を図る ため、FAVOR との比較計算を実施した。PASCALと FAVOR の機能の相違を確認した上で、単一亀裂に対 する CPI及び CPF の比較、RPV 炉心領域に対する亀 裂進展頻度 (FCI) 及び亀裂貫通頻度 (TWCF) の比 較等を実施した。PASCAL と FAVOR とでは開発機 関が異なるため、個々の機能には違いがあるが、本 比較計算を通じて、RPV 寸法や化学成分、応力拡大 係数等の条件を合わせた場合にはほぼ同等の結果が 得られることを確認した (図1)。 4. 結論 PASCAL に整備した個々の機能検証、及び PASCAL と FAVOR のベンチマーク解析を通じて、 PASCALの信頼性を検証し、実用性の向上を図った。 5. 謝辞 本報告は、原子力規制庁からの受託事業「平成28年度高経年化技術評価高度化事業(原子炉一 次系機器の健全性評価手法の高度化)」で得られた成果である。関係各位に謝意を表する。

Development of Probabilistic Evaluation Models of Fracture Toughness KIc and KIa for Japanese RPV Steels

Probabilistic fracture mechanics (PFM) analysis code PASCAL3 has been developed to apply the PFM analysis to the structural integrity assessment of domestic reactor pressure vessels (RPVs). In this paper, probabilistic evaluation models of fracture toughness KIc and KIa which have the largest scatter among the associated factors based on the database of Japanese RPV steels are presented. We developed probabilistic evaluation models for KIc and KIa based on the Weibull and lognormal distributions, respectively. The models are compared with the existing lower bound of fracture toughness in the Japanese code and probabilistic model in USA. As the results, the 5% confidence limits of the models established in present work corresponded to lower bounds of fracture toughness in the Japanese code. The comparison in the models between present work and USA showed significant differences that may have an influence on fracture probability of RPV.Copyright

Study on Application of PFM Analysis Method to Japanese Code for RPV Integrity Assessment Under PTS Events

A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.Copyright

Estimation of Through-Wall Cracking Frequency of RPV Under PTS Events Using PFM Analysis Method for Identifying Conservatism Included in Current Japanese Code

To assess the structural integrity of reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events, the deterministic fracture mechanics approach prescribed in Japanese code JEAC 4206-2007 [1] has been used in Japan. The structural integrity is judged to be maintained if the stress intensity factor (SIF) at the crack tip during PTS events is smaller than fracture toughness KIc. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of pressure components has become attractive recently because uncertainties related to influence parameters can be incorporated rationally. A probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated taking frequencies of event occurrence and crack arrest after crack initiation into consideration.In this study, in order to identify the conservatism in the current RPV integrity assessment procedure in the code, probabilistic analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007, that a semi-elliptic axial crack is postulated on the inside surface of RPV wall, is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.Copyright

Study on Structural Integrity Assessment of Reactor Pressure Vessel Based on Three-Dimensional Thermal-Hydraulics and Structural Analyses

In the structural integrity assessment of a pressurized water reactor pressure vessel (RPV) during pressurized thermal shock (PTS) events, the thermal history of the coolant water and the heat transfer coefficient between the coolant water and RPV are dominant factors. These values can be determined on the basis of thermal-hydraulics (TH) analysis simulating PTS events and Jackson-Fewster correlation. Subsequently, using these values, structural integrity assessments of RPV are performed by structural analysis; e.g., loading that affects crack propagation is evaluated. Three-dimensional TH and structural analyses are recommended for precise assessments of the structural integrity of RPV. In this study, we performed TH and structural analyses simulating typical PTS events using three-dimensional models of cold-leg, downcomer and RPV to more accurately assess the structural integrity of RPV. From these analyses, we obtained loading histories from the reactor core region of RPV in which a crack is postulated in the structural integrity assessment. We discuss the conservativeness of current analysis methods on the structural integrity assessment of RPV through the comparison of loading conditions due to PTS events.Copyright

Benchmark Analysis on Probabilistic Fracture Mechanics Analysis Codes Considering Multiple Cracks and Crack Initiation in Aged Piping of Nuclear Power Plants

In recent years, cracks have been detected in piping systems of nuclear power plants. Many of them are multiple cracks in the same welded joints. Therefore, structural integrity evaluation and risk assessment considering multiple cracks and crack initiation in aged piping have become increasingly important. Probabilistic fracture mechanics (PFM) is a rational methodology in structural integrity evaluation and risk assessment of aged piping in nuclear power plants. Two PFM codes, PASCAL-SP and PRAISE-JNES, have been improved or developed in Japan for the structural integrity evaluation and risk assessment considering the age related degradation mechanisms of pipes. Although the purposes to develop these two codes are different, both have almost the same basic functions to obtain the failure probabilities of pipes. In this paper, a benchmark analysis was conducted considering multiple cracks and crack initiation, in order to confirm their reliability and applicability. Based on the numerical investigation in consideration of important influence factors such as crack number, crack location, crack distribution and crack detection probability of in-service inspection, it was concluded that the analysis results of these two codes are in good agreement.Copyright