Kazuya Osakabe

Development of Stress Intensity Factors for Deep Surface Cracks in Plates and Cylinders

A number of deep surface cracks have been detected in components of nuclear power plants in recent years. The depths of these cracks are even greater than the half of crack lengths. When a crack is detected during in-service inspections, methods provided in the ASME Boiler and Pressure Vessel Code Section XI or JSME Rules on Fitness-for-Service for Nuclear Power Plants can be used to assess the structural integrity of cracked components. The solution of the stress intensity factor is very important in the assessment of structural integrity. However, in the current codes, the solutions of the stress intensity factor are provided for semi-elliptical surface cracks with a limitation of a/l ≤ 0.5, where a is the crack depth, and l is the crack length.In this study, in order to assess the structural integrity in a more rational manner, the solutions of the stress intensity factor were calculated using finite element analysis with quadratic hexahedron elements for deep semi-elliptical surface cracks in plates, and for axial and circumferential semi-elliptical surface cracks in cylinders. The crack dimensions were focused on the range of a/l = 0.5 to 4.0. Solutions were provided at both the deepest and the surface points of the cracks. Furthermore, some of solutions were compared with the available existing studies and with solutions obtained using finite element analysis with quadratic tetrahedral elements and the virtual crack closure-integral method. As the conclusion, it is concluded that the solutions proposed in this paper are applicable in engineering applications.Copyright

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

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

Structural Integrity Evaluation of Reactor Pressure Vessels During PTS Events Using Deterministic and Probabilistic Fracture Mechanics Analysis

To assess the structural integrity of reactor vessels (RVs) during pressurized thermal shock (PTS) events, a deterministic fracture mechanics (DFM) approach has been widely used such as the procedure in JEAC4206-2007. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of RV has become attractive recently because uncertainties related to input parameters can be incorporated rationally. The probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. In this paper, in order to verify the applicability of a PFM method to JEAC4206-2007, deterministic and probabilistic analyses have been performed, and the effects of initial crack size defined in JEAC4206-2007 on the temperature margin obtained from DFM and the probability of crack initiation obtained from PFM have been evaluated. With regard to the PTS event variation, a stuck open valve scenario (SO) has been considered in addition to large- and small-break loss of coolant accident (LBLOCA, SBLOCA) and main steam line break (MSLB).Copyright

Fully Automated SCC and Fatigue Crack Propagation Analyses on Deep Semi-Elliptical Flaws

In this paper, some results of crack propagation analyses of deep initially semi-elliptical flaws under assumed residual stress fields are presented. The crack propagation analyses were performed by using a software system that has been developed by Okada and his colleagues. It is based on a conventional finite element program but uses the quadratic tetrahedral finite elements to model the structure with the crack. The finite element model with the crack can be generated in an automated manner. The stress-intensity factor computations are performed by using the virtual crack closure-integral method (VCCM) for the quadratic tetrahedral finite element which was also proposed by Okada and his colleagues. The automatic meshing scheme for the crack propagation analyses has also been developed by the authors.By the authors’ previous publication, it was shown that the stress intensity factor of deep semi-elliptical surface flaw under assumed residual stress field reached its maximum value at the mid-depth of the crack. Hence, in present study, in order to investigate the feature of the crack propagation of deep surface cracks, we are conducting crack propagation analyses that can predict the crack extension from each point along the crack front for an arbitrary shaped surface flaw. It can also account for material anisotropy in the crack propagation behavior. Then, the SCC crack propagation analyses for a deep semi-elliptical surface flaw in a plate under assumed residual stress fields are being conducted.The results of the crack propagation analyses suggest that the shapes of the crack after the SCC crack propagation may not be exact semi-elliptic in its shape. In this paper, the analytical procedures and some results are presented. The same analytical procedures can be adopted to perform fatigue crack propagation analyses.Copyright

Stress Intensity Factors for Cracks With Large Aspect Ratio in Cylinder

A number of surface cracks with large aspect ratio have been detected in components of nuclear power plants in recent years. The depths of these cracks are even larger than the half-lengths. The solution of the stress intensity factor is very important for the structural integrity assessment of such cracked components. However, in the current codes, such as ASME Boiler and Pressure Vessel Code Section XI and the JSME Rules on Fitness-for-Service for Nuclear Power Plants, solutions of the stress intensity factors are provided for semi-elliptical surface cracks with a limitation of a/l ≤ 0.5, where a is the crack depth and l is the crack length. In order to assess structural integrity in a more rational way, the authors previously developed solutions of the stress intensity factor for semi-elliptical surface cracks in flat plates with a/l = 0.5 to 4 and a/t = 0.0 to 0.8, where t is the wall thickness. In this study, the solutions of the stress intensity factors were calculated for circumferential and axial surface semi-elliptical cracks with large aspect ratios in cylinders. The geometrical dimensions focused on were in the ranges of a/l = 0.5 to 4, a/t = 0.0 to 0.8 and t/Ri = 0 to 1/2, where t is the wall thickness and Ri is the inner radius of the cylinder. Some solutions were compared with the available existing solutions in order to confirm their applicability.Copyright