Study on Life Extension of Aged RPV Material Based on Probabilistic Fracture Mechanics: Japanese Round Robin

1995 ◽  
Vol 117 (1) ◽  
pp. 7-13 ◽  
Author(s):  
G. Yagawa ◽  
S. Yoshimura ◽  
N. Handa ◽  
T. Uno ◽  
K. Watashi ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Surface Crack ◽  
Round Robin ◽  
Life Extension ◽  
Loading Conditions ◽  
Pressurized Water ◽  
Unique Problem ◽  
Probabilistic Fracture Mechanics ◽  
Failure Probabilities

This paper is concerned with round-robin analyses of probabilistic fracture mechanics (PFM) problems of aged RPV material. Analyzed here is a plate with a semi-elliptical surface crack subjected to various cyclic tensile and bending stresses. A depth and an aspect ratio of the surface crack are assumed to be probabilistic variables. Failure probabilities are calculated using the Monte Carlo methods with the importance sampling or the stratified sampling techniques. Material properties are chosen from the Marshall report, the ASME Code Section XI, and the experiments on a Japanese RPV material carried out by the Life Evaluation (LE) subcommittee of the Japan Welding Engineering Society (JWES), while loads are determined referring to design loading conditions of pressurized water reactors (PWR). Seven organizations participate in this study. At first, the procedures for obtaining reliable PFM solutions with low failure probabilities are examined by solving a unique problem with seven computer programs. The seven solutions agree very well with one another, i.e., by a factor of 2 to 5 in failure probabilities. Next, sensitivity analyses are performed by varying fracture toughness values, loading conditions, and pre and in-service inspections. Finally, life extension simulations based on the PFM analyses are performed. It is clearly demonstrated from these analyses that failure probabilities are so sensitive to the change of fracture toughness values that the degree of neutron irradiation significantly influences the judgment of plant life extension.

Benchmarking Study on Probabilistic Fracture Mechanics Analysis Codes xLPR and PASCAL-SP Considering Primary Water Stress Corrosion Cracking

2021 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Water Stress ◽  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Structural Integrity ◽  
Corrosion Cracking ◽  
Pressurized Water ◽  
Probabilistic Fracture Mechanics ◽  
Primary Water ◽  
Failure Probabilities

Abstract Probabilistic fracture mechanics (PFM) is expected as a more rational methodology for the structural integrity assessments of nuclear power components because it can consider the inherent probabilistic distributions of various influencing factors and quantitatively evaluate the failure probabilities of the components. The Japan Atomic Energy Agency (JAEA) has developed a PFM analysis code, PASCAL-SP, to evaluate the failure probabilities of piping caused by aging degradation mechanisms, such as fatigue and stress corrosion cracking in the environments of both pressurized water and boiling water reactors. To improve confidence in the analysis results obtained from PASCAL-SP, a benchmarking study was conducted together with the PFM analysis code, xLPR, which was developed jointly by the U.S. Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute. The benchmarking study was composed of deterministic and probabilistic analyses related to primary water stress corrosion cracking in a dissimilar metal weld joint in a pressurized water reactor surge line. The analyses were conducted independently by NRC staff and JAEA using their own codes and under common analysis conditions. In the present paper, the analysis conditions for the deterministic and probabilistic analyses are described in detail, and the analysis results obtained from the xLPR and PASCAL-SP codes are presented. It was confirmed that the analysis results obtained from the two codes were in good agreement.

Probabilistic Fracture Mechanics Application for Alloy 82/182 Welds in PWRS

2010 ◽  
Author(s):  
Jong-Dae Hong ◽  
Changheui Jang
Keyword(s):  
Fracture Mechanics ◽  
Water Chemistry ◽  
Power Plants ◽  
Structural Integrity ◽  
Boundary Carbide ◽  
Pressurized Water ◽  
Butt Weld ◽  
Monte Carlo Simulation Technique ◽  
Probabilistic Fracture Mechanics ◽  
Failure Probabilities

In operating PWRs (Pressurized Water Reactors), incidents of Alloy 82/182 cracking increased the concern for structural integrity of butt weld locations recently, because of high weld residual stresses. Studies on PWSCC (Primary Water Stress Corrosion Cracking) have been mainly performed using deterministic approaches by controlling parameters, but a quantitative evaluation is difficult because of large uncertainties in each parameter and test results. The purposes of this paper are to provide a probabilistic fracture mechanics (PFM) analysis methodology and quantify failure probabilities for Alloy 82/182 welds in primary piping systems of nuclear power plants. To calculate failure probabilities, Monte Carlo simulation technique was used. To estimate the time to crack initiation, material susceptibility was quantified considering the effects of various processing, grain boundary carbide coverage, water chemistry including zinc addition, and so on. In crack growth analysis, crack orientation and the effects of water chemistry including dissolved hydrogen concentration were considered. And the effects of weld repair were evaluated.

Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP With Regardto PWSCC

2019 ◽  
Vol 5 (3) ◽  
Author(s):  
Akihiro Mano ◽  
Yoshihito Yamaguchi ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Failure Probability ◽  
Power Plants ◽  
Crack Detection ◽  
Structural Integrity ◽  
Energy Agency ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Probabilistic Fracture Mechanics ◽  
Failure Probabilities

Probabilistic fracture mechanics (PFM) analysis is expected to be a rational method for structural integrity assessment because it can consider the uncertainties of various influence factors and evaluate the quantitative values such as failure probability of a cracked component as the solution. In the Japan Atomic Energy Agency, a PFM analysis code PASCAL-SP has been developed for structural integrity assessment of piping welds in nuclear power plants (NPP). In the past few decades, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Thus, structural integrity assessments considering PWSCC have become important. In this study, PASCAL-SP was improved considering PWSCC by introducing several analytical functions such as the models for evaluation of crack initiation time, crack growth rate (CGR), and probability of crack detection. By using the improved version of PASCAL-SP, the failure probabilities of pipes with a circumferential crack or an axial crack due to PWSCC were numerically evaluated. Moreover, the influence of leak detection and nondestructive examination (NDE) on failure probabilities was detected. Based on the obtained numerical results, it was concluded that the improved version of PASCAL-SP is useful for evaluating the failure probability of a pipe considering PWSCC.

Effects of Fracture Toughness Curves of ASME Section XI: Appendix G on a Reactor Pressure Vessel Under Pressure–Temperature Limit Operation

2013 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Kuan-Rong Huang ◽  
Ru-Feng Liu
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Temperature Limit ◽  
Reactor Pressure Vessel ◽  
Probabilistic Fracture Mechanics ◽  
Sufficient Stability ◽  
Reactor Pressure ◽  
Failure Probabilities

After the Code Case N-640 was issued in 1999, the fracture toughness curve of reactor pressure vessel materials in ASME Section XI-Appendix G was amended to the KIC curve. In Taiwan, the present pressure-temperature limit curves of normal reactor startup (heat-up) and shut-down (cool-down) for the reactor pressure vessel is still calculated per KIA curve in 1998 or earlier editions. In this paper, the failure risks of a Taiwan domestic reactor pressure vessel under various pressure-temperature limit operations are analyzed. First, the pressure-temperature limit curves of the Taiwan domestic reactor pressure vessel based on KIA and KIC curves, and various levels of embrittlement, are calculated. Then, the ORNL’s probabilistic fracture mechanics code, FAVOR, and the PNNL’s flaw model are utilized to assess the failure probabilities of the reactor pressure vessel under such pressure-temperature limit transients. Further, the deterministic analyses of FAVOR code are also conducted. It is found that under the pressure-temperature limit transients based on KIC curves, the reactor pressure vessel presents higher failure probabilities, but are all below the allowable risk. The present results indicate that using the KIC curve the pressure-temperature limits can either increase the operational margin or still maintains the sufficient stability of the analyzed reactor pressure vessel.

Interaction of a Crack in the Plasma-Sprayed Ceramic Coating with the Metal Substrate

2005 ◽  
Vol 482 ◽  
pp. 223-226
Author(s):  
Luboš Náhlík ◽  
Zdeněk Knésl ◽  
F. Kroupa
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Surface Crack ◽  
Ceramic Coating ◽  
Ceramic Coatings ◽  
Metal Substrate ◽  
High Density ◽  
Initial Surface ◽  
Young’S Moduli ◽  
Plasma Sprayed

Plasma-sprayed ceramic coatings contain a high density of intrasplat microcracks which are responsible for small Young’s moduli and low fracture toughness. The extension of an initial surface crack in the direction to the interface, where the crack is repelled by the metal substrate with higher Young’s modulus, is studied using the methods of fracture mechanics. It is shown that high tensile stresses induced by the crack in the interface can lead to a local decohesion along the interface so that the crack can deviate into the interface.

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

2020 ◽  
Author(s):  
Akihiro Mano ◽  
Jinya Katsuyama ◽  
Yinsheng Li
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Corrosion Cracking ◽  
Energy Agency ◽  
Pressurized Water ◽  
Water Reactor ◽  
Probabilistic Fracture Mechanics

Abstract A probabilistic fracture mechanics (PFM) analysis code, PASCAL-SP, has been developed by Japan Atomic Energy Agency (JAEA) to evaluate the failure probability of piping within nuclear power plants considering aged-related degradations such as stress corrosion cracking and fatigue for both pressurized water reactor and boiling water reactor environments. To strengthen the applicability of PASCAL-SP, a benchmarking study is being performed with a PFM analysis code, xLPR, which has been developed by U.S.NRC in collaboration with EPRI. In this benchmarking study, deterministic and probabilistic analyses are undertaken on primary water stress corrosion cracking using the common analysis conditions. A deterministic analysis on the weld residual stress distributions is also considered. These analyses are carried out by U.S.NRC and JAEA independently using their own codes. Currently, the deterministic analyses by both xLPR and PASCAL-SP codes have been finished and probabilistic analyses are underway. This paper presents the details of conditions and comparisons of the results between the two aforementioned codes for the deterministic analyses. Both codes were found to provide almost the same results including the values of stress intensity factor. The conditions and results of the probabilistic analysis obtained from PASCAL-SP are also discussed.

scholarly journals Probabilistic Fracture Mechanics Round Robin Analysis of Reactor Pressure Vessels during Pressurized Thermal Shock

2010 ◽  
Vol 47 (12) ◽  
pp. 1131-1139 ◽  
Author(s):  
Myung Jo JHUNG ◽  
Seok Hun KIM ◽  
Young Hwan CHOI ◽  
Yoon Suk CHANG ◽  
Xiangyuan XU ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Round Robin ◽  
Pressure Vessels ◽  
Probabilistic Fracture Mechanics ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

Advanced Probabilistic Fracture Mechanics Using the R6 Procedure

2010 ◽  
Author(s):  
David W. Beardsmore ◽  
Karen Stone ◽  
Huaguo Teng
Keyword(s):  
Fracture Toughness ◽  
Monte Carlo ◽  
Fracture Mechanics ◽  
Monte Carlo Methods ◽  
Probability Of Failure ◽  
Defect Size ◽  
Probabilistic Fracture Mechanics ◽  
Safety Margins ◽  
Reliability Method ◽  
Failure Assessment

Deterministic Fracture Mechanics (DFM) assessments of structural components (e.g. pressure vessels and piping used in the nuclear industry) containing defects can usually be carried out using the R6 procedure. The aim of such an assessment is to demonstrate that there are sufficient safety margins on the applied loads, defect size and fracture toughness for the safe continual operation of the component. To ensure a conservative assessment is made, a lower-bound fracture toughness, and upper-bound defect sizes and applied loads are used. In some cases, this approach will be too conservative and will provide insufficient safety margins. Probabilistic Fracture Mechanics (PFM) allow a way forward in such cases by allowing for the inherent scatter in material properties, defect size and applied loads explicitly. Basic Monte Carlo Methods (MCM) allow an estimate of the probability of failure to be calculated by carrying out a large number of fracture mechanics assessments, each using a random sample of the different random variables (loads, defect size, fracture toughness etc). The probability of failure is obtained by counting the proportion of simulations which lead to assessment points that lie outside the R6 failure assessment curve. This approach can give good results for probabilities greater than 10−5. However, for smaller probabilities, the calculation may be inefficient and a very large number of assessments may be necessary to obtain an accurate result, which may be prohibitive. Engineering Reliability Methods (ERM), such as the First Order Reliability method (FORM) and the Second Order Reliability Method (SORM), can be used to estimate the probability of failure in such cases, but these methods can be difficult to implement, do not always give the correct result, and are not always robust enough for general use. Advanced Monte Carlo Methods (AMCM) combine the two approaches to provide an accurate and efficient calculation of probability of failure in all cases. These methods aim to carry out Importance Sampling so that only assessment points that lie close to or outside the failure assessment curve are calculated. Two methods are described in this paper: (1) orthogonal sampling, and (2) spherical sampling. The power behind these methods is demonstrated by carrying out calculations of probability of failure for semi-elliptical, surface breaking, circumferential cracks in the inside of a pressure vessel. The results are compared with the results of Basic Monte Carlo and Engineering Reliability calculations. The calculations use the R6 assessment procedure.

Parametric Study on Pressure-Temperature Limit Curve

2016 ◽  
Author(s):  
Shin-Beom Choi ◽  
Han-Bum Surh ◽  
Jong-Wook Kim
Keyword(s):  
Sensitivity Analysis ◽  
Fracture Mechanics ◽  
Crack Depth ◽  
Temperature Limit ◽  
Round Robin ◽  
Robin Problem ◽  
Limit Curve ◽  
Element Analysis ◽  
Probabilistic Fracture Mechanics ◽  
Key Parameter

The final goal of this study is to solve the round-robin problem for the safety of a reactor pressure vessel by adopting a finite element analysis and probabilistic fracture mechanics. To do so, a sensitivity analysis and a deterministic analysis should be conducted. This paper contains the results of the sensitivity analysis as intermediate results of a round-robin problem. Key parameters such as the initial Reference Temperature for Nil Ductility Transition, Ni contents, Cu contents, fluence, and input transient were chosen to conduct the sensitivity analysis. In addition, different values of crack depth to the thickness ratio are considered to develop FE models. Moreover, a series of FE analyses are carried out. As a result, each key parameter has an influence on RTNDT and KIc. This means that the P-T limit curve is shifted. If the value of each key parameter is increased, the P-T limit curve is moved to the right side. Therefore, the operating area of the P-T limit curve should be reduced. The results of this paper will be very helpful in enhancing our understanding of the P-T limit curve. In addition, it will be used to adjust the probabilistic fracture mechanics and solve the round-robin problem.

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