Study on Probabilistic Fracture Mechanics Analysis Codes for Pipes With Stress Corrosion Cracks

2008 ◽  
Author(s):  
Hideo Machida ◽  
Manabu Arakawa ◽  
Norimichi Yamashita ◽  
Shinobu Yoshimura
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Nuclear Power ◽  
Power Plants ◽  
Reliability Assessment ◽  
Multiple Cracks ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis ◽  
Integrity Management ◽  
Technical Basis

Risk-Informed integrity management methodologies have been developed in Japanese nuclear power plants. One of the issues of concern is the reliability assessment of piping with flaws due to stress corrosion cracking (SCC). Therefore, the probabilistic fracture mechanics analysis code have been developed, which can perform the reliability assessment for the austenitic stainless steel piping with flaws due to SCC. This paper describes technical basis of this code. This method is based on Monte-Carlo technique considering many sample cases in a piping section, where the initiation and growth of cracks are calculated and piping failures, including leaks and rapture, are evaluated. A notable feature is that multiple cracks can be treated, consequently, assessment of coalescence of cracks and intricate break evaluation of piping section have been included. Moreover, the in-service inspection (ISI) and integrity evaluation by Fitness-for-Service (FFS) code are integrated into the analysis, and the contribution to failure probability decrease can be assessed. Key parameters are determined on a probability basis with the designated probability type throughout the procedure. Size, location and time of crack initiation, coefficients of crack growth due to SCC and factors for piping failure are included in those parameters. With this method the reliability level of the piping through the operation periods can be estimated and the contribution of various parameters including ISI can be quantitatively evaluated.

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.

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

2014 ◽  
Author(s):  
Yinsheng Li ◽  
Kazuya Osakabe ◽  
Genshichiro Katsumata ◽  
Jinya Katsuyama ◽  
Kunio Onizawa ◽  
...  
Keyword(s):  
Risk Assessment ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Multiple Cracks ◽  
Benchmark Analysis ◽  
Probabilistic Fracture Mechanics

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.

Allowable Flaw Size of Japanese Cast Stainless Steel Pipe Using Probabilistic Fracture Mechanics Method

2017 ◽  
Author(s):  
Shotaro Hayashi ◽  
Mayumi Ochi ◽  
Kiminobu Hojo ◽  
Takahisa Yamane ◽  
Wataru Nishi
Keyword(s):  
Stainless Steel ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Thermal Ageing ◽  
Probabilistic Fracture Mechanics ◽  
Stainless Steel Pipe ◽  
Technical Basis ◽  
Cast Stainless Steel

The cast austenitic stainless steel (CASS) that is used for the primary loop pipes of nuclear power plants is susceptible to thermal ageing during plant operation. The Japanese JSME rules on fitness-for-service (JSME rules on FFS)[1] for nuclear power plants specify the allowable flaw depths. However, some of these allowable flaw sizes are small compared with the smallest flaw sizes, which can be detected by nondestructive testing. ASME Section XI Code Case N-838[2] recently specified the maximum tolerable flaw depths for CASS pipes determined by probabilistic fracture mechanics (PFM). In a similar way, the allowable flaw depths of CASS pipes were calculated by PFM analysis code “PREFACE”[3] which considers uncertainty of the mechanical properties of Japanese PWR CASS materials. In order to confirm the validity of PREFACE, the allowable flaw depths calculated by PREFACE were compared with the maximum tolerable flaw depths in the technical basis of Code Case N-838. As a result, although the J calculation method and the embrittlement prediction model of CASS are different, these were qualitatively consistent. In addition, the sensitivity of ferrite content to the allowable flaw depths was investigated.

Strategies for Treating Weld Residual Stresses in Probabilistic Fracture Mechanics Codes

2011 ◽  
Author(s):  
Frederick W. Brust ◽  
R. E. Kurth ◽  
D. J. Shim ◽  
David Rudland
Keyword(s):  
Fracture Mechanics ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Degradation Mechanism ◽  
Welding Process ◽  
Corrosion Cracking ◽  
Uncertainty Sources

Risk based treatment of degradation and fracture in nuclear power plants has emerged as an important topic in recent years. One degradation mechanism of concern is stress corrosion cracking. Stress corrosion cracking is strongly driven by the weld residual stresses (WRS) which develop in nozzles and piping from the welding process. The weld residual stresses can have a large uncertainty associated with them. This uncertainty is caused by many sources including material property variations of base and welds metal, weld sequencing, weld repairs, weld process method, and heat inputs. Moreover, often mitigation procedures are used to correct a problem in an existing plant, which also leads to uncertainty in the WRS fields. The WRS fields are often input to probabilistic codes from weld modeling analyses. Thus another source of uncertainty is represented by the accuracy of the predictions compared with a limited set of measurements. Within the framework of a probabilistic degradation and fracture mechanics code these uncertainties must all be accounted for properly. Here we summarize several possibilities for properly accounting for the uncertainty inherent in the WRS fields. Several examples are shown which illustrate ranges where these treatments work well and ranges where improvement is needed. In addition, we propose a new method for consideration. This method consists of including the uncertainty sources within the WRS fields and tabulating them within tables which are then sampled during the probabilistic realization. Several variations of this process are also discussed. Several examples illustrating the procedures are presented.

Verification of Probabilistic Fracture Mechanics Analysis Code PASCAL

2017 ◽  
Author(s):  
Yinsheng Li ◽  
Genshichiro Katsumata ◽  
Koichi Masaki ◽  
Shotaro Hayashi ◽  
Yu Itabashi ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Working Group ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Energy Agency ◽  
Probabilistic Fracture Mechanics ◽  
Integrity Assessments

Probabilistic fracture mechanics (PFM) has been recognized as a promising methodology in structural integrity assessments of aged pressure boundary components of nuclear power plants because it can rationally represent the influencing parameters in their inherent probabilistic distributions without over conservativeness. In Japan, a PFM analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed by the Japan Atomic Energy Agency (JAEA) to evaluate the through-wall cracking frequencies of Japanese reactor pressure vessels (RPVs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) transients. In addition, efforts have been made to strengthen the applicability of PASCAL to structural integrity assessments of domestic RPVs against non-ductile fracture. On the other hand, unlike deterministic analysis codes, the verification of PFM analysis codes is not easy. A series of activities has been performed to verify the applicability of PASCAL. In this study, as a part of the verification activities, a working group was established in Japan, with seven organizations from industry, universities and institutes voluntarily participating as members. 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 including the verification plan, approaches and results.

scholarly journals Development of Probabilistic Fracture Mechanics Analysis Code for Pipes with Stress Corrosion Cracks

2009 ◽  
Vol 3 (1) ◽  
pp. 103-113 ◽  
Author(s):  
Hideo MACHIDA ◽  
Manabu ARAKAWA ◽  
Norimichi YAMASHITA ◽  
Shinobu YOSHIMURA
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Probabilistic Fracture Mechanics ◽  
Mechanics Analysis

Application of Failure Event Data to Benchmark Probabilistic Fracture Mechanics Computer Codes

2007 ◽  
Author(s):  
F. A. Simonen ◽  
S. R. Gosselin ◽  
B. O. Y. Lydell ◽  
D. L. Rudland ◽  
G. M. Wilkowski
Keyword(s):  
Fracture Mechanics ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Operating Experience ◽  
Corrosion Cracking ◽  
Operating Conditions ◽  
Intergranular Stress Corrosion ◽  
Probabilistic Fracture Mechanics ◽  
Computer Codes

This paper describes an application of data on cracking, leak and rupture events from nuclear power plant operating experience to estimate failure frequencies for piping components that had been previously evaluated using the PROLOCA and PRAISE probabilistic fracture mechanics (PFM) computer codes. The calculations had addressed the failure mechanisms of stress corrosion cracking, intergranular stress corrosion cracking and fatigue for materials and operating conditions that were known to have failed components. The first objective was to benchmark the calculations against field experience. A second objective was a review of uncertainties in the treatments of the data from observed failures and in the structural mechanics models. The database PIPExp-2006 was applied to estimate failure frequencies. Because the number of reported failure events was small, there were also statistical uncertainties in the estimates of frequencies. Comparisons of predicted and observed failure frequencies showed that PFM codes correctly predicted relatively high failure probabilities for components that had experienced field failures. However, the predicted frequencies tended to be significantly greater than those estimated from plant operating experience. A review of the PFM models and inputs to the models showed that uncertainties in the calculations were sufficiently large to explain the differences between the predicted and observed failure frequencies.

Uncertainties in Probabilistic Fracture Mechanics Calculations

2010 ◽  
Author(s):  
F. A. Simonen
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Power ◽  
Power Plants ◽  
Flaw Detection ◽  
Pressure Vessels ◽  
Operating Conditions ◽  
Statistical Distributions ◽  
Probabilistic Fracture Mechanics ◽  
Crack Growth Rates ◽  
Failure Probabilities

This paper addresses uncertainties in probabilistic fracture mechanics (PFM) calculations for pressure boundary components at commercial nuclear power plants. Such calculations can predict the probability that a component will have failed after a specified period of operation, but with large uncertainties that are difficult to quantify. PFM models only approximate details of as-built components as well as actual operating conditions over the lifetime of the component. Statistical distributions used as inputs to the calculations are subject to uncertainties, which also results in large uncertainties in calculated failure probabilities. This paper describes from the author’s perspective various uncertainties that are associated with PFM calculations. Efforts to quantify PFM uncertainties are described along with their impacts on calculated failure probabilities. Many uncertainties are explicitly addressed by statistical distributions for input parameters to the PFM models (e.g. crack growth rates, material strengths, probabilities of flaw detection, etc.). Other calculations have gone further by estimating uncertainties in the parameters of these statistical distributions along with uncertainties in parameters treated as deterministic inputs to the PFM models. Examples from the author’s experience with uncertainty analyses for pressure vessels and piping components are described.

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