Stephen R. Gosselin

Probabilistic Fracture Mechanics Evaluation of Selected Passive Components – Technical Letter Report

This report addresses the potential application of probabilistic fracture mechanics computer codes to support the Proactive Materials Degradation Assessment (PMDA) program as a method to predict component failure probabilities. The present report describes probabilistic fracture mechanics calculations that were performed for selected components using the PRO-LOCA and PRAISE computer codes. The calculations address the failure mechanisms of stress corrosion cracking, intergranular stress corrosion cracking, and fatigue for components and operating conditions that are known to make particular components susceptible to cracking. It was demonstrated that the two codes can predict essentially the same failure probabilities if both codes start with the same fracture mechanics model and the same inputs to the model. Comparisons with field experience showed that both codes predict relatively high failure probabilities for components under operating conditions that have resulted in field failures. It was found that modeling assumptions and inputs tended to give higher calculated failure probabilities than those derived from data on field failures. Sensitivity calculations were performed to show that uncertainties in the probabilistic calculations were sufficiently large to explain the differences between predicted failure probabilities and field experience.

Life Prediction and Monitoring of Nuclear Power Plant Components for Service-Related Degradation

This paper describes industry programs to manage structural degradation and to justify continued operation of nuclear components when unexpected degradation has been encountered due to design materials and/or operational problems. Other issues have been related to operation of components beyond their original design life in cases where there is no evidence of fatigue crack initiation or other forms of structural degradation. Data from plant operating experience have been applied in combination with inservice inspections and degradation management programs to ensure that the degradation mechanisms do not adversely impact plant safety. Probabilistic fracture mechanics calculations are presented to demonstrate how component failure probabilities can be managed through augmented inservice inspection programs.

Enhanced ASME section XI Appendix L flaw tolerance procedure

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code has worked since the early 1990s to develop guidelines for the nuclear power industry to evaluate the serviceability of components that are adversely subjected to fatigue stresses. Results of this work formed the basis for a non-mandatory Appendix L that became part of the 1996 Addenda to the 1995 Edition of Section XI [1]. A key part of this appendix was the introduction of a damage tolerance based examination strategy designed to assure that the component will operate reliability between subsequent inspections. Since the issuance of Appendix L, new data on the flaw detection capabilities have become available, and these data show that ultrasonic inspections can detect flaws much smaller than assumed during the original development of Appendix L. This information allows for significantly smaller sizes for initial postulated flaws. Additionally, Appendix L did not address the potential for multiple fatigue crack initiation sites for a given location on the component. In 1999 the ASME Working Group on Operating Plant Criteria (WGOPC) re-established the Task Group on Operating Plant Fatigue Assessments (TGOPFA) to address these concerns. This paper summarizes the research results, supporting computations, and technical bases for TGOPFA recommended Appendix L improvements. A detailed sample calculation is presented for a PWR charging nozzle-to-pipe weld.Copyright

Performance Demonstration Based Probability of Detection (POD) Curves for Fatigue Cracks in Piping

This paper evaluates non-destructive examination (NDE) detection capabilities for fatigue cracks in piping. Industry performance demonstration initiative (PDI) data for fatigue crack detection were used to develop a matrix of statistically based probability of detection (POD) curves that consider various NDE performance factors. Seven primary performance factors were identified — Material, Crack Geometry/Type, NDE Examination Access, NDE Procedure, Examiner Qualification, Pipe Diameter, and Pipe Wall Thickness. A database of 16,181 NDE performance observations, with 18 fields associated with each observation, was created and used to develop statistically based POD curves for 42 stainless steel and 14 carbon steel performance cases. Subsequent comparisons of the POD fits for each of the cases showed that excellent NDE performance for fatigue cracks can be expected for ferritic materials. Very little difference was observed between the POD curves for the 14 carbon steel performance cases considered in this study and NDE performance could therefore be represented by a single POD curve. For stainless steel, very good performance can also be expected for circumferential cracks located on the same side of the weld from which the NDE examination is made. POD depended primarily on component thickness. Three POD curves for stainless steel were prepared. Best estimate and the associated 95% confidence bounds for POD versus through-wall depth logistic regression digital data are provided. Probabilistic fracture mechanics (PFM) calculations were performed to compare best estimate leak probabilities obtained from both the new performance-based POD curves and previous PFM models. This work was performed under joint funding by EPRI and the U.S. Department of Energy (DOE), Office of Nuclear Energy Science and Technology’s Nuclear Energy Plant Optimization (NEPO) program.Copyright

Technical Letter Report Development of Flaw Size Distribution Tables Including Effects of Flaw Depth Sizing Errors for Draft 10CFR 50.61a (Alternate PTS Rule) JCN-N6398, Task 4

This document describes a new method to determine whether the flaws in a particular reactor pressure vessel are consistent with the assumptions regarding the number and sizes of flaws used in the analyses that formed the technical justification basis for the new voluntary alternative Pressurized Thermal Shock (PTS) rule (Draft 10 CFR 50.61a). The new methodology addresses concerns regarding prior methodology because ASME Code Section XI examinations do not detect all fabrication flaws, they have higher detection performance for some flaw types, and there are flaw sizing errors always present (e.g., significant oversizing of small flaws and systematic under sizing of larger flaws). The new methodology allows direct comparison of ASME Code Section XI examination results with values in the PTS draft rule Tables 2 and 3 in order to determine if the number and sizes of flaws detected by an ASME Code Section XI examination are consistent with those assumed in the probabilistic fracture mechanics calculations performed in support of the development of 10 CFR 50.61a.

Application of Failure Event Data to Benchmark 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.Copyright

Flaw Tolerance for Multiple Fatigue Cracks

Appendix L of Section XI provides for serviceability assessments of piping components that are subject to fatigue stresses. This appendix introduced a damage tolerance examination strategy to assure that components perform reliably throughout operating periods between inspections. Operating periods are based on fatigue crack growth analyses of postulated pre-existing cracks. Evidence from service experience shows that fatigue cracking occurrences at operating nuclear power plants often result from mechanisms that cause cracks to initiate and then grow at multiple locations on the inside surface of a pipe, becoming longer and deeper, and eventually linking to form a single long crack. This paper documents important details of the technical bases for changes to Appendix L. Calculations identified aspect ratios for equivalent single cracks (ESC) between the extremes of a 6:1 ratio and a full circumferential crack that can be used in Appendix L flaw tolerance assessments to account for the initiation, growth, and linking of multiple fatigue cracks. Probabilistic fracture mechanics (PFM) calculations determined ESC aspect ratios that result in the same through-wall crack probability as multiple small cracks (0.02 inch depth) that initiate and coalesce. The computations considered two materials (stainless and low alloy steels), three pipe diameters, five cyclic membrane-to-gradient stress ratios and a wide range of primary loads. Subsequent deterministic calculations identified the ESC aspect ratio for the hypothetical reference flaw depth assumptions in Appendix L. This paper also describes computations that compare the Appendix L flaw tolerance allowable operating period for the ESC models with results obtained when a single default 6:1 aspect ratio reference flaw.Copyright