Technical Letter Report: Evaluation and Analysis of a Few International Periodic Safety Review Summary Reports

At the request of the United States (U.S.) government, the International Atomic Energy Agency (IAEA) assembled a team of 20 senior safety experts to review the regulatory framework for the safety of operating nuclear power plants in the United States. This review focused on the effectiveness of the regulatory functions implemented by the NRC and on its commitment to nuclear safety and continuous improvement. One suggestion resulting from that review was that the U.S. Nuclear Regulatory Commission (NRC) incorporate lessons learned from periodic safety reviews (PSRs) performed in other countries as an input to the NRC’s assessment processes. In the U.S., commercial nuclear power plants (NPPs) are granted an initial 40-year operating license, which may be renewed for additional 20-year periods, subject to complying with regulatory requirements. The NRC has established a framework through its inspection, and operational experience processes to ensure the safe operation of licensed nuclear facilities on an ongoing basis. In contrast, most other countries do not impose a specific time limit on the operating licenses for NPPs, they instead require that the utility operating the plant perform PSRs, typically at approximately 10-year intervals, to assure continued safe operation until the next assessment. The staff contracted with Argonne National Laboratory (Argonne) to perform a pilot review of selected translated PSR assessment reports and related documentation from foreign nuclear regulatory authorities to identify any potential new regulatory insights regarding license renewal-related topics and NPP operating experience (OpE). A total of 14 PSR assessment documents from 9 countries were reviewed. For all of the countries except France, individual reports were provided for each of the plants reviewed. In the case of France, three reports were provided that reviewed the performance assessment of thirty-four 900-MWe reactors of similar design commissioned between 1978 and 1988. All of the reports reviewed were the regulator’s assessment of the PSR findings rather than the original PSR report, and all but one were English translations from the original language. In these reviews, it was found that most of the countries base their regulatory guidance to some extent (and often to a large extent) on U.S. design codes and standards, NRC regulatory guidance, and U.S. industry guidance. In addition, many of the observed operational technical issues and OpE events reported for U.S. reactors are also cited in the PSR reports. The PSR reports also identified a number of potential technical material/component performance issues and OpE events that are not commonly reported for U.S. plants

Evaluation of Effects of LWR Coolant Environments on Fatigue Life of Carbon and Low-Alloy Steels,

Abstract Fatigue tests have been conducted on Types 304 and 3 16NG stainless steels to evaluate the effects of various material and loading variables, e.g., steel type, strain rate, dissolved oxygen (DO) in water, and strain range, on the fatigue lives of these steels. The results conhn significant decreases in fatigue life in water. Unlike the situation with ferritic steels, environmental effects on Types 304 and 316NG stainless steel are more pronounced in low-DO than in high-DO water. Experimental results have been compared with estimates of fatigue life based on a statistical model. The formation and growth of fatigue cracks in air and water environments are discussed

Statistical models for estimating fatigue strain-life behavior of pressure boundary materials in light water reactor environments

The existing fatigue strain versus life (S-N) data for materials used in nuclear power plant components have been compiled and categorized according to material, loading and environmental conditions. Statistical models have been developed for estimating the effects of the various service conditions on the fatigue life of these materials. The results have been used to estimate the probability of initiating a fatigue crack. Data in the literature were reviewed to evaluate the effects of the size, geometry and surface finish of a component on its fatigue life. Fatigue S-N curves for components have been determined by adjusting the probability distribution curves of smooth test specimens for the effect of mean stress and then applying design margins to account for the uncertainties that arise because of component size, geometry and surface finish. The significance of the effect of the environment on the current code design curve and on the proposed interim design curves published in NUREG/CR-5999 is discussed. Estimations of the probability of fatigue cracking in sample components from boiling water reactors and pressurized water reactors are presented

Statistical models for estimating fatigue strain-life behavior of pressure boundary materials in light water reactor environments

The existing fatigue strain versus life (S-N) data for materials used in nuclear power plant components have been compiled and categorized according to material, loading and environmental conditions. Statistical models have been developed for estimating the effects of the various service conditions on the fatigue life of these materials. The results have been used to estimate the probability of initiating a fatigue crack. Data in the literature were reviewed to evaluate the effects of the size, geometry and surface finish of a component on its fatigue life. Fatigue S-N curves for components have been determined by adjusting the probability distribution curves of smooth test specimens for the effect of mean stress and then applying design margins to account for the uncertainties that arise because of component size, geometry and surface finish. The significance of the effect of the environment on the current code design curve and on the proposed interim design curves published in NUREG/CR-5999 is discussed. Estimations of the probability of fatigue cracking in sample components from boiling water reactors and pressurized water reactors are presented