CONSIDERATION OF SEVERE ACCIDENTS IN REGULATORY FRAMEWORK
Sections 7.2 to 7.7 of this chapter describe the phenomena associated with severe accidents. They also describe the state of the knowledge associated with mitigating and managing these phenomena, so that the loads on containment system can be reduced or eliminated. This section describes the regulatory actions taken by the NRC to minimize the adverse effects of severe accidents on the plant structures, systems, and components (SSCs).
On August 8, 1985, NRC issued a Policy Statement on severe accidents regarding future Designs and existing Plants (50 Federal Register Notice 32138) that introduced the Commission’s plan to address severe accident issues for existing commercial nuclear power plants. In the following years, the Commission developed an approach to implement this plan for the existing plants and issued a Generic Letter 88-20  that communicated this plan to all licensees. Each nuclear power plant licensee was requested to perform a plant examination that would look for vulnerabilities to severe accidents and cost-effective safety improvements that could reduce or eliminate the potential vulnerabilities. The specific objectives for these Individual Plant Examinations (IPEs) for each licensee consisted of (1) develop an overall appreciation of severe accident behavior; (2) understand the most likely severe accident sequences that could occur at its plant; (3) gain a more quantitative understanding of the overall probability of core damage and radioactive material releases; and (4) if necessary, reduce the overall probability of core damage and radioactive material release by appropriate modifications to procedures and hardware that would help prevent or mitigate severe accidents. Upon completion of the examination, the licensee was required to submit a report to NRC describing the results and conclusions of the examination.
In order for the licensees to systematically organize the submittals, the NRC (and its contractors) provided administrative and technical guidance through the following documents (partial listing).
- • NUREG 1335 : “IPE Submittal Guidance Document” established format and content guidelines for the licensees’ submittals. The appendices to this document contain: (1) an approach to the back-end portion of a PRA, (2) references to PRAs performed or reviewed by the NRC, (3) NRC responses to questions and comments raised at the IPE workshop, and (4) staff review guidance.
- • NUREG/CR-5132 : “Severe Accident Insights Report,” This report describes the conditions and events that nuclear power plant personnel may encounter during the latter stages of a severe core damage accident and what the consequences might be of actions they may take during these latter stages. The report also describes what can be expected of the performance of the key barriers to fission product release (primarily containment systems), what decisions the operating staff may face during the course of a severe accident, and what could result from these decisions based on our current state of knowledge of severe accident phenomena.
- • NUREG/CR-4920, Volumes 1-5 : “Assessment of Severe Accident Prevention and Mitigation Features,” This series of reports describes plant features and operator actions found to be important in either preventing or mitigating severe accidents in LWRs with five different types of containments.
- • NUREG/CR-2300 : “PRA Procedures Guide,” This report is a guide to the performance of probabilistic risk assessments (PRAs) for nuclear power plants.
- • NUREG/CR-3485 : “PRA Review Manual,” This report describes an approach for reviewing a Level 1 type PRA (a PRA that carries the accident analysis up to the point of calculating the probability of core damage or core melt).
The NRC, Office of Research, evaluated the licensees’ submittals for IPE and documented the results of its evaluations in NUREG-1560 . The report provides perspectives gained by reviewing 75 IPE submittals pertaining to 108 nuclear power plant units (existing in 1997). The NRC review was focused on gaining perspectives in three major areas: (1) improvements made to individual plants as a result of the IPE program, (2) plant specific design and operational features and modeling assumptions that significantly affect the estimates of core damage frequency (CDF) and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. In particular, these results are assessed in relation to the design and operational characteristics of the various reactor and containment types, and by comparing the IPEs to PRA characteristics. In addition to the IPE program, the NRC also implemented a program, “Individual Plant Examination for External Events” (IPEEE). This program is discussed in Chapter 8 of this book.