INSIGHTS FROM MELCOR INDEPENDENT CONFIRMATORY ANALYSES FOR NEW REACTOR DESIGN CERTIFICATION

摘要

The United States (U.S.) Title 10 Code of Federal Regulations (CFR), 10 CFR 52.47(a)(23), requires applicants seeking a design certification (DC) to submit a description and analysis of design features for the prevention and mitigation of SAs (e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass). Applicants perform severe accident analyses for the more likely severe accident scenarios to meet this regulation. To obtain insights on the results of DC applicant s severe accident analyses, the NRC uses the MELCOR computer code to perform independent confirmatory analyses. MELCOR is a fully integrated, engineering-level computer code developed by Sandia National Laboratories for the NRC to model the progression of severe accidents in nuclear power plants. Such accidents involve long-term loss of core cooling; core uncovery, heat-up and degradation; reactor vessel bottom head failure; core relocation to the containment; high temperature and high pressure challenges to the containment; and airborne radionuclide release to the environment. To perform the analyses, the staff selects scenarios to analyze, determines the appropriate modeling approach to perform the simulations, and compares the results to the applicant's simulations. Outcomes from the independent confirmatory analyses for both at-power accidents and shutdown accidents occurring during mid-loop operations are presented. The United States (U.S.) Title 10 Code of Federal Regulations (CFR), 10 CFR 52.47(a)(27), also requires applicants seeking a DC to submit a description of the design-specific probabilistic risk assessment (PRA) and its results. A DC applicant's final safety analysis report (FSAR) is expected to contain a qualitative description of PRA insights and uses, as well as some quantitative PRA results, such that the U.S. Nuclear Regulatory Commission (NRC) staff can perform its safety review. As referenced in the NRC Standard Review Plan (SRP) (NUREG-0800) Chapter 19, the staff compares the design against the Commission's goals of less than 1 ×10~(-4) per year for core damage frequency and less than 1 × 10~(-6) per year for large release frequency. The staff expects that risk is assessed for all modes of operation, including shutdown modes. Although core decay power is lower when the reactor is shut down and the fraction of the time the reactor is shutdown could be small, the risk associated with shutdown accidents could be comparable to at-power accidents. This is especially the case for pressurized water reactors due to factors such as a lower initial water level during mid-loop operations.