Document: NUREG-0800
Document ID: 817bb0ad-6fd7-433e-aa2b-b4b3e1166f9c
Document Type: srp
Title: PROBABILISTIC RISK ASSESSMENT AND SEVERE ACCIDENT EVALUATION FOR
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0717/ML071700652.pdf
Revision Date: 2023-06
Chapter: 19
Section ID: 19.0
CFR Part: 
CFR Title: 

Content:
vere accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high-pressure core melt ejection, hydrogen combustion, and containment bypass. 3. 10 CFR 52.47(a)(27) - A description of the design-specific probabilistic risk assessment (PRA) and its results. For a COL 4. 10 CFR 52.79(a)(17) - Information with respect to compliance with a number of the technically relevant positions of the Three Mile Island requirements in 10 CFR 50.34(f), specfically 10 CFR 50.34(f)(1)(i). 5. 10 CFR 52.79(a)(38) - For light-water reactor designs, a description and analysis of design features for the prevention and mitigation of severe accidents, e.g., challenges to containment integrity caused by core-concrete interaction, steam explosion, high- pressure core melt ejection, hydrogen combustion, and containment bypass. 6. 10 CFR 52.79(a)(46) - A description of the plant-specific PRA and its results. 7. 10 CFR 52.79(c)(1), (d)(1), and (e)(1) - If a COL application references a standard design approval, standard DC, or the use of one or more manufactured nuclear power reactors licensed under Subpart F of 10 CFR Part 52, then, the plant-specific PRA information must use the PRA information for the design approval, design certification, or manufactured reactor, respectively, and must be updated to account for site-specific design information and any design changes or departures. The PRA staff review should also support (1) the expectation, as stated in 10 CFR 52.47(a)(2) and 10 CFR 52.79(a)(2), that reactors will reflect through their design, construction, and operation an extremely low probability of accidents that could result in the release of significant quantities of radioactive fission products and (2) the objective, as stated in 10 CFR 52.47(a)(4) and 10 CFR 52.79(a)(5), to assess the risk to public health and safety resulting from facility operation and ensure the adequacy of plant SSCs that are provided to prevent accidents and