Document: NUREG-0800
Document ID: 7916b088-fb90-4163-84fe-027bd315bcc5
Document Type: srp
Title: REVIEW OF RISK INFORMATION USED TO SUPPORT PERMANENT PLANT-
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0717/ML071700658.pdf
Revision Date: 2023-06
Chapter: 19
Section ID: 19.2
CFR Part: 
CFR Title: 

Content:
e review process. However, it is not intended to suggest that these considerations exhaust the technical issues that affect the potential for large early release. For example, where plant-specific PRA Level 2 analyses exist, these could provide further insights into LERF considerations for that plant. For each major containment type, the factors that most strongly affect the potential for large early release (given that a core damage sequence is underway) are as follows: PWR Large Dry: Containment bypass Containment isolation RCS depressurization Emergency core cooling (ECC) restoration before vessel failure PWR Ice Condenser: Containment isolation Containment bypass Hydrogen igniters RCS depressurization ECC restoration before vessel failure BWR Mark I and II: Containment isolation Containment bypass Venting Containment heat removal: decay heat Containment heat removal: ATWS RCS depressurization ECC restoration before vessel failure BWR Mark III: All Mark I and Mark II issues Igniters It should be noted that, at some BWRs, many sequences that result in vessel breach have a significant probability of also failing the containment. Also, the reader should note that a loss of containment heat removal may significantly contribute to CDF. In reviewing the calculation of change in LERF for a given plant type, reviewers should consider the following factors: Containment Bypass: • Whether the proposed change affects systems that are credited in the prevention of, or in response to, an initiating event involving a steam generator tube rupture (SGTR) or an ISLOCA. • Whether the proposed change affects the frequency or severity of transients that could result in induced steam generator tube ruptures (ISGTR) (i.e., tube rupture in the course of an accident, caused by elevated temperatures and/or elevated pressure differentials). If the proposed change does not directly affect steam generator tube integrity, and the steam generators in the plant are not experiencing