Document: NUREG-0800
Document ID: 8940dd69-2a01-4786-bbe6-f71794a60643
Document Type: srp
Title: LOSS-OF-COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF POSTULATED PIPING
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052350156.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.6.5
CFR Part: 
CFR Title: 

Content:
g-lived radioactivity. 2. The radiological consequences of the most severe LOCA are within the guidelines of 10 CFR Part 100. Appendices A, B and D to this SRP section provides the results of the LOCA analysis. 3. The TMI Action Plan (Ref. 6 and 7) requirements for II.E.2.3, II.K.2.8, II.K.3.5, IL.K.3.25, II.K.3.30, II.K.3.31, and II.K.3.40 have been met. 111. REVIEW PROCEDURES The procedures below are used during both the construction permit (CP) and operating license (OL) reviews. During the CP.review, the values of system parameters and setpoints used in the analysis will be preliminary In nature and subject to change. At the OL review, final values should be used in the analysis and the reviewer compares these to the limiting safety system settings included in the proposed technical specifications. For the review of the ECCS performance analysis, as presented in the applicant's safety analysis report (SAR), the reviewer verifies the following: 15.6.5-4 Rev. 2 - July 1981 1. The calculations were performed using an approved evaluation model. The application should clearly state this and properly reference the evaluation model. If the analysis is done with a new evaluation model, a generic review of the new model is required. 2. An adequate failure mode analysis has been performed to justify the selection of the most limiting single active failure. This analysis is reviewed in.part under SRP Section 6.3. If the design has been changed from that presented in previous applications, changes in the reactor coolant system, reactor core, and ECCS are reviewed with respect to the most limiting single failure. 3. A variety of break locations and the complete spectrum of break sizes were analyzed. If part of the evaluation is done by referencing earlier work, design differences (ECCS, reactor coolant system, reactor core, etc.) between the facilities in question are reviewed. If there are significant differences, sensitivity studies on the important parameters should have