Document: NUREG-0800
Document ID: 1085af08-a4b7-4fe7-bfed-ca7a60f2022c
Document Type: srp
Title: LOSS-OF-COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070734.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.6.5
CFR Part: 
CFR Title: 

Content:
demonstrate that the offsite doses resulting from various accidents presented in the SAR are within the guideline values. Meeting the guideline doses is achieved by a combination of engineered safety features installed in the nuclear facility, an effective emergency core cooling system, and siting the nuclear plant in an area that does not exceed population density requirements. Meeting the nuclear power plant siting criteria provides a level of assurance that the plant will pose no undue risk to the public as a result of the consequences of loss-of-coolant accidents.70 III. REVIEW PROCEDURES The procedures below are used during both the construction permit (CP), and standard design certification, combined license (COL), and operating license (OL) reviews. During the CP 71 review, the values of system parameters setpoints used in the analysis will be preliminary in nature and subject to change. At the OL, COL, or the standard design certification review, 72 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: 1. The calculations were performed using an approved evaluation model as specified in 10 73 CFR 50.46 following the guidance of Appendix K, Section I, or Regulatory Guide 1.157. The application should clearly state this and properly reference the evaluation 74 DRAFT Rev. 3 - April 1996 15.6.5-8 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