Document: NUREG-0800
Document ID: 65bf963b-2ae7-4caf-aee2-1ceabe52e775
Document Type: srp
Title: INITIAL PLANT TEST PROGRAM - FINAL SAFETY ANALYSIS REPORT1
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070650.pdf
Revision Date: 2023-06
Chapter: 14
Section ID: 14.2
CFR Part: 
CFR Title: 

Content:
tems, containment systems, the electrical power systems, and the emergency core cooling systems, security systems and related features, or those identified for design-specific or unique plant features (first-of-a-kind) are reviewed by those branches responsible for reviewing the design of that system and/or design feature. The PTRBHQMB is responsible for ensuring that all initial 16 17 DRAFT Rev. 3 - April 1996 14.2-4 plant tests are reviewed in accordance with this SRP section and will provide the coordination and supplementary review necessary to accomplish a complete review of all initial plant tests including those that may be referenced in a standard plant design. For those areas of review identified above, additional acceptance criteria and/or review methods, beyond those described in this section, are specified in other SRP sections and are used in the overall evaluation of issues related to the initial test program such as the adequacy of testing proposed for specific SSCs and/or design features, the design parameters, characteristics, and performance criteria that should be satisfactorily demonstrated by test, etc.18 II. ACCEPTANCE CRITERIA PTRBHQMB acceptance criteria are based on meeting the relevant requirements of the 19 following regulations: 1A. 10 CFR Part 30, §30.53 as it relates to testing radiation detection equipment and 20 monitoring instruments. 2B. 10 CFR Part 50, §50.34(b)(6)(iii) as it relates to the licenseeapplicant providing information associated with preoperational testing and initial startup operations. 21 3C. 10 CFR Part 50, Appendix B, Section XI as it relates to test programs to demonstrate that structures, systems, and componentsSSCs will perform satisfactorily. 22 4D. 10 CFR Part 50, Appendix J, Section III.A.4 as it relates to the preoperational leakage 23 rate testing of the reactor primary containment building.24 Regulatory Guide 1.68 provides information, recommendations and guidance, and in general describes a basis