Document: NUREG-0800
Document ID: ef80f444-3259-4db1-b721-096a73435ff7
Document Type: srp
Title: CONTAINMENT LEAKAGE TESTING
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070464.pdf
Revision Date: 2023-06
Chapter: 6
Section ID: 6.2.6
CFR Part: 
CFR Title: 

Content:
e contained as proposed by applicants using this kind of containment are reviewed by CSBSCSB (see SRP Section 6.2.3). 6 II. ACCEPTANCE CRITERIA The reactor containment leakage rate testing program, as described in the safety analysis report (SAR), will be acceptable if it meets the requirements stated in Appendix J to 10 CFR Part 50. 7 Appendix J provides the test requirements and acceptance criteria for preoperational and periodic leak testing of the reactor containment and of systems and components which penetrate the containment. Exceptions to Appendix J requirements will be reviewed on a case-by-case basis. Conformance with the requirements of Appendix J constitutes an acceptable basis for satisfying the requirements of the following General Design Criteria applicable to containment leakage rate testing: (a) General Design Criterion 52 (GDC 52), "Capability for Containment Leakage Rate 8 Testing," General Design Criterion 52 as it relates to the reactor containment and 9 exposed equipment being designed to accommodate the test conditions for the containment integrated leak rate test (up to the containment design pressure). (b) General Design Criterion 53 (GDC 53), "Provisions for Containment Testing and 10 Inspection," General Design Criterion 53 as it relates to the reactor containment being 11 designed to permit appropriate inspection of important areas (such as penetrations), an appropriate surveillance program, and leak testing at the containment design pressure of penetrations having resilient seals and expansion bellows. (c) General Design Criterion 54 (GDC 54), "Piping System Penetrating Containment," 12 General Design Criterion 54 as it relates to piping systems penetrating primary reactor 13 containment being designed with a capability to determine if valve leakage is within acceptable limits. 10 CFR Part 100, § 100.11 requires that, as an aid in evaluating a proposed nuclear power plant 14 site, an applicant should assume the expected demonstrable