Document: NUREG-0800
Document ID: 048d2078-58a2-440a-a58f-f456e6c8dec8
Document Type: srp
Title: CONTAINMENT LEAKAGE TESTING
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052340706.pdf
Revision Date: 2023-06
Chapter: 6
Section ID: 6.2.6
CFR Part: 
CFR Title: 

Content:
n 52, "Capability for Containment Leakage Rate Testing" General Design Criterion 52 as it relates to the reactor containment and exposed equipment being designed to accommodate the test conditions for the containment integrated leak rate test (up to the containment design pres- sure). (b) General Design Criterion 53, "Provisions for Containment Testing and Inspection" General Design Criterion 53 as it relates to the reactor containment being designed to permit appropriate inspection of important areas (such as pene- trations), 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, "Piping System Penetrating Containment" General Design Criterion 54 as it relates to piping systems penetrating primary reactor 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 site, an applicant should assume the expected demonstrable leak rate from the containment. Nuclear power plant leak testing experience shows that a design leak rate of 0.1% per day provides adequate margin above typically measured containment leak rates and is compatible with current leak test methods and test acceptance criteria. Therefore, the minimum acceptable design containment leakage rate shall not be less than 0.1% per day. 10 CFR Part 100, § 100.1 addresses factors to be considered when evaluating nuclear power plant sites, and includes the safety features that are engineered into the facility. The secondary containment of dual-type containments, which provide for a controlled, filtered release to the environs of leakage from the 6.2.6-2 Rev. 2 - July 1981 primary reactor containment, is such an engineered safety feature, whose effec- tiveness must be periodically verified as required by Appendix J in Section IV.B. In so doing, the