Document: NUREG-0800
Document ID: 1da20040-18d5-4c75-9863-5f578a9d73a5
Document Type: srp
Title: Revision 5 - March 2007
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0705/ML070550078.pdf
Revision Date: 2023-06
Chapter: 7
Section ID: 7
CFR Part: 
CFR Title: 

Content:
Std. 279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations." Clause 4.4 of IEEE Std. 603-1991 requires identification of the analytical limit associated with each variable. Clause 6.8.1 requires that allowances for uncertainties between the analytical limit and device setpoint be determined using a documented methodology. Clause 3(6) of IEEE 279-1971 requires identification of the levels that, when reached, will necessitate protective action. 10 CFR Part 50, Appendix B, Criterion XI, “Test Control,” and Criterion XII, "Control of Measuring and Test Equipment," provide requirements for tests and test equipment used in maintaining instrument setpoints. 10 CFR 50 Appendix A, General Design Criterion (GDC) 13, “Instrumentation and Control,” requires, in part, that instrumentation be provided to monitor variables and systems, and that controls be provided to maintain these variables and systems within prescribed operating ranges. GDC 20, “Protection System Functions,” requires, in part, that the protection system be designed to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences. 10 CFR 50.36(c)(1)(ii)(A), “Technical Specifications,” requires that, where a limiting safety system setting (LSSS) is specified for a variable on which a safety limit has been placed, the setting be so chosen that automatic protective action will correct the abnormal situation before a safety level is exceeded. LSSSs are settings for automatic protective devices related to variables with significant safety functions. Setpoints found to exceed technical specification limits are considered as malfunctions of an automatic safety system. Such an occurrence could challenge the integrity of the reactor core, reactor coolant pressure boundary, containment, and associated systems. 10 CFR