Document: NUREG-0800
Document ID: 14c7085e-8b87-4b50-97fb-a095ea003ae8
Document Type: srp
Title: describes the general procedures to be followed in reviewing any I&C system. Procedures for
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0525/ML052500511.pdf
Revision Date: 2023-06
Chapter: 7
Section ID: 7.1
CFR Part: 
CFR Title: 

Content:
reviewing each acceptance criterion of 10 CFR 50 and 10 CFR 52 are provided in Appendix 7.1-A. Therefore, review procedures specific to any given diverse I&C system can be synthesized from Appendix 7.1-A. Note that while compliance with ANSI/IEEE Std 279, "Criteria for Protection Systems for Nuclear Power Generating Stations," is required only for protection systems, the criteria of ANSI/IEEE Std 279 and Reg. Guide 1.153, "Criteria for Power, Instrumentation, and Control Portions of Safety Systems" (which endorses IEEE Std 603, "IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations") address considerations that may be used as guidance, where appropriate, for reviewing any diverse I&C application. This part of Section 7.8 provides a review procedure for conformance of diverse I&C systems with the requirements of 10 CFR 50.62 and the SRM regarding SECY-93-087. This part of Section 7.8 highlights specific topics that should be emphasized in the application of the Appendix 7.1-A review procedures to diverse I&C systems. Major design considerations that should be emphasized in the review of any diverse I&C system are identified below. • Design basis — Design bases should be described in the SAR for each diverse I&C system. The design basis should, as a minimum, address the following topics: – The specific design requirements identified in 10 CFR 50.62. Rev. 4 — June 1997 SRP 7.8-5 – Identification of conditions which require protective action by the diverse I&C systems. For DAS these events are identified in the applicant/licensee's D-in-D&D analysis. For ATWS mitigation systems these events are limited to anticipated operational occurrences, defined in the Definitions and Explanations section of 10 CFR 50 Appendix A as those conditions of normal operation which are expected to occur one or more times during the life of the nuclear power unit, and include but are not limited to loss of power to all recirculation pumps, tripping of the