Document: NRC Regulatory Guide
Document ID: 30e6fa68-c7c1-4266-ba45-e06d1bbc07a0
Document Type: regulatory_guide
Title: Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants (Rev. 5)
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML1708/ML17083A134.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.97
CFR Part: 
CFR Title: 

Content:
ons by expressly expanding the applicability of RG 1.97 to holders of, or applicants for, power reactor design certifications or combined licenses under 10 CFR Part 52, and by adding references to the NRC’s 10 CFR Part 52 regulations and related NRC guidance documents. Background Revision 4 of RG 1.97, endorses IEEE Std. 497-2002, “IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations” (Ref. 13). Since the release of Revision 4 of RG 1.97, the IEEE revised Std. 497 in 2010 and again in 2016 to reflect a more technology- neutral approach and to bring the IEEE standard more in line with the international standards referenced in the IEEE Std. 497-2016 and the Harmonization section below. In addition to the Types A, B, C, D, and E variables defined in the previous revisions, IEEE Std. 497-2016 adds a new Type F variable, which provides primary information to indicate fuel damage and the effects of fuel damage. In March 1979, an accident occurred in Unit 2 of the Three Mile Island Nuclear Station. In the aftermath, the nuclear industry and the NRC adopted a more rigorous approach to accident monitoring. In May 1983, the NRC issued Revision 3 of RG 1.97, “Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident” (Ref. 14). The RG prescribed a detailed list of variables to monitor, and specified a comprehensive list of design and qualification criteria to be met. Because of its prescriptive nature, RG 1.97 quickly became the de facto standard for accident monitoring. With the increased use of digital instrumentation systems in advanced nuclear power plant designs, the nuclear industry recognized a need to develop a consolidated standard that was more flexible. Instead of prescribing the instrument variables to be monitored (as was the case in Revision 3 of RG 1.97), the industry developed performance-based criteria for use in selecting