Document: NRC Regulatory Guide
Document ID: fc586342-92f7-4c77-ae76-79e3674cf288
Document Type: regulatory_guide
Title: Criteria for Programmable Digital Devices in Safety-Related Systems of  Nuclear Power Plants + HISTORY –HISTORY 02/2023 – DG-1374, Proposed Revision 4 Prior to issuance of DG-1374, RG 1.152 was entitled, “Criteria for Use of Computers in Safety-Systems of Nuclear Power Plants” 06/2010 – DG-1249, Proposed Revision 3 – Revise 12/2004 – DG-1130, Proposed Revision 2 – Revise 05/1995 – DG-1039, Proposed Revision 1 03/1983 – DG-1130, Proposed Revision 2 – Revise (Rev. 4)
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML2301/ML23012A242.pdf
Revision Date: 2023-05
Chapter: 
Section ID: RG-1.152
CFR Part: 
CFR Title: 

Content:
1, 1971, 10 CFR 50.55a(h)(2) requires compliance with their DG-1374, Page 2 plant-specific licensing basis or IEEE Std 603-1991 and the correction sheet dated January 30, 1995. For applicants for CPs, operating licenses, combined licenses, standard design approvals, design certifications, or manufacturing licenses filed after May 13, 1999, 10 CFR 50.55a(h)(3) requires compliance with IEEE Std 603-1991 and the correction sheet dated January 30, 1995. Although 10 CFR 50.55a(h)(2) and (h)(3) and IEEE Std 603-1991 (incorporated by reference) use the term “safety system,” consistent with the NRC’s definition of safety-related systems in 10 CFR 50.2, “Definitions,” this RG uses the term “safety-related system” in lieu of the term “safety system.” o 10 CFR Part 50, Appendix A, “General Design Criteria for Nuclear Power Plants,” General Design Criterion (GDC) 13, “Instrumentation and control,” requires, in part, that operating reactor licensees provide instrumentation to monitor variables and systems over their anticipated ranges for normal operation, anticipated operational occurrences, and accident conditions as appropriate to ensure adequate safety. o 10 CFR Part 50, Appendix A, GDC 21, “Protection system reliability and testability,” requires, in part, that protection systems be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. It also requires that protection systems be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred. o 10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” Criterion III, “Design Control,” requires, in part, that licensees specify quality standards and provide design control measures for verifying or checking the adequacy of