Document: NRC Regulatory Guide
Document ID: 4d46a966-d280-43da-9b03-8b0abe7b29ce
Document Type: regulatory_guide
Title: Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (Rev. 1)
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML2120/ML21204A065.pdf
Revision Date: 2023-05
Chapter: 
Section ID: RG-1.183
CFR Part: 
CFR Title: 

Content:
ain Steamline Break Until cold shutdown is established Fuel Damage or Preaccident Spike 0.25 Sv (25 rem) 0.05 Sv (5.0 rem) Coincident Iodine Spike 0.025 Sv (2.5 rem) 0.05 Sv (5.0 rem) PWR Locked Rotor Accident 0.025 Sv (2.5 rem) 0.05 Sv (5.0 rem) Until cold shutdown is established PWR Control Rod Ejection Accident 0.063 Sv (6.3 rem) 0.05 Sv (5.0 rem) 30 days for containment pathway; until cold shutdown is established for secondary pathway Fuel Handling Accident 0.063 Sv (6.3 rem) 0.05 Sv (5.0 rem) 2 hours The column labeled “Analysis Release Duration” summarizes the assumed radioactivity release durations identified in the individual appendices to this guide. Refer to these appendices for complete descriptions of the release pathways and durations. 17 For PWRs with steam generator alternative repair criteria, different dose criteria may apply to steam generator tube rupture and main steamline break analyses. 18 The control room exposure period is 30 days for all accidents. 19 Tube rupture in the affected steam generator may result in the need to control steam generator water level using steam dumps. These releases may extend the duration of the release from the affected steam generator beyond the initial isolation. DG-1389, Page 29 5. Analysis Assumptions and Methodology 5.1 General Considerations 5.1.1 Analysis Quality The analyses discussed in this guide are re-analyses of the design basis safety analyses required by 10 CFR 50.67 or evaluations required by 10 CFR 50.34, 10 CFR Part 52, and GDC 19. These analyses are considered to be a significant input to the evaluations required by 10 CFR 50.92 or 10 CFR 50.59 and 10 CFR Part 52. The licensee should prepare, review, and maintain these analyses in accordance with quality assurance programs that comply with Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR Part 50. These design basis analyses were structured to provide a conservative set of assumptions