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:
ession No. ML031530505). 2 Copies of International Commission on Radiological Protection (ICRP) documents may be obtained through its Web site: http://www.icrp.org/; 280 Slater Street, Ottawa, Ontario K1P 5S9, CANADA; Tel: +1(613) 947-9750, Fax: +1(613) 944-1920. 3 Copies of EPA Library Services may be obtained through its Web site: https://www.epa.gov/libraries. DG-1389, Page 37 39. U.S. NRC, “Management of Backfitting, Forward Fitting, Issue Finality, and Information Requests,” Management Directive 8.4, September 2019, (ADAMS Accession No. ML18093B087). DG-1389, Appendix A, Page A-1 APPENDIX A ASSUMPTIONS FOR EVALUATING THE RADIOLOGICAL CONSEQUENCES OF LIGHT-WATER REACTOR MAXIMUM HYPOTHETICAL LOSS-OF-COOLANT ACCIDENTS The assumptions in this appendix are acceptable to the staff of the U.S. Nuclear Regulatory Commission (NRC) for evaluating the radiological consequences of maximum hypothetical accident (MHA) loss-of-coolant accidents (LOCAs) at light-water reactors. These assumptions supplement the guidance in the main body of this guide. Appendix A, “General Design Criteria for Nuclear Power Plants,” to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, “Domestic Licensing of Production and Utilization Facilities” (Ref. A-1), defines LOCAs as those postulated accidents that result from a loss-of-coolant inventory at rates that exceed the capability of the reactor coolant makeup system. Leaks up to a double-ended rupture of the largest pipe of the reactor coolant system (RCS) are included. The MHA LOCA, as with all design basis accidents (DBAs), is a conservative surrogate accident that is intended to challenge aspects of the facility design. Separate mechanistic analyses are performed using a spectrum of break sizes to evaluate fuel and emergency core cooling system performance for conformance with 10 CFR 50.46, “Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors.” With regard to radiological