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:
n all regulatory matters in a manner that complements the NRC’s deterministic approach and supports the traditional defense in depth philosophy, which continues to be an effective way to account for uncertainties in equipment and human performance. In 1995, the NRC published NUREG-1465, “Accident Source Terms for Light-Water Nuclear Power Plants” (Ref. 8), which provides estimates of the accident source term that are more physically 1 The maximum hypothetical accident (MHA) (also referred to as the maximum credible accident) is that accident whose consequences, as measured by the radiation exposure of the surrounding public, would not be exceeded by any other accident whose occurrence during the lifetime of the facility would appear to be credible. As used in this guide, the term “LOCA” refers to any accident that causes a loss of core cooling. The MHA LOCA refers to a loss of core cooling resulting in substantial meltdown of the core with subsequent release into containment of appreciable quantities of fission products. These evaluations assume containment integrity with offsite hazards evaluated based on design basis containment leakage. 2 The MHA should be modeled with the deterministic substantial fuel melt source term being injected or overlaid into the radiological consequence analysis notwithstanding the operation of safety-related equipment designed to preclude significant fuel failure. The purpose of this approach would be to test the adequacy of the containment and other safety-related systems. Safety-related systems may be credited as described in Regulatory Position 5.1.2, as this designation ensures reliability in performing their safety function. DG-1389, Page 6 based and that could be applied to the design of advanced LWRs. NUREG-1465 presents a representative accident source term for a BWR and for a pressurized-water reactor (PWR). These source terms are characterized by the composition and magnitude of the radioactive material, the