Document: NRC Regulatory Guide
Document ID: 0cfc3978-32da-4370-9f31-8a3547d82846
Document Type: regulatory_guide
Title: Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Rev. 0)
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML0833/ML083300022.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.25
CFR Part: 
CFR Title: 

Content:
el, the fuel transfer equipment, the fuel pool, and the methods used to handle discharged fuel should all be considered in arriving at the number of fuel assemblies or rods assumed to be damaged, this guide rather than being addressed to this determination is addressed to the determination of the radiological consequences of a handling accident once an assumption as to the number of assemblies or rods damaged has been made. A conservative approach to determining the quantity of radioactive material available for release from a fuel assembly is to assume that the assembly with the peak inventory is the one damaged. The inventory for the peak assembly represents an upper limit value and is not expected to be exceeded. The inventory should be calculated assuming maximum full power operation at the end of core life immediately preceding shutdown and such calculation should include an appropriate radial peaking factor. Only that fraction of the fission products which migrates from the fuel matrix to the gap and plenum regions during normal operation would be available for immediate release into the water in the event of clad damage. (Migration of fission products is a function of several variables including operating temperature, burnup, and isotopic half life taken in to consideration in establishing the release fractions listed in this guide.) As compared to the quantity immediately released, the quantity of radioactive material released subsequent to the immediate release is considered for the purposes of this guide to be negligible. The assumptions set forth in this guide are based on engineering judgment and results from safety research programs conducted by the AEC and the nuclear industry and are believed to be appropriately conservative. In some cases unusual site characteristics, plant design features, or other factors may require different assumptions which will be considered on an individual case basis. Major changes in fuel composition or management may also