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:
haracteristics, 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 require alterations of these assumptions. C. REGULATORY POSITION 1. The assumptions1 related to the release of radioactive material from the fuel and fuel storage facility as a result of a fuel handling accident are: a. The accident occurs at a time after shutdown identified in the technical specifications as the earliest time fuel handling operations may begin. Radioactive decay of the fission product inventory during the interval between shutdown and commencement of fuel handling operations is taken into consideration. 1 The assumptions given are valid only for oxide fuels of the types currently in use and in cases where the following conditions are not exceeded: a. Peak linear power density of 20.5 kW/ft for the highest power assembly discharged. b. Maximum center-line operating fuel temperature less than 4500°F for this assembly. c. Average burnup for the peak assembly of 25,000 MWD/ton or less (this corresponds to a peak local burnup of about 45,000 MWD/ton). RG-1.25, Page 3 b. The maximum fuel rod pressurization2 is 1200 psig. c. The minimum water depth2 between the top of the damaged fuel rods and the fuel pool surface is 23 feet. d. All of the gap activity in the damaged rods is released and consists of 10% of the total noble gases other than Kr-85,30% of the Kr-85, and 10% of the total radioactive iodine in the rods at the time of the accident. For the purpose of sizing filters for the fuel handling accident addressed in this guide, 30% of the I-127 and I-129 inventory is assumed to be released from the damaged rods. e. The values assumed for individual fission product inventories are calculated assuming full power operation at the end of core life immediately preceding shutdown and such calculation should include an appropriate radial peaking factor. The