Document: NUREG-0800
Document ID: 855b2438-2ddf-48e9-8762-e39097109e12
Document Type: srp
Title: FUEL SYSTEM DESIGN
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052340660.pdf
Revision Date: 2023-06
Chapter: 4
Section ID: 4.2
CFR Part: 
CFR Title: 

Content:
staff concludes that the applicant has described methods of adequately predicting fuel rod failures during postulated accidents so that radioactivity releases are not underestimated and thereby meets the related requirements of 10 CFR Part 100. In meeting these requirements, the applicant has (a) used the fission-product release assumptions of Regulatory Guides 1.3 (or 1.4), 1.25, and 1.77 and (b) performed the analysis for fuel rod failures for the rod ejection accident in accordance with the guidelines of Regulatory Guide 1.77 or with methods that the staff has reviewed and found to be an acceptable alternative to Regulatory Guide 1.77. V. IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section. Except in those cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations. Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regulatory guides and NUREGs. VI. REFERENCES 1. 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants." 4.2-14 Rev. 2 - July 1981 2. 10 CFR Part 100, "Reactor Site Criteria." 3. 10 CFR Part 50, §50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors." 4. "Rules for Construction of Nuclear Power Plant Components," ASME Boiler and Pressure Vessel Code, Section III, 1977. 5. W. J.-O'Donnel and B. F. Langer, "Fatigue Design Basis for Zircaloy Components," Nucl. Sci. Eng. 20, 1 (1964). 6. Regulatory Guide 1.77, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors." 7. "Standard Specification for Sintered Uranium Dioxide Pellets," ASTM Standard C776-76, Part 45, 1977. 8. K. Joon, "Primary