Document: NUREG-0800
Document ID: 3ea2f0ac-4d7e-464a-b1c4-390c3970f642
Document Type: srp
Title: provides specific thermal-hydraulic criteria.  The available radioactive fission product
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0707/ML070740002.pdf
Revision Date: 2023-06
Chapter: 4
Section ID: 4.4
CFR Part: 
CFR Title: 

Content:
s at the appropriate temperature. Other proposed limits must be justified. Results from the seismic and LOCA analysis (see Appendix A to this SRP section) may show that failures by this mechanism will not occur for less severe events. C. Fuel Coolability This subsection applies to postulated accidents, and Chapter 15 of the safety analysis report will contain most of the information to be reviewed. Item 1.C.v below addresses the combined effects of two accidents, and Section 4.2 of the safety analysis report should include that information. To meet the requirements of GDC 27 and 35 as they relate to control rod insertability and core coolability 4.2-11 Revision 3 - March 2007 for postulated accidents, fuel coolability criteria should be provided for all severe damage mechanisms. Coolability, or coolable geometry, has traditionally implied that the fuel assembly retains its rod-bundle geometry with adequate coolant channels to permit removal of residual heat. Reduction of coolability can result from cladding embrittlement, violent expulsion of fuel, generalized cladding melting, gross structural deformation, and extreme coplanar fuel rod ballooning. This subsection also addresses control rod insertability criteria. Complete criteria should address the following: i. Cladding Embrittlement. The ECCS performance analysis must satisfy the fuel design criteria specified within 10 CFR 50.46(b). These criteria ensure a coolable core geometry by preserving adequate postquench ductility in the fuel rod cladding. The current criteria require that (1) the peak cladding temperature remains below 2200 EF and (2) the peak cladding oxidation remains below 17 percent ECR. These criteria were originally developed on the basis of unirradiated Zircaloy test specimens. Zirconium alloy composition, manufacturing process, and in-reactor corrosion alter the postquench characteristics of the fuel cladding material. Rulemaking pursuant to 10 CFR 50.46 is planned to implement a