Document: NUREG-0800
Document ID: ff5838f8-986a-4d8b-a039-fb26df280426
Document Type: srp
Title: – 15.5.2
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070725.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.5.1
CFR Part: 
CFR Title: 

Content:
which an estimate of the number of potential fuel failures shall be provided for radiological dose calculations. For such accidents, fuel failure must be assumed for all rods for which the 36 DNBR or CPR falls below those valvesvalues cited above for cladding integrity unless 37 it can be shown, based on an acceptable fuel damage model (see SRP Section 4.2) that 15.5.1-5 DRAFT Rev. 2 - April 1996 fewer failures occur. There shall be no loss of function of any fission product barrier other than the fuel cladding. e. To meet the requirements of General Design Criteria 10, 15 and 26 the guidelines of Regulatory Guide 1.105, "Instrument Spans and SetpointsInstrument Setpoints for Safety-Related Systems," are used with regard to their impact on the plant response to 38 the type of transient addressed in this SRP section. f. The most limiting plant systems single failure, as defined in the "Definitions and Explanations" of Appendix A to 10 CFR Part 50, shall be identified and assumed in the analysis and shall satisfy the guidelines stated in Regulatory Guide 1.53, "Application of the Single-Failure Criterion to Nuclear Power Plant Protection Systems." (Ref. 14).39 The applicant's analysis of events leading to an increase of reactor coolant inventory should be performed using an acceptable analytical model. The equations, sensitivity studies, and models described in References 5 8, 10, and 12 through 8 15 are acceptable. If other analytical 40 methods are proposed by the applicant, these methods are evaluated by the staff for acceptability. For new generic methods, the reviewer initiates an evaluation. The values of parameters used in the analytical model should be suitably conservative. The following values are considered acceptable for use in the model: a. The initial power level is taken as the licensed core thermal power for the number of loops initially assumed to be operating plus an allowance of 2% to account for power measurement uncertainties, unless a lower