Document: NUREG-0800
Document ID: 145ec0f9-012b-4669-9627-ed1b1d0cce95
Document Type: srp
Title: THERMAL AND HYDRAULIC DESIGN
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070412.pdf
Revision Date: 2023-06
Chapter: 4
Section ID: 4.4
CFR Part: 
CFR Title: 

Content:
ence level, that the hot rod in the core does not experience a departure from nucleate boiling or boiling transition condition during normal operation or anticipated operational occurrences; or b. For DNBR, CHFR or CPR correlations, the limiting (minimum) value of DNBR, CHFR, of CPR is to be established such that at least 99.9% of the fuel rods in the core would not be expected to experience departure from nucleate boiling or boiling transition during normal operation or anticipated operational occurrences. Correlations of critical heat flux are continually being revised as a result of additional experimental data, changes in fuel assembly design, and improved calculational techniques involving coolant mixing and the effect of axial power distributions. As guidance to the reviewer, the correlations listed below have been found acceptable for previously reviewed plants. 4.4-5 DRAFT Rev. 2 - April 1996 a. BWRs - The value of the minimum CPR calculated with the GETAB analysis (Ref. 2Reference 17) will vary for different plants and/or fuel types. Typical 29 values are 1.06 and 1.07. b. PWRs - The value of the minimum DNBR calculated with due allowance for mixing grids (Refs. 3, 4, and 5References 18, 19 and 20) is typically 1.30 using 30 the BAW-2 correlation (Ref. 6Reference 21) or the W-3 correlation (Ref. 31 7Reference 22) . Much lower values, depending upon the test data base and fuel 32 design, are acceptable for more recent correlations such as the WRB-1, CE-1, and BWC. 2. Problems affecting DNBR or CPR limits, such as fuel densification or rod bowing, are accounted for by an appropriate design penalty which is determined experimentally or analytically. Subchannel hydraulic analysis codes such as those described in References 823 and 924 , should be used to calculate local fluid conditions within fuel assemblies 33 for use in PWR DNB correlations. The acceptability of such codes must be demonstrated by measurements made in large lattice experiments or power