Document: NUREG-0800
Document ID: 8da52d2b-9980-4076-8056-2cafebb25ed6
Document Type: srp
Title: THERMAL AND HYDRAULIC DESIGN
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0705/ML070550060.pdf
Revision Date: 2023-06
Chapter: 4
Section ID: 4.4
CFR Part: 
CFR Title: 

Content:
are accounted for by an appropriate design penalty which is determined experimentally or analytically. Subchannel hydraulic analysis codes, such as those described in “TEMP- Thermal Enthalpy Mixing Program,” BAW-10021, Babcock and Wilcox Company, 4.4-6 Revision 2 - March 2007 April 1970 and “THINC-IC-An Improved Program for Thermal-Hydraulic Analysis Of Rod Bundle Cores,” WCAP-7956, Westinghouse Electric Corporation, June 1973, should be used to calculate local fluid conditions within fuel assemblies for use in PWR DNB correlations. The acceptability of such codes must be demonstrated by measurements made in large lattice experiments or power reactor cores. The review should include the effects of radial pressure gradients in the core flow distribution. The reviewer should also confirm that calculations of BWR fluid conditions for use in CHF correlations have been made in accordance with the models specified in “Loss of Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors,” NEDO-10329, Appendix C, General Electric Company, April 1971 and “General Electric Company Analytical Model for Loss of Coolant Accident Analysis in Accordance with 10 CFR Part 50, Appendix K, “NEDO-20566, General Electric Company, November 1975. 3. The design should address core oscillations and thermal-hydraulic instabilities as described in SRP Section 15.9. 4. Methods for calculating single-phase and two-phase fluid flow in the reactor vessel and other components should include classical fluid mechanics relationships and appropriate empirical correlations. For components of unusual geometry, such as those listed below, these relationships should be confirmed empirically using representative databases from approved reports: A. Reactor vessel (“Reactor Vessel Model Flow Tests,” BAW-10037 (nonproprietary version of BAW-10012), Rev. 2, Babcock and Wilcox Company, September 1968). B. Jet pump (“Design and Performance of General