Document: NRC Regulatory Guide
Document ID: 5f799693-27fd-4e13-a5e1-4c02f393d90a
Document Type: regulatory_guide
Title: Best-Estimate Calculations of Emergency Core Cooling System Performance + HISTORY –HISTORY 04/2013 – Periodic Review of Revision 0 – Reviewed with issues identified for future consideration 03/1987 – Draft RS 701-4, Proposed Revision 0
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML0037/ML003739584.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.157
CFR Part: 
CFR Title: 

Content:
. (Available in the NRC Public Document Room.) 4. Pacific Northwest Laboratory, "COBRA/TRAC - A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Pri mary Coolant Systems," NUREG/CR-3046, 5 Vols. (PNL-4385), March 1983. 5. L. J. Siefken et al., "FRAP-T6: A Computer Code for the Transient Analysis of Oxide Fuel Rods," NUREG/CR-2148 (EG&G, EGG-2104), May 1981. 6. G. A. Berna et al., "FRAPCON-2: A Com puter Code for the Calculation of Steady State Thermal-Mechanical Behavior of Oxide Fuel Rods," NUREG/CR-1845, January 1981. 7. "Compendium of ECCS Research for Realistic LOCA Analysis," NUREG-1230, December 1988. 8. D. Lanning and M. Cunningham, "Trends in Thermal Calculations for Light Water Reactor Fuel (1971-1981)," in Ninth Water Reactor Safety Research Information Meeting, USNRC, NUREG/CP-0024, Vol. 3, March 1982. 9. Idaho National Engineering Laboratory, "MATPRO Version 11 (Revision 2): A Hand book of Materials Properties for Use in the Analysis of Light-Water Reactor Fuel Rod Be havior," NUREG/CR-0497, Rev. 2, August 1981. 10. American Nuclear Society, "American Na tional Standard for Decay Heat Power in Light Water Reactors," ANSI/ANS-5.1-1979, August 1979. (ANS, 555 North Kensington Avenue, La Grange Park, Illinois 60525.) 11. J. V. Cathcart et al., "Zirconium Metal-Water Oxidation Kinetics: IV Reaction Rate Studies," Oak Ridge National Laboratory, ORNL/ NUREG-17, August 1977. (Available from NTIS.) 12. H. J. Richter, "Separated Two-Phase Flow Model: Application to Critical Two-Phase Flow," EPRI Report NP-1800, Electric Power Research Institute, Palo Alto, CA, April 1981. 13. D. Abdollahian et al., "Critical Flow Data Re view and Analysis," Report NP-2192, Electric Power Research Institute, Palo Alto, CA, Janu ary 1982. 14. USNRC, "The Marviken Full Scale Critical Flow Tests, Summary Report," (Joint Reactor Safety Experiments in the Marviken Power Sta tion, Sweden), NUREG/CR-2671, May 1982. 15. M. Reocreux, "Contribution to the Study