Document: NUREG-0800
Document ID: 070a558a-dd64-4d4c-b104-780f0d1c49a3
Document Type: srp
Title: CONTAINMENT FUNCTIONAL DESIGN
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0523/ML052340652.pdf
Revision Date: 2023-06
Chapter: 6
Section ID: 6.2.1
CFR Part: 
CFR Title: 

Content:
shment Project in Japan for Period Ending June 1965 (No. 1)," prepared for the National Reactor Testing Station, February 28, 1966 (unpublished work). 11. H. Uchida, A. Oyama, and Y. Toga, "Evaluation of Post-Incident Cooling Systems of Light-Water Power Reactors," Proc. Third International Conference on the Peaceful Uses of Atomic Energy, Volume 13, Session 3.9, United Nations, Geneva (1964). 12. "FLOOD/MODOO2 - A Code to Determine the Core Reflood Rate for a PWR Plant with Two Core Vessel Outlet Legs and Four Core Vessel Inlet Legs," Interim Report, Aerojet Nuclear Company, November 2, 1972. 13. "FLOOD/MODOO1 - A Code to Determine the Core Reflood Rate for a PWR Plant with Two Core Vessel Outlet Legs and Two Core Vessel Inlet Legs," Interim Report, Aerojet Nuclear Company, October 11, 1972. 14. "COMPARE: A Computer Program for the Transient Calculation of a System of Volumes Connected by Flowing Vents," LA-NUREG-6488-MS, September 1976. 15. RELAP4 MODS, A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems Users Manual, ANCR-NUREG-1335, September 1976. 16. F. J. Moody, "Maximum Flow Rate of a Single Component, Two-Phase Mixture," Jour. of Heat Transfer, Trans. Am. Soc. of Mechanical Engineers, Vol. 87, No. 1, February 1965. 17. "CRAFT-2 Fortran Program for Digital Simulation of a Multinode Reactor Plant During a Loss-of-Coolant Accident," BAW-10092, Babcock and Wilcox Company, December 1974. 6.2.1-5 Rev. 2 - July 1981 18. "Westinghouse Mass and Energy Release Data for Containment Design," WCAP-8312, Westinghouse Electric Corporation, March 1974. 19. "NRC Safety Evaluation Report - Standard Reference System, CESSAR System 80," Combustion Engineering Inc., December 1975. 20. F. C. Cadek et al., "PWR FLECHT (Full-Length Emergency Cooling Heat Transfer), Final Report," WCAP-7665, Westinghouse Electric Corporation, April 1971. 21. Final Safety Analysis Report for Donald C. Cook Nuclear Plant, Units 1 and 2, Appendices M and