Document: NUREG-0800
Document ID: b486e8d3-c8b0-4990-b7ec-211a3aae25c3
Document Type: srp
Title: CONTAINMENT FUNCTIONAL DESIGN
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070456.pdf
Revision Date: 2023-06
Chapter: 6
Section ID: 6.2.1
CFR Part: 
CFR Title: 

Content:
port - Standard Reference System, CESSAR System 80," Combustion Engineering Inc., December 1975. 29 2126. Final Safety Analysis Report for Donald C. Cook Nuclear Plant, Units 1 and 2, Appendices M and N, American Electric Power Company, and the Staff Safety Evaluation Report. AEC Docket Nos. 50-315/316. 30 2427. Branch Technical Position CSB 6-1, "Minimum Containment Pressure Model for PWR ECCS Performance Evaluation," attached to SRP Section 6.2.1.5. 328. ASME Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NE, "Class MC Components," American Society of Mechanical Engineers. 31 5. Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors." 32 629. C. F. Carmichael and S. A. Marks, "CONTEMPT-PS, A Digital Computer Code for Predicting the Pressure-Temperature History Within a Pressure Suppression Containment Vessel in Response to a Loss-of-Coolant Accident," IDO-17252, Phillips Petroleum Company, April 1969. 33 730. L. C. Richardson, L. J. Finnegan, R. J. Wagner, and J. M. Waage, "CONTEMPT, A Computer Program for Predicting the Containment Pressure-Temperature Response to a Loss-of-Coolant Accident," IDO-17220, Phillips Petroleum Company, June 1967. 34 831. R. J. Wagner and L. L. West, "CONTEMPT-LT Users Manual," Interim Report I-214-74-12.1, Aerojet Nuclear Company, August 1973. 932. R. I. Miller, "Evaluation of the Predictive Capabilities of the CONTEMPT-PS Computer Code by Comparison of Calculated Results with the Humboldt Bay and Bodega Bay Pressure Suppression Tests," Interim Report 4.2.1.1, Idaho Nuclear Corporation, September 1970. 1033. T. Tagami, "Interim Report on Safety Assessments and Facilities Establishment Project in Japan for Period Ending June 1965 (No. 1)," prepared for the National Reactor Testing Station, February 28, 1966 (unpublished work). 1134. H. Uchida, A. Oyama, and Y. Toga, "Evaluation of Post-Incident Cooling Systems of