Document: 10 CFR Part 52
Document ID: 9805c5dc-79e7-4f14-88c7-530acc92c17e
Document Type: cfr
Title: Contents of applications; technical information in final safety analysis report.
Source: 10 CFR Part 52
Source URL: https://www.ecfr.gov/current/title-10/part-52/section-52.79
Revision Date: 
Chapter: 
Section ID: 52.79
CFR Part: 52
CFR Title: 10

Content:
design criteria for the facility. Appendix A to part 50 of this chapter , “General Design Criteria for Nuclear Power Plants,” establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have previously been issued by the Commission and provides guidance to applicants in establishing principal design criteria for other types of nuclear power units; ( ii ) The design bases and the relation of the design bases to the principal design criteria; ( iii ) Information relative to materials of construction, arrangement, and dimensions, sufficient to provide reasonable assurance that the design will conform to the design bases with adequate margin for safety. ( 5 ) An analysis and evaluation of the design and performance of structures, systems, and components with the objective of assessing the risk to public health and safety resulting from operation of the facility and including determination of the margins of safety during normal operations and transient conditions anticipated during the life of the facility, and the adequacy of structures, systems, and components provided for the prevention of accidents and the mitigation of the consequences of accidents. Analysis and evaluation of ECCS cooling performance and the need for high-point vents following postulated loss-of-coolant accidents shall be performed in accordance with the requirements of §§ 50.46 and 50.46a of this chapter ; ( 6 ) A description and analysis of the fire protection design features for the reactor necessary to comply with 10 CFR part 50, appendix A , GDC 3, and § 50.48 of this chapter ; ( 7 ) A description of protection provided against pressurized thermal shock events, including projected values of the reference temperature for reactor vessel beltline materials as defined in §§ 50.60 and 50.61(b)(1) and (b)(2) of this chapter ; ( 8 ) An analysis and description of the equipment