Document: 10 CFR Part 52
Document ID: e9c3b096-4c13-4284-8ed2-18c06d0500ef
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-52.157
Revision Date: 
Chapter: 
Section ID: 52.157
CFR Part: 52
CFR Title: 10

Content:
hat the manufactured reactor complies with the earthquake engineering criteria in appendix S to 10 CFR part 50; ( 15 ) Information sufficient to demonstrate compliance with the applicable requirements regarding testing, analysis, and prototypes as set forth in § 50.43(e) of this chapter ; ( 16 ) The technical qualifications of the applicant to engage in the proposed activities in accordance with the regulations in this chapter; ( 17 ) A description of the quality assurance program applied to the design, and to be applied to the manufacture of, the structures, systems, and components of the reactor. Appendix B to 10 CFR part 50, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” sets forth the requirements for quality assurance programs for nuclear power plants. The description of the quality assurance program must include a discussion of how the applicable requirements of appendix B to 10 CFR part 50 have been and will be satisfied; and ( 18 ) Proposed technical specifications applicable to the reactor being manufactured, prepared in accordance with the requirements of §§ 50.36 and 50.36a of this chapter ; ( 19 ) The site parameters postulated for the design, and an analysis and evaluation of the reactor design in terms of those site parameters; ( 20 ) The interface requirements between the manufactured reactor and the remaining portions of the nuclear power plant. These requirements must be sufficiently detailed to allow for completion of the final safety analysis; ( 21 ) Justification that compliance with the interface requirements of paragraph (f)(20) of this section is verifiable through inspections, testing, or analysis. The method to be used for verification of interface requirements must be included as part of the proposed ITAAC required by § 52.158(a) ; ( 22 ) A representative conceptual design for a nuclear power facility using the manufactured reactor, to aid the NRC in its review of the final safety analysis