Document: NUREG-0800
Document ID: 94ab38ac-ddfc-4ba1-ae38-bc25ffa6e976
Document Type: srp
Title: – 15.4.5
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070716.pdf
Revision Date: 2023-06
Chapter: 15
Section ID: 15.4.4
CFR Part: 
CFR Title: 

Content:
tions of the reactor coolant pressure boundary are not exceeded because the protection system operates to maintain the maximum pressure within the reactor coolant and main steam system pressures below 110% of the design values. 3. The applicant has met the positions of Regulatory Guide 1.53 as related to the single-failure criterion and Regulatory Guide 1.105 as related to instrument actuations of systems and components important to safety. For design certification reviews, the findings will also summarize, to the extent that the review is not discussed in other safety evaluation report sections, the staff’s evaluation of inspections, tests, analyses, and acceptance criteria (ITAAC), including design acceptance criteria (DAC), site interface requirements, and combined license action items that are relevant to this SRP section.40 V. IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section. 15.4.4-11 DRAFT Rev. 2 - April 1996 This SRP section will be used by the staff when performing safety evaluations of license applications submitted by applicants pursuant to 10 CFR 50 or 10 CFR 52. Except in those 41 cases in which the applicant proposes an acceptable alternative method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its evaluation of conformance with Commission regulations. The provisions of this SRP section apply to reviews of applications docketed six months or more after the date of issuance of this SRP section.42 Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regulatory guides. VI. REFERENCES 1. Regulatory Guide 1.70, "Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants." 2. ASME Boiler and Pressure Vessel Code, Section III, "Nuclear Power Plant Components," Article NB-7000, "Protection Against