Document: NUREG-0800
Document ID: a1ba9094-2225-4b8b-aadf-92e70968c29a
Document Type: srp
Title: REACTOR VESSEL INTEGRITY
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070432.pdf
Revision Date: 2023-06
Chapter: 5
Section ID: 5.3.3
CFR Part: 
CFR Title: 

Content:
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.37 V. IMPLEMENTATION The following is intended to provide guidance to applicants and licensees regarding the NRC staff's plans for using this SRP section. 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 38 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.39 Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regulations and regulatory guide. 40 VI. REFERENCES 1. Standard Review Plan Section 5.2.3, "RCPB Materials." 2. Standard Review Plan Section 5.2.4, "RCPB Inservice Inspection and Testing." 3. Standard Review Plan Section 5.3.1, "Reactor Vessel Materials." 4. Standard Review Plan Section 5.3.2, "Pressure-Temperature Limits."41 5.3.3-9 DRAFT Rev. 2 - April 1996 91. 10 CFR Part 50, Section 50.55a, "Codes and Standards." 42 2. 10 CFR 50.60, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation."43 3. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events."44 54. 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants." (Criterion 1, "Quality Standards and Records;."45 5. 10 CFR 50, Appendix A, General Design