Document: NUREG-0800
Document ID: 4a96f7f8-16d2-41a4-a436-12b528d783d7
Document Type: srp
Title: REACTOR VESSEL INTEGRITY
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0631/ML063190008.pdf
Revision Date: 2023-06
Chapter: 5
Section ID: 5.3.3
CFR Part: 
CFR Title: 

Content:
�Pressure-Temperature Limits and Pressurized Thermal Shock.” (5) SRP Section 3.13, “Threaded Fasteners - ASME Code Class 1, 2, and 3.” We have reviewed all factors contributing to the structural integrity of the reactor vessel and conclude there are no special considerations that make it necessary to consider potential reactor vessel failure for this plant. The bases for our conclusion are that the design, materials, fabrication, inspection, and quality assurance requirements for the plant will conform to applicable NRC regulations and regulatory guides, and to the rules of the ASME Boiler and Pressure Vessel Code, Section III. The stringent fracture toughness requirements of the regulations and ASME Code Section III will be met, including requirements for surveillance of vessel material properties throughout service life, in accordance with Appendix H of 10 CFR Part 50. Also, operating limitations on temperature and pressure will be established for this plant in accordance with Appendix G, “Protection Against Non-ductile Failure,” of ASME Code Section III, Appendix G to 10 CFR Part 50, and 10 CFR 50.61 (for PWRs). The integrity of the reactor vessel is assured because the vessel (1) will be designed and fabricated to the high standards of quality required by the ASME Boiler and Pressure Vessel Code and any pertinent Code Cases; (2) will be made from materials of controlled and demonstrated high quality; (3) will be subjected to extensive preservice inspection and testing to provide assurance that the vessel will not fail because of material or fabrication deficiencies; (4) will be operated under conditions and procedures and with protective devices that provide assurance that the reactor vessel design conditions will not be exceeded during normal reactor operation, maintenance, testing, and anticipated operational occurrences; (5) will be subjected to periodic inspection to demonstrate that the high initial quality of the reactor vessel has not