Document: NUREG-0800
Document ID: c9c204f0-a162-491c-8c25-ee0418212f29
Document Type: srp
Title: PRESSURE-TEMPERATURE LIMITS, UPPER-SHELF ENERGY, AND PRESSURIZED
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0703/ML070380185.pdf
Revision Date: 2023-06
Chapter: 5
Section ID: 5.3.2
CFR Part: 
CFR Title: 

Content:
ust be submitted to the staff for review and approval on an individual case basis at least 3 years prior the date on which the predicted Charpy USE will no longer satisfy the requirements of paragraph IV.A.1.a, or on a schedule approved by the Director, Office of Nuclear Regulation. In addition to the ASME Code, Regulatory Guide 1.161 provides an acceptable methodology for the performance of analyses intended to meet the provisions for additional analysis specified in paragraph IV.A.1.a. C. Pressurized Thermal Shock in Pressurized-Water Reactors. The reviewer will evaluate the projected values for RTPTS, including the calculational methods and assumptions, and compare the projected values with the screening criteria in 5.3.2-13 Revision 2 - March 2007 10 CFR 50.61. For each PWR where the RTPTS value for any material in the beltline is projected to exceed the PTS screening criterion before the expiration date of the operating license, the licensee should submit an analysis and schedule for the implementation of flux reduction programs that are reasonably practical to avoid exceeding the PTS screening criterion. If the analysis indicates that no reasonably practical flux reduction program will prevent the value of RTPTS from exceeding the PTS screening criterion before the expiration date of the operating license, the licensee can choose between the two options in 10 CFR 50.61 to meet PTS requirements. The licensee can submit a safety analysis to determine the modifications necessary to prevent potential failure of the reactor vessel as a result of postulated PTS events if continued operation beyond the screening criterion is allowed. The staff will review these safety analyses against the requirements of 10 CFR 50.61 and the guidance of Regulatory Guide 1.154. Alternatively, the licensee can perform a thermal-annealing treatment of the reactor vessel pursuant to 10 CFR 50.61(b)(7) to recover fracture toughness. In accordance with 10 CFR 50.61, the licensee must