Document: NRC Regulatory Guide
Document ID: 2704425a-c58a-45c4-93ab-8761721c3e7a
Document Type: regulatory_guide
Title: Evaluation of Reactor Pressure Vessels with Charpy Upper-Shelf Energy Less Than 50 Ft-Lb
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML0037/ML003740038.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.161
CFR Part: 
CFR Title: 

Content:
he 50 ft-lb limit of Appendix G to 10 CFR Part 50. The analysis methods in the Regulatory Position are based on methods developed for the ASME Code, Section XI, Appendix K ( 8). The staff has reviewed the analysis methods in Appendix K and finds that they are echnically acceptable but are not complete, because Appendix K does not provide information on the selection of transients and gives very little detail on the selection of material properties. In this regulatory guide, specific guidance is provided on selecting transients for consideration and on appropriate material properties to be used in the analyses. Ductile tearing is the dominant fracture process in the upper-shelf region of the Charpy impact energy versus a .aim fnveor RPV materials. The conditions govern ing cleavage mode-conversion of the ductile tearing process in materials with low Charpy upper-shelf energy are still not well understood and are not considered in this regulatory guide. The material property needed to characterize ductile taring in the analysis methods in this regatory guide is the material's J-integral fracture resistance, the J-R curve. This curve is a function of the material, the irradiation condition, the loading rate, and the material temperature. The curve is detrmined by testing the specific material, uider the condi tions of interest, in accordance with the American Society for 1.161-1 Testing and Materials Standard Test Method E 1152-87, "Standard Test Method for Determining J-R Curves' (Ref. 9). Unfortunately, the specific material of interest (Le., the material from the beltline region of the reactor vessel under operation) is seldom available for testing. Thus, testing programs have used generic materials that are expected to represent the range of actual materials used in fabricating reactor pressure vessels in the United States. Statistical analyses of these generic data have been performed and reported in NUREG/CR-5729, -Multivariable Modeling of Pressure Vessel and