Document: NRC Regulatory Guide
Document ID: 5a810f5d-d35a-4bd8-bde7-205519300c40
Document Type: regulatory_guide
Title: Thermal Shock to Reactor Pressure Vessels
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML1221/ML12216A018.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.2
CFR Part: 
CFR Title: 

Content:
11/2/70 (Reprinted 12/1/70) SAFETY GUIDE 2 THERMAL SHOCK TO REACTOR PRESSURE VESSELS A. Introduction Proposed General Design Criterion 35 speci- fies design and operating conditions necessary to assure that the reactor coolant pressure boundary will behave in a nonbrittle manner. To provide protection against loss of coolant accidents, present designs provide for the in- jection of large quantities of cold emergency coolant into the reactor coolant system. The effect on the reactor pressure vessel of this cold water injection is of concern because the reac- tor vessel is subjected to greater irradiation than other components of the reactor coolant pressure boundary and, thus, has a greater po- tential for becoming brittle. A suitable program which may be used to implement General Design Criterion 35 to assure that the reactor pressure vessel will behave in a nonbrittle manner under loss of coolant accident conditions is described in this guide. B. Discussion The injection of cold water by the emergency core cooling system into a hot reactor pressure vessel after a loss of coolant accident raises the possibility that a vessel embrittled by irradia- tion and having a small internal defect could fail suddenly as a result of the large thermal gradient imposed and the resulting high stresses. Analyses by the reactor vendors indi- cate that cold water injected into a hot reactor pressure vessel toward the end of the vessel's service life could cause incipient defects of the maximum size expected to grow; however, the maximum crack depth is predicted to be no more than 30 to 60 percent of vessel wall thick- ness. The vessel is not expected to fail under these conditions. The maximum crack depth expected cannot be firmly established since the vessel material fracture toughness properties assumed in the analyses have not yet been com- pletely confirmed. The additional data needed to resolve the uncertainties in the. fracture toughness prop- erties of reactor vessel material