Document: NRC Regulatory Guide
Document ID: da2e0703-3488-44b0-b6d0-089aac7cae3d
Document Type: regulatory_guide
Title: Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML0037/ML003740028.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.154
CFR Part: 
CFR Title: 

Content:
plant-specific Emergency Response Guidelines. However, a sum- mary discussion re1 ati ng the referenced material to the overall subject should be provided. 8.3 Inservice Inspection and Nondestructive Examination Program The use of state-of-the-art nondestructive examination (NDE) techniques could provide an opportunity to decrease any conservatism that might exist in the flaw density value used in the analysis. This decrease in conservatism, however, may be less important than the decrease in uncertainty in the actual flaw density that may result from an examination of this type. Existing inservice inspection programs should be reevaluated to consider incorporation of state-of-the-art examination techniques for inspecting the clad-base metal interface and the near-surface area. This includes plant-unique consideration of the clad surface conditions. Considerati on should be given to increased frequency of inspections. The reliability of the NDE method selected to detect small flaws should be documented. 8.4 Plant Modifications All plant modifications should be evaluated and optimized in light of any com- peting risks that might arise from events other than PTS events to ensure that overall plant safety is appropriately balanced. PI ant modifications that may be considered include the following: 1. Instrumentation, Controls, and Operation a. Reactor vessel downcomer water temperature monitor. b. Instantaneous and integrated reactor coolant system cooldown rate monitors. c. Steam dump interlock. d. Feedwater i sol ation/f low control 1 ogic. e. Reactor coolant system .pressure and temperature monitors. f. Control system to prevent repressurization of the reactor primary coolant system during overcooling events. g. Monitor to measure margin between vessel inner-surface temperature and current RTNDT at that location. h. Diagnostic instrumentation and displays. i. Primary coolant system pump trip logic.. j. Automatic isolation of auxiliary feedwater to broken steam 1 i