Document: NRC Regulatory Guide
Document ID: ad61f8a3-1cce-4446-9542-dcdda55c1ec6
Document Type: regulatory_guide
Title: Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing + HISTORY - HISTORY 07/2015 – DG-1323 , Proposed Revision 4 03/2013 – Periodic Review of Revision 3 – No Issues Identified 11/2006 – DG-1163 , Proposed Revision 3 (Rev. 4)
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML1508/ML15083A390.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.20
CFR Part: 
CFR Title: 

Content:
se flow effects at BWR and PWR nuclear power plants, including SMRs, during design, construction, and operation, including situations when power uprates or major plant modifications are proposed. This program includes the analytical methodologies, assumptions, computer programs, and code and code edition for the evaluation of the plant components, including the method of determining the load definition and the uncertainties and bias errors of analytical and measurement procedures. The program also includes a comparison of component stresses against code allowable limits. Finally, the program includes testing methods, instrumentation, and measurements. Adverse effects in reactors caused by FIV, AR, AIV and MIV can be sensitive to minor changes in arrangement, design, size, and operating conditions. For two nominally identical nuclear power plants, one might experience significant adverse flow effects, such as valve and steam dryer failures, while the other does not. Relatively small changes in operating conditions can cause a previously small adverse flow effect to be magnified, leading to structural failures. For example, severe acoustic excitation occurred in the steam system of one BWR nuclear power plant when flow was increased by 16 percent for EPU operation. Also, a steam dryer in another BWR plant experienced fatigue cracking caused by the reactor pump excitation at its vane passing frequency (VPF). Specific guidance for these assessments, both predictive and measurement-based, is provided in this regulatory guide. In developing a suitable measurement program, it is essential that the selected locations for vibration and acoustic monitoring instrumentation be evaluated for potential effects on the component and system dynamic response. Operating experience has revealed failures of steam dryers and main steam system components (including relief valves) in BWR nuclear power plants following EPU implementation. These failures have demonstrated the importance of