Document: NUREG-0800
Document ID: 228b9b6e-c81e-4d76-9b8e-12f6c3bd7fa1
Document Type: srp
Title: and Regulatory Guide 1.92,(Reference 10) "Combining Modal Responses
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML0520/ML052070336.pdf
Revision Date: 2023-06
Chapter: 3
Section ID: 3.7.2
CFR Part: 
CFR Title: 

Content:
ons, test acceptance criteria and bases, and permissible deviations from the criteria should be provided before the test. g. Visual and nondestructive surface inspections should be performed after the completion of the vibration tests. The inspection program description should include the areas subject to inspection, the methods of inspection, the design access provisions to the reactor internals, and the equipment to be used for performing such inspections. These inspections should be conducted preferably following the removal of the internals from the reactor vessel. Where removal is not feasible, the inspections should be performed by means of equipment appropriate for in situ inspection. The areas inspected should include all load-bearing interfaces, core restraint devices, high stress locations, and locations critical to safety functions. For internals of subsequent reactors that have the same design, size, configuration, and operating conditions as the prototype reactor internals, the vibration test program should conform to the requirements of the appropriate non-prototype program as specified in Regulatory Guide 1.20. 5. Relevant requirements of GDC General Design Criteria 2 and 42, 4, 14, and 15 are met 45 as given below. Dynamic system analyses should be performed to confirm the structural design adequacy of the reactor internals and the reactor coolant piping (unbroken loops) to withstand the dynamic loadings of the most severe LOCA in combination with the SSE. Where a substantial separation between the forcing frequencies of the LOCA (or SSE) loading and the natural frequencies of the internal structures can be demonstrated, the analysis may treat the loadings statically. The most severe dynamic effects from LOCA loadings are generally found to result from a postulated double-ended rupture of a primary coolant loop near a reactor vessel inlet or outlet nozzle with the reactor in the most critical normal operating mode. However, all other postulated break