Document: NUREG-0800
Document ID: 5e3dbec9-a39f-4531-9eb9-705982190066
Document Type: srp
Title: REACTOR PRESSURE VESSEL INTERNALS
Source: NUREG-0800
Source URL: https://www.nrc.gov/docs/ML1613/ML16134A059.pdf
Revision Date: 2023-06
Chapter: 3
Section ID: 3.9.5
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CFR Title: 

Content:
h the design details to be finalized as part of the as-built analysis and frequency response testing when demonstrating compliance with the applicable ITAAC. Therefore, when reviewing a DC application for a BWR nuclear power plant design, the NRC staff reviewer should verify that the applicant’s FSAR or DCD specifies as Tier 2* information critical aspects of the methodology for completing the final design of the steam dryer, such that it may not be modified without NRC staff review and approval. Review Interfaces Other SRP sections interface with this section as follows: 1. Evaluation of rupture locations, rupture loads, and dynamic effects of postulated rupture of piping is performed under SRP Section 3.6.2, “Plant Design for Protection against Postulated Piping Failures in Fluid Systems Outside Containment.” 2. Evaluation of the adequacy of analyses justifying exclusion of certain postulated pipe ruptures from design bases is performed under SRP Section 3.6.3, “Leak-Before-Break Evaluation Procedures.” 3. Evaluation of the adequacy of analysis methods for seismic Category I RPV internals and system dynamic analysis, identification of design transients and of service lifetime transient cyclic loadings to be reflected in the design and fatigue analyses of RPV internals is performed under SRP Section 3.9.1, “Special Topics for Mechanical Components.’ 4. Evaluation of the adequacy of dynamic analyses under steady state and operational flow transient conditions and the proposed program for preoperational and startup testing of RPV internals against various vibration excitation mechanisms (such as FIV, AR, AIV, and MIV) is performed under SRP Section 3.9.2. 5. Evaluation of the adequacy of the structural integrity design of the RPV internals, including adequacy of the design fatigue curves for the reactor internals materials to account for cumulative service-related and environmental usage factor effects, and consideration of each combination of design,