Document: NRC Regulatory Guide
Document ID: 3b4c5897-7060-42e3-83a9-cf0790b7dc60
Document Type: regulatory_guide
Title: Acceptability of ASME Code, Section XI, Division 2, Requirements for RIM Programs for NPPs, for Non-LWRs
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML2112/ML21120A185.pdf
Revision Date: 2023-05
Chapter: 
Section ID: RG-1.246
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CFR Title: 

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unless the document requesting or requiring the collection displays a currently valid OMB control number. DG-1383, Page 4 B. DISCUSSION Reason for Issuance The NRC staff is issuing this RG to provide applicants and licensees of non-LWRs an acceptable method for developing and implementing a PSI and ISI program. This RG endorses with conditions ASME Code, Section XI, Division 2 for use by non-LWR licensees. Background NRC regulations in 10 CFR 50.34(b)(6)(iv) and 52.79(a)(29)(i) require all applicants for operating and combined licenses to include plans for conducting normal operations, including maintenance, surveillance, and periodic testing of SSCs. However, the regulations prescribe specific preservice and inservice inspection program requirements only for boiling and pressurized water-cooled nuclear power reactors.1 Nevertheless, as described below, the GDCs in 10 CFR Part 50, Appendix A, as applicable to non-LWR designs, indicate the importance of an adequate preservice and inservice inspection program. RG 1.232 provides guidance on how the GDCs in 10 CFR Part 50, Appendix A may be adapted for non-LWR designs.2 Appendix A to RG 1.232 contains the advanced reactor design criteria (ARDC). These criteria are generally applicable to six different types of non-LWR technologies (i.e., sodium- cooled fast reactors, lead-cooled fast reactors, gas-cooled fast reactors, modular high-temperature gas- cooled reactors, fluoride high-temperature reactors, and molten salt reactors). Within Appendix A to RG 1.232 are several ARDC that relate to SSC testing: • ARDC-1 states that SSCs important to safety are to be tested to quality standards commensurate with the importance of the safety functions to be performed. • ARDC-14 states that the reactor coolant boundary shall be tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. • ARDC-30 indicates that the components that are part of the reactor coolant