ML25105A229
| ML25105A229 | |
| Person / Time | |
|---|---|
| Site: | Nuclear Energy Institute |
| Issue date: | 04/11/2025 |
| From: | Voelsing K Nuclear Energy Institute |
| To: | Office of Administration |
| References | |
| NRC–2024–0183, 90FR12184 00001 | |
| Download: ML25105A229 (1) | |
Text
PUBLIC SUBMISSION As of: 4/15/25, 9:04 AM Received: April 11, 2025 Status: Pending_Post Tracking No. m9d-7gv0-lcxy Comments Due: April 19, 2025 Submission Type: Web Docket: NRC-2024-0183 Guidelines for Inservice Testing at Nuclear Power Plants - Inservice Testing of Pumps and Valves and Inservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants Comment On: NRC-2024-0183-0001 Draft NUREG: Guidelines for Inservice Testing at Nuclear Power Plants: Inservice Testing of Pumps and Valves and Inservice Examination and Testing of Dynamic Restraints (Snubbers) at Nuclear Power Plants Document: NRC-2024-0183-DRAFT-0001 Comment on FR Doc # 2025-04048 Submitter Information Email:atb@nei.org Organization:Nuclear Energy Institute General Comment See attached file(s)
Attachments 04-11-25_NRC_Draft NUREG-1482 Comments 4/15/25, 9:52 AM NRC-2024-0183-DRAFT-0001.html file:///C:/Users/TAW/Downloads/NRC-2024-0183-DRAFT-0001.html 1/1 SUNI Review Complete Template=ADM-013 E-RIDS=ADM-03 ADD: Nick Hansing, Mary Neely Comment (1)
Publication Date: 3/14/2025 Citation: 90 FR 12184
Phone: 202.468.6673 Email: khv@nei.org Kelli Voelsing Director, Engineering April 11, 2025 Office of Administration Mail Stop: TWFN-7-A60M U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTN: Program Management, Announcements and Editing Staff
Subject:
NEI Comments on Draft NUREG-1482, Revision 4, Guidelines for Inservice Testing at Nuclear Power Plants, Docket ID: NRC-2024-0183 Submitted via Regulations.gov Project Number: 689
Dear Program Management,
Announcements, and Editing Staff:
The Nuclear Energy Institute (NEI),1 on behalf of our members, appreciates the opportunity to comment on the Draft NUREG-1482, Revision 4, Guidelines for Inservice Testing at Nuclear Power Plants. We commend the NRCs ongoing efforts to update and clarify guidance related to inservice testing (IST) of pumps, valves, and dynamic restraints.
NEIs principal concern with the draft guidance centers on the inclusion of guidance containing NRC positions related to the implementation of 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. We think that the inclusion of this guidance, specifically Appendix B and Appendix A-4, in NUREG-1482 is inappropriate for the following reasons:
Mismatch in Scope: NUREG-1482 is focused on IST program requirements, whereas 10 CFR 50.69 provides an alternative treatment (AT) pathway for low safety-significant (LSS) components.
1 The Nuclear Energy Institute (NEI) is responsible for establishing unified policy on behalf of its members relating to matters affecting the nuclear energy industry, including the regulatory aspects of generic operational and technical issues. NEIs members include entities licensed to operate commercial nuclear power plants in the United States, nuclear plant designers, major architect and engineering firms, fuel cycle facilities, nuclear materials licensees, and other organizations involved in the nuclear energy industry.
Office of Administration April 11, 2025 Page 2 Nuclear Energy Institute Integrating 50.69 content into an IST-focused document may create confusion regarding regulatory expectations for these components.
Licensee Authority over ATs: The regulation explicitly states that licensees are responsible for determining ATs, rather than the NRC. NRC guidance suggesting acceptable approaches to ATs is inconsistent with the 10 CFR 50.69 rule.
NEI strongly recommends that Appendix B and Appendix A.4 be removed from the final version of NUREG-1482 to avoid misinterpretation and maintain clarity.
In the event that Appendix B is retained in any form, NEI has provided detailed comments in Attachment A identifying specific proposed revisions to align the document with the regulatory framework and intent of 10 CFR 50.69.
We appreciate the NRCs continued consideration of stakeholder comments on regulatory matters and remain available to support further clarification or discussion as needed. Please contact me at khv@nei.org if further information is required.
Thank you for your consideration.
Sincerely, Kelli Voelsing Director, Engineering Attachment A: NEI Comments on Draft NUREG-1482, Rev. 4 cc:
Nicholas Hansing, NRC, NRR Stew Bailey, NRC, NRR Tom Scarborough, NRC, NRR Meena Khanna, NRC, NRR
Appendix A NEI Comments on Draft NUREG-1482, Rev. 4 Page A-1 Section Text Comment Recommendation Section 8.2 Page 8-2 Lines 4-4 In either case, the NRC staff expects the licensees proposal to address the principles described in RG 1.175, including those related to implementation and monitoring The either cases which are referred to in this sentence may confuse the reader. It might mean the case of proposing an alternative to the entire IST program or proposing alternatives of more limited scope. A reader might also think that the either cases referred to licensees pursuing alternatives under 10 CFR 50.55a9(z)(1), the 2020 Code Edition, or 10 CFR 50.69.
This sentence should be clarified to read, When proposing risk-informed alternatives under 10C FR 520.55a(z)(1), the NRC staff expects the licensees proposal to address the principles described in RG 1.175, including those related to implementation and monitoring.
Appendix A.4 Entire
- Section, Pages A.4-1 and A.4-2 Appendix A.4 is intended to identify considerations for the categorization of snubbers under 10 CFR 50.69 NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, has been endorsed by the NRC in RG 1.201, Revision 1 as an acceptable method for performing categorization of SSCs, including snubbers, under 10 CFR 50.69. Additionally, the NRC explicitly approves the categorization processes to be used under 10 CFR 50.69 in each licensees SER.
Because the NRC already has endorsed guidance for performing categorization under 10 CFR 50.69, and because the NRC explicitly reviews and approves the categorization process for each licensee seeking approval for 10 CFR 50.69, there is no need for additional guidance on this topic. This section should either be removed completely or should reference NEI 00-04.
Appendix B
Entire Appendix Appendix B provides guidance on the treatment of pumps, valves, and dynamic restraints under 10 CFR 50.69.
10 CFR 50.69 establishes a framework for licensees to apply alternative treatments (ATs) to low-safety-significant (LSS) components in lieu of complying with inservice testing (IST) requirements. The inclusion of Appendix B in this NRC IST guidance document is inappropriate for the following reasons:
- Mismatch in Scope - The primary purpose of NUREG-1482 is to provide guidance on IST, whereas 10 CFR 50.69 explicitly allows It is requested that the NRC remove Appendix B from the final version of NUREG-1482 to maintain clarity and consistency with the regulatory framework established by 10 CFR 50.69.
Page A-2 Section Text Comment Recommendation licensees to apply ATs as an alternative to IST requirements for LSS components. Including guidance on 50.69 within an IST-focused document creates confusion regarding the regulatory expectations for LSS components.
- Licensee Responsibility for ATs - The 10 CFR 50.69 regulation explicitly places the responsibility for determining ATs on the licensee, not the NRC. The NRC does not have approval authority over ATs selected under 50.69. Any guidance within NUREG-1482 describing what the NRC find acceptable for ATs under 50.69 contradicts the regulations intent and could create unnecessary regulatory burden and misinterpretation.
In the event that the NRC will not accept this recommendation for the complete removal of Appendix B from NUREG-1482, the remaining comments in this table address specific concerns with the wording in Appendix B.
B-4 Lines 21-28 Note that two of the STRs listed in 10 CFR 50.69(b)(1) are 10 CFR Part 21, Reporting of 21 Defects and Noncompliance, and 10 CFR Part 50, Appendix B.
RISC-3 SSCs might have been purchased or dedicated as safety-related components and, as such, were at one time subject to the full range of STRs.
Other components of the same nameplate and/or treatment might be categorized as RISC-1 at the plant or other plants or might be considered safety-related at plants that have not implemented 10 CFR 50.69. Licensees should ensure that operating experience from RISC-3 SSCs is shared, as The draft text states that licensees should ensure that OE from RISC-3 SSCs is shared to support the intent of 10 CFR Part 21 requirements. However, we respectfully disagree with the implied expectation that additional reporting mechanisms are necessary beyond those already established. Specific concerns are as follows:
- The statements of consideration for the final 10 CFR 50.69 Rule (ML022630018 - SECY 02-0176 Attachment 1 -
https://www.nrc.gov/docs/ML0226/ML022630018.pdf) provides a definitive discussion on the Commissions consideration of applicability of Part 21 to RISC-1, 2, 3, and 4 SSCs under 10 CFR 50.69. The Commissions decision making on this matter is clear. Specifically, the text states, The Commission concludes that Part 21 reporting requirements extend only to RISC-1 SSCs because they are important in ensuring public health and safety.
RISC-2 SSCs are not subject to reporting because they play a lesser role than RISC-1 SSCs in protection of public health and safety and with the significant changes in treatment allowed under
§ 50.69, no regulatory purpose would be served by Part 21 reporting (as previously discussed). Individually, RISC-3 and RISC-4 SSCs have little or no risk significance and no regulatory purpose would be served by subjecting RISC-3 and RISC-4 SSCs to Part 21 and § 50.55(e).
- Consistent with the Commissions deliberations above, it is clear that the categorization process under 10 CFR 50.69 is designed to To align with the intent of 10 CFR 50.69 and avoid unnecessary regulatory burden, we request that the NRC remove the statement regarding additional OE expectations for RISC-3 SSCs, outside of the standard industry OE-sharing processes.
Page A-3 Section Text Comment Recommendation applicable, to support the intent of 10 CFR Part 21 requirements by alerting affected licensees of potential issues.
ensure that RISC-3 SSCs are of very low safety significance and that if categorization is performed correctly, identified non-conformances associated with RISC-3 SSCs should not meet the Part 21 threshold of a defect that could create a substantial safety hazard. As a result, RISC-3 SSCs are unlikely to generate Part 21 reportable events in the first place.
- Existing CAP and OE Mechanisms - 10 CFR 50.69 requires, Inspection and testing, and corrective action shall be provided for RISC-3 SSCs. Therefore, all of the standard aspects of the Corrective Action Program including, as applicable, cause determination, extent of condition investigation, tracking, trending, etc. will continue to be applied to RISC-3 SSCs. The Institute of Nuclear Power Operations (INPO) maintains the Industry Reporting and Information System (IRIS), which tracks OE using multiple criteria, including SSC make and model numbers.
Procurement engineers use this system to search for relevant OE as part of their standard procurement processes. This existing system effectively ensures that the necessary OE is shared among utilities regardless of the safety-related designation or RISC categorization of a component.
- It is important to note that a suppliers obligations under Part 21 to make notifications for any items procured and supplied under Appendix B are not impacted by 50.69. If a supplier has information concerning a [potential] Part 21 concern, they are required to make notification to any customers who procured that part under a safety related procurement process invoking Appendix B, so notifications/reports for other licensees who may have similar components installed in RISC-1 applications will continue to occur.
- No Need for a Parallel Reporting Process - The draft NUREG text appears to imply the need for a shadow reporting process outside of 10 CFR Part 21, which is neither required by regulation nor supported by the 50.69 framework. The 50.69 rule explicitly maintains Part 21 applicability for RISC-1 SSCs while removing unnecessary requirements for RISC-3 SSCs. Introducing an additional expectation for OE reporting undermines this risk-informed approach.
Page A-4 Section Text Comment Recommendation B-3 Lines 18-21 The treatment of RISC-3 SSCs must be consistent with the categorization process. One way to achieve this could be the application of consensus standards where the application of such standards meets the 10 CFR 50.69(d)(2) requirements for RISC-3 SSCs.
The section implies that the NRC has authority to determine the appropriateness of alternative treatments (ATs) for RISC-3 SSCs by stating that the use of consensus standards is an "appropriate" approach. This is inconsistent with 10 CFR 50.69, which explicitly places the responsibility for selecting ATs on the licensee. The NRC does not have the authority to define, approve, or dictate what constitutes an acceptable AT in advanceits role is limited to oversight after implementation.
Revise the text to remove any language suggesting that the NRC determines the appropriateness of specific ATs. Remove text stating that the NRC suggests or finds the use of consensus standards for RISC-3 SSCs appropriate.
B-4 Lines 43-50 and B-5 Lines 1-2 The regulations in 10 CFR 50.69(b)(2)(iv) remove RISC-3 SSCs from the scope of certain provisions of 10 CFR 50.55a. The provisions being removed include those that relate to inspection and testing in 10 CFR 50.55a(f) (10 CFR 50.55a(g) for snubbers).
The regulations in 10 CFR 50.69 allow licensees to use alternative treatment instead of the American Society of Mechanical Engineers (ASME)
Operation and Maintenance of Nuclear Power Plants (OM Code).
Nevertheless, the NRC staff considers the ASME Code Cases endorsed in 10 CFR 50.55a and listed in Regulatory Guide 1.192, The section incorrectly suggests that ASME Code Cases endorsed in 10 CFR 50.55a are an "acceptable" method for establishing alternative treatment (AT) for RISC-3 SSCs. This implies NRC approval of specific AT methods, which is inconsistent with 10 CFR 50.69. The regulation explicitly removes RISC-3 SSCs from the scope of 10 CFR 50.55a(f) and (g), allowing licenseesnot the NRCto determine appropriate ATs. The NRC's role is limited to oversight after implementation, not pre-approval of specific treatment methods.
Revise the text to remove any language suggesting that the NRC determines the appropriateness of specific ATs. Remove text stating that the NRC suggests or finds the use of endorsed or listed ASME Code Cases for RISC-3 SSCs acceptable.
Page A-5 Section Text Comment Recommendation Operation and Maintenance Code Case Acceptability, ASME OM Code, to be one acceptable method of establishing treatment of RISC-3 SSCs, where applicable, in that those endorsed Code 1 Cases adjust treatment based on the safety significance of the components.
B-5 Line 34-39 As stated at 69 FR 68042, under 10 CFR 50.69, most STRs will be removed from RISC-3 SSCs, which will typically comprise a large percentage of safety-related SSCs in a nuclear power plant. These STRs will be replaced with the high-level treatment requirements in 10 CFR 50.69(d)(2) that will allow significant reduction in the treatment applied to RISC-3 SSCs. This reduction in treatment can introduce common-cause concerns and weaken defenses against them.
The statement asserting that 10 CFR 50.69 leads to "significant reduction in treatment," introduces "common-cause concerns," and "weakens defenses" is speculative, unsupported by operating experience, and inappropriate for a NUREG document. There is no documented evidence showing that "most" special treatments (STRs) are removed from RISC-3 SSCs, nor that the application of ATs results in widespread degradation in performance or increased common-cause failures. 10 CFR 50.69(d)(2) establishes high-level treatment requirements to ensure RISC-3 SSCs perform their required functions, and provides the licensees with flexibility to determine appropriate ATs based on robust categorization and evaluation processes. In addition, the regulation requires that the performance of the SSCs be monitored to ensure that the categorization process remains valid.
Remove or revise this statement to align with the regulatory intent of 10 CFR 50.69, which is a risk-informed process that maintains necessary treatment based on safety significance and performance. Any discussion of potential impacts should be supported by data or operating experience, rather than speculation.