ML25057A384
| ML25057A384 | |
| Person / Time | |
|---|---|
| Site: | Nuclear Energy Institute |
| Issue date: | 02/26/2025 |
| From: | Facemire J Nuclear Energy Institute |
| To: | Office of Administration |
| References | |
| 90FR8782 00004, NRC-2024-0203, DG-1436 | |
| Download: ML25057A384 (1) | |
Text
PUBLIC SUBMISSION As of: 2/26/25, 1:00 PM Received: February 26, 2025 Status: Pending_Post Tracking No. m7m-3w6d-1zfv Comments Due: February 26, 2025 Submission Type: Web Docket: NRC-2024-0203 Acceptability of ASME Code,Section III, Division 5, "High Temperature Reactors" Comment On: NRC-2024-0203-0003 Draft Regulatory Guide: Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Document: NRC-2024-0203-DRAFT-0010 Comment on FR Doc # 2025-02036 Submitter Information Email:atb@nei.org Organization:Nuclear Energy Institute General Comment See attached file(s)
Attachments 02-26-25_NRC_Addl Comments on DG-1436 2/26/25, 1:02 PM NRC-2024-0203-DRAFT-0010.html file:///C:/Users/BHB1/Downloads/NRC-2024-0203-DRAFT-0010.html 1/1 SUNI Review Complete Template=ADM-013 E-RIDS=ADM-03 ADD: Ramon Gascot Lozada, Trish Walker Webb, Mary Neely Comment (4)
Publication Date: 2/3/2025 Citation: 90 FR 8782
Jon Facemire Senior Project Manager, New Nuclear Phone: 202.256.0190 Email: jwf@nei.org February 26, 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 Additional Comments on Draft Regulatory Guide DG-1436 [Docket ID: NRC-2024-0203]
Project Number: 689 Submitted via Regulations.gov
Dear Program Management,
Announcements, and Editing Staff:
The Nuclear Energy Institute (NEI)1, on behalf of its members, appreciates the opportunity to provide comments on Draft Regulatory Guide DG-1436, Acceptability of ASME Code,Section III, Division 5, High-Temperature Reactors. We commend the NRCs efforts in updating this guidance to support the safe and effective deployment of high-temperature reactor technologies.
During the re-opened comment period, NEI has conducted additional reviews of the proposed conditions outlined in DG-1436 and is providing Attachment 1 with several additional areas related to the CC-N-940 where further clarification or revision would enhance the regulatory framework. Also, we have provided two additional clarifying comments related to Table A-1 and Appendix HBB-Y in Attachment 2, based on the discussion in the public meeting to discuss the comments on this DG which was held on February 21, 2025. In particular, we would like to highlight the following key comments:
- 1. Comment 2 on CC-N-940 - Ultrasonic Examination Qualification: The proposed requirement for ultrasonic examination qualification using ASME Section V, Article 14, Intermediate Rigor is not consistent with existing requirements for this scope of components. The industrial and nuclear construction codes used to form the basis for this requirement have established general performance demonstration approaches that are appropriate and sufficient for this application. We recommend aligning the qualification requirements with endorsed codes and standards as additional specificity imposed through conditions is not necessary.
- 2. Comment 4 on CC-N-940 - Initial Sample Size for Progressive Sampling: The proposed approach to progressive sampling imposes a significantly higher initial sampling requirement than what is 1 The Nuclear Energy Institute (NEI) is responsible for establishing unified policy on behalf of is 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 February 26, 2025 Page 2 Nuclear Energy Institute specified in other industrial piping codes such as ASME B31.3 for similar applications. The current approach in N-940 is already consistent with industrial standards, and we believe additional conditions are unnecessary. We recommend maintaining alignment with established industry practices to ensure efficiency while maintaining safety.
- 3. Comment 6 on CC-N-940 - Initial Service Leak Testing: The condition requiring initial service leak tests to be implemented consistent with ASME BPVC B31.1 paragraph 137.7.1 unnecessarily limits the application of system leak testing as an alternative to hydrostatic testing. Code Case N-940 already limits the use of this alternative to moderate energy piping systems, consistent with ASME B31.3, Category D fluid systems. Additionally, Comment 5 states that the definition of moderate energy piping systems for advanced non-light water technologies is subject to NRC review. Therefore, additional conditions are not needed. We recommend removing this condition to ensure regulatory consistency with existing industrial codes.
Additional detailed comments addressing conditions on CC-N-940 in DG-1436 are included in Attachment 1 for the NRCs consideration. The proposed clarifications on Comments 2, 4 and 6 support the use of Section III Design Rules as a special treatment in instances where industrial construction codes may otherwise be justified.
As requested by the NRC during the public meeting, additional perspectives on Table A-1 wording and the basis for not endorsing Appendix HBB-Y are also provided in Attachment 2.
NEI appreciates the NRCs commitment to ensuring a robust regulatory framework that supports the safe deployment of advanced reactor technologies while maintaining alignment with industry best practices. We look forward to continued engagement on this important matter and welcome further discussions. If you have any questions or require additional information, please contact me at jwf@nei.org. Thank you for your time and consideration.
Sincerely, Jon Facemire Senior Project Manager, New Nuclear, Additional Comments on Related to CC-N-940 in Draft Regulatory Guide DG 1436, Acceptability of ASME Code,Section III, Division 5, High-Temperature Reactors, Additional Comments on Draft Regulatory Guide DG 1436, Acceptability of ASME Code,Section III, Division 5, High-Temperature Reactors c:
Ramon Gascot, NRC Matthew Hiser, NRC Meraj Rahimi, NRC Raj Iyengar, NRC Meg Audrain, NRC Joseph Bass, NRC Greg Oberson, NRC
- Additional Comments on Related to CC-N-940 in Draft Regulatory Guide DG-1436, Acceptability of ASME Code,Section III, Division 5, High-Temperature Reactors Page 1 No.
DG-1436 Condition DG-1436 Basis Comment 1-CC-N-940 The ultrasonic examination should be performed using encoded ultrasonic examination technology that produces an electronic record of the ultrasonic responses indexed to the probe position, permitting off-line analysis of images built from the combined data.
The use of encoded ultrasonic testing (UT) allows data analysis by additional qualified examiners and permits future examination results to be compared to original UT examination results.
This process also provides a permanent record of the examination data. This is consistent with the requirements in ASME B31.1 (Ref. 15) paragraph 136.4.6(a)(1) and Code Case N-831-1 (Ref. 16) paragraph (g).
Industry stakeholders understand this position and will discuss if this clarification should be included in a revision to CC-N-940.
2-CC-N-940 The ultrasonic examination should be qualified using ASME Section V, Article 14 Intermediate Rigor on test specimens that are of the same materials and similar size and thickness for the welds being examined.
This position provides qualification requirements to demonstrate the effectiveness of the UT examination systems, which is reasonable for safety-significant SSCs.Section V, Article 14 (Ref. 17) Intermediate Rigor requires limited performance demonstration. This is consistent with the requirement in ASME B31.1 paragraph 136.4.6(a)(2) that the procedures and equipment used to collect and analyze UT data need to be demonstrated.
ASME B31.1 and other codes, including Section III, provide general requirements for performance demonstration and do not explicitly invoke Section V, Article 14.
Therefore, this condition is not consistent with the requirements for this scope of components and goes beyond requirements in endorsed codes and standards for industrial and nuclear use.
3-CC-N-940 Progressive sampling under Nonmandatory Appendix B should not be applied to elevated temperature service (temperature >
750°F).
This position is consistent with industrial code ASME B31.1 Table 136.4.1-1, which requires 100% radiographic testing (RT) or UT for piping over NPS 2 that operates at temperatures over 750°F.
Industry stakeholders understand this position and will discuss if this clarification should be included in a revision to CC-N-940.
4-CC-N-940 For applications of progressive sampling under Nonmandatory Industrial codes, such as ASME B31.3 (Ref. 18), generally The premise of the Section III alternate requirements is
- Additional Comments on Related to CC-N-940 in Draft Regulatory Guide DG-1436, Acceptability of ASME Code,Section III, Division 5, High-Temperature Reactors Page 2 Appendix B, the initial sample size should be either (1) at least 50% of the population of welds or (2) a population justified statistically to provide 95% confidence that 5% or fewer of the welds contain defects.
Justification for (3) a lesser initial sample than one of the options in the preceding sentence may be provided subject to NRC review and approval on a case-by-case basis based on the risk significance of the SSC.
require 5% or more inspection sampling. Given that NSRST SSCs should provide increased assurance beyond normal industrial practices, the sampling size of 5% in N-940 does not provide reasonable confidence of performance.
Positions (1) and (2) in condition 4-CC-N-940 in Table 5, are intended to provide a baseline on inspection sampling to provide reasonable confidence of performance, while meaningfully lowering from the 100%
inspection required for SR SSCs under regular ASME BPVC Section III rules. (1) would more likely be limiting in cases where there are fewer welds to inspect, while (2) could be applied in cases with a larger population of welds that could statistically justify less than 50% sampling.
that the special treatment is applying the nuclear design rules with other construction requirements consistent with industrial codes. Therefore, the base technical requirement consistent with industrial requirement in ASME B31.3 for Normal Fluid Service systems (B31.3 paragraph 341.4.1(b)) is sufficient for this scope of components and additional conditions are not required.
5-CC-N-940 The definition of a moderate energy piping system for fluid systems with a service fluid other than water under -
7000(b)(2) of this Code Case is subject to NRC review and approval.
The justification for a definition of moderate energy piping for a technology should consider the potential impacts on other SSCs from failure of the piping system based on the operating conditions and characteristics of the fluid. Similar considerations for light-water reactors can be found in Branch Technical Position (BTP) 3-3, Revision 3 (ML070800027).
Given the uncertainty of the characteristics of fluid systems other than water, the condition in 5-CC-N-940 is needed to ensure the definition of moderate energy piping is appropriately considered for each technology.
Industry stakeholders understand this position and recognize that the definition of moderate energy piping systems will be specific to the advanced reactor design.
- Additional Comments on Related to CC-N-940 in Draft Regulatory Guide DG-1436, Acceptability of ASME Code,Section III, Division 5, High-Temperature Reactors Page 3 6-CC-N-940 Nonmandatory Appendix D should be implemented consistent with paragraph 137.7.1 from ASME BPVC B31.1:
137.7.1 When specified by the owner, an initial service test and examination is acceptable when other types of tests are not practical or when leak tightness is demonstrable due to the nature of the service. One example is piping where shutoff valves are not available for isolating a line and where temporary closures are impractical. Others may be systems where during the course of checking out of pumps, compressors, or other equipment, ample opportunity is afforded for examination for leakage prior to full scale operation.
In particular, an initial service leak test shall be used only when other types of tests are not practical or when leak tightness is demonstrable due to the nature of the service.
The condition in 6-CC-N-940 is consistent with industrial code ASME B31.1, which only allows initial service leak tests instead of pressure tests when other types of tests are not practical or when leak tightness is demonstrable due to the nature of the service.
The premise of the Section III alternate requirements is that the special treatment is applying the nuclear design rules with other construction requirements consistent with industrial codes. Therefore, the base technical requirement consistent with industrial requirement in ASME B31.3 for Category D systems (B31.3 paragraph 345.1(a)) is sufficient for this scope of components and additional conditions are not required.
- Additional Comments on Draft Regulatory Guide DG-1436, Acceptability of ASME Code,Section III, Division 5, High-Temperature Reactors Page 1 Item Comment/Basis Addition to Comment 2 of January 27, 2025, comment letter on DG-1436 DG-1436 states: The NRC staff did not review Nonmandatory Appendix HBB-Y and therefore is not endorsing it.
Advanced reactor developers have indicated that they plan to use Nonmandatory Appendix HBB-Y as part of material qualification. Endorsement of this appendix will streamline future license applications for advanced reactor developers.
As a result of the public meeting on February 21, 2025, it is understood that NRC did not review this appendix since it is a non-mandatory appendix and since it describes the documentation package that would be needed to qualify a new material under the code. The NRC stated that since code qualification of the new materials would ultimately be required, they did not see the need to review or endorse this non-mandatory appendix.
New reactor developers and vendors generally plan to seek addition of new materials to the code. However, they also point out that there is a significant timeline associated with development of the necessary documentation, qualification per the code, publication of the new code version with the new material, and subsequent endorsement of the code version by the NRC. This process can take many years before a new material is code qualified and the code is endorsed by the NRC. The new reactor developers and vendors would like to have a documented position stating that the documentation requirements in Appendix HBB-Y are acceptable to the NRC. Without early NRC feedback, there is a risk that industry will invest significant time and resources into developing documentation packages only to later find that NRC has additional requirements. A documented NRC position on Appendix HBB-Y would ensure alignment between industry documentation and NRC expectations, potentially reducing future NRC review efforts and enabling more timely deployment of advanced reactors.
We recognize that NRC does not typically review non-mandatory appendices. However, NRC currently endorses portions of non-mandatory Appendices HBB-T and HBB-Z in DG-1436, so there is precedent for NRC endorsement of a non-mandatory appendix. Because this appendix outlines documentation requirements rather than introducing new technical criteria, its endorsement would streamline the qualification process without setting a precedent for endorsing all non-mandatory appendices. Endorsing Appendix HBB-Y would not replace the need for full code qualification but would provide early confidence that the documentation package meets NRC expectations, thereby reducing regulatory uncertainty for new reactor developers. If endorsement of Appendix HBB-Y is not appropriate, NRC could consider documenting its reason for not reviewing the Appendix and clarifying its expectations for documentation packages used in material qualification. This would provide industry with the necessary confidence while maintaining NRCs current endorsement practices.
- Additional Comments on Draft Regulatory Guide DG-1436, Acceptability of ASME Code,Section III, Division 5, High-Temperature Reactors Page 2 Item Comment/Basis Addition to comment 3 of January 27, 2025, comment letter on DG-1436 Table A-1 discusses standards applicable to SR and NSRST SSCs. The information in this table appears to be inconsistent with the intent of NSRST as discussed in NEI 18-04 (and endorsed by NRC in RG 1.233). NSRST SSCs should be designed using commercial grade techniques, with special treatments applied as identified as necessary by an applicant. Table A-1 implies that NSRST SSCs should be designed using SR codes and standards, with reductions needing to be justified.
Suggested re-wording for the relevant text in several rows in column Quality Group C of Table A-1 is as follows:
ASME Code,Section III, Division 5 or Industrial Codes. with appropriate justification. 7 or 8