ML25188A048

Public Comment Table for Regulatory Guide (RG) 1.87, Revision 3, Acceptability of ASME Code Section III, Division 5, High Temperature Reactors

ML25188A048
Person / Time
Issue date:	07/25/2025
From:	Margaret Audrain, Bass J, Gascot-Lozada R NRC/NRR/DANU, NRC/RES/DE
To:
References
RG 1.87, Rev 3
Download: ML25188A048 (60)

Text

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 On December 10, 2024, the U.S. Nuclear Regulatory Commission (NRC) published a notice in the Federal Register (87 FR 69052) announcing that Draft Regulatory Guide (DG)-1436 (proposed Revision 3 of Regulatory Guide 1.87) was available for public comment. The public comment period closed on January 27, 2025. On February 3, 2025, a second window for public comment was open (89 FRN 100921), that period closed on February 26, 2025, the NRC staff received the following comments:

Anthony J. Schoedel Westinghouse Telephone: (412) 374-6118 Email: schoedaj@westinghouse.com ADAMS Accession No.: ML25028A204 Doug Kalinousky X-Energy Telephone: (301) 358-5600 Email: dkalinousky@x-energy.com ADAMS Accession No.: ML25028A205 Jon Facemire NEI Telephone: (202) 256-0190 Email: jwf@nei.org ADAMS Accession No.: ML25028A206 &

ML25057A384 Ian Gifford TerraPower Telephone: (425) 324-2888 Email: igifford@terrapower.com ADAMS Accession No: ML25028A209 Connor Nicol Telephone: Not provided Email: nico8021@vandals.uidaho.edu ADAMS Accession No: ML25036A006 Mark Alphonso-Waters Radiant Telephone: Not provided Email: mark.alphonsowaters@radiantnuclear.com ADAMS Accession No: ML25057A265 The comments below are repeated from the comment letters as stated, unless noted otherwise.

No. Commenter Comment NRC Resolution 1

Anthony J.

Schoedel Westinghouse DG-1436, page 20 states: "The NRC staff did not review Nonmandatory Appendix HBB-Y and therefore is not endorsing it." It is requested that NRC review and endorse this appendix as part of the final issuance of this regulatory guide, Response: The NRC staff understands there is significant industry interest in an NRC endorsement of Nonmandatory Appendix HBB-Y. The staff is aware that this Nonmandatory Appendix provides guidance

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 as endorsement of this appendix will support material qualification for future advanced reactor license applications for developing a data package for ASME committees to use to incorporate new materials into HB under ASME Boiler and Pressure Vessel Code (BPVC)

Section III, Division 5. It is not within the NRCs regulatory authority to endorse a procedure as acceptable for submitting materials to and meeting requirements of a third-party entity such as the ASME.

Although HBB-Y is not part of the staffs endorsement, it provides useful guidance to ASME Code committees and code users with respect to the adequacy of the data for elevated temperature nuclear applications. In addition, the NRC staff notes that it participates on ASME Code Committees in accordance with Management Directive 6.5 (ML18073A164).

Change in response to this comment:

The staff has revised the RG to establish a position relevant to HBB-Y as shown below:

The NRC staff is not endorsing Nonmandatory Appendix HBB-Y because it is for information only.

Additionally, it is not within the NRCs regulatory authority to endorse a procedure as acceptable for submitting materials to and meeting requirements of a third-party entity such as the ASME.

2 Anthony J.

Schoedel Westinghouse DG-1436, Table A-1 discusses standards applicable to safety-related (SR) and non-safety-related with special treatment (NSRST) structures, systems, and components (SSCs). The information in this table appears to be inconsistent with the intent of classifying SSCs as NSRST as discussed in NEI 18-04 (and endorsed by NRC in Regulatory Guide 1.233).

Response: The NRC staff agree with the commenters interpretation of NEI 18-04. It is not staffs intent to imply that SR codes and standards should be used for designing NSRST SSCs. Rather, the update to Table A-1 is only intended to clarify that the endorsement of ASME Section III, Division 5 extends to NSRST components, and therefore, if proposed by an

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 Based on NEI 18-04, NSRST SSCs should be designed using commercial grade techniques, with special treatments applied as identified as necessary by an applicant. DG-1436, Table A-1 implies that NSRST SSCs should be designed using SR codes and standards, with reductions needing to be justified. It is requested that this is clarified as part of the final issuance of this regulatory guide.

applicant, NRC would accept this code according to the same criteria that would be applied to its use for SR components.

Based on this and other comments, the staff has revised the footnotes to clarify what is expected in terms of a justification and to make clear that the endorsement of ASME Section III, Division 5 extends to NSRST components.

Change in response to comment: Footnotes 7 and 8 to Table A-1 are modified as shown below:

7 These standards may include ASME Code,Section III, Division 5, which is endorsed by NRC.

Codes that have not been endorsed by NRC, such as ASME Code,Section VIII, Division 1 and Division 2, or other alternate standards, may be used with appropriate justification. The applicant should justify how codes that have not been endorsed by the NRC as well as any special treatments are appropriate for the SSC.

8 These standards may include ASME Code,Section III, Division 5, which is endorsed by NRC.

Codes that have not been endorsed by NRC, such as, ASME B31.1/B31.3 or other alternate standards, may be used with appropriate justification. The applicant should justify how codes that have not been endorsed by the NRC as well as any special treatments are appropriate for the SSC.

3 Doug Kalinousky X-Energy Regarding Staff Regulatory Guidance, C.1.e. HAB-3220, Categories of the Owners Responsibilities Response: The NRC staff agree with this comment.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 In ASME Section III, Division 1 and Division 2 2023 Edition, NCA-3220 does not exist. Per ASME Section III Division 1 and Division 2 2021 Edition, NCE-3200 was revised in its entirety. The NCA-3220 content referenced in this section are from ASME Section Division 1 and Division 2 2019 Edition.

Therefore, X-energy requests clarification on the comparative sections in ASME Section III Division 1 and Division 2, 2023 Edition.

Change in response to comment: Based on a comparison with the 2023 edition of ASME Section III Subsection NCA, the NRC staff agrees with the comment and regulatory position C.1.e in DG-1436 will be deleted.

4 Doug Kalinousky X-Energy Regarding Staff Regulatory Guidance, C.1.f. HAB-3842.2, Evaluation of the Qualified Material Organizations Program by GC Certificate Holders In ASME Section III Division 1 2023 Edition, NCA-3842.2(h) and NCA-3842.2(i) are not applicable references since the NCA-3800, Metallic Material Organization's Quality System Program, was deleted in the 2023 version. X-energy would like clarification if the regulatory guide intends to reference NCA-3315.2(h) and NCA-3315.2(i) instead.

Similarly, in ASME Section III Division 1 2023 Edition, HAB-3859.1(a) through (e) do not exist. Does the regulatory guide intend to reference HAB-4559.1(a) through HAB-4559.1(e) instead?

HAB-3842.2 section g requires annual audits of the MO being evaluated, and users of this regulatory guide will be required to develop performance assessments. X-energy feels these requirements are overly burdensome and unnecessary for the GC certificate holder qualifying a material organization. X-energy requests these requirements be removed from the regulatory guide.

Response: The NRC staff agree with this comment.

Change in response to comment: Based on a comparison with the 2023 edition of ASME Section III Subsection NCA, the NRC staff agrees with the comment and regulatory position C.1.f in DG-1436 will be deleted.

5 Doug Kalinousky X-Energy Regarding Staff Regulatory Guidance, C.1.g. HAB-4000, Quality Assurance When comparing HAB-4133 and NCA-4133, the requirements are essentially the same. X-energy feels that applying NCA-4133 does not add value beyond HAB-4133.

Response: The NRC staff agree with this comment.

Change in response to comment: Based on a comparison with the 2023 edition of ASME Section III Subsection NCA, the NRC staff agrees with the

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 X-energy requests the staff remove the reference to NCA-4133 from the regulatory guidance because it is unnecessary.

comment and regulatory position C.1.g in DG-1436 will be deleted.

6 Doug Kalinousky X-Energy Regarding Staff Regulatory Guidance, C.1.h. HAB-4555.3, Approval and Control of Suppliers of Subcontracted Services When comparing HAB-4555.3(b) and NCA-4255.3(b), the requirements are essentially the same. When comparing HAB-4555.3(b) to NCA-4255.3(b) they are essentially the same with HAB-4555.3(b) referencing the GC Certificate Holder's Program rather the Certificate Holder's Program referenced NCA-4255.3(b). X-energy requests the staff remove the reference to NCA-4255.3(b) because it is unnecessary.

Response: The NRC staff agree with this comment.

Change in response to comment: Based on a comparison with the 2023 edition of ASME Section III Subsection NCA, the NRC staff agrees with the comment and regulatory position C.1.h in DG-1436 will be deleted.

7 Doug Kalinousky X-Energy Regarding Staff Regulatory Guidance, C.1.i. HAB-5000, Authorized Inspection When comparing HAB-5256 and NCA-5256, the requirements are essentially the same. X-energy feels that applying NCA-5256 does not add value beyond HAB-5256.

X-energy requests the staff remove the reference to NCA-5256 because it is unnecessary Response: The NRC staff agree with this comment.

Change in response to comment: Based on a comparison with the 2023 edition of ASME Section III Subsection NCA, the NRC staff agrees with the comment and regulatory position C.1.i in DG-1436 will be deleted.

8 Doug Kalinousky X-Energy Regarding Staff Regulatory Guidance, C.1.j. HAB-5125, Duties of Authorized Nuclear Inspector Supervisor (Graphite)

From the 2023 editions of Section III Division 1 and Section III Division 5, NCA-5125(h) and NCA5125(i) are essentially the same as HAB-5125(g) and HAB-5125(h). X-energy feels that applying NCA-5125(h) and NCA-5125(i) does not add value beyond HAB-5125(g) and HAB-5125(h). X-energy requests the staff remove the references to NCA-5125(h) and NCA-5125(i) because they are unnecessary.

Response: The NRC staff agree with this comment.

Change in response to comment: Based on a comparison with the 2023 edition of ASME Section III Subsection NCA, the NRC staff agrees with the comment and regulatory position C.1.j in DG-1436 will be deleted.

9 Doug Kalinousky X-Energy Regarding Staff Regulatory Guidance, C.1.m. HAB-8180, Renewal Response: The NRC staff agree with this comment.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 When comparing 2023 editions of Section III Division 1 and Section III Division 5, NCA-8182(a) and NCA-8182(b) and HAB-8182(a) and HAB-8182(b) are essentially the same. X-energy feels that applying NCA-8182(a) and NCA-8182(b) does not add value beyond HAB-8182(a) and HAB-8182(b).

X-energy requests the staff remove the references to NCA-8182(a) and NCA-8182(b) because they are unnecessary.

Change in response to comment: Based on a comparison with the 2023 edition of ASME Section III Subsection NCA, the NRC staff agrees with the comment and regulatory position C.1.m in DG-1436 will be deleted.

10 Doug Kalinousky X-Energy Regarding C.1.cc. Mandatory Appendix HHA-III-4200, Irradiated or Oxidized Graphite Regarding regulatory guidance on HHA-III-4200 (Page 21),

the samples are required to be tested at environmental condition that are consistent with the qualification envelope defined in the Design Specification. Limited extrapolations are allowed per code (HHA-II-4100), so X-energy requests the staff add a clarification to the guidance to explicitly mention that limited extrapolations are allowed.

Response: The NRC staff disagree with the need for clarification. The limitation in the RG on HHA-III-4200 does not limit extrapolation, with justification, because HHA-III-4200 does not discuss extrapolation.

The NRC has not placed any limitations or conditions on HHA-II-4100 and therefore does not prohibit limited extrapolations with appropriate justification.

The NRC will not address extrapolation in HHA-III-4200 in the RG but encourages ASME Code to address this topic. No changes will be made based on this comment.

Change in response to comment: None.

11 Doug Kalinousky X-Energy Regarding C.1.q. Mandatory Appendix, HBB-I-14 Tables and Figures Compared to Regulatory Guide 1.87 Revision 2, Revision 3 appears to have added a new requirement to position C.u.(1),

Mandatory Appendix HBB-I-14 Tables and Figures:

When using welded product forms listed in Table HBB-I-14.1(a), applicants and licensees should justify in the design report the use of the stress values in Tables HBB-I-14.2, HBB-I-14.3A, HBB-I-14.3B, HBB-I-14.3D, HBB-I-14.4A, HBB-I-14.4B, HBB-I-14.4D HBB-I-14.6A, HBB-I-14.6B, HBB-I-14.6D.

Response: The NRC staff disagree with this comment.

During the review of the 2023 code edition, it came to the attention of the NRC that it is non-conservative to assign time-dependent allowable stresses from Tables HBB-I-14.x as referenced in C.1.q in DG-1436 (C.1.i in RG 1.87 Rev. 3) that were developed from wrought products to welded products. Welded products include SA-249, SA-312, SA-358 and SA-403, Grade WP, Class W for Type 304 and 316 stainless steels, and SA-234 Grade WP22 welded fittings and SA-691, Grade 21/4 CR for 21/4Cr-1Mo steel. The creep strengths of welded products are lower than those for wrought products.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 This regulatory guidance requires justification of data used for stress limits for welds made of 304/316 SS and 21/4Cr-1Mo material that are in elevated temperature service. Regulatory Guide 1.87, Revision 2, did not appear to require users to provide additional justification of data. X-energy would like the staff to elaborate on the rationale for this addition.

Condition C.1.q in DG-1436 (C.1.i in RG 1.87 Rev. 3) was developed to provide designers the option to use these welded products, instead of disallowing their use, for HBB Class A construction if lower allowable stresses that can be justified by data from welded products are used in the HBB Class A design procedures.

The NRC encourages ASME to address this topic.

Change in response to comment: Welded product forms include SA-249, SA-312, SA-358 and SA-403, Grade WP, Class W for Type 304 and 316 stainless steels, and SA-234 Grade WP22 welded fittings and SA-691, Grade 21/4 CR for 21/4Cr-1Mo steel. shall be added to the basis for C.1.q in DG-1436 (C.1.i in RG 1.87 Rev. 3) to clarify which base materials are welded product forms.

12 Doug Kalinousky X-Energy Regarding Table A-1, Classification and Standards Applicable to Components in High Temperature Reactors Table A-1, Classification and Standards Applicable to Components in High Temperature Reactors, provides guidance for components based on classification. For non-safety related with special treatment (NSRST) components, the table instructs that users of the Regulatory Guide may apply ASME Code,Section III, Division 5 or Industrial Codes with appropriate justification. Furthermore, footnotes on this guidance instruct that:

7 These standards may include ASME Code,Section VIII, Division 1 and Division 2 with appropriate justification.

Applicants may propose alternate standards with appropriate justification.

Response: The NRC staff confirms X-energy's understanding of the updates to Table A-1. Based on this and other comments, the staff has revised the footnotes to clarify what is expected in terms of justification and to make clear that Section III, Division 5 may be used without further justification.

Change in response to comment: Footnotes 7 and 8 to Table A-1 are modified as shown below:

7 These standards may include ASME Section III, Division 5, which is endorsed by NRC. Codes that have not been endorsed by NRC, such as ASME Section VIII, Division 1 and Division 2, or other alternate standards, may be used with appropriate

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 8 These standards may include ASME B31.1/B31.3 with appropriate justification. Applicants may propose alternate standards with appropriate justification.

X-energys interpretation of the update to Table A-1 is that the regulatory guidance explicitly clarifies ASME III Div. 5 as an endorsed standard for designing NSRST components that does not require users to provide additional justification. The updated guidance is not intended to be more restrictive than the equivalent guidance in Regulatory Guide 1.87, Revision 2.

Please provide clarification.

justification. The applicant should justify how codes that have not been endorsed by the NRC as well as any special treatments are appropriate for the SSC.

8 These standards may include ASME Section III, Division 5, which is endorsed by NRC. Codes that have not been endorsed by NRC, such as ASME B31.1/B31.3 or other alternate standards, may be used with appropriate justification. The applicant should justify how codes that have not been endorsed by the NRC as well as any special treatments are appropriate for the SSC.

13 Jon Facemire NEI Regarding the following statement in DG-1436:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20).

This use would be subject to NRC review and approval in order to confirm that the risk significance of the NSRST SSC is sufficiently below that of safety-related SSCs.

NEI believes this regulatory position is inappropriate.

As described in Regulatory Guide 1.233 Guidance for a Technology-Inclusive, Risk-Informed, and Performance-Based Methodology to Inform the Licensing Basis and Content of Applications for Licenses, Certifications, and Approvals for Non-Light-Water Reactors:

The selection of licensing-basis events (LBEs); classification and special treatments of structures, systems, and components (SSCs); and assessment of defense in depth (DID) are fundamental to the safe design of non-LWRs. These activities Response: The NRC staff partially agree with this comment and understands the concern about how this language related to confirming that the risk significance of the NSRST SSC is sufficiently below that of safety-related SSCs may create confusion about the SSC classification methodology from RG 1.233 and NEI 18-04. The staff also appreciates the suggested wording from NEI.

Based on this comment and others, the staff will revise the position to incorporate the suggested wording from NEI to clarify the relationship with the approved SSC classification methodology under RG 1.233 and NEI 18-04.

Change in response to comment:

NRC staff revised the position from:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 also support identifying the appropriate scope and depth of information non-LWR designers and applicants should provide in applications for licenses, certifications, and approvals. This RG endorses Nuclear Energy Institute (NEI) 18-04, Revision 1, Risk-Informed Performance-Based Guidance for Non-Light Water Reactor Licensing Basis Development, (Ref. 3) as one acceptable method for non-LWR designers to use when carrying out these activities and preparing their applications. The methodology in NEI 18-04 provides a process by which the content of applications will permit understanding of the system designs and their relationship to safety evaluations for a variety of non-LWR designs.

For licensees who are using NEI-18-04, or other accepted methods, to identify licensing bases events (LBEs), classify SSCs, establish special treatments, identify programmatic controls, and assess DID for non-LWRs, these methods have already been accepted by the NRC and are not subject to an additional or separate review and approval as implied by the current wording in DG-1436. The concern raised in DG-1436 about risk-significance of NSRST SSCs is adequately addressed in the NEI 18-04 methodology with checks on risk-significance and appropriate consideration of uncertainties and cliff-edge effects.

Suggested wording is as follows:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20).

The technical justification for use of alternate requirements in these sections is subject to NRC approval. The NRC may review classification of NSRST SSCs in accordance with the approved methodology.

appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20). This use would be subject to NRC review and approval in order to confirm that the risk significance of the NSRST SSC is sufficiently below that of safety-related SSCs and is consistent with the reliability and capability targets specified for the NSRST SSC.

to:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20). The justification for use of alternate requirements in these sections as a special treatment to achieve the reliability and capability targets specified for the NSRST SSC is subject to NRC review and approval. The NRC may review classification of NSRST SSCs in accordance with the approved methodology in RG 1.233.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 14 Jon Facemire NEI 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.

Industry notes that this is a rather short appendix, and endorsement would be beneficial to future license applications.

It is requested that NRC review and endorse this appendix as part of this Regulatory Guide.

If the staff cannot review and endorse this appendix, then specific reasons for non-endorsement should be provided to this basis to determine path forward.

Addition to original comment 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 Response: The NRC staff understands there is significant industry interest in an NRC endorsement of Nonmandatory Appendix HBB-Y. The staff is aware that this Nonmandatory Appendix provides guidance for developing a data package for ASME committees to use to incorporate new materials into HB under ASME Boiler and Pressure Vessel Code (BPVC)

Section III, Division 5. It is not within the NRCs regulatory authority to endorse a procedure as acceptable for submitting materials to and meeting requirements of a third-party entity such as the ASME.

Although HBB-Y is not part of the staffs endorsement, it provides useful guidance to ASME Code committees and code users with respect to the adequacy of the data for elevated temperature nuclear applications. In addition, the NRC staff notes that it participates on ASME Code Committees in accordance with Management Directive 6.5 (ML18073A164).

Change in response to this comment:

The staff has revised the RG to establish a position relevant to HBB-Y as shown below:

The NRC staff is not endorsing Nonmandatory Appendix HBB-Y because it is for information only.

Additionally, it is not within the NRCs regulatory authority to endorse a procedure as acceptable for submitting materials to and meeting requirements of a third-party entity such as the ASME.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 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.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 15 Jon Facemire NEI 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.

Addition to original comment.

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 Response: The NRC staff agree with the commenters interpretation of NEI 18-04. It is not the staffs intent to imply that SR codes and standards should be used for designing NSRST SSCs. Rather, the update to Table A-1 is only intended to clarify that the endorsement of ASME Section III, Division 5 extends to NSRST components, and therefore, if proposed by an applicant, NRC would accept this code according to the same criteria that would be applied to its use for SR components.

Based on this and other comments, the staff has revised the footnotes to clarify what is expected in terms of a justification and to make clear that the endorsement of ASME Section III, Division 5 extends to NSRST components.

Change in response to comment: Footnotes 7 and 8 to Table A-1 are modified as shown below:

7 These standards may include ASME Section III, Division 5, which is endorsed by NRC. Codes that have not been endorsed by NRC, such as ASME Section VIII, Division 1 and Division 2, or other alternate standards, may be used with appropriate justification. The applicant should justify how codes that have not been endorsed by the NRC as well as any special treatments are appropriate for the SSC.

8 These standards may include ASME Section III, Division 5, which is endorsed by NRC. Codes that have not been endorsed by NRC, such as ASME B31.1/B31.3 or other alternate standards, may be used with appropriate justification. The applicant should

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 justify how codes that have not been endorsed by the NRC as well as any special treatments are appropriate for the SSC.

16 Jon Facemire NEI Appendix A, Page A-2, Footnote 1 contains duplicate text.

Delete duplicate text: include important to safety SSCs and also include SSCs that do not perform any safety function required by NRC regulations or credited in the safety analysis.

Response: The NRC staff agree that the identified text is duplicative.

Change in response to comment: The identified text is removed.

17 Jon Facemire NEI The Code Cases cited in scope of Reg Guide 1.87 Rev 3 do not agree with CC listing in 2023 Edition of CC (C&S Connect).

This revision also endorses, with exceptions and limitations, the Code CasesN-940. CC-N940 is not listed in 2023 edition of Nuclear Component Code Cases.

Also, CC N-898 listed was amended in 2022 as N-898-1.

Response to CC N-940 comment: The NRC staff agree that CC N-940 is not in the base edition of 2023 nuclear code case publication. The Code Case N-940 is published in "2023 BPVC Code Cases: Nuclear Components Supplement 7."

Response to CC N-898 comment: The NRC staff recognize that CC N-898 was updated. The only references to CC N-898 within DG-1436 are related to describing the endorsement in RG 1.87 Rev.2 which does endorse CC N-898. All references to the Code Case for the updated endorsement in the proposed RG 1.87 Rev. 3 refer to CC N-898-1.

Change in response to comment: No change.

18 Jon Facemire NEI The statement notes that ASME III Division 1 rules apply to time-independent material strength and deformation, with a maximum allowable temperature of 370 degrees Celsius (°C) (700 degrees Fahrenheit [°F]) for some materials and 425 °C (800 °F) for others. The NRC incorporates by reference portions of the ASME Code,Section III, Division 1, in 10 CFR 50.55a.

10 CFR 50.55a includesSection III condition: Subsection NH that allows temperatures to 900 °F for PWR pressurizer 316 Response: The NRC staff agree with this recommendation and has revised this paragraph from the background section to remove any potential conflict between the statement in the RG and the condition in 10 CFR 50.55a related to Subsection NH, which is not applicable to the 2015 Edition and later editions.

Change in response to comment:

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 SS heater sleeves. Consider modifying the background statement in Reg Guide 1.87 to address this condition.

The provisions in Subsection NH, Class 1 Components in Elevated Temperature Service, 1995 Addenda through all editions and addenda up to and including the 2013 Edition incorporated by reference in paragraph (a)(1) of this section, may only be used for the design and construction of Type 316 stainless steel pressurizer heater sheaths where service conditions do not cause the components to reach temperatures exceeding 900 °F. This condition is not applicable to the 2015 Edition and later editions.

The following paragraph is revised to add the bold text:

ASME Code,Section III, Division 1, Rules for Construction of Nuclear Power Plant Components (Ref. 8), contains the rules of construction of ASME Class 1, 2, 3, metal containment components and their supports, and core support structures. These rules generally apply to time-independent material strength and deformation, with a maximum allowable temperature of 370 degrees Celsius (°C) (700 degrees Fahrenheit [°F]) for some materials and 425 °C (800 °F) for others. The NRC incorporates by reference portions of the ASME Code,Section III, Division 1, in 10 CFR 50.55a.

19 Jon Facemire NEI This basis content is provided in Section C STAFF REGULATORY GUIDANCE a(1)

The term items commensurate with their contribution to safety or risk is vague and there is a need to clarify which SSCs are appropriate to use in these sections of code.

This content is provided in C. STAFF REGULATORY GUIDANCE 1. ASME Code,Section III, Division 5 (a) 1) not Regulatory Guidance Position b Response: The NRC staff agree that position a (1) corresponds to the Basis for Regulatory Guidance Position b.

Change in response to comment: Basis for Regulatory Guidance Position b shall be changed to Basis for Regulatory Guidance Position a (1).

20 Jon Facemire NEI

q. Mandatory Appendix HBB-I-14 Tables and Figures Should the welded product form specified in item (1) reference Table HBB-I-14.1(b) Permissible Weld Materials, and not Table HBB-I-14.1(a) Permissible Base Materials for Structures Other Than Bolting?

Response: The NRC staff disagree with this comment.

The NRC is putting a condition on welded product forms which are included in Table HBB-I-14.1 (a). The appropriate table is referenced.

Change in response to comment: No change.

21 Jon Facemire NEI Regulatory positions C.1.e, f, g, h, i, j, k, l and m Response: The NRC staff partially agree with this comment. Regulatory positions C.1.e, f, g, h, i, j, and m

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 In Staff Regulatory Guidance do not seem to consider that Division 5 Subpart HAB is applicable specifically to Graphite and Composite Core Components and Assemblies and is intended (with the exception of the glossary) to be applied independent of NCA (which is referenced in HAA for the metallic parts of the reactor including the vessel containing the Graphite and Composite Core Components and Assemblies.

Additional specifics are included in Comments 10 and 11.

in DG-1436 are addressed in the NRCs response to comments 3,4,5,6,7,8, and 9 above, respectively.

With respect to regulatory position C.1.k in DG-1436 (C.1.e in RG 1.87 Rev. 3), NCA-5230(d) requires the inspectors to verify that design calculations have been prepared for those components and supports not requiring Design Reports and it clarifies that the inspectors are not responsible for the accuracy of the calculations. The NRC staff did not take any exceptions to NCA-5230(d) in the incorporation by reference of the 2021 edition of ASME Section III Subsection NCA. Further, the comment seems to suggest that these two requirements would not be applicable for graphite and composite materials. No changes will be made to this regulatory position.

With respect to regulatory position C.1.l in DG-1436 (C.1.f in RG 1.87 Rev. 3), NCA-5290(c)(1) and (c)(2) require the N-3 Data Report to be reviewed and signed by the inspector only after (1) it has been certified by the Owner and (2) the inspector has reviewed and N-3 form and verified the data reports are appropriately referenced and are on file, etc. The NRC did not take any exceptions to NCA-5290(c)(1) and (c)(2) in the incorporation by reference of the 2021 edition of ASME Section III Subsection NCA. Further, the comment seems to suggest that these two requirements would not be applicable for graphite and composite materials. No changes will be made to this regulatory position.

Change in response to comment: No change.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 22 Jon Facemire NEI Regulatory positions C.1.e(1) - This would be an inappropriate reference. HAB-3220 specifically addresses the Owner's responsibilities regarding Graphite and Composite Core Components and Assemblies and includes proper Code references for such.

NCA-3220 does not exist in the 2023 Edition of Section III NCA. Responsibilities of the Owner. It has been replaced Table NCA-3200-1, which identifies the types of Certificate Holder or Designer and paragraph numbers. Where an X appears in the column under the type of Certificate Holder, the paragraph reference on the left side of the Table is required for the associated Certificate of Authorization.

HAA-1110(a) states, The rules of Subsection HA, Subpart A are contained in Divisions 1 and 2, Subsection NCA, except for those paragraphs or subparagraphs (with numbered headers) replaced by corresponding numbered HAA paragraphs or subparagraphs in this Subpart or new numbered HAA paragraphs or subparagraphs added to this Subpart.

Hence the requirements of NCA-3200 that were relocated to Table NCA-3200-1 apply for everything outside of the Graphite or Composite Components or Assemblies.

Compared with NCA-3220-2017, which appears in the Staff Regulatory Guidance C1(e), to be what is in the Regulatory Guidance, the only Owner responsibilities eliminated in HAB-3220 are those that have no applicability to Graphite or Composite Components or Assemblies, specifically overpressure protection and Division 2 requirements:

(m) designating the overpressure protection requirements for each component or system, including the Class of overpressure protection rules assigned to each component or Response: The NRC staff agree with this comment.

Change in response to comment: Based on a comparison with the 2023 edition of ASME Section III Subsection NCA, the NRC staff agrees with the comment and regulatory position C.1.e in DG-1436 will be deleted.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 system and the location of the overpressure protection devices 7

(n) providing and filing the Overpressure Protection Report (NCA-3270) required for the nuclear power system (o) reviewing and approving the Construction Specification, Design Drawings, and Construction Report for Division 2 construction (Table NCA-3200-1)

(q) designating Designers responsibilities with respect to construction surveillance for Division 2 construction (NCA-3252) 23 Jon Facemire NEI Regulatory positions C.1.f(1), g(1), h(1), i(1), j(1), k(1), l(1),

m(1) - Section III, Division 5 Subsection HA General Requirements is divided into two parts. Unlike Subpart A, which references Subsection NCA General Requirements for Metallic Materials, Subpart B is intended to be a stand-alone General Requirements provision specifically applicable to Graphite and Composite Core Components and Assemblies only. Numerous references cited apply only to the metallic portions of the reactor, which are covered in Subpart A. In addition, there are multiple references to provisions in NCA-3800, which was deleted from NCA in the 2023 Edition.

Subpart HAB has been reconciled, as appropriate, to the requirements of NCA in the 2021 Edition. Those items that vary are mostly stylistic editorial differences since the 2017 Edition of NCA or are requirements omitted because they do not apply to Graphite and Composite Core Components and Assemblies.

Response: The NRC staff agree with this comment.

Regulatory positions C.1.f (1), g (1), h (1), i (1), j (1),

and m (1) in DG-1436 are addressed in the NRCs response to comments 4, 5, 6, 7, 8, and 9 above, respectively.

Regulatory positions C.1.k(1) and m (1) in DG-1436 are addressed in the NRCs response to comment 21.

24 Jon Facemire NEI Regarding regulatory guidance on HHA-III-4200 (Page 21), the samples are required to be tested at environmental condition that are consistent with the qualification envelope defined in the Design Specification. Limited extrapolations are allowed per code (HHA-II-4100), so NEI members request Response: The NRC staff disagree with the need for clarification. The limitation in the RG on HHA-III-4200 does not limit extrapolation, with justification, because HHA-III-4200 does not discuss extrapolation.

In addition, limited extrapolations are allowed with appropriate justification as specified in HHA-II-4100.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 the staff add a clarification to the guidance to explicitly mention that limited extrapolations are allowed.

The NRC has not placed any limitations or conditions on HHA-II-4100 and therefore does not prohibit limited extrapolations with appropriate justification.

The NRC will not address extrapolation in HHA-III-4200 in the RG but encourages ASME Code to address this topic. No changes will be made based on this comment.

Change in response to comment: No change.

25 Jon Facemire NEI Compared to Regulatory Guide 1.87 Revision 2, Revision 3 appears to have added a new requirement to position C.u.(1),

Mandatory Appendix HBB-I-14 Tables and Figures: When using welded product forms listed in Table HBB-I-14.1(a),

applicants and licensees should justify in the design report the use of the stress values in Tables HBB-I-14.2, HBB-I-14.3A, HBB-I-14.3B, HBB-I-14.3D, HBB-I-14.4A, HBB-I-14.4B, HBB-I-14.4D HBB-I-14.6A, HBB-I14.6B, HBB-I-14.6D.

This regulatory guidance requires justification of data used for stress limits for welds made of 304/316 SS and 21/4Cr-1Mo material that are in elevated temperature service. Regulatory Guide 1.87, Revision 2, did not appear to require users to provide additional justification of data.

NEI members would like the staff to elaborate on the rationale for this addition or remove the additional requirement.

• Staff consider this comment to be on position C.1.q.

in DG-1436 instead of C.u.(1) as stated in the comment.

Response: The NRC staff understand the request of clarification. During the review of the 2023 code edition, it came to the attention of the NRC that it is non-conservative to assign time-dependent allowable stresses from Tables HBB-I-14.x that were developed from wrought products to welded products. They include SA-249, SA-312, SA-358 and SA-403, Grade WP, Class W for Type 304 and 316 stainless steels, and SA-234 Grade WP22 welded fittings and SA-691, Grade 21/4 CR for 21/4Cr-1Mo steel. The creep strengths of welded products are lower than those for wrought products.

Condition C.1.q. in DG-1436 (C.1.i in RG 1.87 Rev.

3) was developed to provide designers the option to use these welded products, instead of disallowing their use, for HBB Class A construction if lower allowable stresses that can be justified by data from welded products are used in the HBB Class A design procedures.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 The NRC encourages ASME to address this topic.

There is no change to C.1.q in DG-1436 (C.1.i in RG 1.87 Rev. 3).

Change in response to comment: No change.

26 Ian Gifford TerraPower General Comment The body of Regulatory Guide (RG) 1.87 describes an approach that is acceptable to NRC staff to assure the mechanical/structural integrity of components that operate in elevated temperature environments and that are subject to time-dependent material properties and failure modes.

Appendix A, "High Temperature Reactor Quality Group Classification," provides guidance for quality group classification. Quality group classification is a separate topic that should not be paired with the endorsement of ASME BPVC Section III, Division 5. Furthermore, Appendix A appears to apply deterministic quality group classification guidance to the risk-informed, performance-based (RIPB) RG 1.233 process. This is not compatible with the RIPB implementation of RG 1.233, which selects special treatments based on the required functional performance of SSCs.

Recommendations in order of preference:

1. Remove Appendix A from RG 1.87.

2. Remove the deterministic quality group classification guidance from Appendix A.

3. Move Appendix A to an independent guidance document.

Response: The NRC staff disagree with this comment.

The purpose of this guidance update was primarily to update the staffs endorsement of ASME Section III, Division 5 from the 2017 to 2023 Code edition. The staff understands TerraPowers perspective on the relationship between the body of RG 1.87 and Appendix A. However, there is value and clarity in Appendix A, which provides guidance on how to identify quality standards based on SSC classification across the three main classification approaches. Based on the purpose of this guidance update and the importance of the guidance in Appendix A, the staff will not incorporate any of the recommendations in this comment.

However, in response to this and other comments regarding the LMP approach under RG 1.233 focusing on the functional performance of SSCs, the staff has added a paragraph to the discussion of the LMP Approach (RG 1.233) in Section A-2 of Appendix A to clarify this position.

Change in response to comment: The following text is added to the discussion of the LMP Approach (RG 1.233) in Section A-2 of Appendix A:

The LMP approach under RG 1.233 focuses on the functional performance of SSCs, which may in

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 some cases lead to specific SSCs possessing both SR and NSRST functions (e.g. SR for one function and NSRST for a different function). In these cases, the overall classification for an SSC is determined by its function with the highest safety classification. Accordingly, the staff guidance in Table A-1 is based on the overall classification for an SSC. However, an SSC with functions of varying safety classification may be able to justify a different code and standard than the application of Table A-1 to the overall classification based on the specific details of the plant design and functions of that SSC within the plant.

27 Ian Gifford TerraPower Related Guidance First Bullet Page 2 This paragraph notes that NUREG-2245 exceptions and limitations on the 2017 Edition remain in this revision.

Reliance on consensus standards without additional regulatory restrictions should be expected, particularly since the NRC staff is actively participating and votes on Code changes.

Recommendation:

Remove exceptions and restrictions based on a technical review of a prior Code Edition (2017).

Response: The NRC staff disagree with this comment.

The staff has reviewed the changes since the 2017 edition of Section III, Division 5 and removed limitations which have been addressed by the updates in the 2023 edition of the Code. NUREG-2245 is referenced as the technical basis for the limitations which have remained unchanged from 2017 edition endorsement.

The NRC staff are active participants in ASME Code, vote on many actions, and frequently provide comments from its perspective as a safety regulator.

However, the NRC is one member in the consensus committees and language changes and additions are often made without total agreement in the committees.

In those cases, the staff has a regular practice of clearly identifying its concerns during Code meeting discussions and in its votes.

Change in response to this comment: None.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 28 Ian Gifford TerraPower Related Guidance Second Bullet Page 3 RG 1.26 provides deterministic-related guidance for light water reactors (LWRs) regarding quality groups. References to LWR technology guidance should be minimized as they are not technology inclusive and are often based on deterministic principles and not on risk-informed safety-significant functions. The quality group classification philosophy of RG 1.26 is not compatible with most advanced reactor technologies which operate at near atmospheric conditions, therefore lacking the pressure retaining function typical for LWRs.

Recommendation:

Remove references to LWR technology guidance.

Specifically, RG 1.26.

Response: The NRC staff disagree with this comment.

RG 1.26 is included here based on its discussion in Appendix A under the traditional approach to SSC classification. Please see the response to comment #49 below from TerraPower on the discussion of RG 1.26 in Appendix A.

Change in response to this comment: None.

29 Ian Gifford TerraPower BRGP q(1) Page 5 and corresponding C.1.q(1) and C.1.q(2) Page 15 Basis for Regulatory Guidance Position (BRGP) q (1): In this paragraph the staff requires additional justification for using stress values in HBB-I-14.2, I-14.3, I-14.4, I-14.6 on the basis that these values were established for non-welded products.

Further in this paragraph, the staff stipulates that other non-nuclear codes use reduction factors for stress developed from non-welded products. ASME.BPVC.III.5 2023 Edition HBB-3221(a)(2), "Weldments," uses a reduction factor of 0.8 on the stress allowable limits. The NRC's basis statement does not appear to take this reduction factor into account and effectively adds additional margin. Furthermore, recent code actions including testing and benchmarking of code rules on time-dependent material properties and creep by Argonne National Laboratory indicated excessive margin greater than a factor of ten in existing code rules. Considering that most Response: The NRC staff disagree with this comment.

The staff replies assuming the commenter is referring to HBB-3221(b)(2), Weldments.

The term weldment used in this paragraph, and throughout HBB, refers to fusion welds that join similar parent materials in wrought product forms. In contrast, HBB also discusses welded product forms, such as welded tube or welded pipe from Table HBB-I-14.1(a). HBB treats welded products the same as wrought products, which essentially disregards the presence of the weld in the welded product and does not provide any guidance or appropriate allowable stresses for a welded product.

Paragraph HBB-3221(b)(2) provides the criteria in determining the allowable stresses for weldments. The allowable stresses from HBB for weldments are only

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 advanced reactors operate at or near atmospheric pressures, the additional justification required does not appear justified.

1. The new limitation added in C.1.q(1) is not supported by the basis, as the Code uses weld reduction factors.

2. The existing limitation in C.1.q(2) is not provided with a basis to justify why the limitation is necessary.

Unless there is a clear and justified basis the staff should consider removing this limitation.

Recommendation:

1. The new limitation imposed by C.1.q(1) should be removed. 2. Remove the existing limitation in C.1.q(2) unless sufficient justification can be made, and provided, for its inclusion.

applicable to the list of permissible weld metals in HBB for fusion welding of wrought product forms and they are not intended for use with the weld in a welded product form.

During the review of the 2023 code edition, it came to the attention of the NRC that it is non-conservative to assign time-dependent allowable stresses from Tables HBB-I-14.x as referenced in C.1.q in DG-1436 (C.1.i in RG 1.87 Rev. 3) that were developed from wrought products to welded products because the creep strengths of welded products are lower than those for wrought products. Therefore, there is no change to C.1.q(1) in DG-1436 (C.1.i(1) in RG 1.87 Rev. 3).

The welded product forms include:

SA-249, SA-312, SA-358 and SA-403, Grade WP, Class W for Type 304 and 316 stainless steels, SA-234, Grade WP22 welded fittings and SA-691, Grade 21/4 CR for 21/4Cr-1Mo steel.

The NRC encourages ASME to address this topic.

The limitation C.1.q(2) in DG-1436 (C.1.i(2) in RG 1.87 Rev. 3) was carried over from RG 1.87 Revision 2 and the technical justification was provided therein. It is noted that ASME has revised the values of Type 304 and 316 stainless steels in C&S Record 24-1574.Therefore, there is no change to C.1.q(2) in DG-1436 (C.1.i(2) in RG 1.87 Rev. 3).

Change in response to this comment: None.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 30 Ian Gifford TerraPower BRGP v Page 5 and corresponding C.1.v Page 20 BRGP v is used to add code limitation C.1.v without a proper basis. The basis provided for BRGP v is insufficient and does not support the restriction on code paragraph HBB-Z-1212.3.

As HBB-Z-1212.3 explains "...The rationale is that the procedure for creep-fatigue design by inelastic analysis, per Nonmandatory Appendix HBB-T, explicitly and separately accounts for the accumulation of damage in the material using the Code rules. As such, representing damage development in the constitutive model double counts damage and may lead to an overly conservative design..." Imposing this limitation results in undue burden on advanced reactor applications Recommendation:

Remove code limitation C.1.v. The limitation double-counts the conservatism.

Response: The NRC staff disagree with the recommendation. There is no double counting within the framework of the HBB creep-fatigue assessment procedure.

The role of a constitutive model in the inelastic analysis approach of HBB for creep-fatigue assessment is to provide the behavioral trends of the stress and strain histories due to deformation. The stress relaxation history during hold is used to calculate the creep damage using creep rupture data and the strain range is used to calculate the fatigue damage using fatigue failure data.

Subparagraph HBB-Z-1212.3 conflated the notion of deformation as determined from constitutive model with the code procedure in the creep damage and fatigue damage evaluation that leverage failure data.

Change in response to this comment: None.

31 Ian Gifford TerraPower BRGP w Page 5 and corresponding C.1.w Page 20 BRGP w is used to add code limitation C.1.w without consideration of a complete Code basis. HBB-T-1300 limits inelastic strains to 1%, 2%, and 5% for average, bending, and local strains respectively; also referenced in Figure HBB-3221-1 for strain and deformation limits. The additional basis provided in Position w stipulates that non-physical behavior may be exhibited by the constitutive model for stainless steel (SS) 316 in HBB-Z due to the internal variable omega. Given the limitations on strain limits in HBB-T-1300, the proposed additional limitation on the internal variable, omega, is only of interest from a theoretical perspective but in practice the limits Response: The NRC staff disagree with the recommendation.

The role of an inelastic material model when used in the inelastic analysis approach for design is to capture the key deformation responses that are deemed important in addressing the structural failure modes covered by HBB. The effect of elastic follow-up (HBB-3138) on the local stress redistribution near stress raisers is an important attribute for creep-fatigue assessment.

The inelastic material model in subparagraph HBB-Z-1322 for Type 316 stainless steel has the deficiency

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 of HBB-T-1300 make it redundant and unnecessary.

Extracting the one limit fails to consider the code as a whole.

Recommendation:

Remove the HBB-T-1300 limits.

that parameter continuously increases with time and the formulation leads to an effective elastic modulus of 1, with 0 < 1, where is the Youngs modulus as tabulated in the code. This results in the continuous reduction of the effective elastic stiffness.

Locally, this affects the slopes of the loading and unloading curves in the stress-strain space under cycling. Apart from not being representative of the elastic behavior of metals, it would not allow the elastic follow-up effect on stress redistribution to be captured adequately when the artificial stiffness reduction in the structure globally is significant.

The condition in C.1.w in DG-1436 (C.1.o in RG 1.87 Rev. 3) provides some allowance on the effective elastic stiffness reduction in capturing the elastic follow-up effect.

The NRC encourages the ASME to address this topic.

The HBB design rules were developed to guard against different structural failure modes with reasonable assurance. The strain limits check guards against excessive sectional (or through wall) deformation/ratcheting response and the creep-fatigue check guards against the local, or pointwise, creep-fatigue failure at stress raisers caused by geometric or material discontinuities. Per code intent, both checks are necessary and there is no double counting.

The proposed removal of the HBB-T-1300 limits has no relevance in the context of the condition C.1.w in

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 DG-1436 (C.1.o in RG 1.87 Rev. 3) imposed on HBB-Z-1322.

Change in response to this comment: None.

32 Ian Gifford TerraPower BRGP x Page 6 and corresponding C.1.x Page 20 BRGP x is used to add code limitation C.1.x without consideration of complete Code basis. Similar to the C.1.w comment above, HBB-T-1300 strain limits may already provide adequate and sufficient limitation such that the constitutive equations in HBB-Z-1325 may be reasonably acceptable and well within the accuracy required for practical engineering. Given the inherent margin, as demonstrated by the Argonne National Laboratory benchmark tests (see comment on Background / page 5, and corresponding C.1.q(1) and C.1.q(2)), the staff should consider whether all "edge" cases of theoretical arguments would need to be limited which creates an impractical domain for practicing engineers.

Recommendation:

Reconsider limitations on all "edge" cases.

Response: The NRC staff disagrees with the recommendation.

An inelastic model that gives contradictory predictions is not acceptable for use per the basis in the RG.

The NRC encourages ASME to address this topic.

Change in response to this comment: None.

33 Ian Gifford TerraPower General Dozens of additional limitations (C.1.z through C.1.rr) on nonmetallic Class SN core components are added. This increases the burden on license applicants that utilize nonmetallic components.

Recommendation:

Reconsider the value of adding limitations on nonmetallic Class SN core components.

Response: The NRC staff disagree with this recommendation. The NRC staff places conditions and limitations which are deemed necessary to ensure structural integrity.

Change in response to this comment: None.

34 Ian Gifford TerraPower BRGP 1-CC-N-898-1 BRGP 2-CC-N-898-1 Pages 8 & 9 and Table 5. N-898-1 Page 24 Response: The NRC staff disagree with the recommendation.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 BRGPs 1-and 2-CC-N-898-1 are used to add code limitations in Table 5 without potential consideration of complete Code basis. The same basis as used in "Background Position w /

page 5, and C.1.w / page 20", and "Background Position x /

page 6, and C.1.x / page 20" are used here. The comment on this limitation is the same as noted for those two positions above Recommendation:

Remove the HBB-T-1300 limits.

CC-N-898-1 Position 1 There is no double counting within the framework of the HBB creep-fatigue assessment procedure.

The role of a constitutive model in the inelastic analysis approach of HBB for creep-fatigue assessment is to provide the behavioral trends of the stress and strain histories due to deformation. The stress relaxation history during hold is used to calculate the creep damage using creep rupture data and the strain range is used to calculate the fatigue damage using fatigue failure data.

Subparagraph HBB-Z-1212.3 conflated the notion of deformation as determined from constitutive model with the code procedure in the creep damage and fatigue damage evaluation that leverage failure data.

There is no change to CC-N-898-1 Position 1.

CC-N-898-1 Position 2 As stated in the basis for this position, there are multiple deficiencies in the inelastic model for Alloy 617. Due to the complexity of the Alloy 617 model, the NRC was not able to find reasonable conditions to be imposed on the model.

The NRC encourages ASME to address this topic.

The NRC disagrees with the proposed removal of the HBB-T-1300 limits. It has no relevance in the context of the positions in CC-N-898-1.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 There is no change to CC-N-898-1 Position 2.

Change in response to this comment: None.

35 Ian Gifford TerraPower BRGP 1-CC-N-940 Page 9 and Table 5. 1-CC-N-940 Page 24 The condition in 1-CC-N-940 imposes an additional restriction on ultrasonic testing (UT) technology without consideration of the entirety of the construction Code.

The position references ASME B31.1 for additional requirements without consideration of other measures provided by the construction code. In ASME-III there are many requirements that are above and beyond B31.1, such as Certified Material Test Reports, design and fabrication rules for time independent properties, and restrictions on fabrication and welding methods/configurations. In addition, CC-N-940 invokesSection V article 4 in its entirety for UT. Therefore, if Section V is acceptable to the regulator, then no additional requirements should be applied on this code case with respect to allowable UT technologies. It is not appropriate to select the most restrictive requirements and apply them without consideration to safety-significance and the overall framework of the construction code, which includes design, materials, fabrication, testing, examination, and installation.

Recommendation:

Remove the additional restrictions imposed by this condition.

Response: The NRC staff disagree with this recommendation. The staff has considered the entirety of the construction Code in developing this position as well as considering the range of NSRST SSCs to which it could be applied under this endorsement.

Additionally, recent code changes in HBA 2610 and HBB 2610 have removed many of the material related requirements for items commensurate with their contribution to safety or risk.

Consistent with the stated objective of CC N-940 to apply Section III for design with industrial codes for construction, the staff position is that incorporating this provision from B31.1 is appropriate since it provides a record of examination data, and it allows future examination results to be compared to original UT examination results.

Change in response to comment:

Language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5.

Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 performance monitoring approaches, such as surveillance methods, in-service inspection, and continuous monitoring. Continuous monitoring of an SSCs ability to perform its function, such as online monitoring for cracking, pressure, temperature, or chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

36 Ian Gifford TerraPower BRGP 3-CC-N-940 Page 10 and Table 5. 3-CC-N-940 Page 24 The condition in 3-CC-N-940 imposes an additional restriction on progressive sampling without consideration of the entirety of the construction Code.

Similar comment to "Background Position 1-CC-N-940 / page 9, and Table 5 1-CC-N-940 / page 24," when looking at appropriate exams, the entirety of the construction code should be considered. Selecting from B31.1 requirements that are more restrictive disregards the additional measures that exist in ASME-III. The limitation of progressive exams for piping above 750 F should not be relevant for ASME-III due to the additional construction requirements imposed by Division 5, which thoroughly address the time independent effects. Furthermore, Division 5 requires Certified Material Test Reports, and restricts weld configurations. The combination of enhanced construction, design rigor, and a comprehensive examination, supports the safety significance of moderate energy piping, particularly in advanced designs that operate near atmospheric pressures.

Recommendation:

Remove the additional restrictions imposed by this condition Response: The NRC staff disagree with this recommendation. The staff has considered the entirety of the construction Code in developing this position as well as considering the range of NSRST SSCs to which it could be applied under this endorsement.

Consistent with the stated objective of CC N-940 to apply Section III for design with industrial codes for construction, the staff position is that incorporating this provision from B31.1 is appropriate to provide reasonable confidence of NSRST SSC performance on a generic basis.

Change in response to comment:

Language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5:

Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of performance monitoring

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 approaches, such as surveillance methods, in-service inspection, and continuous monitoring.

Continuous monitoring of an SSCs ability to perform its function, such as online monitoring for cracking, pressure, temperature, or chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

37 Ian Gifford TerraPower BRGP 4-CC-N-940 Page 10 and Table 5. 4-CC-N-940 Page 24 The condition in 4-CC-N-940 imposes an additional restriction on progressive sampling without consideration of the entirety of the construction Code and the reduced safety significance of moderate energy piping.

This comment is similar to the other CC-N-940 restrictions.

When looking at appropriate exams, the entirety of the construction code should be considered. Selecting from B31.1 requirements that are more restrictive disregards the additional measures that exist in Section-III. See comments above.

Recommendation:

Remove the additional restrictions imposed by this condition Response: The NRC staff disagree with this recommendation. The staff has considered the entirety of the construction Code in developing this position as well as considering the range of NSRST SSCs to which it could be applied under this endorsement.

Based on the scope of this endorsement being for safety-significant NSRST SSCs that should provide increased assurance beyond normal industrial practices, the staff believe it is necessary to approve the user's sampling plans to assess the adequacy of special treatments. The text and basis for 4-CC-N-940 have been revised to clarify the purpose of the condition and positions (1), (2), and (3) within the condition.

Change in response to comment:

Condition 4-CC-N-940 has been revised to state the following:

4-CC-N-940. For applications of progressive sampling under Nonmandatory Appendix B, the initial sample size should be one of the following:

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 (1) a population justified statistically to provide 95% confidence that 5% or fewer of the welds contain defects, or (2) a lesser initial sample justified as an alternative approach as described in footnote 6 below, or (3) for instances where a designer does not prefer to use the statistical justification in (1) or develop an alternative approach in (2), 50%

random sampling is acceptable.

The basis for 4-CC-N-940 has been revised to state the following:

The initial sample requirement in commercial codes (e.g., B31.1 and B31.3) for SSCs that are most analogous to those for nuclear service within the scope of this Code Case generally varies between 5% and 100%, depending on their design of the component and service condition. Given this variance and without specific information about the SSC to which the Code Case may be applied, staff cannot conclude that a 5% sample is generically appropriate. Position (1) in condition 4-CC-N-940 in Table 5 is 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 and under commercial codes in certain situations.

Position (2) provides explicit clarity that a CC user may propose what it believes to be an appropriate

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 sample percentage based on the specific design and service condition of the SSC. Meanwhile, position (3) provides a simpler approach that does not require additional analysis or justification and is more likely to be applied in cases where there are fewer welds to inspect. Consistent with a graded approach, NRC staff finds 50% random sampling to be an acceptable approach based on the generally lower safety significance of NSRST SSCs relative to SR SSCs.

In addition, language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5.

Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of performance monitoring approaches, such as surveillance methods, in-service inspection, and continuous monitoring. Continuous monitoring of an SSCs ability to perform its function, such as online monitoring for cracking, pressure, temperature, or chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

38 Ian Gifford TerraPower BRGP 6-CC-N-940 Page 10 and Table 5. 6-CC-N-940 Page 24 Response: The NRC staff disagree with this recommendation. The staff has considered the entirety of the construction Code in developing this position as

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 The condition in 4-CC-N-940 imposes an additional restriction on leak testing without consideration of the entirety of the construction Code and the reduced safety significance of moderate energy piping.

This comment is similar to the other CC-N-940 restrictions.

When looking at appropriate leak tests, the entirety of the construction code should be considered. Selecting from B31.1 requirements that are more restrictive disregards the additional measures that exist in Section-III. See comments above.

Recommendation:

Remove the additional restrictions imposed by this condition.

well as considering the range of NSRST SSCs to which it could be applied under this endorsement.

Regarding leak testing, both B31.1 and B31.3 require hydrostatic testing. B31.3 allows the owner to specify initial service leak testing as an alternative to hydrostatic testing only for Category D fluid. Other fluid categories including normal process fluid still require hydrostatic testing.

Consistent with the stated objective of CC N-940 to apply Section III for design with industrial codes for construction, the staff position is that incorporating this provision from B31.1 is appropriate to provide reasonable confidence of NSRST SSC performance on a generic basis.

Change in response to comment:

Language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5:

Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of performance monitoring approaches, such as surveillance methods, in-service inspection, and continuous monitoring. Continuous monitoring of an SSCs ability to perform its function, such as online monitoring for cracking, pressure, temperature, or

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

39 Ian Gifford TerraPower C.1.a(1) Page 12 Position C.1a(1), as stated, is not appropriate. This position adds an interpretation that NSRST SSCs should consider the most restrictive application of ASME BPVC Section III. This position applies safety-related (SR) equivalent application of codes and standards to NSRST SSCs, which is not consistent with the RG 1.233 risk-informed, performance-based philosophy. The most restrictive application of ASME BPVC Section III may not be warranted as a special treatment application to non-safety-related SSCs. The addition of this paragraph creates an undue burden on license applications to prove a "negative," (i.e., why ASME BPVC Section III does not apply). This is not consistent with risk-informed design.

Under the application of RG 1.233, the risk significance of an NSRST SSC is not required to be below the risk significance of SR SSCs. RG 1.233 does not require SR equivalent treatment for NSRST SSCs whose risk significance exceeds that of SR SSCs. The special treatments for NSRST SSCs are selected to provide reasonable confidence that SSCs will perform their functions reflected in the licensing basis events, which may not require the application of SR equivalent treatment.

Recommendation:

Position C.1.a(1) should be revised to remove the confirmation that the risk significance of the NSRST SSC is sufficiently below that of safety-related SSCs.

Response: The NRC staff partially agree with this recommendation. The staff does not agree that this position indicates that NSRST SSCs must consider ASME BPVC Section III or apply SR codes and standards to NSRST SSCs. The purpose of this position is to clarify what scope of SSCs the term items commensurate with their contribution to safety or risk may be applied to. NSRST SSCs may have a range of risk and safety significance depending on their function within a specific design. This position is intended to provide clarity for what scope of SSCs this term may be applied to, while ensuring the staff has the ability to confirm this is appropriate on an application-specific basis.

Based on this comment and others, the staff will revise the position to remove the statement related to confirming that the risk significance of the NSRST SSC is sufficiently below that of safety-related SSCs.

Change in response to comment: NRC staff revised the position from:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20). This use would be subject to NRC review and approval in order to confirm that the risk significance of the NSRST SSC is sufficiently below that of safety-related SSCs and is consistent

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 with the reliability and capability targets specified for the NSRST SSC.

to:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20). The justification for use of alternate requirements in these sections as a special treatment to achieve the reliability and capability targets specified for the NSRST SSC is subject to NRC review and approval. The NRC may review classification of NSRST SSCs in accordance with the approved methodology in RG 1.233.

40 Ian Gifford TerraPower C.1.b Page 12 The paragraph restricts qualified chartered engineers from fulfilling ASME certifying engineer roles. Chartered engineers in ASME-III App. XXIII are internationally recognized professionals who are subject to the same qualification requirements as United States registered professional engineers. This restriction disregards international programs and regulations.

Recommendation:

Remove this restriction Response: The NRC staff disagree with this recommendation. The following condition is already in 50.55a(b)(1)(xii):

(xii)Section III condition: Certifying Engineer.

When applying the 2017 and later editions of ASME BPV Code Section III, the NRC does not permit applicants and licensees to use a Certifying Engineer who is not a Registered Professional Engineer qualified in accordance with paragraph XXIII-1222 for Code-related activities that are applicable to U.S. nuclear facilities regulated by the NRC. The use of paragraph XXIII-1223 is prohibited.

Since this condition already exists in 10 CFR 50.55a, the NRC does not agree with the proposed

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 recommendation to remove this condition. No changes will be made to this regulatory position.

Change in response to comment: None.

41 Ian Gifford TerraPower C.1.u Page 20 In this revision 3, the staff states that "The NRC staff did not review. Nonmandatory Appendix HBB-Y and therefore is not endorsing it." The same basis was prevalent in revision 2 of RG 1.87. Nonmandatory Appendix HBB-Y, Guidelines for Design Data Needs for New Materials, should be reviewed as part of this RG revision update.

Recommendation:

This restriction should be removed Response: The NRC staff understands there is significant industry interest in an NRC endorsement of Nonmandatory Appendix HBB-Y. The staff is aware that this Nonmandatory Appendix provides guidance for developing a data package for ASME committees to use to incorporate new materials into HB under ASME Boiler and Pressure Vessel Code (BPVC)

Section III, Division 5. It is not within the NRCs regulatory authority to endorse a procedure as acceptable for submitting materials to and meeting requirements of a third-party entity such as the ASME.

Although HBB-Y is not part of the staffs endorsement, it provides useful guidance to ASME Code committees and code users with respect to the adequacy of the data for elevated temperature nuclear applications. In addition, the NRC staff notes that it participates on ASME Code Committees in accordance with Management Directive 6.5 (ML18073A164).

Change in response to this comment:

The staff has revised the RG to establish a position relevant to HBB-Y as shown below:

The NRC staff is not endorsing Nonmandatory Appendix HBB-Y because it is for information only.

Additionally, it is not within the NRCs regulatory authority to endorse a procedure as acceptable for submitting materials to and meeting requirements of a third-party entity such as the ASME.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 42 Ian Gifford TerraPower C.1.t Page 20 Nonmandatory Appendix HBB-T-1710, Special Strain Requirements at Welds, provides a non-specific requirement without a basis. Stress relaxation is already limited by the procedures in HBB-T that were deemed sufficient to address the staff's concerns.

Stress relaxation cracking is geometry specific, and each specific geometry is required to be subjected to the stress relaxation rules in HBB-T if used. The limitation as written is open to interpretation and should be removed or revisited to include a more defined scope and purpose.

Recommendation:

This restriction should be removed Response: The NRC staff disagree with the recommendation.

HBB does not provide any provision in addressing the phenomenon of stress relaxation cracking that has been observed in operating high temperature nuclear and non-nuclear components and considered a structural integrity issue.

This condition was carried over from RG 1.87, Revision 2. Technical justification was provided in NUREG-2245 (ML23030B636) and the referenced contractor report from NUMARK (ML20349A003.)

From the open literature, the cracking phenomenon is attributable to the relaxation of weld residual stress during operation. This causes stress redistribution and deformation enhancement under certain constraint conditions, resulting in cracking around the weld.

The relaxation procedure of HBB-T is related to stress redistribution of the secondary stresses during the hold part of a thermal transient that contributes to creep damage under the creep-fatigue condition.

The two kinds of relaxation are not related.

Change in response to comment: None.

43 Ian Gifford TerraPower Background Page 4 The following statement was unchanged from RG 1.87 Revision 2: "Appendix A to this RG provides guidance for the Response: The NRC staff agree with this recommendation. Appendix A does not provide methods for safety classification but rather for the selection of quality standards.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 quality group classification of components in non-LWR designs. It provides one method that is acceptable to the NRC staff for the safety classification of components for non-LWR nuclear power plants." However, the second sentence, "It provides one method that is acceptable to the NRC staff for the safety classification of components for non-LWR nuclear power plants," misrepresents the contents of Appendix A. The quality group classifications in RG 1.87 Appendix A do not provide a method for the safety classification of components, they provide a method for the selection of quality standards Recommendation:

Revise the background statement on page 4 as follows:

"Appendix A to this RG provides guidance for the quality group classification of components in non-LWR designs. It provides a method for the selection of quality standards.

Change in response to comment:

The subject text has been revised to add the bold text below:

Appendix A to this RG provides guidance for the quality group classification of components in non-LWR designs. In addition, it provides one method that is acceptable to the NRC staff for the selection of quality standards with respect to the safety classification of components for non-LWR nuclear power plants.

44 Ian Gifford TerraPower BRGP b Page 5 There appears to be a typo on page 5. "Basis for Regulatory Guidance Position b" The language under this heading seems to describe the basis for Regulatory Guidance Position C.1.a(1)

Recommendation:

Change the basis heading to read: Basis for Regulatory Guidance Position C.1.a(1)

Response: The NRC staff agree that position C.1.a (1) corresponds to the Basis for Regulatory Guidance Position C.1.b.

Change in response to comment: Basis for Regulatory Guidance Position b shall be changed to Basis for Regulatory Guidance Position a (1).

45 Ian Gifford TerraPower C.1.b(2) Page 12 There appears to be an editorial error in the numbering for Position C.2.b(2). There is no Position C.2.b(1).

Recommendation:

Renumber Position C.2.b(2) as C.2.b(1).

Response: The NRC staff agree that there is a numbering error.

Change in response to comment: Position C.1.b (2) shall be renumbered to C.1.b (1).

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 46 Ian Gifford TerraPower A-2 Pages A-2, -3 The following paragraph appears under the heading "Traditional Approach:" "This appendix addresses pressure-retaining components and supports of high-temperature reactors. The guidance in RG 1.26 should be used for pressure-retaining components containing water, steam, or radioactive material in light-water-cooled nuclear power plants. Other systems not covered by this RG 1.87, such as instrument and service air; diesel engines, their generators, and auxiliary support systems; diesel fuel; emergency and normal ventilation; fuel handling; and radioactive waste management systems, should be designed, fabricated, erected, and tested to quality standards commensurate with the safety function to be performed." For clarity, the scope of systems covered by RG 1.87 Appendix A should be identified at the beginning of the appendix, rather than under the "Traditional Approach" safety classification category header. Additionally, the scope of Appendix A should be clarified to include only components under the jurisdiction of the ASME Code.

Recommendation:

Clarify the scope of Appendix A at the beginning of the appendix. Include clarification that the scope only includes components under the jurisdiction of the ASME Code.

Response: The NRC staff agree with this comment that the scope of systems covered by RG 1.87 Appendix A should be identified at the beginning of the appendix and has made a corresponding change.

The staff does not agree that the scope of Appendix A should be modified to state that it includes only components under the jurisdiction of the ASME Code, which is not a clearly defined term.

Change in response to comment: The following text is removed from the discussion of the Traditional Approach in Section A-2 of Appendix A:

Other systems not covered by this RG 1.87, such as instrument and service air; diesel engines, their generators, and auxiliary support systems; diesel fuel; emergency and normal ventilation; fuel handling; and radioactive waste management regulations or credited in the safety analysis include important to safety SSCs and also include SSCs that do not perform any systems, should be designed, fabricated, erected, and tested to quality standards commensurate with the safety function to be performed.

The following text is added to Section A-1 of Appendix A:

This appendix addresses pressure-retaining components, core components, and supports of high-temperature reactors. Other systems not covered by this RG 1.87, such as instrument and service air; diesel engines, their generators, and auxiliary support systems; diesel fuel; emergency

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 and normal ventilation; fuel handling; and radioactive waste management regulations or credited in the safety analysis include important to safety SSCs and also include SSCs that do not perform any systems, should be designed, fabricated, erected, and tested to quality standards commensurate with the safety function to be performed.

47 Ian Gifford TerraPower Appendix A DG-1436 Appendix A provides a method for the selection of quality standards under the application of RG 1.233 that is based on the safety classification of components. However, the RG 1.233 process is implemented based on the functions performed by SSCs. In the NEI 18-04 process, as endorsed by RG 1.233, special treatments, capability targets, and reliability targets are selected to provide assurance that SSCs perform their functions identified in the licensing basis events.

Therefore, it is better aligned with RG 1.233 to provide guidance that codes and standards should be selected based on the safety significance of the functions performed by SSCs within ASME Code jurisdiction, rather than based on the overall safety classification of the SSC.

Recommendation:

Clarify that the Appendix A guidance for RG 1.233 implementation is based on the safety significance of functions performed by SSCs within ASME Code jurisdiction.

Response: The NRC staff partially agree with this comment. The staff understands this comment is focused on the distinction between the safety significance of the functions performed by SSCs and the overall classification of the SSCs. RG 1.233 does introduce a new aspect to the process of classifying SSCs and identifying the appropriate quality standards with the emphasis on SSC functional performance. The staff position is that the overall classification for an SSC is determined by its function with the highest safety classification. The staff guidance in Table A-1 is based on the overall classification for an SSC.

However, the staff also notes that an SSC with functions of varying safety classification may be able to justify a different code and standard than the overall classification would indicate based on the specific design details and functions of that SSC. The staff has added a paragraph to the discussion of the LMP Approach (RG 1.233) in Section A-2 of Appendix A to clarify this position. The staff does not agree that the scope of Appendix A should be modified to state that it includes only components within ASME Code jurisdiction, which is not a clearly defined term.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 Change in response to comment: The following text is added to the discussion of the LMP Approach (RG 1.233) in Section A-2 of Appendix A:

The LMP approach under RG 1.233 focuses on the functional performance of SSCs, which may in some cases lead to specific SSCs possessing both SR and NSRST functions (e.g. SR for one function and NSRST for a different function). In these cases, the overall classification for an SSC is determined by its function with the highest safety classification. Accordingly, the staff guidance in Table A-1 is based on the overall classification for an SSC. However, an SSC with functions of varying safety classification may be able to justify a different code and standard than the application of Table A-1 to the overall classification based on the specific details of the plant design and functions of that SSC within the plant.

48 Ian Gifford TerraPower Table A-1 Page A-8 The proposed revision modifies Table A-1 Quality Group C column to indicate that ASME Code,Section III, Division 5 should be used for NSRST SSCs DG-1436 appears to contain an underlying perspective that for each NSRST SSC, an applicant using RG 1.233 should justify deviating from each special treatment that would be applied to an equivalent SR component. This perspective is counter to the risk-informed, performance-based NEI 18-04 process endorsed by RG 1.233. NSRST SSCs are not safety-related and should not be viewed as subject to the same treatment as SR SSCs by default. NEI 18-04 and RG 1.233 allow for the specification of right-sized requirements for NSRST SSCs to Response: The NRC staff agree with the commenters interpretation of NEI 18-04. It is not staffs intent to imply that SR codes and standards should be used for designing NSRST SSCs. Rather, the update to Table A-1 is only intended to clarify that the endorsement of ASME Section III, Division 5 extends to NSRST components, and therefore, if proposed by an applicant, NRC would accept this code according to the same criteria that would be applied to its use for SR components.

Based on this and other comments, the staff has revised the footnotes to clarify what is expected in terms of a justification and to make clear that the endorsement of

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 leverage risk information. Special treatments for NSRST SSCs refer to requirements that provide increased assurance beyond normal industrial practices that SSCs perform their design-basis functions. Special treatments can be selected as additions to normal industrial practices, rather than justified eliminations from SR practices. The safety classification of an NSRST SSC should generally be seen as sufficient justification for not applying ASME BPVC Section III and should not require a case-by-case justification.

Recommendation:

Table A-1 should be revised to indicate that industrial codes such as ASME BPVC Section VIII, ASME B31.1, and ASME B31.3 are appropriate for NSRST SSCs if they cover the component service and design conditions, unless a specific special treatment for an NSRST component is determined to require the elevated assurances of ASME BPVC Section III.

ASME Section III, Division 5 extends to NSRST components.

Change in response to comment: Footnotes 7 and 8 to Table A-1 are modified as shown below:

7 These standards may include ASME Code,Section III, Division 5, which is endorsed by NRC.

Codes that have not been endorsed by NRC, such as ASME Code,Section VIII, Division 1 and Division 2, or other alternate standards, may be used with appropriate justification. The applicant should justify how codes that have not been endorsed by the NRC as well as any special treatments are appropriate for the SSC.

8 These standards may include ASME Code,Section III, Division 5, which is endorsed by NRC.

Codes that have not been endorsed by NRC, such as, ASME B31.1/B31.3 or other alternate standards, may be used with appropriate justification. The applicant should justify how codes that have not been endorsed by the NRC as well as any special treatments are appropriate for the SSC.

49 Ian Gifford TerraPower Section A-2 (Traditional Approach) Pages A-2, -3 Table A-1 Page A-8 Section A-2 and Table A-1 information on RG 1.26 and the "Traditional Approach." This information was unchanged from RG 1.87 Revision 2.

The use application for the "Traditional Approach" information in Section A-2 is not clear because RG 1.26 and ASME BPVC Section III, Division 5 are not compatible. It is Response: The NRC staff disagree with this comment.

Appendix A describes three options for how to approach the safety classification of SSCs: the traditional approach, the risk-informed approach under 10 CFR 50.69, and the Licensing Modernization Project (LMP) Approach under RG 1.233. The appendix also emphasizes that it is not possible to know all the design details associated with future designs and the NRC staff will evaluate an applicants implementation of this appendix on a case-

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 unclear whether Appendix A is defining quality groups that are an alternate to RG 1.26 for a non-LWR applicant not utilizing 10 CFR 50.69 or RG 1.233 for the safety classification of SSCs. The RG 1.26 quality groups and the Section A-2 discussion of the types of functions in each RG 1.26 quality group is specific to LWRs and is not technology inclusive.

Additionally, the inclusion of the RG 1.26 quality groups in Table A-1 is misleading. Table A-1 appears to correlate RG 1.26 quality groups to RG 1.233 safety classifications. This correlation is inappropriate because the RG 1.26 quality groups are defined based on the component functions and radionuclide inventory release potential, not on the component safety classification. RG 1.26 Quality Group C may contain safety-related and non-safety-related components, as explained on DG-1436, page A-2.

Recommendation:

1. Remove the RG 1.26 quality groups from Table A-1.

2. Clarify the intended application of the RG 1.26 information in Section A-2 (Traditional Approach) or remove.

3. Consider making Appendix A nonmandatory.

by-case basis to determine if the proposals are appropriate for the specific design.

Under the traditional approach, the primary guidance that is available is in RG 1.26, which is specific to LWRs. However, the traditional approach is still an option for non-LWRs and the intent of including a description of RG 1.26 is to provide the appropriate context for potential non-LWR designs that may choose to use the traditional approach.

Table A-1 is intended to provide general guidance on what quality standards may be appropriate for applicants using any of the three classification methods, not to correlate between classification methods. The staff would like to emphasize some of the text preceding Table A-1 in providing additional context to Table A-1:

Table A-1 represents the design standards that the NRC has determined are appropriate for the different categorization methods described in this appendix without having specific design information available for a reactor design. This does not mean that other codes or standards are not acceptable, but the NRC has not generically evaluated other codes or standards at this time.

There may be instances where deviations from the recommendations in Table A-1 can be justified based on the specifics of the design.

To address the recommendation to make Appendix A nonmandatory, the staff notes that regulatory guidance

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 is not a requirement and provides one acceptable method for meeting the regulatory requirements.

Change in response to this comment: None.

50 Ian Gifford TerraPower Appendix A Terminology The terminology used in the Appendix A text and Table A-1 (carried over from Revision 2) is inconsistent and could be improved for clarity. Specifically, "Safety Classification Categories" and "License Modernization Project Approach" are used in Section A-2 and "Classification Method" and "Risk-Informed (RG 1.233)" are used in Table A-1 for the same concepts.

Recommendation:

Align the usage of terms to be consistent.

Response: The NRC staff partially agree with this recommendation. The title of Section A-2 is Safety Classification Categories, which refers to the categories of SSCs (e.g. SR or NSR for the traditional approach; RISC-1, RISC-2, RISC-3, or RISC-4 for 10 CFR 50.69; and SR, NSRST, and NST for RG 1.233) that may result from the various classification methods.

The term Classification Method in Table A-1 is referring to the three main approaches to classifying SSCs: traditional approach, risk-informed (10 CFR 50.69), and Licensing Modernization Project (LMP).

Therefore, no change will be made to the title of Section A-2 or the term Classification Method in Table A-1 because they are not referring to the same concept.

To clarify the terminology related to "License Modernization Project Approach" in Section A-2 and "Risk-Informed (RG 1.233)" are in Table A-1, the staff has revised the text in both locations to be consistent.

Change in response to this comment:

The staff revised the text "License Modernization Project (LMP) Approach" in Section A-2 and "Risk-Informed (RG 1.233)" in Table A-1 to both state Licensing Modernization Project (LMP) Approach (RG 1.233)."

51 Ian Gifford TerraPower C.1 and C.1.c Page 12 A-4 Pages A-6, -7 Response: The NRC staff disagree with this comment.

The staff would like to clarify that those sections

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 The first paragraph of C.1 states, "The NRC staff is unable to review those sections identified as being in the course of preparation to determine whether they are acceptable, and therefore, the staff does not endorse them." C.1 makes a similar statement stating that NRC staff does not endorse sections it is unable to review.

In contrast to this, A-4 states, "Table A-1 represents the design standards that the NRC has determined are appropriate for the different categorization methods described in this appendix without having specific design information available for a reactor design. This does not mean that other codes or standards are not acceptable, but the NRC has not generically evaluated other codes or standards at this time. There may be instances where deviations from the recommendations in Table A-1 can be justified based on the specifics of the design."

The statement used in A-4 seems a better approach when addressing the NRC staff's view regarding documents it has not yet reviewed.

Recommendation:

1. Revise C.1 to state, "The NRC staff is unable to review those sections identified as being in the course of preparation to determine whether they are acceptable. This does not mean that those sections are not acceptable. There may be instances where those portions can be justified based on the specifics of the design."

2. Revise C.1.c to state, "Where ASME identifies portions of ASME Code,Section III, Division 5, as being in the course of preparation, the NRC staff is unable to review those sections to determine whether or not they are acceptable. This does not mean that those sections are not acceptable. There may be identified as being in the course of preparation refers to sections that are placeholders and currently have no contents within ASME Code. Therefore, the statements in C.1 and C.1.c are not a reasonable comparison to the statement in A-4 that is discussing other codes or standards that exist but have not been endorsed specifically by NRC.

In addition, the staff notes that regulatory guidance is not a requirement and provides one acceptable method for meeting the regulatory requirements. Therefore, a statement that the staff does not endorse a code or standard (or portion of one) on a generic basis does not mean it may not be justified on a more specific basis.

Change in response to this comment: None.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 instances where those portions can be justified based on the specifics of the design."

52 Ian Gifford TerraPower General Comment The quantity and scope of exceptions and limitations on the endorsement of the ASME Code,Section III, Division 5 is likely to result in excessive burden in implementation of the Code. In many instances, the exceptions and limitations consider only select, isolated portions of the Code and neglect to consider implementation of the Code as a whole. The NRCs engagement with the industry during the development and update of the various codes such as ASME Code,Section III, Division 5, should identify areas in which the staff dissents from the industry during the conduct of the code meetings when productive dialogue might occur instead of issuing new or proposed revisions to existing guidance after the fact. It is anticipated that better engagement and recognition of differences during code meetings would yield closer consensus, thereby gaining efficiencies over the current process and possibly reducing the scope of exceptions and limitations in lieu of the creation of additional exceptions and limitations as this proposed revision contains.

Recommendation:

Consider revising internal guidance to stress to NRC staff the importance of identifying and addressing consensus gaps during code meetings.

Response: The NRC staff disagree with this comment.

The staff note that it is an active participant in ASME Code, votes on many actions, and frequently provides comments from its perspective as a safety regulator.

However, the NRC is one member in the consensus committees and language changes and additions are often made without total agreement in the committees.

In those cases, the staff has a regular practice of clearly identifying its concerns during Code meeting discussions and in its votes.

Change in response to this comment: None.

53 Connor Nicol Criteria for SSCs under items commensurate with their contribution to safety or risk -

The guide lacks a clear methodology for evaluating SSCs (structures, systems, and components) under this standard.

While the NRC expresses openness to applying alternate rules Response: The NRC staff disagree with this comment.

This RG is not intended to provide an SSC classification methodology. For NSRST SSCs, the SSC classification methodology is provided in RG 1.233 and the associated report NEI 18-04.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 to non-safety-related with special treatment (NSRST) SSCs, specific criteria should be established for determining acceptable risk thresholds. Without this, the categorization could be inconsistently applied across different reactor designs Based on other comments, the staff will revise the position to incorporate the suggested wording from NEI to clarify the relationship with the approved SSC classification methodology under RG 1.233 and NEI 18-04.

Change related to comment:

NRC staff revised the position from:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20). This use would be subject to NRC review and approval in order to confirm that the risk significance of the NSRST SSC is sufficiently below that of safety-related SSCs and is consistent with the reliability and capability targets specified for the NSRST SSC.

to:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20). The justification for use of alternate requirements in these sections as a special treatment to achieve the reliability and capability targets specified for the NSRST SSC is subject to NRC review and approval. The NRC may review classification of NSRST SSCs in accordance with the approved methodology in RG 1.233.

54 Connor Nicol Welded Product Justification -

Response: The NRC staff disagree with this comment

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 The requirement that applicants justify the use of stress values for welded products (Table HBB-I-14.2, etc.) is necessary but lacks guidance on what constitutes an acceptable justification.

A reduction factor similar to those in non-nuclear codes should be provided as a baseline rather than relying on ad hoc applicant justifications.

The HBB design approach is based on variable design lifetimes, and it accounts for the time-dependence of key design parameters such as the allowable stresses.

These factors are highly dependent on the reactor design.

The NRC encourages ASME to address this topic.

Change in response to this comment: None.

55 Connor Nicol Limitations on Oxidation and Weight Loss Criteria for Composites -

The 1% weight change threshold for oxidation in HHB-3141 is arbitrary and not well-supported by existing literature.

Likewise, the removal of material experiencing 30% weight loss from structural calculations lacks a strong technical basis.

The NRC should provide clear technical justifications for these limits or require applicants to demonstrate the appropriateness of their own selected values based on experimental data.

Response: The NRC staff agree with this comment.

The conditions C.1.gg (1) in DG-1436 (C.1.y(1) in RG 1.87 Rev. 3), which is on the 1% weight change threshold, and C.1.gg (3) in DG-1436 (C.1.y(3) in RG 1.87 Rev. 3), which is on the 30% weight loss limit, both specify that the Designer should determine the appropriate limits and provide justification for those limits.

Change in response to this comment: None 56 Connor Nicol Buckling Criteria for Composite Core Components -

The condition that applicants justify the applicability of strain factors in HGB-III-2000-1 assumes that buckling is purely strain-controlled. However, many high-temperature reactor designs involve significant elastic follow-up, which could alter the failure mechanism. The NRC should mandate additional validation studies for any assumptions of strain-controlled buckling.

Response: The NRC staff disagree with this comment but understands the nature of the comment. Note that the buckling criteria in HGB-III-2000 are for metallic components.

However, no change to condition C.1.y in DG-1436 (C.1.q in RG 1.87 Rev. 3) nor C.1.s in DG-1436 (C.1.k in RG 1.87 Rev. 3) is made as designer should use the load factor if there is significant elastic follow-up in the case of strain-controlled buckling. This is discussed in the ASME Companion Guide to the ASME Boiler &

Pressure Vessel Codes, Volume 1, K.R. Rao, ed.,

ASME Press, New York, NY

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 Change in response to this comment: None 57 Connor Nicol Non-Endorsement of Key Material Models -

The non-endorsement of inelastic models for 9Cr-1Mo-V steel and Alloy 617 (HBB-Z-1325, HBB-Z-1326) creates uncertainty for applicants seeking to use these materials. If these models are inadequate, the NRC should explicitly require an alternative validated model rather than leaving it entirely to the applicants discretion.

Response: The NRC staff disagree with this comment.

The code permits alternate material models to be used by designers as stated in paragraph HBB-Z-1130.

The NRC staff understands that it is a major undertaking to develop a constitutive model suitable for using the inelastic analysis approach to address the HBB design requirements.

It is outside the scope of this RG to provide further guidance on constitutive model development.

The NRC encourages ASME to address this topic.

Change in response to this comment: None.

58 Mark Alphonso-Waters Radiant ASME Code for Nuclear Graphite - Key Limitations Limited Practical Use - ASME Section III, Division 5 graphite design rules were developed by researchers and industry experts but have never been successfully used to build and operate a high-temperature gas-cooled reactor (HTGR). Many past reactor designs (U.S., U.K., Germany, Japan, China) followed different codes, and some existing graphite components would not meet the new ASME requirements. The current top-down approach to quality and technical standards may not be effective.

Lack of Industry Support - The non-metallic supply chain (graphite, ceramics, etc.) is not interested in adopting NQA-1 Response: Thank you for your comment. The NRC staff are aware of the topics discussed in your comment. As this comment does not identify technical inaccuracies within the ASME Code or DG-1436 the NRC staff will not make changes to the RG.

Change in response to comment: None.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 for compliance with ASME HAB requirements. Unlike metals, which have an existing supply chain for nuclear applications, non-metallic material organizations rely on ISO-9001 and AS9100Dalready used for safety-critical military and aerospace components. Implementing a separate, complex QA system with uncertain financial returns discourages suppliers from participating. Additionally, there are no accredited test labs (ISO 17025) for most ASTM graphite tests, making compliance more difficult.

Ongoing Development - The requirements are still evolving as designers and manufacturers identify issues. The non-metallic working group of the ASME high temperature reactor committee is actively working to resolve conflicts, with participation from companies like Radiant and other HTGR developers.

Considerations for reactor size: The requirements apply equally to large power reactors (1000 MW) with tens of thousands of graphite parts and to microreactors (1-5 MW) with hundreds of graphite parts. The differences in safety approach and risk associated with these reactor types are substantial. The code does not currently provide for a risk informed design approach. The same limitation applies to metallic components.

59 Mark Alphonso-Waters Radiant Regarding: Section C. Staff Regulatory Guidance

1. ASME Code,Section III, Division 5 (1) The phrase items commensurate with their contribution to safety or risk appears in Code Case N-940, HBA-2610, HBB-2610 and within ASME Code,Section III, Division 1. In the application of ASME Code,Section III, Division 5 design rules, there is a need to clarify for which SSCs these rules are appropriate. The alternate requirements in these sections should not be applied to Response: The NRC staff disagree with this comment.

The staff position is that the alternate requirements under CC N-940 should not be applied to SR SSCs on a generic basis but may be appropriate for NSRST SSCs. RG 1.87 clearly endorses ASME BPVC Section III, Division 5 for use with SR SSCs.

Change in response to comment:

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20).

This use would be subject to NRC review and approval in order to confirm that the risk significance of the NSRST SSC is sufficiently below that of safety-related SSCs and is consistent with the reliability and capability targets specified for the NSRST SSC.

NRC should not preclude the application of code case N-940 to strictly safety-related SSCs. This limits the flexibility of the owners and designers to assess risk and classify components appropriately. N-940 was previously approved, and further modifications/restrictions are not necessary.

Based on other comments, NRC staff revised the position from:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20). This use would be subject to NRC review and approval in order to confirm that the risk significance of the NSRST SSC is sufficiently below that of safety-related SSCs and is consistent with the reliability and capability targets specified for the NSRST SSC.

to:

The alternate requirements in these sections should not be applied to safety-related SSCs but may be appropriate for use for SSCs categorized as NSRST under RG 1.233 (Ref. 20). The justification for use of alternate requirements in these sections as a special treatment to achieve the reliability and capability targets specified for the NSRST SSC is subject to NRC review and approval. The NRC may review classification of NSRST SSCs in accordance with the approved methodology in RG 1.233.

60 Mark Alphonso-Waters Radiant Regarding: Section C. Staff Regulatory Guidance

d. HAB-3126, Subcontracted Calibration Services; HAB-3127, Subcontracted Testing Services; and HAB-4555.3, Approval and Control of Suppliers of Subcontracted Services The HAB subsection was previously written and approved considering inherent differences between metallic (subsection Response: The NRC staff disagree with this comment and does not have jurisdiction to develop supply chain for the advanced reactor industry.

Change in response to comment: None.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 HAA) and non-metallic (subsection HAB) components.

Several comments in DG-1436 replace portions of HAB with NCA, imposing additional QA requirements on subpart HAB without justification for why the original HAB provisions were inadequate. In most of these cases, the additions from NCA are not necessary or appropriate for non-metallic, non-pressure boundary components. When possible, the NRC should focus on eliminating unnecessary requirements rather than introducing additional ones.

Radiant recommends that NRC take action to enable and help develop a potential supply chain. There is no current advance reactor industry, and no associated supply chain. Providing considerations for developing industries to meet NQA-1 requirements by providing flexibility in applications of HAB, HHA, and HHB as needed in the near term to accelerate industry and development in developing quality engineering.

61 Mark Alphonso-Waters Radiant Regarding: Section C. Staff Regulatory Guidance

u. Nonmandatory Appendix HBB-Y, Guidelines for Design Data Needs for New Materials (1)

The NRC staff did not review Nonmandatory Appendix HBB-Y and therefore is not endorsing it.

One of the most severe limitations of BPVC III.5 is the lack of qualified materials available for high temperature applications.

Every advanced reactor developer is actively planning or considering test programs that will support future code cases to qualify materials. These efforts will be most productive if the NRC is engaged in the process. NRC can signal this readiness by reviewing, commenting and endorsing HBB-Y.

Radiant recommends NRC review and endorse section HBB-Y.

Response: The NRC staff understands there is significant industry interest in an NRC endorsement of Nonmandatory Appendix HBB-Y. The staff is aware that this Nonmandatory Appendix provides guidance for developing a data package for ASME committees to use to incorporate new materials into HB under ASME Boiler and Pressure Vessel Code (BPVC)

Section III, Division 5. It is not within the NRCs regulatory authority to endorse a procedure as acceptable for submitting materials to and meeting requirements of a third-party entity such as the ASME.

Although HBB-Y is not part of the staffs endorsement, it provides useful guidance to ASME Code committees and code users with respect to the adequacy of the data for elevated temperature nuclear applications. In addition, the NRC staff notes that it

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 participates on ASME Code Committees in accordance with Management Directive 6.5 (ML18073A164).

Change in response to this comment:

The staff has revised the RG to establish a position relevant to HBB-Y as shown below:

The NRC staff is not endorsing Nonmandatory Appendix HBB-Y because it is for information only.

Additionally, it is not within the NRCs regulatory authority to endorse a procedure as acceptable for submitting materials to and meeting requirements of a third-party entity such as the ASME.

62 Mark Alphonso-Waters Radiant With regards to safety classification and Appendix A, NEI 18-04 guidelines should be utilized with consistency.

Special consideration should be placed for applications in which accident tolerant fuel (such as TRISO) minimizes the consequence to the public. Certain features of a non-light water reactor, such as the primary coolant as a pressure boundary, may no longer be significant while fuel safety limits are maintained.

Response: The NRC staff agree with this comment.

This is addressed via the following text in Section A-1 of Appendix A:

In establishing standards acceptable to the NRC staff, it is not possible to know all the design details associated with future designs.

There may be some instances where the standards established in this appendix may be overly conservative or possibly require supplementation for a specific design. As such, the NRC staff will evaluate an applicants implementation of this appendix on a case-by-case basis to determine if the proposals are appropriate for the specific design.

Change in response to comment: None.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 63 Jon Facemire NEI 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.

DG-1436 Basis: 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.

Response: The NRC staff understand your comment.

Change in response to comment:

Language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5:

Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of performance monitoring approaches, such as surveillance methods, in-service inspection, and continuous monitoring. Continuous monitoring of an SSCs ability to perform its function, such as online monitoring for cracking, pressure, temperature, or chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

64 Jon Facemire NEI 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.

DG-1436 Basis: 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 Response: The NRC staff disagree with this comment.

While Section III and ASME B31.1 do not explicitly invoke Section V, Article 14 for performance demonstration, both require the NDE systems to be demonstrated. Since code case N-940 is allowing the use of UT in lieu of RT for NSRST SSCs, it is important that the UT system be demonstrated in its capability to detect and size flaws in order to have confidence that UT can be used in lieu of RT. For CC N-940 to be applied to NSRST SSCs,Section V,

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 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.

Article 14 Intermediate Rigor provides reasonable qualification requirements to demonstrate the effectiveness of the UT examination systems.

The intent of the condition, however, is not to require conformance with Section V, Article 14, but rather to indicate that staff would accept that methodology for an NSRST component without further justification.

Change in response to comment:

Language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5:

Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of performance monitoring approaches, such as surveillance methods, in-service inspection, and continuous monitoring. Continuous monitoring of an SSCs ability to perform its function, such as online monitoring for cracking, pressure, temperature, or chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 65 Jon Facemire NEI 3-CC-N-940. Progressive sampling under Nonmandatory Appendix B should not be applied to elevated temperature service (temperature > 750°F).

DG-1436 Basis: 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 CCN-940.

Response: The NRC staff understand your comment.

Change in response to comment:

Language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5:

Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of performance monitoring approaches, such as surveillance methods, in-service inspection, and continuous monitoring. Continuous monitoring of an SSCs ability to perform its function, such as online monitoring for cracking, pressure, temperature, or chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

66 Jon Facemire NEI 4-CC-N-940. For applications of progressive sampling under Nonmandatory 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.

Response: The NRC staff disagree with this comment.

The staff has considered the entirety of the construction Code in developing this position as well as considering the range of NSRST SSCs to which it could be applied under this endorsement.

Based on the scope of this endorsement being for safety-significant NSRST SSCs that should provide increased assurance beyond normal industrial

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 DG-1436 Basis: Industrial codes, such as ASME B31.3 (Ref.

18), generally 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.

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 Normal Fluid Service systems (B31.3 paragraph 341.4.1(b)) is sufficient for this scope of components and additional conditions are not required.

practices, the staff believe it is necessary to approve the user's sampling plans to assess the adequacy of special treatments. The text and basis for 4-CC-N-940 have been revised to clarify the purpose of the condition and positions (1), (2), and (3) within the condition.

Change in response to comment:

Condition 4-CC-N-940 has been revised to state the following:

4-CC-N-940. For applications of progressive sampling under Nonmandatory Appendix B, the initial sample size should be one of the following:

(1) a population justified statistically to provide 95% confidence that 5% or fewer of the welds contain defects, or (2) a lesser initial sample justified as an alternative approach as described in footnote 6 below, or (3) for instances where a designer does not prefer to use the statistical justification in (1) or develop an alternative approach in (2), 50%

random sampling is acceptable.

The basis for 4-CC-N-940 has been revised to state the following:

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 The initial sample requirement in commercial codes (e.g., B31.1 and B31.3) for SSCs that are most analogous to those for nuclear service within the scope of this Code Case generally varies between 5% and 100%, depending on their design of the component and service condition. Given this variance and without specific information about the SSC to which the Code Case may be applied, staff cannot conclude that a 5% sample is generically appropriate. Position (1) in condition 4-CC-N-940 in Table 5 is 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 and under commercial codes in certain situations. Position (2) provides explicit clarity that a CC user may propose what it believes to be an appropriate sample percentage based on the specific design and service condition of the SSC. Meanwhile, position (3) provides a simpler approach that does not require additional analysis or justification and is more likely to be applied in cases where there are fewer welds to inspect.

Consistent with a graded approach, NRC staff finds 50% random sampling to be an acceptable approach based on the generally lower safety significance of NSRST SSCs relative to SR SSCs.

In addition, language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5:

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of performance monitoring approaches, such as surveillance methods, in-service inspection, and continuous monitoring. Continuous monitoring of an SSCs ability to perform its function, such as online monitoring for cracking, pressure, temperature, or chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

67 Jon Facemire NEI 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).

DG-1436 Basis: 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.

Response: The NRC staff understand your comment.

Change in response to comment:

Language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5:

Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of performance monitoring approaches, such as surveillance methods, in-service inspection, and continuous monitoring. Continuous monitoring of an SSCs ability to perform its

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 function, such as online monitoring for cracking, pressure, temperature, or chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

68 Jon Facemire NEI 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.

DG-1436 Basis: 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 Response: The NRC staff disagree with this comment.

The staff has considered the entirety of the construction Code in developing this position as well as considering the range of NSRST SSCs to which it could be applied under this endorsement.

Regarding leak testing, both B31.1 and B31.3 require hydrostatic testing. B31.3 allows the owner to specify initial service leak testing as an alternative to hydrostatic testing only for Category D fluid systems.

Other fluid categories including normal process fluid still require hydrostatic testing.

Consistent with the stated objective of CC N-940 to apply Section III for design with industrial codes for construction, the staff position is that incorporating this provision from B31.1 is appropriate to provide reasonable confidence of NSRST SSC performance on a generic basis.

Change in response to comment:

Language has been added to explicitly provide greater flexibility for Conditions 1-CC-N-940 through 6-CC-N-940. The following language is included as a footnote to Table 5:

Alternative approaches to those specified in conditions 1-CC-N-940 through 6-CC-N-940 may be proposed with justification subject to

Response to Public Comments on Draft Regulatory Guide (DG)-1436, Acceptability of ASME Code,Section III, Division 5, High Temperature Reactors Proposed Revision 3 of Regulatory Guide 1.87 systems (B31.3 paragraph 345.1(a)) is sufficient for this scope of components and additional conditions are not required.

NRC review and approval. Justification should be based on the ability to meet the reliability and capability targets for the SSC. The justification can also include consideration of performance monitoring approaches, such as surveillance methods, in-service inspection, and continuous monitoring. Continuous monitoring of an SSCs ability to perform its function, such as online monitoring for cracking, pressure, temperature, or chemistry changes to indicate a leak or loss of boundary integrity, could bolster such a justification.

69 Anonymous I recommend no nuclear development until nuclear fusion is a safe and viable energy surce owned and controlled by the public utilities not private.

Response: Thank you for the comment. The NRC staff consider this comment to be out of the scope of the current action.

Change in response to comment: None.