ML22101A267

DG-1380 (RG 1.87 Rev 2) Public Comment Resolution Table

ML22101A267
Person / Time
Issue date:	01/30/2023
From:	Jeffrey Poehler, Robert Roche-Rivera NRC/RES/DE
To:
Poehler, J; Roche-Rivera, R
Shared Package
ML22101A233	List: ML22101A254 ML22101A263 ML22101A267
References
DG-1380, NRC-2021-0117 RG 1.87 Rev 2
Download: ML22101A267 (13)

Text

Response to Public Comments on Draft Regulatory Guide (DG)-1380 Acceptability of ASME Code Section III, Division 5, High Temperature Reactors and NUREG-2245, Technical Review of the 2017 Edition of ASME Section III, Division 5, `High Temperature Reactors' On August 20, 2021, the NRC published a notice in the Federal Register (86 FR 46888) that Draft Regulatory Guide, DG-1380 (Proposed Revision 2 of Regulatory Guide [RG] 1.87; Agencywide Documents Access and Management System [ADAMS] Accession No. ML21091A276) and NUREG-2245 (ADAMS Accession No. ML21223A097) were available for public comment. The Public Comment period ended on October 19, 2021. Subsequent to the public comment period for DG-1380, the NRC staff completed its review of Code Cases N-872 and N-898, related to the use of Nickel-Based Alloy 617. On March 1, 2022, the NRC staff issued a supplemental Federal Register notice (87 FR 11490) to DG-1380 requesting public comment on the staffs proposed endorsement of Code Cases N-872 and N-898. The public comment period for the supplemental Federal Register notice ended on March 31, 2022. The NRC received comments from the organizations listed below. The NRC has combined the comments and NRC staff responses in the following table.

Comments were received from the following:

Letter ADAMS Accession No. Commenter Affiliation Commenter Name No.

1 ML21286A738 -- N. Prasad Kadambi 3 ML21292A289 Nuclear Energy Institute (NEI) Mark A. Richter 4 ML21292A291 GE-Hitachi Michael Arcaro Comments on DG-1380:

Commenter Specific Comments NRC Staff Resolution N. Prasad It is unfortunate that the NRC staff has chosen to use the old- The staff disagrees with this comment.

Kadambi fashioned guidance structure of updating a LWR Regulatory Guide supported by a NUREG for advanced reactor high-temperature The comment implies that the NRCs endorsement of the components. It would have been more beneficial to use a structure ASME Section III, Division 5 code (Section III-5) is that would be amenable to accomplish the aspirations of the inconsistent with the use of risk-informed, performance-Commissions direction in SRM-SECY-98-0144, White Paper on based approaches for licensing of advanced non-light Risk-Informed and Performance-Based Regulation (White Paper). water reactors (ANLWR). The staffs simple aim in Such a structure would seek to accomplish outcomes consistent with revising RG 1.87, however, was to determine whether the January 2023

recent statements by the staff regarding regulating toward ASME BPV Code,Section III, Division 5, is an reasonable assurance of adequate protection. Clearly such a acceptable method for assuring the integrity of structures, structure would focus on safety decision-making rather than systems, and components under specified conditions, specifying a process for compliance with a prescriptive set of rules including temperatures higher than those specified in that is codified in the ASME Section III, Division 5 (S-III-5) Section III, Division 1.

standard.

The staff is not mandating the use of ASME Boiler and The decision-making would take as input information from results Pressure Vessel Code,Section III, Division 5 (Section produced by application of S-III-5 to a set of components. This set of III-5, or S-III-5 in the terminology used in the comment).

components would be functional contributors to some significant DG-1380 states at the beginning of Section C.1 that feature of an advanced reactor design. The application of S-III-5 to [t]he NRC staff endorses the 2017 Edition of the ASME the components would produce information which characterizes the Code,Section III, Division 5, as a method acceptable to capabilities of systems that support the design feature. The designer the NRC staff for the materials, mechanical/structural incorporates these system capabilities to achieve functional purposes design, construction, testing, and quality assurance of to be provided by the design feature. The functional requirements mechanical systems and components and their supports associated with design feature would be met by systems that perform of high-temperature reactors, with the exceptions and to set criteria to deliver physical needs of the design. Ideally, the limitations stated below.

functional requirements and criteria would be demonstrably fit-for-purpose with no unnecessary requirements. Achieving all this The staffs endorsement with exceptions and limitations would be in keeping with the Commissions White Paper. in DG-1380 of Section III-5 does not preclude the use of performance-based or risk-informed approaches.

The range of application for S-III-5 is vast when liquid-metal-cooled, gas-cooled, and molten-salt fueled or cooled reactors are The comment states that the range of application for S-considered. The multitude of possibilities of materials, construction III-5 is vast when liquid-metal-cooled, gas-cooled, and methods, and service environments make sensible prescriptive molten-salt fueled or cooled reactors are considered. The approaches almost impossible. Yet the combination of NUREG- multitude of possibilities of materials, construction 2235 and DG-1380 that the NRC staff has chosen to employ uses methods, and service environments make sensible just such an approach. It appears that the regulated community is so prescriptive approaches almost impossible.

much in need of S-III-5 that no negative comments have come forth so far even though the cost impacts are likely to be substantial and The staff agrees that ANLWRs may have a wide range of sub-optimal. It appears that this community is not paying attention to coolants, materials and construction methods. Section the costs of implementing S-III-5. Under the circumstances, it falls III-5 provides methods to prevent certain failure modes, to the NRC to meet its obligations under the Nuclear Energy such as overload, stress rupture, creep and creep-fatigue, Innovation and Modernization Act (NEIMA) to find a risk-informed but does not address the effects of the coolant and performance-based (RIPB) approach to fulfill its role in environment on materials. Aspects such as corrosion and reducing the costs of advanced reactors. irradiation will have to be addressed by applicants for 2

ANLWR designs by means other than those provided in S-III-5 requires the designer to provide a complete Design Section III-5, such as environment-specific materials Specification which fulfills Owner/Operator responsibilities while qualification programs, and in-situ surveillance programs also complying with whatever the local regulatory authority requires. during operation, or other strategies to provide reasonable The combination of NUREG-2235 and DG-1380 shows scant assurance of component reliability. The NRC is also recognition of the fact that Design Specifications that draw only on reviewing for endorsement ASME BPV Code Section XI, S-III-5 would be quite incomplete. S-III-5 does look to ASME Division 2, Reliability Integrity Management (Section Section III, Division 1 rules for many needs. However, this only XI-2), which allows the use of diverse strategies to makes the process more prescriptive and convoluted. The pursuit of ensure component reliability. Section XI-2 includes a less prescriptive approach needs to look to what the Commission risk-informed and performance based approaches, such has explicitly offered by way of remedies for this type of situation in as suggested by the comment. One of the strategies the White Paper. provided by Section XI-2 for reliability integrity management is design practices, so the design process for It should be clear to the staff at this point of its rulemaking that the ANLWRs need not operate in a silo as implied by the needs for the 10 CFR Part 53 would motivate the NRC staff to seek comment.

RIPB solutions for advanced reactors. RIPB solutions will require that the design function not operate in a silo, ignoring construction The requirement for the designer to provide a complete and operation needs as has been the practice in the past with the Design Specification as required by Section III-5 does existing LWR fleet. The combination of NUREG-2235 and not preclude designers from addressing other design DG-1380 essentially continues existing practices by ignoring the aspects that are not addressed by Section III-5, and would Commissions recognition that NRC should offer flexibility to not hinder the user of risk-informed and performance-determine how to meet the established performance criteria in ways based approaches to address regulatory requirements.

that will encourage and reward improved outcomes. In the context of No changes were made to the RG or NUREG-2245 as a 10 CFR Part 53, the improved outcomes clearly relate to functional result of this comment.

success and not just avoidance of component failure.

The NRC staff has immediate access to a number of guidance documents and research products that could address an RIPB approach to S-III-5. NUREG/BR-0303, Guidance for Performance-Based Regulation offers guidance for employing risk-informed, performance-based, and RIPB approaches to all NRC regulated activities. It also offers a methodology for using a structured set of performance objectives capable of dealing with the integrated decision-making necessary to roll-up component performance capabilities into functional performance success. Early research related to alternatives to prescriptive regulation is documented in 3

NUREG/CR-5392, Elements of an Approach to Performance-Based Regulatory Oversight. Additional research related to decision-making under uncertainty is documented in NUREG/CR-6833, Formal Methods of Decision Analysis Applied to Prioritization of Research and Other Topics. The American Nuclear Societys Risk-Informed, Performance-Based Principles and Policy Committee (RP3C) has considerable information on application of RIPB methods for advanced reactors.

In summary, the combination of NUREG-2235 and DG-1380 falls short of providing guidance that would accomplish the objectives of NEIMA. The NRC staff should recognize this as part of the NRC-2021-0177 comment resolution process. Additionally, there is an opportunity to propose activity that focuses on achieving the Commissions objectives in the White Paper.

NEI Should add a statement that Code Cases may be implemented upon The staff disagrees with this comment. It appears that ASME Committee approval. NEI is suggesting that Code Cases should be automatically approved by NRC when the Code Cases are approved by the ASME Code Committee. The NRC must independently review code cases that have been approved by ASME to determine if the code cases would comport with NRC requirements prior to approving them for use.

No changes were made to the RG or NUREG-2245 as a result of this comment.

NEI Should add a statement that deviations from Code Case may be It is not clear from the comment whether the comment is made with appropriate 50.59 analysis or equivalent analysis. intended to be restricted to Code Cases or to the Code, broadly. Nevertheless, the NRC staff agrees that in some instances 10 CFR 50.59, Changes, tests and experiments, may be available to make changes of this sort. Because the NRC staff anticipates that applicants or licensees would incorporate this RG into their licensing basis in different ways, it is premature at this point for the 4

NRC to establish whether 10 CFR 50.59 or other change control processes would be available or applicable.

This RG is a guidance document, not a regulation.

Applicants and licensees are always free to use other means to demonstrate that reactor designs meet the applicable regulations.

No changes were made to the RG or NUREG-2245 as a result of this comment.

NEI Section C.1.z(1); The staff agrees with this comment, and has revised the RG to remove the limitation related to extrapolation in Extrapolation to determine the allowable time for use-fractions is an HGB-3224(d) and made conforming changes in the intended use of the Code, both to obtain tib in HGB-3224(d) and in NUREG report.

other portions of the Code, including those referenced by the staff in the discussion of NUREG-2245 page 3-193. Extrapolation is not prohibited elsewhere in the Code; the Code is silent on extrapolation in the referenced paragraphs, which does not prohibit extrapolation as indicated in the Foreword to the Code, the Code does not address all aspects of these activities and those aspects that are not specifically addressed should not be considered prohibited.

Prohibiting extrapolation for determining allowable times may place an economic penalty on designs by restricting component design life or requiring significant overdesign to obtain the required life. It is noted that HGB-1124 restricts the time at elevated temperature to the maximum time associated with Smt; extrapolation does not permit increasing the operating time at elevated temperature beyond the restriction of HGB-1124, but rather allows for calculation of the use-fraction in conditions of low operating stress relative to the allowables.

Restricting extrapolation for a component with a specified 300,000-hour design life at elevated temperature results in a use-fraction of greater than or equal to 1.0 regardless of the specified Service Loadings; this would occur because the denominator in the use-5

fraction summation would always be less than or equal to 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. To achieve a time fraction of 1.0 in this case, all Service Level A, B, and C loadings would be required to have a stress less than or equal to St at 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at the appropriate temperature, even if the Service Loading duration was much shorter, with higher stresses permitted by HGB-3224(c) equation (10).

The most significant contributors to the use-fraction summation will be Service Loadings where the stresses are relatively high, and the allowable times have limited or no extrapolation. The Code margins for these Service Loadings are not at risk of being degraded by extrapolation. Lower stress Service Loadings, where tib is extrapolated to longer times, would be smaller overall contributions to the use-fraction summation since the total duration of all elevated temperature service loadings is limited to the time associated with Smt. Since the low stress Service Loadings would have small overall contribution to the use-fraction, extrapolation error in these cases would not have a significant impact on the overall margins.

NEI Appendix A - General; The NRC staff agrees with this comment to the extent that an applicant or licensee need not comply with There are numerous places within Appendix A that are inconsistent specified 10 CFR Part 50 requirements with respect to with 10 CFR 50.69. See comments below for specific examples of RISC-3 components in accordance with 10 CFR 50.69. In where Appendix A is inconsistent with 10 CFR 50.69. addition to changes made to address subsequent specific examples, the staff clarified that the definition of safety-related in 10 CFR 50.2 used in both traditional and 10 CFR 50.69 SSC classification processes may not be fully applicable to high temperature reactor designs.

Further, Quality Group D was removed from the Appendix as it relates to NSR SSCs that are not important to safety and the designers and owners are responsible for assigning the appropriate standards for these SSCs.

NEI Appendix A, A-2. Safety Classification Categories - Traditional The NRC staff agrees with the first part of this comment, Approach, Page A-2 (19 of 26); but the staff disagrees with the proposed new paragraph.

Accordingly, the staff significantly revised Appendix A, 6

It is important to point out that in RG 1.26 Quality Group D is Section A-2. The staff reworded the second paragraph of applied only to water- and steam-containing components that are Section A-2 to clarify the applicability of Quality Group not part of the reactor coolant pressure boundary or included in D (as defined in RG 1.26) and RG 1.143. The staff also Quality Groups B or C but are part of systems or portions of systems clarified that RG 1.26 and RG 1.143 endorsed standards that contain or may contain radioactive material. for components within the scope of Quality Group D and the scope of RG 1.143. Finally, the staff stated that The first two full paragraphs should be combined into one paragraph certain endorsed standards include high temperature and re-written as shown below. operating conditions within their scope that may be appropriate for high temperature reactor applications, and Proposed New paragraph: the adequacy of these standards may be addressed during SSCs that are NSR may function to prevent a radiological release to the review of an application for a specific design.

the public by ensuring that no dose to the public is beyond the regulatory limits of 0.1 rem total effective dose equivalent (TEDE) set by 10 CFR Part 20, Domestic Licensing of Production and Utilization Facilities, (Ref. A-5). While such SSCs do not meet the criteria for an SR SSC, there is still a need to ensure component integrity. RG 1.26 assigns Quality Group D to components that contain or may contain radioactivity but are not part of the reactor coolant pressure boundary or included in Quality Groups B or C.

Refer to RG 1.26 for more information on this traditional approach.

RG 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, (Ref. A-10) provides information related to the classification of radioactive waste management systems that fall within the scope of that RG. SSCs that are NSR and do not meet the criteria for special treatment are left to the applicant to specify any standards for design and fabrication.

NEI Appendix A, A-2. Safety Classification Categories - Traditional The staff disagrees with this comment. There may be Approach, Page A-2 (19 of 26); additional reasons that special treatment may be appropriate, not just because a system contains Last full paragraph states: NSR mechanical components that need radioactive material. The RG was revised to provide an special treatment, such as for systems containing high levels of additional example, defense-in-depth, for the application radioactive material of special treatment to NSR mechanical equipment. The staff also clarified the discussion of NSR SSCs.

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Change to: NSR mechanical components that need special treatment, such as for systems containing high levels of radioactive material as this part of section A-2 only applies to the Traditional Approach for Safety Classification Categories.

NEI Appendix A, A-2. Safety Classification Categories - Risk Informed The staff agrees with this comment to the extent the draft Approach, Page A-3 (20 of 26); language was confusing or could have been understood as overly restrictive. 10 CFR 50.69(d) requires, in part, Second full paragraph is inconsistent with 10 CFR 50.69. that, if the risk-informed categorization process is voluntarily adopted by an applicant, the applicant shall Needs to be re-written so that for RISC-2 components the owner has ensure that RISC-1 and RISC-2 SSCs perform their the flexibility allowed by 10 CFR 50.69 and that for RISC-3 safety-related functions consistent with the categorization components,Section III and Appendix B are not required. process assumptions, and that the treatment of RISC-3 SSCs is consistent with the categorization process. The RG was revised to indicate standards that may be used with appropriate justification to demonstrate categorization process assumptions are satisfied for RISC-2 and RISC-3 components and to indicate that the NRC endorsed the ASME Code,Section III, Division 5 standard as an appropriate standard to meet regulatory requirements applicable to RISC-1 components.

NEI Appendix A, A-4 Quality Group Classifications, The staff disagrees with this comment. Section 50.69 Pages A-5 and A-6 (22 and 23 of 26); does not control the quality group for an SSC classified using the traditional or LMP approaches. The staff Should be re-written to be consistent with 10 CFR 50.69 (i.e., for nonetheless revised the RG to clarify how to determine Group B and C the owner defines these requirements). For Group B SSC quality group depending on the classifications that the owner also needs to provide reasonable confidence. For Group result from the three approaches (addressed in the C, the requirements need to be consistent with the categorization Appendix) an applicant could take to SSC classification process. and the safety significance of the SSCs functions. The staff also revised the RG to indicate that the standards acceptable to the staff are based on staff judgement without full knowledge of the design details and that other standards and quality assurance aspects may be appropriate depending on the design details.

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NEI Appendix A, Table A-1, Page A-7 (24 of 26); The staff disagrees with this comment. Section 50.69 does not control the quality group for an SSC classified Should be re-written to be consistent with 10 CFR 50.69 (i.e., for using the traditional or LMP approaches. The staff Group B and C the owner defines these requirements / applicable nonetheless revised the RG to clarify how to determine codes and standards). For Group B the owner also needs to provide SSC quality group depending on the classifications that reasonable confidence. For Group C, the requirements need to result from the three approaches (addressed in the be consistent with the categorization process. Appendix) an applicant could take to SSC classification and the safety significance of the SSCs functions. The staff also revised the RG to indicate that the standards acceptable to the staff are based on staff judgement without full knowledge of the design details and that other standards and quality assurance aspects may be appropriate depending on the design details.

Table A-1 is the staffs recommendation based on limited knowledge of the design for a reactor. The information in the table is guidance, not a requirement. The RG was revised to annotate Table A-1 to indicate that alternatives may be appropriate, and to clarify that alternative treatment under 10 CFR 50.69 for SSCs categorized as RISC-1, RISC-2, RISC-3, or RISC-4 requires NRC review and approval in accordance with 10 CFR 50.69.

NEI Appendix A, Table A-1, Page A-7 (24 of 26); The staff does not agree with this comment. Table A-1 is the staffs recommendation based on limited knowledge Table A-1 is not consistent with 10 CFR 50.69. The interpretation of of the design for a reactor. The information in the table Table A-1 is such that the user is required to use the codes and is guidance, not a requirement. The staff nonetheless standards as defined in the table for the specified quality groups. rewrote Table A-1 to clarify the guidance.

However, there may be alternative design and construction codes Class A and B applicable and acceptable for Quality Group A, B, C and D components.

Table A-1 should be re-written.

Comments on NUREG-2245:

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Commenter Specific Comments NRC Staff Resolution NEI Page 3-107, lines 16-20; The staff agrees with this comment, and notes that the same limitation is in DG-1380. After further review, the

[NUREG-2245 Text]: The NRC staff is not endorsing Mandatory staff revised the limitations on Smt, St, and Sr for Type Appendix HBB-I-14 for: (a) Type 304 stainless steel (Type 304 SS) 304 stainless steel in Section C.1.u(1)(a) of the RG, to be values of Smt, St, and Sr for temperatures greater than 1300 °F or dependent on both temperature and time.

700 °C.

Conforming changes were made in NUREG-2245, As the basis for the above restriction, NUREG-2245 Sections 3.7.5, Section 3.7.

3.7.6, and 3.7.9 utilized comparisons in ANL/AMD-21/1, Tables 3 and 4. The staff proposed a cutoff at temperatures where the difference is -10% or greater. Review of Tables 3 and 4 of ANL/AMD-21/1 indicates that typically the difference does not reach -10% until longer times, for example at 725°C St in Table 3 does not drop below the 10% criteria until 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Has the staff considered use of both temperature and time to set this limit and allow short duration conditions at temperatures greater than 1300 °F or 700 °C, where the 2017 Code allowable stresses meet the 10% criteria?

NEI Page 3-107, lines 21-22; The staff agrees with this comment, and notes that the same limitation is in DG-1380. After further review, the

[NUREG-2245 Text]: The NRC staff is not endorsing Mandatory staff revised the limitation on the Sr values for Type 316 Appendix HBB-I-14 for: [] (b) Type 316 stainless steel (Type 316 stainless steel in C.1.u(1)(b) of DG-1380 to be SS) Sr values for temperatures greater than 1300 °F or 700 °C. dependent on both temperature and time.

As the basis for the above restriction, NUREG-2245 Section 3.7.9 Conforming changes have been made in NUREG-2245, utilized comparisons in ANL/AMD-21/1 Table 6. The staff Section 3.7.

proposed a cutoff at temperatures where the difference is -10% or greater. Review of Tables 6 of ANL/AMD-21/1 indicates that typically the difference does not reach -10% until longer times, for example, at 725°C Sr in Table 6 does not drop below the 10%

criteria until 1,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Has the staff considered use of both temperature and time to set this limit and allow short duration 10

conditions at temperatures greater than 1300°F or 700°C, where the 2017 Code allowable stresses meet the 10% criteria?

NEI Page 3-107, lines 24-25; The staff agrees with this comment, and notes that the same limitation is in DG-1380. After further review, the

[NUREG-2245 Text]: The NRC staff is not endorsing Mandatory staff revised the limitation on the Smt, St, and Sr values Appendix HBB-I-14 for: [] (c) 2-1/4Cr-1Mo material Smt, St, for 2-1/4Cr-1Mo in C.1.u(1)(c) of the RG to be and Sr values for temperatures greater than 950 °F or 510°C. dependent on both temperature and time.

As the basis for the above restriction, NUREG-2245 Sections 3.7.5, Conforming changes have been made in NUREG-2245, 3.7.6, and 3.7.9 utilized comparisons in ANL/AMD-21/1, Tables 10 Section 3.7.

and 11 and Figure 4. Review of Tables 10 and 11 indicates that the 2017 Code allowable stresses were conservative at 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> up to 550°C. Has the staff considered use of both temperature and time to set this limit and allow short duration conditions at temperatures greater than 950 °F or 510°C?

GE-Hitachi Abstract, Page iii, line 3; The comment seems to imply that newer versions of codes and standards available 6 months before the NUREG 2245 has evaluated 2017 edition of ASME III Division 5. docketing of an application should be used, regardless of New reactor designs will need to evaluate reactor design against NRCs review and endorsement of the revised codes and applicable codes and standards in effect 6 months before the standards. The NRC does not have such a requirement.

docketed date of the licensing application. There are over 20 An applicant or licensee that chooses to use RG 1.87, changes (records) proposed against ASME III Division 5 2021 Revision 2 (final version of DG-1380), will use the edition. Are there plans to update NUREG 2245 to current issued editions of the Code endorsed in the RG. An applicant ASME code? that chooses to use a version of the ASME Code,Section III, Division 5, that is not endorsed in RG 1.87, Revision 2, will need to justify any deviations from the provisions of the 2017 and 2019 Editions of the Code endorsed in RG 1.87, Rev. 2. There are currently no plans to update NUREG-2245. The NRC staff will likely consider the need to review later editions of ASME Section III, Division 5 for endorsement as needed and would consider a request by ASME and/or industry stakeholders to do so.

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No changes were made to the RG or NUREG-2245 as a result of this comment.

GE-Hitachi Section 1.4, Pages 1-3 and 1-4; Since the issuance of DG-1380 the NRC staff completed its review of two additional Code Cases, namely N-872 Section 1.4, Review of ASME Code Section III Division 5 and and N-898. The staffs endorsement of these additional Associated Code Cases, page 1-4, line 5. The only current code code cases is included in the final version of the RG 1.87 cases in scope of NUREG 2245 are N-861 and N-862. There are Revision 2. The review of other Section III, Division 5 numerous CC listed against ASME III D5 such as N-290-3, N-812- code cases is out of the scope of the current effort, but the 1, N-822-4, N-872, N-875, N-898. Is there a future plan to endorse staff anticipates these code cases could be addressed in the active code cases? future revisions of RG 1.87.

No changes were made to the RG or NUREG-2245 as a result of this comment.

GE-Hitachi Section 3, Page 3-1 lines14-15; The comment refers to documents that are copyrighted by ASME. Copies of these documents can be purchased The assessments of NUREG 2245 reviews HBB sections with CC from ASME, Two Park Avenue, New York, NY 10016-1592, 1593, 1594, 1595, 1596. Can these legacy CC be provided as 5990; telephone (800) 843-2763. Purchase information is public documents on NRC webpage [Home Nuclear Reactors New available through the ASME Web-based store at Reactors Advanced Reactors (non-LWR designs) Endorsement https://www.asme.org/publications-Review of ASME B&PV Code Section III, Division 5, "High submissions/publishing-information.

Temperature Reactors]?

No changes were made to the RG or NUREG-2245 as a result of this comment.

GE-Hitachi Section 4, Pages 4 4-7; The comment refers to documents that are copyrighted by ASME. Copies of these documents can be purchased Section 4, Technical Review of Code Cases N-861 and N-862. Will from ASME, Two Park Avenue, New York, NY 10016-these code cases be available as public documents in NRC webpage 5990; telephone (800) 843-2763. Purchase information is

[Home Nuclear Reactors New Reactors Advanced Reactors (non- available through the ASME Web-based store at LWR designs) Endorsement Review of ASME B&PV Code Section https://www.asme.org/publications-III, Division 5, "High Temperature Reactors]? submissions/publishing-information.

No changes were made to the RG or NUREG-2245 as a result of this comment.

GE-Hitachi Section 5, Pages 5 5-5; The NRC staff will continue to actively participate in the Section III, Division 5 code committees. It is possible 12

Section 5 of NUREG 2245-Draft Report for Comment lists that the NRC limitations and exceptions could be exceptions and or limitations to ASME Section III Division 5 2017 removed if (1) The ASME code committees revise edition. Is there future scope between NRC and ASME to update Section III, Division 5 to address the NRC limitations ASME III-D5 to disposition the NRC exceptions and limitations and and exceptions, or (2) provide additional technical bases re-assess future revision of ASME III Division 5 under revision to supporting the existing code provisions on which the NUREG 2245? NRC is imposing limitations.

No changes were made to the RG or NUREG-2245 as a result of this comment.

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