ML24053A058
| ML24053A058 | |
| Person / Time | |
|---|---|
| Issue date: | 07/30/2024 |
| From: | NRC/NMSS/DREFS/RRPB |
| To: | |
| References | |
| NRC-2018-0289, RIN 3150-AK21, ASME 2021-2022 | |
| Download: ML24053A058 (14) | |
Text
[ML24053A058]
NRC Responses to Public Comments Final Rule:
American Society of Mechanical Engineers 2021-2022 Code Editions NRC-2018-0289; RIN 3150-AK21 U.S. Nuclear Regulatory Commission July 2024
ii Table of Contents Abbreviations and Acronyms................................................................................................................... iii Introduction.............................................................................................................................................. 1 Overview of Public Comments.................................................................................................................. 1 Comment Categorization.......................................................................................................................... 1 Category A: Comments on the ASME OM Code....................................................................................... 2 Category B: Comments on the ASME BPV Code,Section III..................................................................... 2 Category C: Comments on the ASME BPV Code,Section XI..................................................................... 4 Category D: Comments on the NRCs Specific Requests for Comment.................................................... 9
iii Abbreviations and Acronyms ADAMS Agencywide Documents Access and Management System ASME American Society of Mechanical Engineers ASTM ASTM International, formerly known as American Society for Testing and Materials BPV Boiler and Pressure Vessel CFR Code of Federal Regulations FR Federal Register ISI inservice inspection NRC U.S. Nuclear Regulatory Commission OM Operation and Maintenance OWOL optimized weld overlay PSI preservice inspection PWSCC primary water stress corrosion cracking
1 Introduction This comment response document contains summaries of all public comments received by the NRC on the proposed rule American Society of Mechanical Engineers 2021-2022 Code Editions, published in the Federal Register (FR) on August 8, 2023 (88 FR 53384) for public comment with a 75-day public comment period. The NRC published corrections to the proposed rule on August 21, 2023 (88 FR 56780), and August 29, 2023 (88 FR 59471). This comment response document provides the NRCs response to public comments received on the proposed rule.
Overview of Public Comments The NRC received three public comment submissions on the proposed rule.
A comment submission is a communication or document submitted to the NRC by an individual or entity, with one or more individual comments addressing a subject or issue. If a response to a public comment resulted in a change to the rule language or the supporting preamble, the NRCs comment response indicates the change made and where the change occurred.
Comments and the NRCs responses for each category are presented below. The individual comments are identified in the form [XX-YY], where XX represents the comment submittal number in table 1, and YY represents the individual comment number as enumerated in the annotated comment submissions compiled by the NRC (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24053A089).
Individual public comment submissions are available online in the NRC Library at https://www.nrc.gov/reading-rm/adams.html. From this page, the public can access ADAMS, which supplies text and image files of the NRCs public documents. If you do not have access to ADAMS, or if there are problems in accessing the documents located in ADAMS, contact the NRCs Public Document Room reference staff at 1-800-397-4209 or 301-415-4737, or by sending an email to pdr.resource@nrc.gov. In addition, public comments and supporting materials related to this final rule can be found at https://www.regulations.gov by searching for Docket ID NRC-2018-0289.
Table 1: Comment Submissions Comment Submission ID Commenter Affiliation ADAMS Accession No.
Regulations.gov Submission ID 1
Betsy Moenkedick bamoenke@southernco.com Inservice Testing Owners Group ML23269A236 NRC-2018-0289-0003 2
Kathryn Hyam hyamk@asme.org ASME ML23296A087 NRC-2018-0289-0004 3
Jim Cirilli jcirilli@epri.com Electric Power Research Institute ML23299A274 NRC-2018-0289-0005 Comment Categorization This document places each public comment into one of the following categories:
2 Category A: Comments on the American Society of Mechanical Engineers (ASME)
Operation and Maintenance (OM) Code Category B: Comments on the ASME Boiler and Pressure Vessel (BPV) Code,Section III Category C: Comments on the ASME BPV Code,Section XI Category D: Comments responding to the Specific Requests for Comment in the proposed rule.
Within each category, the NRC either reproduced comments as written by the commenter or summarized the comments for conciseness and clarity. At the end of the comment or comment summary, the NRC referenced the source of the comment.
Category A: Comments on the ASME OM Code A-1 10 CFR 50.55a(b)(3)(x)
Comment Summary A-1: A commenter suggested revising the proposed language in Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(b)(3)(x) for OM condition: Class 1 Pressure Relief Valve Sample Expansion to be more concise and to possibly split the owner-established acceptance criteria into two sentences to avoid confusion. The commenter provided this proposed revision:
(x) OM condition: Class 1 Pressure Relief Valve Sample Expansion. When implementing paragraph I-1320(c)(1) in Appendix I, Inservice Testing of Pressure Relief Devices in Water-Cooled Reactor Nuclear Power Plants, of the editions and addenda of the ASME OM Code, incorporated by reference in paragraph (a)(1)(iv) of this section, the requirement for sample expansion of Class 1 Pressure Relief Valves shall be implemented such that for each valve tested for which the as-found set-pressure (first test actuation) exceeds the plus/minus tolerance limit of the Owner-established design set-pressure acceptance criteria of paragraph I-1310(e), two additional valves shall be tested from the same valve group. If the Owner has not established design set-pressure acceptance criteria, then for each valve tested for which the as-found set-pressure (first test actuation) exceeds +/-3 percent of valve nameplate set-pressure, two additional valves shall be tested from the same valve group.
(1-1)
NRC Response: The NRC agrees with this comment and finds that the proposed revision clarifies the condition in 10 CFR 50.55a(b)(3)(x). Therefore, the NRC has revised the final rule as suggested in the comment.
Category B: Comments on the ASME BPV Code,Section III B-1 10 CFR 50.55a(b)(1)(xi)
Comment Summary B-1: A commenter suggested that, in the event of breakage in a specimen that is away from the fusion zone, the test yield plot strength be evaluated before committing to a retest. The commenter suggests a retest be required only if the base material failed at less than the minimum required base material yield strength. The commenter proposed a revision for 10 CFR 50.55a(b)(1)(xi)(B). (2-2)
3 NRC Response: The NRC agrees with this suggestion to reword the first sentence. The NRC finds that the suggested rewording improves clarity. Therefore, the condition is revised as shown below.
(B) Mandatory Appendix XXVI: Second provision. When performing procedure qualification for high speed tensile impact testing of butt fusion joints in accordance with XXVI-2300 or XXVI-4330 of the 2015 through 2021 Editions of BPV Code Section III, breaks in the specimen that are away from the fusion zone require the test plot yield strength to be evaluated to confirm sound base material. If the base material failed (broke) at less than minimum required base material yield strength, a retest is required.
Comment Summary B-2: A commenter recommended that the NRC remove the condition to perform preservice inspections (PSIs) of steam generator tubes in 10 CFR 50.55a(b)(1)(xiii).
The requirement was implemented when the ASME BPV Code,Section XI, included steam generator tubes in the category of items requiring PSI before initial plant startup. This requirement was removed from Section XI, and the commenter recommends that the corresponding requirement in Section III be removed. (2-3)
NRC Response: The NRC disagrees with the comment. These conditions were added in the final rule that incorporated by reference the 2019 Edition of the ASME Code (87 FR 65128;10/27/2022), and the current rulemaking is only extending the applicability of the conditions through the latest edition of the ASME Code. The NRC added the conditions to ensure that newly installed steam generator tubes have an adequate baseline examination, and that flaws be dispositioned using the criteria in the design specifications to ensure the steam generator tubings structural integrity and capability to perform its intended safety function. The conditions were added because the 2019 Edition of Section XI and Section III of the ASME Code removed the requirement for PSI of steam generator tubes. Specifically, in the 2019 Edition of Section III, the provisions of NB-5360 removed the requirements for eddy current preservice examination of installed steam generator tubes and the criteria for evaluating flaws found during the preservice examination. However, since the steam generator tubes are designated as Class 1 reactor coolant pressure boundary and are fabricated in accordance with Section III of the ASME Code, the PSI of these fabricated items is required before initial plant startup. A preservice examination is important because it ensures that the steam generator tubes, which are part of the reactor coolant pressure boundary, are acceptable for initial operation. In addition,Section XI of the ASME Code still requires subsequent inservice examinations, and therefore preservice examination is required to establish the baseline condition of the steam generator tubes, which is essential in assessing the nature of indications found in the tubes during these subsequent inservice examinations. These PSIs must be performed with the objective of finding and characterizing the types of preservice flaws that may be present in the tubes and flaws that may occur during operation.
No revisions were made as a result of this comment.
B-3 General Support Comment Summary B-3: A commenter expressed support for the NRCs proposed revisions for 10 CFR 50.55a(b)(1)(vi) and 10 CFR 50.55a(b)(1)(xiv). (2-1, 2-4)
NRC Response: A response is not required for these comments.
4 Category C: Comments on the ASME BPV Code,Section XI C-1 10 CFR 50.55a(b)(2)(xxvi)
Comment Summary C-1: A commenter recommended that the 10 CFR 50.55a(b)(2)(xxvi) condition be deleted. The commenter stated that the Owners procedures, developed and implemented in accordance with the Owners quality assurance program, are sufficient to ensure that the reassembly of mechanical joints is performed satisfactorily. In addition, the Owners procedures typically contain requirements for verifying leaktightness, so there is little need for the NRC to specify this condition. (2-8)
NRC Response: The NRC disagrees with this comment. This language was developed from public comments received on the proposed rule to incorporate by reference the 2019 Edition and discussed in public meetings on June 25, 2020, August 21, 2020, and May 6, 2021.The NRC position that leaktightness verifications be performed when reassembling mechanical joints is consistent with industry practice and was part of the basis for removing the pressure test requirements. The final rule to incorporate by reference the 2019 Edition replaced the requirement for a Section XI pressure test and VT-2 examination with the requirement for the owner to perform a leak check to the standards of its quality assurance program under Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, to 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, to demonstrate the joints leaktightness. This leak check was part of the NRCs risk determination for removing the pressure test. The NRC notes that besides this 10 CFR 50.55a(b)(2)(xxvi) condition, there is no requirement to perform the leaktightness verifications after mechanical joints are reassembled. Also, the commenter notes that typically owners procedures provide leaktightness verifications, but this typical practice is not a requirement for performing this verification after reassembly of mechanical joints. In addition, the preamble to the final rule to incorporate by reference the 2019 Edition clarified that this would be a requirement, even if not part of the licensees Appendix B program:
To be clear, this condition requires licensees to verify that these mechanical joints are leak tight even under circumstances a licensees program under appendix B to 10 CFR part 50 would not require such verification. However, licensees need not define a new leak test or personnel qualifications; instead, licensees will apply the quality standards of their appendix B programs.
Therefore, this condition will not be changed because it provides the requirement to perform the verification to ensure that the owners procedures are consistent throughout the industry. No changes in the 2021 Edition were presented that would result in the NRC needing to update the current condition.
No revisions were made as a result of this comment.
C-2 10 CFR 50.55a(b)(2)(xxxiv)(A)(2)
Comment Summary C-2: A commenter expressed opposition to this proposed condition.
Owners currently using the 2013 Edition through the 2019 Edition of ASME Section XI are permitted to use Code Case N-513-4 or N-513-5, as approved for use in table 1 of Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 19, issued October 2019, or Revision 20, issued December 2021, at the time this case was listed in their inspection program. The proposed condition, if not modified, would prohibit the continued use of these revisions of Code Case N-513 and would require update to the latest revision approved for use in Regulatory Guide 1.147. This would be in conflict with the guidance
5 in Regulatory Guide 1.147 that contains provisions to allow continued use of a revised or annulled Code Cases through the end of the Owners current inservice inspection (ISI) interval.
The commenter provided a proposed revision of the condition. (2-9)
NRC Response: The NRC agrees with the comment. Licensees are permitted through 10 CFR 50.55a(b)(5) to continue using an earlier version of a Code Case that was already in the licensees ISI program for a prescribed length of time. The NRC has also extended the applicability of using Code Case N-513 in lieu of paragraph U-S1-4.2.1(c) of Appendix U from the 2015 Edition through the 2019 Edition. Because of this change, the dates listed in 10 CFR 50.55a(b)(2)(xxxiv)(A)(2) were deleted as they were redundant to the dates listed in 10 CFR 50.55a(b)(2)(xxxiv)(A).
The NRC revised this condition, in part, to read as follows:
(2) In lieu of the appendix referenced in paragraph U-S1-4.2.1(c) of Appendix U, an approved revision of the ASME BPV Code Case N-513 must be used in accordance with NRC Regulatory Guide 1.147 at the time the case was incorporated into the licensees program.
Comment Summary C-3: A commenter suggested that the NRC-approved equation to index RTTo to the KIa curve was adopted into Article-4200 of the 2017 Edition of Section XI, and therefore, the condition in 10 CFR 50.55a(b)(2)(xxxv) is no longer required. (2-17, 2-18)
NRC Response: The NRC disagrees with the comment. The conditions on the KIa curve are still valid for the 2013 and 2015 Editions of Section XI since the NRC-approved equation was not incorporated until the 2017 Edition. No such conditions are placed on the 2017 Edition or later.
No revisions were made as a result of this comment.
C-4 10 CFR 50.55a(b)(2)(xlv)
Comment Summary C-4: A commenter stated that the new proposed condition in 10 CFR 50.55a(b)(2)(xlv) is unnecessary and should not be retained in the final rule. The IWA-4540(e) requirement contains specific provisions to confirm that leakage is not throughwall leakage that might compromise the containment function. IWA-4540(e) requires that additional testing be performed if leakage is detected during the Appendix J to 10 CFR part 50 test. This additional testing is performed specifically to confirm that there is no leakage through the brazed joints or welds made in the course of the repair/replacement activity. The commenter mentioned that, if the staff believes there is a technical basis to retain this condition, the tests specified in IWA-5242(a) should be allowed for pneumatic tests where applicable, in lieu of the proposed condition to require a VT-2 examination. (2-10)
NRC Response: The NRC disagrees with the comment. The condition is specifying that a VT-2 examination be performed during the Appendix J test to detect leakage. This is not additional testing but is part of the Appendix J test to determine if leakage is present during the walkdown VT-2 examination. If leakage is detected during the Appendix J test with the VT-2 examination, then the additional testing mentioned by the commenter would apply in order to confirm where the leakage is occurring, such as at a welded or brazed joint. In regard to the applicability for pneumatic tests, this condition is only specifying that a VT-2 examination is performed during the Appendix J test.
Therefore, if a pneumatic test is allowed by the ASME Code, then the VT-2 examination requirements shall apply, including the requirements for IWA-5240, Visual Examination.
6 IWA-5242(a) is part of IWA-5240 and therefore would be allowed for pneumatic tests where applicable.
No revisions were made as a result of this comment.
C-5 10 CFR 50.55a(b)(2)(xlvi)
Comment Summary C-5: A commenter recommended deleting the newly proposed condition in 10 CFR 50.55a(b)(2)(xlvi). The commenter stated there are no technical or safety concerns with allowing items to be fabricated at facilities other than that of the owner, provided the work is performed by the owner or owners contracted repair/replacement organization with a quality assurance program that complies with IWA-4142, and that the services of an inspector are used in accordance with the requirements of IWA-4170. (2-11)
NRC Response: The NRC disagrees with the comment. The proposed change in the 2021 Edition of the ASME Code would allow an owner to procure ASME Code,Section III, items with no ASME stamp/certification mark from a repair/replacement organization that does not have an ASME Certificate of Authorization and conducts fabrication activities off site of the owners facility. This contradicts NCA-8330 in ASME Code,Section III, which only allows an item with no ASME stamp/certification mark applied to the item for an organization with an ASME Certificate of Authorization, since the organization with an ASME Certificate of Authorization is required to follow additional controls of the item in NCA-8330 to ASME Code,Section III.IWA-4131 in the 2021 Edition of the ASME Code,Section XI, does not provide controls of these items through completion of installation for an organization that does not have an ASME Certificate of Authorization. In addition, this proposed change does not provide the requirements, such as in NCA-5000 of Section III, for how the Section XI authorized nuclear inservice inspector will inspect the offsite fabrication of an item under Section III of the ASME Code. Without these additional controls or an ASME Certificate of Authorization, the proposed change would not adequately ensure the items ability to perform its safety function.
No revisions were made as a result of this comment.
C-6 10 CFR 50.55a(b)(2)(xlvii)
Comment Summary C-6: A commenter recommended revision of the proposed condition in 10 CFR 50.55a(b)(2)(xlvii). The commenter recommends revising this condition to require stress corrosion crack growth analysis only if the weld overlay material is not resistant to stress corrosion cracking. (2-12)
NRC Response: The NRC disagrees with the comment. The purpose of the condition is to require flaw evaluation of the more resistant material to ensure that the design is sufficient to meet ASME Code requirements. If no growth is claimed, a sufficient technical basis should be included as part of the flaw analysis package.
No revisions were made as a result of this comment.
Comment Summary C-7: A commenter suggested revision of the rule text in 10 CFR 50.55a(g)(4)(iv) to allow owners to use later editions of Section XI without having to obtain NRC approval, provided the later Code edition is incorporated by reference in 10 CFR 50.55a(a) and is used in its entirety, and the applicable conditions in 10 CFR 50.55a(b)(2) are met. (2-13)
7 NRC Response: The NRC agrees with the comment. The NRC received a similar public comment on a proposed rule (88 FR 13717; 3/6/2023) that also affected 10 CFR 50.55a, and on July 17, 2024 issued a final rule (89 FR 58039) that addressed this comment. The NRC is not taking additional action on this matter in this final rule.
No revisions were made as a result of this comment.
C-8 10 CFR 50.55a(g)(6)(ii)(B)
Comment Summary C-8: A commenter recommends deleting the regulation in 10 CFR 50.55a(g)(6)(ii)(B), claiming that the ASME Code,Section XI, no longer contains requirements for submitting program documentation to the NRC, regardless of component class, so there is no need to retain this provision in the regulation. (2-14)
NRC Response: The NRC disagrees with the comment. The licensees are not required to submit to the NRC for approval their containment ISI programs that were developed to satisfy the requirements of Subsection IWE and Subsection IWL with specified conditions, but the NRC does require that the program elements and the required documentation be maintained on site for audit.
No revisions were made as a result of this comment.
C-9 10 CFR 50.55a(g)(6)(ii)(F)(8)
Comment Summary C-9: A commenter recommended that the NRC remove the proposed condition on Code Case N-770-7 inspection items C-2 and F-2 in the final rule. The commenter provided three specific reasons to support this recommendation:
(1)
In the Federal Register notice for the proposed rule, the NRC expressed concern that optimized weld overlays (OWOLs) still structurally rely on 25 percent of the primary water stress corrosion cracking (PWSCC) material of the original buttweld to provide structural integrity for the weld. The commenter stated that this concern was addressed in the approval process of the technical basis for OWOL, MRP-169, Revision 1A, Alternative Design/Analysis Requirements for OWOL Subject to Performance Demonstration Initiative Limitations, issued March 2, 2009. MRP-169 postulates a flaw that is 100 percent through the original weld and applies alternative structural factors that would demonstrate structural integrity for the locations analyzed. Based on the NRCs conclusion from the safety evaluation report, the addition of weld material resistant to PWSCC on top of the original weld provides defense in depth to maintain structural integrity and would preclude reactor coolant pressure boundary leakage if an examination were to miss a weld flaw or size it incorrectly.
(2)
The commenter stated that if a flaw was missed and grew to 100 percent throughwall circumferentially, the OWOL design would maintain structural integrity, and the likelihood of cracks developing after an OWOL is applied is very low. Further, the OWOL is designed to reduce residual stress and generate a zero-stress intensity factor at the crack tip so that an existing flaw will not be likely to grow significantly. If this is not met, then the OWOL design acceptability will be demonstrated through an acceptable crack growth analysis, which feeds into the design life of the OWOL and reinspection interval according to Code Case N-770-7. Code Case N-770-7 also would require a scope expansion if a flaw was found during an ISI examination. For welds applying a stress improvement process, 100 percent of the susceptible material is credited structurally and sampling is allowed for that mitigation method.
8 (3)
The commenter stated that structural integrity is demonstrated in the design of the OWOL to provide a favorable residual stress field to prevent new cracking from developing in the weld, to slow or stop the growth of existing flaws, and to provide a structurally credited material resistant to PWSCC that will preclude reactor coolant pressure boundary leakage. All of the OWOL population will be examined once after application, and OWOL will be added to the sampling population only if the examination is acceptable. The commenter stated that periodic examination of the sampling of OWOL is adequate to monitor the performance of these welds.
Based on the above rationale, the commenter recommends removal of the proposed condition in 10 CFR 50.55a(g)(6)(ii)(F)(8). (2-20)
NRC Response: The NRC disagrees with this comment. As indicated in item (1) above, the NRC recognizes that OWOL is an acceptable mitigation method to address PWSCC. However, the inspection frequency of once per ISI interval (nominally 10 years) was intended to address the uncertainty of the calculated inside surface stress factor and nondestructive examination to identify all flaws of concern. The 10-year reinspection frequency is consistent with and integral to the approval of OWOL mitigation to address PWSCC.
Regarding item (2), a flaw analysis would see a circumferential flaw grow through the susceptible material to the resistant material and then slow by a certain factor of growth.
However, the circumferential flaw would continue to grow along the flaw circumferentially around the weld in the susceptible material. This would limit the effectiveness of leak-before-break. Therefore, the NRC continues to find that each weld should be inspected rather than relying on a 25 percent sample inspection to identify any growing flaws in the weld.
In the future, once the Alloy 690/52/152 Primary Water Stress Corrosion Cracking Expert Panel completes its activities, and if the NRC elects to endorse an Alloy 52/152 crack growth rate factor of improvement, specific flaw evaluation calculations, the NRC may consider optionally including probabilistic analysis with a sample inspection frequency to provide an adequate technical basis for a change. However, at this time, absent additional technical information that is not yet available, the NRC is maintaining the inspection frequency of once per ISI interval.
Regarding item (3), the NRC recognizes that the OWOL design and installation process is an effective repair or mitigation to address PWSCC. However, within that analysis is the NRC requirement that each weld be inspected once per ISI interval. While the NRC agrees that the OWOL will provide a favorable residual stress field if applied appropriately and all the original construction welding process, including repairs, is included in the analysis, the NRC notes that some uncertainty remainshence, the initial examination recommendation. However, PWSCC is an age-affected degradation, and susceptible materials have generally seen additional cracking through component life. Further, volumetric examination of the lower 50 percent of the weld for axial flaws and 75 percent for circumferential flaws cannot be qualified for detection and sizing of PWSCC flaws in OWOLs. Given these uncertainties and the stated concern for effectiveness of leak-before-break in item (2) above, the inspection frequency of once per ISI interval remains appropriate to address these issues. As noted in the response to item (2) above, a probabilistic analysis using an approved improvement factor for the resistant material to show the effectiveness of a sampling inspection frequency may address this issue in the future.
No revisions were made as a result of this comment.
9 C-10 10 CFR 50.55a(g)(4)(ii)
Comment Summary C-10: A commenter stated that there is a typo in 10 CFR 50.55a(g)(4)(ii) related to the 18-month grace period for updating Appendix VIII. The commenter provided the history from multiple NRC final rules, which included discussion of the purpose for updating the dates associated with the grace period in each Code Edition rulemaking. The commenter recommended that the NRC update the final rule effective date referenced in 10 CFR 50.55a(g)(4)(ii) to allow licensees to continue to use the 18-month grace period for updating Appendix VIII. (3-1)
NRC Response: The NRC agrees with the commenter that the 18-month grace period for updating Appendix VIII in accordance with 10 CFR 50.55a(g)(4)(ii) is outdated. The 18-month grace period for implementing updated versions of Appendix VIII to the BPV Code was first mentioned in the final rule ASME Codes and New and Revised ASME Code Cases (76 FR 36232) in 2011 as a result of public comment and applied only to licensees whose next ISI interval was to begin within the next 12 months. These specific licensees were granted an extra 6 months to incorporate the 2007 Edition and 2008 Addenda versions of Appendix VIII in their next interval. In the final rule Incorporation by Reference of ASME Codes and Code Cases (82 FR 32934), the NRC amended 10 CFR 50.55a(g)(4)(ii) to provide the same 18-month grace period for licensees that would need to update their Appendix VIII program.
This change was also the result of public comments received. In the final rule ASME 2015-2017 Code Editions Incorporation by Reference (85 FR 26540), the NRC revised 10 CFR 50.55a(g)(4)(ii) once again to provide licensees with an 18-month grace period to update their Appendix VIII program because the NRC recognizes that updating the Appendix VII program is complex and time consuming and that licensees would face the possibility of needing to maintain multiple Appendix VIII programs if units were to update their ISI programs on different dates. This change was also in response to public comments received. The NRC did not update this provision in the final rule American Society of Mechanical Engineers 2019-2020 Code Editions (87 FR 65128) or in the proposed rule ASME 2021-2022 Code Editions (88 FR 53384), and the commenter states that this provision was not updated in error.
As a result of this comment, the NRC updated 10 CFR 50.55a(g)(4)(ii) to change the commencement of the 18-month grace period for updating to the effective date for this final rule.
C-13 General Support Comment Summary C-11: A commenter expressed support for the NRCs proposed revisions of the provisions in 10 CFR 50.55a(b)(2)(viii), 10 CFR 50.55a(b)(2)(ix),
10 CFR 50.55a(b)(2)(xv), 10 CFR 50.55a(g)(6)(ii)(D)(9), and 10 CFR 50.55a(g)(6)(ii)(F)(1). (2-5, 2-6, 2-7, 2-15, 2-16)
NRC Response: A response is not needed for these comments.
Category D: Comments on the NRCs Specific Requests for Comment The NRC posed the following in the proposed rule:
In the 2021 Edition of the ASME Code,Section XI, ASME removed the IWB-3134 and IWC-3125 requirements for nuclear plant owners to submit analytical evaluations to the regulatory authority having jurisdiction at the plant site. The NRC proposes to condition the 2021 Edition of the ASME Code,Section XI to require that such evaluations be submitted to the NRC, maintaining the status quo for U.S. plants. The analytical evaluation reports provide the NRC
10 with a tool to efficiently inspect and validate flaws identified by a licensee and the activities to address them (e.g., analysis for continued operation or repair/replacement). Furthermore, the reports provide the NRC with valuable operating experience data to monitor degradation trends across the industry to ensure public health and safety. There are other similar reporting requirements in
§50.55a, including §50.55a(b)(2)(xxxii), §50.55a(b)(2)(xliii), and
§50.55a(g)(6)(ii)(F)(6). The NRC is seeking advice and recommendations from the public on the proposed condition and the related requirements to ascertain their perceived value. We are particularly interested in comments and supporting rationale from the public on the following:
(1)
What alternative means are there for the NRC to accomplish the goal of monitoring degradation trends such that the NRC could remove the condition?
(2)
How can the NRC effectively leverage the information provided in flaw evaluations and associated component degradation in a way that is transparent to stakeholders and ensures structural integrity of nuclear components without incurring excessive administrative burden for plant owners?
D-1 10 CFR 50.55a(b)(2)(xliii)(A)
Comment Summary D-1: A commenter stated that the condition is significantly different in two ways from the previous requirements in IWB-3720(c) as follows:
a)
IWB-3720(c) referred to the evaluation procedures themselves and not the analyses performed using the procedures.
b)
The words from IWB-3720(c) said subject to acceptance rather than shall be submitted.
The commenter recommended that the NRC revise 10 CFR 50.55a(b)(2)(xliii)(A) to be consistent with the original wording in IWB-3720(c) in that the procedures must be accepted by the NRC, not the analyses. Further, the commenter recommended that the condition be clarified to state that use of Section XI Nonmandatory Appendix E is acceptable and only procedures other than Appendix E require prior NRC approval. (2-18, 2-19)
NRC Response: The NRC partially agrees with the comment. The NRC agrees that the original wording of IWB-3720(c) was unclear on whether the analysis itself should be submitted and whether review and approval by the regulatory authority was required. The NRC also agrees that Nonmandatory Appendix E provides acceptable procedures for analyzing a violation of reactor pressure vessel operating limits. In an actual case, the NRCs Reactor Oversight Process would be invoked to determine near-term safety impacts of the transient, operability of the reactor coolant system, and authorization of plant startup. In the longer term, however, licensee submission would allow the NRC to review the structural integrity analyses that demonstrate that, even though the excursion caused the plant to violate its pressure-temperature limit and technical specification requirements, the reactor pressure vessel continued to meet its original design requirements and has sufficient integrity for the remaining licensed life.
11 As a result of this comment, the NRC revised the condition in 10 CFR 50.55a(b)(2)(xliii) to make it clear that review and approval of the submission is not required. However, the analysis should be submitted to the NRC for oversight of longer-term safety impacts. This approach provides clarity and information for the NRC to monitor important operating events that impact the integrity of the reactor coolant system. As a result of the changes discussed in comment summary D-2, conditions (B) and (C) in 10 CFR 50.55a(b)(2)(xliii) were deleted and condition (A) was moved to 10 CFR 50.55a(b)(2)(xliii).
D-2 10 CFR 50.55a(b)(2)(xliii)(B) and (C)
Comment Summary D-2: The commenter reviewed the history of the codification of the Master Curve approach to characterizing toughness in ferritic low-alloy steels, including Code Cases N-629 and N-631, Nonmandatory Appendices A and G to Section XI, and Subarticle NB-2330 of Section III. The commenter stated that the NRC did not condition the use of Master Curve concepts in the referenced Code Cases or in Section III. The commenter noted that the NRC voted to approve Code Case N-830-1, which also incorporates Master Curve concepts, at the Section XI Standards Committee. The commenter stated that the NRC began placing conditions on the use of Master Curve concepts in Section XI beginning in 2017. The commenter recommended the following:
a)
Repeal the conditions in 10 CFR 50.55a(b)(2)(xliii) placed on A-4400, A-4200(c) and G-2110(c).
b)
Impose a condition on RTTo in A-4400, A-4200(c) and G-2100(c) that is consistent with the NB-2330 and CC N-830-1 approaches. The condition should require that any value of RTTo used in these articles be determined from a T0 value estimated per [ASTM International, formerly known as American Society for Testing and Materials] ASTM E1921 with twice as much margin adjustment as defined in ASTM E1921 added to the RTTo value.
NRC Response: The NRC agrees with the comment. The NRC clarifies that the condition placed on A-4400 is in 10 CFR 50.55a(b)(2)(xxxvi), rather than 10 CFR 50.55a(b)(2)(xliii). The commenters proposal provides a consistent, technically sound option for dispositioning Master Curve rules in Section XI. As a result of this comment, the NRC revised the condition in 10 CFR 50.55a(b)(2)(xxxvi), deleted the conditions in 10 CFR 50.55a(b)(2)(xliii)(B) and (C), and added a condition, 10 CFR 50.55a(b)(2)(l), that referencesSection III, NB-2331(a)(5), of the 2021 Edition.