ML20099F026
| ML20099F026 | |
| Person / Time | |
|---|---|
| Issue date: | 04/07/2020 |
| From: | Office of Nuclear Regulatory Research |
| To: | |
| michael eudy | |
| Shared Package | |
| ML2009E089 | List: |
| References | |
| DG-1341 RG-1.188 | |
| Download: ML20099F026 (10) | |
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Response to Public Comments on Draft Regulatory Guide (DG)-1341 Standard Format and Content for Applications To Renew Nuclear Power Plant Operating Licenses Proposed Revision 2 of Regulatory Guide (RG) 1.188 On September 17, 2019 the NRC published a notice in the Federal Register (Vol. 84, No. 180, page 48953) that Draft Regulatory Guide, DG-1341 (Proposed Revision 2 of RG 1.188), was available for public comment. The public comment period ended on October 17, 2019. The NRC received comments from the individual and organization listed below. The NRC has combined the comments and NRC staff responses in the following table.
Comments were received from the following:
Samuel Miranda WER89794360 ADAMS Accession No. ML19297G285 Commenter Section of Specific Comments NRC Resolution DG-1341 Sam General (1) On February 25, 2014, the Nuclear Energy Institute The staff disagrees with the comments.
Miranda (NEI) held a forum [1] in Washington, to discuss the prospect of subsequent license renewals (SLRs) that In accordance with the Atomic Energy Act of 1954, as would allow plants to operate for a total lifetime of 80 amended, the NRC regulates commercial nuclear power years. Jennifer Uhle, the NRC's deputy director for in the United States. The criteria for review and approval reactor safety programs, stated that the agency staff plans of subsequent license renewal applications (SLRAs) are to update its regulatory guidance on conducting license contained in the NRCs regulations in 10 CFR Part 54 renewal reviews to prepare for SLR applications. It (Requirements for Renewal of Operating Licenses for appears that DG-1341 is a draft that update. Dr. Uhle Nuclear Power Plants) and the NRCs subsequent also, identified several potential aging effects on reactor license renewal guidance documents, NUREG-2191, pressure vessels, piping, cables and plant concrete Volumes 1 and 2, Generic Aging Lessons Learned for structures that NRC review guidance must consider and Subsequent License Renewal (GALL-SLR) Report industry should address in its applications. We won't (ML17187A031 and ML17187A204) and NUREG-2192, allow subsequent license renewal unless we're assured the Standard Review Plan for Review of Subsequent plants are safe to operate in the extended period." She License Renewal Applications for Nuclear Power also pressed industry to address those issues. She said Plants. (ML17188A158). It is typical for the NRC to Month Year
Commenter Section of Specific Comments NRC Resolution DG-1341 that the NRC staff "is not going to be able to resolve engage all external stakeholders during the development these issues, nor is it our role." of regulatory guidance. At times, NRC guidance will Douglas Walters, vice president for regulatory affairs endorse industry guidance for developing applications at NEI, did not agree that all of Dr. Uhles potential aging when the NRC staff determines that implementation of effects needed to be resolved during the NRCs SLR the industry guidance would permit an applicant to meet reviews. Walters said, "Not everything you need to do for the applicable NRC requirements. In this case, in RG long-term operation is a part of the regulatory process. 1.188, Revision 2, the NRC endorses revision 6 of I don't agree [finding solutions for each aging effect in Nuclear Energy Institute (NEI) 95-10, Industry advance] should be a requirement of getting a new Guideline for Implementing the Requirements of 10 CFR license." Part 54-The License Renewal Rule, and NEI 17-01, In the 5-1/2 years that followed the NEIs forum, NEI Industry Guideline for Implementing the Requirements submitted NEI 17-01 [2], the NRC endorsed it [3], NEI of 10 CFR Part 54 for Subsequent License Renewal.
revised it, and submitted it in its endorsed version [4],
and Dr. Uhle moved from the NRC to the NEI. NEI also Contrary to Mr. Mirandas statement, DG-1341 was not a requested an exemption for the NRCs review and draft of the update to the regulatory guidance referred to endorsement fees. [5] by Jennifer Uhle. Rather, Mrs. Uhle was referring to the It seems that the NEI is leading the NRC in updated guidance for subsequent license renewal that was establishing the criteria for review and approval of issued in July 2017 in NUREG-2191 and NUREG-2192.
SLRAs. Who is the regulator of commercial nuclear These NUREGs were published in draft form for power in the US, the NRC or the NEI? comment and were reviewed by the Advisory Committee Section 1.5 of NEI 17-01 lists four approaches to on Reactor Safeguards (ACRS). The technical bases for resolve open safety issues, the first of which is, If these NUREGs are in NUREG-2221, Technical Bases resolution has been achieved before issuance of a for Changes in the Subsequent License Renewal renewed license, implementation of that resolution could Guidance Documents NUREG-2191 and NUREG-2192, be incorporated within the SLRA. The plant-specific (ML17362A126) and the disposition of public comments implementation information should be provided. is in NUREG-2222, Disposition of Public Comments on This implies that a renewed license could be issued the draft Subsequent License Renewal Guidance without resolution of all the applicable open safety issues. Documents NUREG-2191 and NUREG-2192, What are the potential aging issues that would not be (ML17362A143) both issued December 2017.
resolved during the SLR reviews?
The final rule for 10 CFR Part 54 (60 FR 22491; May 9, 1995), referred to hereafter as the 1995 license renewal rule, focuses on managing the effects of aging rather than identifying each and every aging mechanism. As such, 2
Commenter Section of Specific Comments NRC Resolution DG-1341 the NRC regulatory finding for each license renewal focuses on managing the effects of aging during the period of extended operation.
The GALL-SLR includes (a) new aging management programs for neutron fluence and high voltage insulators, (b) further evaluations for development of new plant-specific programs, as needed, to manage the effects of irradiation on concrete and steel structural components; and (c) revised programmatic criteria for BWR and PWR vessel internals programs to consider higher fluences during the SLR period. The NRC staff continues to interact with industry, the Department of Energy, and the international community to better understand the state of knowledge, ongoing research, and operating experience regarding appropriate technical issues, including: reactor pressure vessel embrittlement; irradiation-assisted stress corrosion cracking of reactor internals; concrete structures and containment degradation; and electrical cable environmental qualification, condition monitoring, and assessment. Additional knowledge gained from these processes will inform future renewals.
Contrary to the commenters suggestion, the NRC will not renew an operating license if an open safety issue exists for the LRA or SLRA. In addition, the staffs final safety evaluation report will address all applicable license renewal issues per the requirements of 10 CFR Part 54.
The staff notes that generic resolutions or methodologies for evaluation of safety issues may be included in topical reports that are referenced by the LRA or SLRA. These topical reports could be used by SLR applicants, as appropriate, for generic solutions or developing plant specific aging solutions, subject to NRC staff approval.
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Commenter Section of Specific Comments NRC Resolution DG-1341 Where no generic solutions exist, plant-specific solutions must be provided by license renewal applicants. These plant-specific solutions could be based on unique plant design configurations, radiation fluence levels. etc. All applicable aging issues within the scope of license renewal are addressed during the staffs review of a license renewal application under 10 CFR Part 54.
The staff did not make any changes to the DG based on these comments.
Sam General (2) Note Walters choice of words. He said new The staff disagrees with this comment.
Miranda license, not renewed license. I believe he was correct.
10 CFR §50.51, Continuation of license states that, In the 1991 license renewal rule statement of Each license will be issued for a fixed period of time to considerations, the Commission explained that renewal be specified in the license but in no case to exceed 40 applications are not initial applications for a facility (i.e.,
years from date of issuance. they are not new license applications) (56 FR 64943, Renewal of operating licenses for nuclear power plants 64970; December 13, 1991) and the extended operation is governed by 10 CFR part 54. Renewal of operating of a nuclear power plant beyond the term of its current licenses for nuclear power plants, according to 10 CFR license is achieved through the issuance of a renewed
§54, would not be license extensions, by amendment. license and not a license amendment (56 FR at 64961-They would be new licenses that expire in 20 years. If 64962). Therefore, the NRC considers the terms these operating licenses were to be renewed, under 10 renewed license and subsequent renewed license to CFR §50, then they would be license amendments that be the appropriate terms for operating licenses renewed authorize extending operations by 20 years, provided that under 10 CFR Part 54.
licensees implement acceptable aging management programs. So, a plants lifetime, operating under a new The term of the renewed license is established in 10 CFR license (10 CFR §54) is only 20 years. A plants lifetime, 54.31 and cannot exceed 20 years (or a maximum of 40 operating under a renewed license (10 CFR §50) is 60 years if 20 years remaining on the existing license is years. (After an SLR, that would be 80 years.) It is included). The period of time for the renewed license confusing to see new licenses labelled renewed term is the sum of the additional amount of time licenses. requested beyond the expiration date of the current operating license plus the remaining number of years on the current operating license.
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Commenter Section of Specific Comments NRC Resolution DG-1341 Once a license renewal or subsequent license renewal is issued per 10 CFR 54.31(c), the renewed license becomes effective immediately, thereby superseding the existing operating (or renewed operating) license. As required by 10 CFR 54.35, each holder of the renewed (or subsequent renewed) license is subject to the regulations in Part 54 as well as the regulations that apply to other holders of operating licenses.
The staff did not make any changes to the DG based on these comments.
Sam General (3) When a plant nears its 40 year design lifetime, it can The staff agrees with the comment that license Miranda be authorized to operate for an additional 20 years via 10 amendments can be accomplished through the 10 CFR CFR §54 renewal process; but this is a new license, not Part 50 license amendment process; however, this does a license amendment. If the plants operators were to not apply to license extension, which is governed by the apply for an extended license expiration date, under the 40-year initial operating license term established by license amendment provisions of 10 CFR §50, then the statute and 10 CFR 50.51; in contrast, the issuance of a operators (or licensees) would have to file a No renewed license is separate from the existing license and Significant Hazards statement, as specified by 10 CFR § license renewal is not accomplished by amending the 50.92, Issuance of amendment. Among other things, existing license.
the No Significant Hazards statement would provide assurance that the proposed license amendment (e.g., a As explained in the preceding NRC comment response, a license extension), would not pose a significant hazard if, license to operate a nuclear power plant beyond the time operation of the facility in accordance with the proposed established in its current operating license is a renewed amendment would not involve a significant increase in license, the term of a nuclear power plant license cannot the probability of an accident previously evaluated. be changed by a license amendment, and the term cannot exceed 20 years (or up to 40 years if the period includes the remaining years on the existing license).
Further, the issues presented in this comment with respect to a No Significant Hazards consideration (NSHC) were addressed by the NRC in its response to a petition 5
Commenter Section of Specific Comments NRC Resolution DG-1341 submitted by the commenter under 10 CFR 2.206 and are not addressed here (see ADAMS Accession No. ML17346A939). In addition, 10 CFR 50.90 NSHC findings do not apply to license renewal because renewed licenses are not amendments.
The staff did not make any changes to the DG based on these comments.
Sam General (4) If the renewal were to be effected as an amendment The staff disagrees with the comment. The staff Miranda to the original 10 CFR §50 license, then it would be very interprets this comment as speculative in nature (i.e.,
difficult, maybe impossible, to show that the amendment what if license renewal were a license amendment).
would not involve a significant increase in the probability of an accident previously evaluated. The As explained in the preceding NRC comment responses, licensing bases of all nuclear plants include analyses or a license to operate a nuclear power plant beyond the evaluations (found in FSARs; usually in Chapter 15), of time established in its current operating license is a various postulated events that are grouped into renewed license, and the term of a nuclear power plant categories, each of which is defined by a range of the license cannot be changed by a license amendment.
expected frequencies of occurrence of its events, and a Therefore, a renewal cannot be effected by an specification of analysis acceptance criteria. One amendment.
category, Condition II or infrequent incidents, is defined as those incidents or events that, may occur 10 CFR Part 54 requires that the current licensing basis during the life of the particular plant. So, an infrequent (CLB) be maintained. A renewed license cant be issued incident would not occur more than once in 40 years. A under 10 CFR 54.29 unless the NRC has reasonable single infrequent incident that is not handled correctly, by assurance that licensed activities will continue to be the plants automatic reactor protection systems, or by its conducted in accordance with the CLB and that any operators, could easily end the plants operating lifetime. changes made to the plants CLB in order to comply with This is what happened at Three Mile Island, in 1979. 10 CFR 54.29(a) are in accord with the AEA and the (That plant had been in operation for only about a year.) Commissions regulations.
If a new license expiration date lengthens the operating lifetime from 40 years to 60 years, then that would Further, the issues presented in this comment with respect significantly increase (e.g., by 50%) the expected to the apparent increase in frequencies of occurrence of frequency of occurrence of infrequent incidents infrequent incidents were addressed by the NRC staff in previously evaluated. An important principle of license its response to a petition submitted by the commenter 6
Commenter Section of Specific Comments NRC Resolution DG-1341 renewal holds that the plant-specific licensing basis must under 10 CFR 2.206 and are not addressed here (see be maintained during the renewal term in the same ADAMS Accession No. ML17346A939).
manner and to the same extent as during the original licensing term. Therefore, in order to maintain the The staff did not make any changes to the DG based on expected frequency of occurrence of infrequent these comments.
incidents at the same value, including the renewed term, it is necessary to make an improvement in plant design and/or operations. The average expected frequency of occurrence of infrequent incidents must become once in 60 years. After an SLR, it must be halved, to once in 80 years.
What is the improvement in plant design and/or operations that is planned, by the licensees, to maintain their plant-specific licensing bases to include their respective renewal terms?
Sam General (5) The first license renewal, effected under 10 CFR The staff agrees that the first license renewal is effected Miranda §54, might not be subject to 10 CFR §50.92, since it under 10 CFR Part 54 and not under 10 CFR 50.92.
would be a new license. However, the SLR would However, the subsequent license renewal is also effected certainly be subject to 10 CFR §50.92, since that would under Part 54 and not under 10 CFR 50.92. The staff be an amendment to the new, extended license. therefore disagrees with the comment that the SLR would Consequently, the SLR should entail some improvement be effectuated under 10 CFR 50.92.
in a plants design and/or operation to maintain the current licensing basis (CLB), particularly with respect to The staff agrees with the comments description of CLB that plants expected frequency of occurrence of as it is consistent with the definition of CLB in 10 CFR infrequent incidents. (The CLB includes the NRC 54.3. In addition, Part 54 requires that the CLB be regulations contained in 10 CFR parts 2, 19, 20, 21, 26, maintained. Renewed licenses cant be issued under 30, 40, 50, 51, 54, 55, 70, 72, 73, 100 and the plant 10 CFR 54.29 unless the NRC has reasonable assurance specific design-basis information defined in 10 CFR 50.2 that licensed activities will continue to be conducted in as documented in the most recent FSAR.) accordance with the CLB and that any changes made to the plants CLB in order to comply with 10 CFR 54.29(a) are in accord with the AEA and the Commissions regulations.
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Commenter Section of Specific Comments NRC Resolution DG-1341 The issues presented in this comment with respect to the apparent increase in frequencies of occurrence of infrequent incidents were addressed by the NRC staff in its response to a petition submitted by the commenter under 10 CFR 2.206 and are not addressed further here (see ADAMS Accession No. ML17346A939).
The staff did not make any changes to the DG based on these comments.
Sam General (6) 10 CFR §54, Requirements for Renewal of The staff disagrees with the comment. As explained in Miranda Operating Licenses for Nuclear Power Plants, governs preceding NRC comment responses, the NRC cannot the issuance of renewed operating licenses. So, the issue a power reactor license for a term that exceeds 40 issuance of renewed operating license is a new license. years, and the Commission determined in the 1991 Would it not be simpler, and less confusing, to issue a license renewal rule that a nuclear power plant licensee license amendment, under 10 CFR §50, which would cannot extend its license term through a license extend the license expiration date, and record a license amendment.
commitment (or condition) to establish and implement an acceptable aging management program? Power upratings The staff did not make any changes to the DG based on are authorized by license amendments to licenses issued these comments.
under 10 CFR §50. Why not license extensions, too?
Sam General (7) Probabilistic risk assessment (PRA) arguments cannot The staff agrees that license renewal under 10 CFR Part Miranda be used to dismiss the occurrence of three infrequent 54 requires plants to maintain their CLB during the incidents, as highly unlikely. This is because 10 CFR §54 period of extended operation. DG-1341 did not discuss requires that plants maintain their current, deterministic the use of PRA; accordingly, the comments regarding licensing bases during the extended terms of operation PRA are beyond the scope of the DG.
that are authorized by their renewed licenses. This is supported by the Statement of Consideration, The The staff did not make any changes to the DG based on Commission reaffirms its previous conclusion (see 56 FR these comments.
64943 - 64956) that PRA techniques are most valuable when they focus the traditional, deterministic-based regulations and support the defense-in depth philosophy.
In this regard, PRA methods and techniques would focus regulations and programs on those items most important 8
Commenter Section of Specific Comments NRC Resolution DG-1341 to safety by eliminating unnecessary conservatism or by supporting additional regulatory requirements. PRA insights would be used to more clearly define a proper safety focus, which may be narrower or may be broader.
In any case, PRA will not be used to justify poor performance in aging management or to reduce regulatory or programmatic requirements to the extent that the implementation of the regulation or program is no longer adequate to credit for monitoring or identifying the effects of aging. --- FR 22468, Vol. 60, No. 88 (May 8, 1995)
Sam General (8) Aging management programs are focused upon The staff agrees with this comment to the extent that Miranda reactor plant systems and components; but they should systems, structures, and components (SSCs) support also account for offsite supporting systems like, for functions within the scope of license renewal per 10 CFR example, emergency power supplies, and cooling water 54.4.
sources. For Peach Bottom, for example, both are found in the Susquehanna River. Exelon Generation Company, 10 CFR Part 54 requires that the CLB be maintained. A LLC (Exelon), applied to the NRC on July 10, 2018 for renewed license cant be issued under 10 CFR 54.29 an SLR for Peach Bottom Units 2 and 3. These reactors, unless the NRC has reasonable assurance that licensed both of which were connected to the grid in 1974, are activities will continue to be conducted in accordance General Electric MK1 BWRs that could operate until with the CLB and that any changes made to the plants 2053 and 2054. Theyre also connected (directly) to the CLB in order to comply with 10 CFR 54.29(a) are in Conowingo Dam, for emergency power. The dam, which accord with the AEA, as amended, and the Commissions was completed in 1928, is now plagued with sediment regulations.
problems. Its FERC license expired on September 1, 2014. Exelons license renewal application (for the next The focus of the license renewal regulations in Part 54 is 50 years, until 2054) is contested. Therefore, Conowingo on appropriate aging management of SSCs and will Dam is currently operating without a license. By 2054, if include evaluations of aging management programs.
licensed, Conowingo Dam would be 126 years old, and Guidance on acceptable aging management programs is the Peach Bottom reactors would be 80 years old. Peach provided in NUREG-2191 and NUREG-2192. If offsite Bottom is offered herein as only one example. What structures or components are credited in the CLB for provisions are contained in DG-1341 to account for performing certain functions that are within the scope of issues like these? Would they be addressed in either the license renewal (e.g., dams that may be credited for 9
Commenter Section of Specific Comments NRC Resolution DG-1341 SLR safety evaluations or the environmental impact providing a source of power during station blackout statements? conditions per 10 CFR 54.4) there is a requirement to include them within the scope of the aging management review. These aging management programs are addressed in the SLR safety evaluations. As an example, Peach Bottom SLR FSER Section 3.5.2.1.2 credits inspections of the Conowingo Dam concrete required by FERC as adequate aging management. Guidance for SLR applicants on how to address situations as described by the commenter has been provided in both the industry guidance in NEI 17-01 being endorsed by this regulatory guide and in NEI 95-10 which was previously endorsed by the NRC and is being updated by NEI 17-01. In addition, acceptance criteria have been established by the NRC and included in the SRP-SLR (NUREG-2192).
Existing guidance already addresses the commenters concerns.
Otherwise, to the extent that the failure of an offsite structure such as a dam, or loss of the power it generates, may affect the safety of a nuclear power plant (under either an initial, renewed, or subsequent renewed license), the licensee is required to consider such potential failure as appropriate, under 10 CFR Part 50.
The staff did not make any changes to the DG based on these comments.
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