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Safety Evaluation Report Related to the License Renewal of Oconee Nuclear Station, Units 1, 2 and 3 (NUREG-1723) |
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Docket Nos. 50-269, 50-270, and 50-287
Duke Energy Corporation
Manuscript Completed: March 2000
Date Published: March 2000
Division of Regulatory Improvement Program
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
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This safety evaluation report (SER) documents the technical review of the Oconee Nuclear Station (ONS), Unit Nos. 1, 2, and 3 license renewal application (LRA) by the U.S. Nuclear Regulatory Commission (NRC) staff. By letter dated July 6, 1998, Duke Energy Corporation (Duke) submitted the license renewal application for the ONS in accordance with Title 10 of the Code of Federal Regulations Part 54 (10 CFR Part 54). Duke is requesting renewal of the operating licenses issued under Section 104 of the Atomic Energy Act of 1954, as amended, for the ONS, Unit Nos. 1, 2, and 3 (license numbers DPR-38, DPR-47, and DPR-55, respectively) for a period of 20 years beyond the current expiration dates: midnight, February 6, 2013, for Unit 1; midnight, October 6, 2013, for Unit 2; and midnight, July 19, 2014, for Unit 3.
The ONS is located in Oconee County in northwestern South Carolina on the shores of Lake Keowee. The three-unit nuclear station was constructed during the period from 1967 to 1974. Each unit consists of a Babcock and Wilcox (B&W) pressurized-water reactor nuclear steam supply system designed to generate 2568 MW thermal, or approximately 860 MW electric.
On the basis of its evaluation of the LRA, the staff concludes that: (1) actions have been identified and have been or will be taken with respect to managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require an aging management review under 10 CFR 54.21(a)(1), and (2) actions have been identified and have been or will be taken with respect to time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c). Accordingly, the staff finds that there is reasonable assurance that the activities authorized by a renewed license will continue to be conducted in accordance with the current licensing basis for the ONS, Unit Nos. 1, 2, and 3 during the period of extended operation.
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This report describes the results of a review by the Nuclear Regulatory Commission (NRC) staff of an application to renew the licenses for the three units of the Oconee Nuclear Station (ONS). Under the Atomic Energy Act (AEA), the NRC issues licenses for commercial power reactors to operate for up to 40 years. The AEA also permits the licenses to be renewed. The NRC established license renewal requirements in the regulations. When those requirements are satisfied, a license can be renewed for up to 20 additional years.
Plant owners are interested in license renewal because they need to know what requirements must be satisfied to permit long-term plant operation. This knowledge helps them to predict the cost of plant operation for long-term energy planning.
The requirements for license renewal are presented in Part 54 of Title 10 to the Code of Federal Regulations (10 CFR Part 54). When those requirements were developed, the NRC concluded that the existing licensing basis and the regulatory process are adequate to maintain safe plant operation, except for the possible effects of aging on passive systems, structures, and components. Therefore, the requirements in 10 CFR Part 54 focus on managing the effects of aging for such passive structures and components as buildings, tanks, and pipes.
The NRC also established requirements for a license renewal environmental report in 10 CFR Part 51. Those requirements establish the scope of a review of environmental impacts, which is one part of the NRC's responsibilities under the National Environmental Policy Act (NEPA). The results of that review are described in Supplement 2 of NUREG-1437, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding the Oconee Nuclear Station."
In a letter dated July 6, 1998, Duke Energy Corporation (Duke) filed an application to renew the licenses for its three-unit Oconee Nuclear Station. Duke requested a 20-year extension in the license term for the three units. The existing licenses expire on midnight, February 6, 2013 (for Unit 1), midnight, October 6, 2013 for (Unit 2), and midnight, July 19, 2014, for (Unit 3). If granted, the renewed licenses would extend to February 6, 2033; October 6, 2033; and July 19, 2034, respectively.
The ONS is located in Oconee County in northwestern South Carolina on the shores of Lake Keowee. The three-unit nuclear station was constructed during the period from 1967 to 1974. Each unit consists of a Babcock and Wilcox (B&W) pressurized-water reactor nuclear steam supply system designed to generate 2568 MW thermal, or approximately 860 MW electric.
In accordance with 10 CFR Part 54, Duke submitted information in its renewal application that identifies all plant systems, structures, and components (SSCs): (1) that are safety-related, (2) whose failure could affect safety-related functions, and (3) that are relied on to demonstrate compliance with the NRC's regulations for fire protection, environmental qualification, pressurized thermal shock, anticipated transients without scram, and station blackout. Duke's application also describes how the effects of aging will be managed in such a way that the intended functions of those structures and components will be maintained for the 20-year period of extended operation. These structures and components include, but are not limited to, the containment building, other safety-related structures, the reactor vessel, the reactor cooling system pressure boundary, steam generators, the pressurizer, piping, pump casings, and valve bodies. The surveillance and maintenance programs for active equipment (for example, motors, diesel generators, air compressors, control rod drives, instruments, cooling fans, and batteries), as well as other aspects of the plant design and licensing basis, are required to be maintained throughout the period of extended operation.
For some passive structures and components within the scope of the renewal evaluation, no additional action was required if Duke demonstrated that the existing programs provide adequate aging management. In other cases, Duke described changes to existing programs and new programs to ensure that applicable aging effects would be adequately managed. These activities include, for example, adding new monitoring programs, increasing inspections, or revising inspection criteria.
Another requirement for license renewal is the identification and updating of time-limited aging analyses. During the design phase for a plant, certain assumptions are made about the length of time the plant will be operated and are incorporated into design calculations for several of the plant's SSCs. These calculations must be shown to be valid for the period of extended operation or be projected to the end of the period of extended operation, or the applicant must demonstrate that the effects of aging on these SSCs will be adequately managed for the period of extended operation.
This report describes the results of the NRC staff's review of the Duke programs to manage aging effects. In this report, we conclude that Duke has demonstrated that aging effects applicable to the required scope of SSCs will be adequately managed for the 20-year period of extended operation. Our evaluation describes the features of the maintenance and inspection programs that we relied on to develop this conclusion. Our evaluation also describes how Duke has resolved our questions about specific aging management concerns. In some cases, our conclusion is based on changes in procedures or actions that will be taken. Duke will update its final safety analysis report, associated with the existing license, to include the changes to the licensing basis reflected in this report, which we relied on to grant a renewed license.
During meetings to gather public comments about the environmental impacts of extending the ONS licenses, we heard several concerns related to plant safety because of aging effects. Interested individuals and groups expressed specific concerns regarding embrittlement of the reactor vessel and other aging effects on plant safety systems and fuel storage facilities. In applicable sections of this report, we describe the particular programs, maintenance activities, and inspection procedures that we have relied on to conclude that those concerns have been adequately addressed.
NRC verified the conclusions in this report by conducting inspections. The scope of the inspections consisted of selected information in the renewal application and information in this report. The inspection results form the basis for a separate recommendation by the administrator of the regional office responsible for the plant.
The bases for the conclusions in this report are also reviewed by the NRC's Advisory Committee on Reactor Safeguards (ACRS). ACRS independently reviews the application and submits its recommendation directly to the Commission; that recommendation is included in the published version of this report (Chapter 5).
In our recommendation for granting a renewed license for the ONS, we have described the programs, maintenance activities, and inspection procedures that we rely on to conclude that there is reasonable assurance that Duke has taken or will take such actions to manage the effects of aging for a 20-year period of extended operation, such that the plant can continue to operate safely.
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This document is a safety evaluation report (SER) on the application for license renewal for the Oconee Nuclear Station (ONS), Unit Nos. 1, 2, and 3, as filed by the applicant, Duke Energy Corporation (Duke or applicant). By a letter dated July 6, 1998, Duke submitted its application to the United States Nuclear Regulatory Commission (NRC) for renewal of the ONS operating licenses for an additional 20 years. The NRC staff prepared this report and reviewed the renewal application for compliance with the requirements of 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants." The NRC license renewal project manager for the ONS is Joseph M. Sebrosky. Mr. Sebrosky may be contacted by calling him at 301-415-1132, or by writing to him at the License Renewal and Standardization Branch, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001.
In its July 6, 1998, submittal, Duke requested renewal of the operating licenses issued under Section 104 of the Atomic Energy Act of 1954, as amended, for ONS, Unit Nos. 1, 2, and 3 (license numbers DPR-38, DPR-47, and DPR-55, respectively) for a period of 20 years beyond the current license expirations of February 6, 2013; October 6, 2013; and July 19, 2014, respectively. The ONS is located in Oconee County in northwestern South Carolina on the shores of Lake Keowee. Each unit consists of a Babcock and Wilcox (B&W) pressurized-water reactor nuclear steam supply system designed to generate 2568 MW thermal, or approximately 860 MW electric. Details concerning the plant and the site are found in the updated Final Safety Analysis Report (UFSAR) for ONS, Unit Nos. 1, 2, and 3.
The license renewal process proceeds along two tracks: a technical review of safety issues and an environmental review. The requirements for these two reviews are stated in NRC regulations 10 CFR Parts 54 and 51, respectively. The safety review for the ONS license renewal is based on Duke's application for license renewal and on the applicant's answers to requests for additional information (RAIs) from the NRC staff. In meetings and docketed correspondence, Duke has also supplemented its answers to the RAIs and submitted answers to the open items identified in the June 16, 1999, version of this SER. The public can review the license renewal application (LRA) and all pertinent information and materials, including the UFSAR mentioned above, at the NRC Public Document Room, 2120 L Street, NW., Washington, D.C. In addition, the application and significant information and material related to the renewal review are available on the NRC Web page at www.nrc.gov.
This SER summarizes the findings of the staff's safety review of the ONS LRA and delineates the scope of the technical details considered in evaluating the safety aspects of its proposed operation for an additional 20 years beyond the term of the current operating license. The staff reviewed the LRA in accordance with the NRC regulations and the guidance presented in the NRC draft "Standard Review Plan (SRP) for the Review of License Renewal Applications for Nuclear Power Plants," dated September 1997.
Chapters 2 through 4 of the SER address the staff's review and evaluation of license renewal issues that have been considered during the review of the application. Chapter 5 contains the report of the Advisory Committee on Reactor Safeguards (ACRS). The conclusions of this report are in Chapter 6.
Appendix A is a chronology of NRC's and Duke's principal correspondence related to the review of the application. Appendix B is a bibliography of the documents used during the course of the review. Appendix C is a list of abbreviations used throughout the report. The NRC staff's principal reviewers and its contractors for this project are listed in Appendix D. Appendix E presents an index of the staff's RAIs and Duke's responses.
In accordance with 10 CFR Part 51, the staff prepared draft and final plant-specific supplements to the generic environmental impact statement (GEIS) that discuss the environmental considerations related to renewing the license for the ONS, Unit Nos. 1, 2, and 3. The draft and final plant-specific supplements to the GEIS were issued separately from this report. Specifically, NUREG-1437 Supplement 2, "Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding the Oconee Nuclear Station" dated December 1999, is the final environmental report for ONS.
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Pursuant to the Atomic Energy Act of 1954, as amended, and NRC regulations, licenses for commercial power reactors to operate are issued for 40 years. These licenses can be renewed for up to 20 additional years. The original 40-year license term was selected on the basis of economic and antitrust considerations--not by technical limitations. However, some individual plant and equipment designs may have been engineered on the basis of an expected 40-year service life.
In 1982, the NRC held a workshop on nuclear power plant aging, in anticipation of the interest in license renewal. That led the NRC to establish a comprehensive program plan for nuclear plant aging research (NPAR). Based on the results of that research, a technical review group concluded that many aging phenomena are readily manageable and do not pose technical issues that would preclude life extension for nuclear power plants.
In 1986, the NRC published a request for comment on a policy statement that would address major policy, technical, and procedural issues related to life extension for nuclear power plants.
In 1991, the NRC published the license renewal rule in 10 CFR Part 54. The NRC participated in industry sponsored demonstration programs to apply the rule to pilot plants and develop experience to establish implementation guidance. To establish a scope of review for license renewal, the rule defined age-related degradation unique to license renewal. However, during the demonstration program, the NRC found that many aging mechanisms occur and are managed during the period of the initial license. In addition, the NRC found that the scope of the review did not allow sufficient credit for existing programs, particularly for the implementation of the maintenance rule, which also manages plant aging phenomena.
As a result, in 1995 the NRC amended the license renewal rule. The amended 10 CFR Part 54 established a regulatory process that is expected to be simpler, more stable, and more predictable than the previous license renewal rule. In particular, 10 CFR Part 54 was clarified to focus on managing the adverse effects of aging rather than on identifying all aging mechanisms. The rule changes were intended to ensure that important systems, structures, and components (SSCs) will continue to perform their intended function in the period of extended operation. In addition, the integrated plant assessment (IPA) process was clarified and simplified to be consistent with the revised focus on passive, long-lived structures and components.
In parallel with these efforts, the NRC pursued a separate rulemaking effort, 10 CFR Part 51, to focus the scope of the review of environmental impacts of license renewal, in fulfilling NRC's responsibilities under the National Environmental Policy Act of 1969 (NEPA).
License renewal requirements for power reactors are based on two key principles:
| (1) | The regulatory process is adequate to ensure that the licensing bases of all currently operating plants provide and maintain an acceptable level of safety, with the possible exception of the detrimental effects of aging on the functionality of certain plant SSCs in the period of extended operation and possibly a few other issues related to safety only during the period of extended operation. |
| (2) | The plant-specific licensing basis must be maintained during the renewal term in the same manner and to the same extent as during the original licensing term. |
In implementing these two principles, the rule in 10 CFR 54.4, defines the scope of license renewal as those plant SSCs (a) that are safety-related, (b) whose failure could affect safety-related functions, and (c) that are relied on to demonstrate compliance with the NRC's regulations for fire protection, environmental qualification, pressurized thermal shock, anticipated transients without scram, and station blackout.
Pursuant to 10 CFR 54.21(a), the applicant must review all SSCs within the scope of the rule to identify structures and components subject to an aging management review (AMR). Structures and components subject to an AMR are those that perform an intended function without a change in configuration or properties and that are not subject to replacement based on qualified life or specified time period. As required by 10 CFR 54.21(a), it must be demonstrated that the effects of aging will be managed in such a way that the intended function or functions of those structures and components will be maintained, consistent with the current licensing basis, for the period of extended operation. Active equipment, however, is considered to be adequately monitored and maintained by existing programs. In other words, the detrimental aging effects that may occur for active equipment are more readily detectable and will be identified and corrected through routine surveillance, performance indicators, and maintenance. The surveillance and maintenance programs for active equipment, as well as other aspects of maintaining the plant design and licensing basis, are required throughout the period of extended operation. 10 CFR 54.21(d) requires that a supplement to the FSAR contain a summary description of the programs and activities for managing the effects of aging.
Another requirement for license renewal is the identification and updating of time-limited aging analyses. During the design phase for a plant, certain assumptions are made about the length of time the plant will be operated and these assumptions are incorporated into design calculations for several of the plant's SSCs. In accordance with 10 CFR 54.21(c)(1), these calculations must be shown to be valid for the period of extended operation or must be projected to the end of the period of extended operation, or the applicant must demonstrate that the effects of aging on these SSCs will be adequately managed for the period of extended operation.
In 1996, the NRC developed and issued draft regulatory guide DG-1047, "Standard Format and Content for Applications To Renew Nuclear Power Plant Operating Licenses." This guide proposes to endorse an implementation guideline prepared by the Nuclear Energy Institute (NEI) as an acceptable method of implementing the license renewal rule. The NEI guideline is NEI 95-10, "Industry Guideline for Implementing the Requirements of 10 CFR Part 54--The License Renewal Rule," which was issued in March 1996. The NRC prepared a draft standard review plan (SRP) for the safety review, which was placed in the Public Document Room in September 1997. The draft regulatory guide will be used, along with the draft SRP, to review applications and to assess technical issue reports involved in license renewal as submitted by industry groups. As experience is gained, NRC will improve the SRP and clarify regulatory guidance.
The staff revised the environmental protection regulations in 10 CFR Part 51 in December 1996 to facilitate the environmental review for license renewal. The staff prepared a "Generic Environmental Impact Statement (GEIS) for License Renewal of Nuclear Plants," NUREG-1437(1), in which it examined the possible environmental impacts associated with renewing licenses of nuclear power plants. For certain types of environmental impacts, the GEIS establishes generic findings that are applicable to all nuclear power plants. These generic findings are identified as Category 1 issues in 10 CFR Part 51, Subpart A, Appendix B. Pursuant to 10 CFR 51.53(c)(3)(i), an applicant for license renewal may incorporate these generic findings in its environmental report. Analyses of those environmental impacts that must be evaluated on a plant-specific basis, Category 2 issues, must be included in the environmental report in accordance with 10 CFR 51.53(c)(3)(ii).
In accordance with NEPA and the requirements of 10 CFR Part 51, the NRC performed a plant-specific review of the environmental impacts of license renewal, including whether there was new and significant information not considered in the GEIS. A public meeting was held on October 19, 1998, near the ONS as part of the NRC's scoping process to identify environmental issues specific to the plant. Results of the environmental review and a preliminary recommendation with respect to the license renewal action were documented in NRC's draft plant-specific Supplement 2 to the GEIS, which NRC issued on May 20, 1999. During the 75-day comment period that followed, another public meeting was held near the site on July 8, 1999, at which the staff described the results of the NRC environmental review and answered questions related to it in order to provide members of the public with information to assist them in formulating any comments they might have regarding the review. On December 9, 1999, the staff issued the final version of Supplement 2 to the GEIS on the ONS, in which it presented its final environmental analysis that considers and weighs the environmental effects of the license renewal, and alternatives available for avoiding adverse environmental effects. The staff considered and addressed the comments that were received during the comment period.
Based on (1) the analysis and findings in the "Generic Environmental Impact Statement for License Renewal of Nuclear Power Plants," NUREG-1437; (2) the Environmental Report submitted by Duke; (3) consultation with other Federal, State, and local agencies; (4) its own independent review; and (5) its consideration of public comments, the staff recommended, in Supplement 2 to NUREG-1437 that the Commission determine that the adverse environmental impacts of license renewal for the ONS Units Nos. 1, 2, and 3 are not so great that preserving the option of license renewal for energy planning decisionmakers would be unreasonable.
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The requirements for renewing operating licenses for nuclear power plants are described in 10 CFR Part 54. The staff performed its technical review of the ONS application for license renewal in accordance with Commission guidance and the requirements of 10 CFR 54.19, 54.21, 54.22, 54.23, and 54.25. The standards for renewing a license are contained in 10 CFR 54.29. This SER describes the results of the staff's technical review.
In 10 CFR 54.19(a), the Commission requires a license renewal applicant to submit general information. Duke submitted this general information in Enclosure 1 to its July 6, 1998, submittal letter regarding the application for renewed operating licenses for the ONS, Unit Nos. 1, 2, and 3. In that enclosure the staff finds that Duke submitted the information required by 10 CFR 54.19(a).
In 10 CFR 54.19(b), the Commission requires that LRAs include "conforming changes to the standard indemnity agreement, 10 CFR 140.92, Appendix B, to account for the expiration term of the proposed renewed license." Duke stated the following in its renewal application regarding this issue:
The current indemnity agreement for Oconee states in Article VII that the agreement shall terminate at the time of expiration of that license specified in Item 3 of the Attachment to the agreement. Item 3 of the Attachment to the indemnity agreement, as revised by Amendment No. 9, lists six license numbers. Duke requested that conforming changes be made to Article VII of the indemnity agreement, and/or Item 3 of the Attachment to that agreement, specifying the extension of agreement until the expiration dates of the renewed Oconee operating licenses as set forth in this Application. Thus, license number DPR-38 would be extended to expire at midnight, February 6, 2033; DPR-47 would be extended to expire at midnight, October 6, 2033; and DPR-55 would be extended to expire at midnight, July 19, 2034. In addition, should the license numbers be changed upon issuance of the renewed licenses, Duke requests that conforming changes be made to Item 3 of the Attachment, and any other section of the indemnity agreement as appropriate.
The staff intends to maintain the license numbers on issuance of the renewed license. Therefore, there is no need to make conforming changes to the indemnity agreement, and the requirements of 10 CFR 54.19(b) have been met.
In 10 CFR 54.21, the Commission requires that each application for a renewal license for a nuclear facility must contain the following information: (a) an integrated plant assessment (IPA), (b) current licensing basis (CLB) changes during NRC review of the application, (c) an evaluation of time-limited aging analyses (TLAAs), and (d) a final safety analysis report (FSAR) supplement. Duke submitted the information to address the license renewal requirements of 10 CFR 54.21(a) and (c) in Exhibit A to the LRA of July 6, 1998. Exhibit A is titled "Oconee Nuclear Station, License Renewal--Technical Information, OLRP-1001." Duke submitted the information to address the license renewal requirements of 10 CFR 54.21(b) in a letter dated September 30, 1999. Duke submitted the information to address the license renewal requirements of 10 CFR 54.21(d) in Exhibit B of its LRA.
In 10 CFR 54.22, the Commission states requirements regarding technical specifications. Duke addressed the requirements of 10 CFR 54.22 in Exhibit C of its LRA.
The staff evaluated the technical information required by 10 CFR 54.21 and 10 CFR 54.22 in accordance with the NRC's regulations and the guidance presented the draft SRP titled "Review of License Renewal Applications for Nuclear Power Plants," which was published in September 1997. The staff's evaluation of the LRA in accordance with 10 CFR 54.21 and 54.22 appears in Chapters 2, 3, and 4 of this SER.
The staff's evaluation of the environmental information required by 10 CFR 54.23 can be found in the draft and final plant-specific supplements to the GEIS (NUREG-1437, Supplement 2), that state the considerations related to renewing the license for ONS, Unit Nos. 1, 2, and 3.
The report by the Advisory Committee on Reactor Safeguards required by 10 CFR 54.25 is in Chapter 5 of this SER. The findings required by 10 CFR 54.29 are in Chapter 6 of this report.
In accordance with 10 CFR 54.17(e), Duke also incorporated by reference several Babcock and Wilcox Owners Group topical reports into the ONS LRA. The topical reports demonstrate generically that the aging effects for reactor coolant system components are adequately managed for the period of extended operation under a renewed license. Specifically, Duke incorporated the following topical reports into its application:
The staff has issued separate safety evaluations for these topical reports. Specifically, the staff issued the final safety evaluations for the following topical reports: BAW-2243 on March 21, 1996; BAW-2244 on August 18, 1997; BAW-2241P on February 18, 1999; BAW-2251 on April 26, 1999; and BAW-2248 on December 9, 1999. In accordance with procedures established in NUREG-0390, "Topical Report Review Status," the staff requested that the Babcock and Wilcox Owners Group publish accepted versions of the reports. The accepted version incorporates the transmittal letter and the staff's safety evaluation between the title page and the abstract. The accepted versions includes an -A (designating accepted) following the report identification symbol.
Each safety evaluation for the topical reports is intended to be a standalone document. An applicant incorporating the topical reports by reference into its LRA must ensure that the conditions of approval contained in the safety evaluations are met. The staff's evaluation of how the topical reports were incorporated into the application is found in Section 3.4 of this SER.
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As a result of its initial review of the LRA for the ONS, including the additional information submitted to the NRC, the staff identified a number of open issues and confirmatory items when this report was issued in June 1999. That report was revised to describe in each applicable section the manner by which those matters have been resolved.
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10 CFR 54.21, "Contents of application -- technical information," requires, in part, that each application for license renewal contains an integrated plant assessment (IPA) that identifies and lists those systems, structures, and components (SSCs) satisfying the criteria in 10 CFR 54.4(a)(1), (a)(2), and (a)(3) that are subject to an aging management review (AMR). 10 CFR 54.4, "Scope," defines the criteria for inclusion of SSCs within the scope of 10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants."
The Oconee Nuclear Station (ONS) IPA was developed along traditional engineering disciplines, that is, mechanical, civil/structural, and electrical. The methodology used by the applicant to identify structures and mechanical systems at the ONS subject to an AMR is generally consistent with the industry guidance in an Nuclear Energy Institute (NEI) document NEI 95-10, "Industry Guideline for Implementing the Requirements of 10 CFR Part 54 -- The License Renewal Rule." However, the applicant developed a process specific to the ONS for identifying electrical components.
Exhibit A, "License Renewal -- Technical Information (OLRP-1001)," to the ONS license renewal application (LRA) contains the technical information required by 10 CFR 54.21(a) and (c), including the methodology used to identify the SSCs at the ONS that are within the scope of license renewal. Exhibit A, Section 2.2, "Identification of Systems, Structures, and Components Within the Scope of License Renewal," describes the process used by the applicant to satisfy the criteria contained in 10 CFR 54.4(a)(1), (a)(2), and (a)(3) for structures and mechanical systems at the ONS. The methodology used to identify electrical components within the scope of license renewal is described in Section 2.6, "Electrical Components," of Exhibit A.
Additionally, Section 2.3.1, "Description of the Process to Identify Reactor Building (Containment) Structural Components"; Section 2.4.1, "Description of the Process to Identify Reactor Coolant System Components and Class 1 Component Supports Subject to Aging Management Review"; Section 2.5.1, "Process Used to Identify Mechanical Components Subject to Aging Management Review"; Section 2.5.2, "Detailed Process Descriptions"; Section 2.6.1, "Description of the Process to Identify Electrical Components Subject to Aging Management Review"; and Section 2.7.1, "Description of the Process to Identify Structural Components Subject to Aging Management Review," contain amplifying information on the process used by the applicant to satisfy the requirements of 10 CFR 54.21(a)(1) and (a)(2) for the ONS structural, mechanical, and electrical components that are subject to an AMR for license renewal.
2.1.2.1 Technical Information for Identifying Systems, Structures, and Components Within the Scope of License Renewal
In OLRP-1001, Subsection 2.2, "Identification of Systems, Structures, and Components Within the Scope of License Renewal" of Exhibit A of the LRA the applicant states the following:
| Because the ONS was licensed before terms such as 'safety-related' were more precisely defined by the NRC, a list of the ONS safety-related SSCs, in and of itself, will not meet the intent of 10 CFR 54.4(a)(1). Because the criteria in 10 CFR 54.4(a)(1) are the scoping criteria of many modern-day, regulatory-required programs, ONS conducted a design study that validated all functions required for the successful mitigation of ONS design-basis events and identified the systems and components relied upon to complete those functions. The individual design-basis event mitigation calculations produced as a result of the study contain a list of the system functions required to successfully mitigate each event. The applicant determined that the systems that perform these functions are within the scope of license renewal. |
During an audit of the ONS license renewal scoping and screening process conducted by the NRC staff on October 27 through 30, 1998, at Duke Energy Corporation's offices in Charlotte, N.C., the audit team learned that the "design study" identified in Subsection 2.2.1.1 and the Oconee Safety-Related Designation Clarification (OSRDC) project developed in response to GL 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events" (July 1983) was one and the same. Specifically, in its November 4, 1983, response to GL 83-28, as supplemented by letters dated January 17, 1984, and June 9, 1987, the applicant described the scope of the ONS operational QA program for safety-related equipment classification. The NRC staff approved the scope of the ONS operational QA program in a safety evaluation dated November 4, 1987.
In a supplemental response to GL 83-28, dated April 12, 1995, the applicant submitted amplifying information on the ONS QA-1 licensing basis, and on information given to the NRC Region II staff during a February 6, 1995, meeting. The QA-1 designation originally applied to ONS SSCs that were relied upon to mitigate a large-break loss-of-coolant accident (LBLOCA) coincident with a loss of offsite power (LOOP) event; the QA-1 designation did not encompass all SSCs which are relied upon to remain functional during and following design-basis events (DBEs) as defined under 10 CFR 54.4(a)(1).
In Attachment 3 to the April 12, 1995, letter, "Supplemental Response to Subpart 1 of Section 2.2.1 of GL 83-28 General Criteria for Classifying QA-1 SSCs," the applicant stated that the list of additional QA-1 SSCs would be developed through the OSRDC project by July 10, 1995. Also, in Attachment 4, "Oconee Licensing Position on Non QA-1 SSCs Which Are Used to Mitigate Accidents," the applicant committed to developing a new QA classification (QA-5) so that these SSCs can be identified for testing and maintenance under selected Appendix B [to 10 CFR Part 50] criteria without procuring the SSCs per Appendix B.
On this basis, and by letter dated December 1, 1998, the staff requested that the applicant do the following:
In its February 17, 1999, response to the staff's request for additional information (RAI 2.2-6), the applicant clarified the role of the OSRDC project in the ONS license renewal process. Subsequent to the applicant's response to the staff's RAI, the staff met with the applicant on March 11, 1999, to obtain clarification and additional insights into the methodology used by the applicant to meet the requirements of 10 CFR 54.4 for identifying the SSCs within the scope of the rule. As a result of the meeting on March 11, 1999, the applicant submitted additional information and clarifications in a letter dated March 18, 1999. In a May 11, 1999, meeting, which is documented in a meeting summary dated May 19, 1999, Duke met with the staff to further discuss the DBEs used by the applicant to determine the safety-related SSCs required by the scoping criteria under 10 CFR 54.4(a)(1). During this meeting Duke agreed to supplement its response to the staff's RAI 2.2-6, to include a description of the process used to identify events for Oconee license renewal scoping consistent with the presentation that was given to the staff.
2.1.2.2 Technical Information for the Structures and Components Subject to an Aging Management Review
During the audit of October 27 through 30, 1998, members of the NRC staff visited the Duke Energy Corporate Office in Charlotte, NC, to review the license renewal scoping and screening methodology and justification for the ONS LRA. The audit team reviewed the site-specific specifications used to identify the structures and components (SCs) subject to an AMR from those identified as being within the scope of the rule. The staff also reviewed other supporting documentation and interviewed applicant staff members as part of its evaluation of the applicant's process for identifying those SCs subject to an AMR. The staff also performed an inspection of the applicant's scoping process from April 26, 1999, to April 30, 1999, and performed another audit of the applicant's scoping methodology during the week of August 16, 1999. In addition, there were numerous public meetings, telecommunications, and docketed correspondence, including RAIs and RAI responses between the staff and the applicant to address specific scoping concerns as discussed below.
Mechanical Components Review
During the week of October 27, 1998, the site-visit team reviewed the methodology used by the applicant to identify and list the mechanical components subject to an AMR, as well as the applicant's technical justification for this methodology. The team also examined the applicant's results from the implementation of this methodology by reviewing an overview of the mechanical systems identified as being within the scope of license renewal, a sample of evaluation boundaries drawn within those systems, the resulting components determined to be within the scope of the rule, the corresponding component-level intended functions, and the resulting list of mechanical components subject to an AMR.
The site-visit team reviewed the methodology described in the LRA, Subsection 2.4 and 2.5, entitled "Reactor Coolant System Mechanical Components and Class 1 Component Supports," and "Mechanical System Components." The site-visit team also reviewed a number of on-site engineering documents not docketed, including Oconee site specification OSS-0274.00-00-0001, "Oconee Mechanical System Scoping for License Renewal"; OSS-274.00-00-0002, "Oconee Mechanical Component Screening for License Renewal"; appropriate portions of the ONS updated final safety analysis report (UFSAR); the ONS flow diagrams that contain the color-coded evaluation boundaries for the systems identified as being within the scope of license renewal; and the mechanical component commodity-type menus developed by the applicant to identify the SCs that are required to be subject to an AMR under 10 CFR 54.21(a)(1)(i) and (a)(1)(ii). The site-visit team found the applicant's process consistent with the scoping process described in the "Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule" (NEI 95-10, Revision 0), and adequate for the purpose of determining the mechanical components requiring an AMR. However, the staff needed to better understand the OSRDC process, which was used to determine the applicable design-basis events defined in the applicant's current licensing basis (CLB), to ensure that all the mechanical systems, as required by 10 CFR 54.4(a)(1) and (a)(2), were identified as being within the scope of license renewal.
As a result of the information reviewed, the staff issued RAI 2.2-6 relating to the design-basis events used to determine the mechanical systems within the scope of the rule and the resulting components requiring an AMR. Duke provided an initial response to the staff's request for additional information in a letter dated February 17, 1999. The RAI response was followed by a technical meeting on March 11, 1999, and a supplemental response from Duke dated March 18, 1999.
During the week of April 26, 1999, the staff performed an inspection of the results of the applicant's scoping activities including the scoping of mechanical SSCs. The results of this inspection were documented in NRC inspection report number IR99-011. On May 11, 1999, the applicant met with NRC staff to further discuss concerns with its list of design-basis events that were used to scope mechanical SSCs for the purpose of license renewal. That meeting led to additional information provided by the applicant in a letter dated June 22, 1999. The staff performed a third site visit during the week of August 16, 1999, to audit additional design documentation relative to the scoping methodology and the design-basis events used in the applicant's scoping process. As a result of a number of additional meetings and docketed correspondence between the staff and the applicant, Duke submitted its final response on November 30, 1999. The response discussed specific events and their inclusion as scoping events for the purpose of license renewal.
Structures and Structural Component Review
The site-visit team reviewed the methodology used by Duke to identify and list the structural components subject to an AMR, as well as the applicant's technical justification for this methodology. The team also examined the applicant's results from the implementation of this methodology by reviewing the structural components identified as being within the scope of license renewal, the corresponding structural-level intended functions, and the resulting list of structural components subject to an AMR.
The site-visit team reviewed the methodology described in the LRA, Subsection 2.3 and 2.7, entitled "Reactor Building Structural (Containment) Components," and "Structures and Structural Components." The site-visit team also reviewed a number of on-site engineering documents including Oconee site specification OSS-0274.00-00-0007, "Oconee Structures and Structural Component Aging Management Review," a number of other ONS specifications relating to structural classifications, appropriate portions of the ONS UFSAR, ONS General Arrangement Drawings, ONS Commodities and Facilities Drawings, and Quality Standards Manual NSD 307. As a result of the information reviewed, the staff issued RAI 2.6.7-1 requesting additional information relating to the validations of the structures determined not to be within the scope of license renewal. Duke provided a response to the staff's request for additional information in a letter dated February 17, 1999. The RAI response was followed by a technical meeting on March 11, 1999. During the week of April 26, 1999, the staff also performed an inspection of the results of the applicant's overall scoping activities including the scoping of structural SSCs.
Electrical Components Review
The site-visit team reviewed the methodology used by Duke to identify and list the electrical components subject to an AMR, as well as the applicant's technical justification for the identification process. The team examined the applicant's results from the implementation of this methodology by reviewing the list of electrical components subject to an AMR.
The site-visit team reviewed the methodology described in the LRA, Subsection 2.6, entitled "Electrical Components." The site-visit team also reviewed a number of on-site engineering documents including Oconee site specification OSS-0274.00-00-0006, "Oconee Electrical Component Aging Management Review for License Renewal," appropriate portions of the ONS UFSAR, ONS Electrical Drawings, and NEI 95-10, Revision 0. The site-visit team found the applicant's process to be significantly different from the scoping process described in the industry guideline, NEI 95-10, and determined that additional information was needed for the staff to adequately assess the applicant's process for scoping electrical SSCs.
As a result of the information reviewed, the staff issued RAI 2.6-1 requesting additional clarification of the applicant's scoping process. Duke provided an initial response to the staff's request for additional information in a letter dated February 17, 1999. The RAI response was followed by a technical meeting on March 11, 1999, and a supplemental response from Duke dated March 18, 1999. As previously noted, during the week of April 26, 1999, the staff also performed an inspection of the results of the applicant's overall scoping activities including the scoping of electrical SSCs.
In Section 2.2, "Identification of Systems, Structures, and Components Within the Scope of License Renewal," of Exhibit A of the LRA, the applicant described the methodology used to identify the mechanical components that are within the scope of license renewal and subject to an AMR. The mechanical components included within the scope of license renewal and subject to an aging management review are described in Section 2.4, "Reactor Coolant System Mechanical Components and Class 1 Component Supports," and Section 2.5, "Mechanical System Components" of Exhibit A of the LRA. The structures included within the scope of license renewal and subject to an aging management review are described in Section 2.3, "Reactor Building (Containment) Structural Components," and Section 2.7, "Structures and Structural Components," of Exhibit A of the LRA. The electrical components included within the scope of license renewal and subject to an aging management review are described in Section 2.6, "Electrical Components," of Exhibit A of the LRA.
2.1.3.1 Evaluation of the Methodology for Identifying Systems, Structures and Components Within the Scope of License Renewal
As indicated above, the applicant stated in its LRA that ONS conducted a design study that was used to validate all the functions required for the successful mitigation of ONS design-basis events and identified the SCs relied upon to complete those functions. On October 27 through 30, 1998, members of the NRC staff visited the Duke Energy Corporate Office in Charlotte, NC, to review the license renewal scoping and screening methodology and justification presented in the ONS LRA. As a result of that review, the staff confirmed that the applicant relied on a design study to identify the SSCs that are needed to satisfy the requirements of 10 CFR 54.4(a).
The site-visit team discussed this design study and the process used to identify the SSCs within the scope of the rule. The basic process, as described by the applicant, involved identifying all the SSCs that met the "safety-related criteria" under 10 CFR 54.4(a)(1). Evaluation boundaries were established for the portions of those systems and structures required to perform the system functions that satisfied the specified criteria. In addition, the applicant stated that it had reviewed the non safety-related SSCs whose failure could prevent the successful completion of the safety functions identified from the review of the safety-related criteria under 10 CFR 54.4(a)(2). Again, evaluation boundaries were established for the portions of those non safety-related systems and structures. The components within those evaluation boundaries that were not already identified were added to the scope of license renewal.
The team found the results of the applicant's design study and subsequent scoping activities to be a reasonable approach for identifying a supplemental list of SSCs to complement the applicant's list of QA-1 SSCs required by the scoping criteria under 10 CFR 54.4 (a)(1) and (a)(2). However, the team concluded that the design study used by the applicant to meet the scoping requirements for license renewal was not fully described in the LRA. Therefore, the staff submitted a request for additional information to obtain the necessary information.
In its February 17, 1999, response to the staff's request for additional information (RAI 2.2-6), the applicant clarified the role of the OSRDC study in the Oconee license renewal scoping process. Specifically, the applicant stated that the "design study" in Exhibit A of the LRA refers only to the second initiative of the OSRDC project. The purpose of the first initiative of the project, identified as a commitment associated with the applicant's response to GL 83-28, was to clarify the ONS QA-1 licensing basis by developing a list of all QA-1 SSCs at ONS.
The purpose of the second initiative of the OSRDC study was to clarify ONS's licensing basis with respect to design-basis-event mitigation requirements, that is, to identify non-QA-1 SSCs credited with accident mitigation functions at ONS and those SSCs whose failure could prevent satisfactory accomplishment of any of the applicable accident mitigation functions. The third and fourth initiatives of the OSRDC project involved identifying and implementing an "augmented" QA (QA-5) program for those SSCs identified as a result of the second initiative and were not relevant to the license renewal scoping process.
Subsequent to the applicant's response to the staff's RAI, the staff met with the applicant on March 11, 1999, to obtain clarification and additional insights into the methodology used by the applicant to justify the scoping results. Specifically, the staff requested that the applicant describe its methodology for identifying the SSCs within the scope of 10 CFR 54.4(a)(1) and (a)(2) as it applies to design-basis events defined under 10 CFR 50.49(b)(1).
During the March 11, 1999, meeting, the discussion focused on which ONS design-basis events (DBEs) were considered in the ONS license renewal scoping process. Specifically, the staff was interested in how the applicant complied with the requirements of 10 CFR 54.4(a)(1) and with the definition of DBEs in 50.49(b)(1). The applicant stated its position that the set of DBEs contained in Chapter 15 of the ONS Updated Final Safety Analysis Report (UFSAR) complies with the requirements of 10 CFR 54.4(a)(1) and meets the definition in 10 CFR 50.49(b)(1). The applicant also stated that in order to be conservative, it considered an additional set of events based on plant-specific insights.
In a letter dated March 18, 1999, the applicant submitted additional information and clarifications as a result of the meeting on March 11, 1999. Specifically, the applicant: (1) amended its original response to RAI 2.2-6 to provide additional clarification in accordance with discussions held during the meeting, (2) amended its response to RAI 2.6-1 to clarify the electrical scoping description and to indicate how the results were validated, and (3) amended its response to RAI 2.6.7-1 to indicate how the validation of structural results was performed.
In a May 11, 1999, meeting, which is documented in a summary dated May 19, 1999, Duke stated that the license renewal "scoping events" included UFSAR Chapter 15 events, natural phenomena criteria, post-Three Mile Island emergency feedwater design basis scenarios, and turbine building floods mitigated by the standby shutdown facility. Duke considered a total of 26 events when initially scoping to comply with 10 CFR 54.4(a)(1) and 10 CFR 54.4(a)(2). Duke also stated that it reviewed an additional 32 events for possible inclusion into the set of scoping events. Duke determined that none of the additional 32 events needed to be considered for purposes of scoping in accordance with 10 CFR 54.4 (a)(1) and 10 CFR 54.4(a)(2). Because of the narrow definition of DBEs used by the applicant, the staff was concerned that the applicant may have overlooked some SCs needed to prevent or mitigate any of the additional 32 events that might have been identified if the applicant used the broader 10 CFR 50.49(b)(1) view of a DBE. Therefore, the applicant was then asked to take the following actions:
The staff reviewed the applicant's response to Open Item 2.1.3.1-1 provided in a letter dated June 22, 1999, and performed an audit of on-site information during August 16-18, 1999. The staff then performed a review of the 32 events, that were originally considered by the applicant but determined not to be DBEs, against the applicant's UFSAR, license conditions, the applicable regulations, Commission orders, and exemptions that are in effect and that define the applicant's design requirements. As a results of these activities, the staff identified 10 events that the staff believed needed additional consideration under the license renewal scoping criteria, 10 CFR 54.4(a) of the rule. The applicant was asked to reevaluate these 10 events for potential SCs that needed to be included within the scope of license renewal. In response to this request the applicant identified seven additional events that needed to be considered for scoping under 10 CFR 54.4(a)(1) and (a)(2). The results of the applicant's review were as follows:
As a result of this review, the applicant did not identify any additional SSCs associated with these ten events, that needed to be added to the scope of license renewal, and therefore, did not add any addition SCs to the list of SCs subject to an AMR.
On the basis of the staff's reviews and the applicant's actions described above, the staff found that there is reasonable assurance that the applicant has considered the necessary DBEs in the implementation of its scoping methodology used to identify the SSCs required by the scoping criteria under 10 CFR 54.4(a)(1) and (a)(2). Open Item 2.1.3.1-1 is closed.
2.1.3.2 Evaluation of Methodology for Identifying Structures and Components Subject to an Aging Management Review
Mechanical Components
The methodology used by the applicant for identifying mechanical component within the scope of the rule included the following steps: identifying all systems and their intended functions that are relied upon to remain functional during and following the design-basis events for which the plant must be designed; identifying all the systems and intended functions whose failure could prevent satisfactory accomplishment of any of the functions identified under 10 CFR 54.4(a)(1); identifying all those systems and intended functions necessary to demonstrate compliance with the regulated events identified under 10 CFR 54.4(a)(3); and identifying all other mechanical systems or portions of systems that contain safety-related and seismically designed components.
The process used by the applicant to identify the mechanical components requiring an AMR included a set of highlighted ONS flow diagrams that were used to define the evaluation boundaries of the license renewal-related equipment. These highlighted drawings identified the flow paths required to be functional during and following design-basis events, and the components necessary for each system to accomplish its intended function(s). Interfacing flow paths, which share a common pressure boundary with the principal path, or non safety-related flow paths whose failure could prevent satisfactory accomplishment of any of the safety-related functions under 10 CFR 54.4(a) were included. The highlighted flow diagrams were color-coded to distinguish between Class 1 and non-Class 1 seismic piping.
In Exhibit A, Section 2.5.2, "Detailed Process Description," the applicant described the process to scope and screen mechanical components within the scope of the rule and subject to an AMR. However, details regarding this methodology that would give the staff an understanding about how the requirements of 10 CFR 54.21 are being met were not provided. In RAI 2.5.2-1, the staff asked the applicant to provide a brief narrative that explained how the screening of mechanical components within the scope of license renewal was performed. In its response to this RAI, Duke stated the following:
| The mechanical component screening is consistent with the guidance provided in NEI 95-10, Rev. 0. Components subject to an AMR are those that are "passive" and "long-lived." A menu of every mechanical component type installed at ONS was developed, going beyond the list of components in NEI 95-10. Using the "passive" and "long-lived" guidance, a determination was made for each of those mechanical component types. The components within the evaluation boundaries shown on the license renewal flow diagrams were "driven" through the menu to determine if they were subject to an AMR. From this exercise, a list of components subject to an AMR was developed. |
The staff notes that the mechanical components subject to an AMR resulting from the applicant's process described in Section 2.5.2 of the LRA, and the mechanical screening process discussed in the response to RAI 2.5.2-1, are provided in Sections 2.5.3 through 2.5.14 of the LRA.
After the evaluation boundaries were established, the process is designed to identify those components within the evaluation boundaries that require an AMR primarily by eliminating those components excluded under 10 CFR 54.21(a)(1)(i). The applicant also identified the component-level intended functions that are required to fulfill the system-level intended functions during the scoping process. The resulting list of components, and groups of component types subject to an AMR was presented in the LRA, Subsections 2.5.3 through 2.5.14 and associated tables. These tables also contained the intended functions and the materials of construction for each of the mechanical components.
On the basis of the above review, the staff finds that the methodology used by the applicant to identify mechanical components that require an AMR is consistent with the requirements of the rule. The evaluation for the specific implementation of this methodology for ONS mechanical components can be found in Section 2.2 of this safety evaluation.
Structures
The screening process for structures began with the development of a list of structural component types from the structures determined to be within the scope of the rule using the requirements of 10 CFR 54.21, and the guidance contained in NEI 95-10, and the "NUMARC Containment and Class I Structures Industry Report." Other structural components were added from the review of the commitments made by the applicant with respect to the "regulated events" identified under 10 CFR 54.4(a)(3). The applicant also reviewed design basis specifications and structural drawings to complete its list of structural components within the scope of the rule. To verify that the list was complete, an independent review was performed by ONS structural experts.
The applicant then identified structural component-level intended functions from information in the UFSAR, ONS site specifications, commitments associated with design-basis events, regulated events, or input from Duke structural experts. This resulted in a list of component-level functions that supported the structural-level intended function plus some additional intended functions unique to individual components. For example, the spent fuel storage racks have a component specific intended function to provide separation to prevent criticality which does not match the Auxiliary building intended functions. The applicant then removed those structural components identified as performing their intended function with moving parts or a change in configuration or properties in the rule and in NEI 95-10, Appendix B. The applicant also removed all structural components that are replaced based on qualified life or specified time period. The remaining components were listed as structural components requiring an AMR.
On the basis of the above review, the staff finds that the methodology used by the applicant to identify the structures and structural components that require an AMR is consistent with the requirements of the rule. The implementation of this methodology and the listing of the structures and structural components for Reactor Building and other structures is evaluated in Section 2.2 of this safety evaluation.
Electrical Component
The methodology used to identify the electrical component requiring an AMR was different from the methodology used for mechanical and structural components. The applicant opted to develop a different process from the industry guidance. During the staff initial review, and the October 27 through 30, 1998 site-visit, the staff found the applicant's methodology unclear. The staff expressed its concern and documented its need for additional information in a letter dated December 1, 1998. In its February 17, 1999, response to the staff's request for additional information (RAI 2.2-6), the applicant provided a written description of its methodology.
The process for determining the electrical components subject to an AMR began with a complete list of all electrical component-types used at ONS. For this list of component types, the applicant identified the intended function(s) and eliminated those component types that required moving parts, or a change in configuration or properties to perform its intended function(s) as allowed by 10 CFR 54.21(a)(1)(i) and staff agreed-upon guidance in NEI 95-10. For those components remaining, the applicant eliminated a selected group of component types that did not meet the scoping criteria under 10 CFR 54.4(a). Finally, the applicant eliminated those components that are replaced based on a qualified life or specified time period. All remaining components are subject to an AMR. The above process describes the basic steps used in the identification of electrical components. Although this process is not consistent with the industry guidance provided in NEI 95-10, it is permitted by the rule and the staff finds it acceptable.
On the basis of the above review, the staff finds that the methodology used by the applicant to identify electrical components that require an AMR is consistent with the requirements of the rule. The evaluation for the specific implementation of this methodology for ONS electrical components can be found in Section 2.2.3.7 of this safety evaluation.
On the basis of the above review, the staff finds that there is reasonable assurance that the applicant's methodology for identifying the SSCs within the scope of license renewal and requiring an AMR is consistent with the requirements of 10 CFR 54.4 and 10 CFR 54.21(a)(1).
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In Sections 2.3 through 2.7 of Exhibit A, "License Renewal--Technical Information," of the LRA, the applicant described the SCs that are subject to an AMR for license renewal. The staff reviewed these sections of the LRA to determine if there is reasonable assurance that the applicant has listed those SCs subject to an AMR to meet the requirements stated in 10 CFR 54.21(a)(1).
The staff reviewed Sections 2.3 through 2.7 of Exhibit A to the LRA to determine if there is reasonable assurance that the applicant has appropriately identified and listed those SCs subject to an AMR to meet the requirements stated in 10 CFR 54.21(a)(1). The statement of considerations (SOC) for the license renewal rule (60 FR 22478) indicates that an applicant has the flexibility to determine the set of SCs for which an AMR is performed, provided that this set encompasses the SCs for which the Commission has determined an AMR is required. Accordingly, the staff focused its review on verifying that the implementation of the applicant's methodology discussed in Section 2.1 of this safety evaluation report (SER) did not result in the omission of SCs subject to an AMR in accordance with 10 CFR 54.21(a)(1). The staff performed the following two-step evaluation:
The staff used the ONS Updated Final Safety Analysis Report (UFSAR) in performing its review. Pursuant to 10 CFR 50.34(b), the FSAR contains " [a] description and analysis of the SSCs of the facility, with emphasis upon performance requirements, the bases, with technical justification therefore, upon which such requirements have been established, and the evaluations required to show that safety functions will be accomplished." The FSAR is required to be updated periodically pursuant to 10 CFR 50.71(e). Thus, the UFSAR contains updated plant-specific licensing-basis information regarding the SSCs and their functions.
The applicant presented its methodology (i.e., the integrated plant assessment (IPA)) to identify the SSCs within the scope of license renewal in Sections 2.1 and 2.2 of Exhibit A of the LRA. This IPA methodology consists of a review of all plant systems and structures to determine those that are within the scope of license renewal in accordance with the requirements of 10 CFR 54.4. The staff reviewed the IPA methodology and presented its evaluation in Section 2.1 of this SER. The applicant documented the implementation of that methodology in Sections 2.3 through 2.7 of Exhibit A of the LRA.
To ensure that the IPA methodology described in Sections 2.1 and 2.2 of Exhibit A of the LRA was implemented properly and identified the systems and structures within the scope of license renewal, the staff performed the following additional review: the staff sampled the contents of the UFSAR to identify systems or structures that may have intended functions meeting the scoping requirements of 10 CFR 54.4 that the applicant did not include within the scope of license renewal; the staff selected several systems, such as the radiation monitors and spent fuel building ventilation; and in a letter to the applicant dated December 2, 1998, the staff requested additional information about these systems.
In their January 25, 1999, response to NRC RAI 2.2-7 on whether radiation monitors were within the scope of license renewal, the applicant stated that the radiation monitors do not support any system intended functions as defined in 10 CFR 54.4(a). The staff agrees that while some radiation monitors do not support system intended functions. However, the staff believed that radiation monitors that detect activity in the control room air supply are credited for initiating certain operator actions. This continuous radiation monitoring was thought to be a safety-related function that cautions the control room operators to manually activate the filtration train of the control room pressurization and filtration system for Units 1, 2, and 3 control rooms under given accident conditions to meet TMI Action Plan Item III.D.3.4, for control room habitability.
The continuous radiation monitoring is described in ONS UFSAR Section 9.4.1.3, which states that "[r]eturn air from the control room is continuously monitored by a radiation monitor before recirculating back to the control room. A high radiation level will alert the operators to energize the outside air filter trains." On April 8, 1999, the staff requested that the applicant clarify its justification for excluding the radiation monitors from within the scope of license renewal. On May 10, 1999, the applicant responded to the staff's April 8, 1999, request for clarification of RAI 2.2-7. In its response, the applicant stated that although radiation monitors RIA-39 for Units 1, 2, and 3 will prompt operators to energize outside filter trains, operation of the monitors is not relied upon for the successful mitigation of any design-basis event and failure of the monitors will not prevent the successful mitigation of any design-basis event. In addition, the applicant stated that the radiation monitors are not relied upon to meet the requirements of any of the regulated events identified in 10 CFR 54.4(a)(3). The staff also requested that the applicant review the functions of the radiation monitors on OLRP-1002 drawings OLRFD-116C-1.1, 124B-1.5, and 133A-1.5 to ensure that these monitors did not have any intended functions that would require the monitors be included within the scope of license renewal. In its May 10, 1999, response, the applicant stated that the radiation monitors identified on the referenced drawing are all non-safety-related and not relied upon for the successful mitigation of a design-basis event. The staff has reviewed the applicant's responses and agrees that the radiation monitors are not within the scope of license renewal.
In NRC RAI 2.2-8, the staff asked the applicant to justify the omission of the spent fuel pool (SFP) ventilation system from within the scope of license renewal. SFP area ventilation is often credited in mitigating fuel handling accidents as well as performing other safety functions. The applicant responded in a letter dated February 17, 1999, that its analyses show that the system is not required to remain functional during or following any design-basis event to ensure any of the functions required by 10 CFR 54.4(a)(1) and does not meet the criteria of 10 CFR 54.4(a)(2) or (3) and is, therefore, not within the scope of license renewal. The staff reviewed the applicant's response and Chapter 15 of the UFSAR, and agreed with the applicant's decision to not include the system in the scope of license renewal.
In a letter to the applicant dated April 16, 1999, the staff requested additional information (RAI 2.5-1) concerning the identification and listing of components associated with instrumentation lines within the scope of license renewal. Rules for highlighting the OLRFD drawings in the front of each OLRP-1002 volume of flow diagrams contain the statement, "[a]ll instrumentation lines normally open to the process flow through, but not including the instrument, are included in license renewal. These lines are not highlighted except for containment penetrations." Section 2.5 of Exhibit A of the LRA lists the mechanical systems within the scope of license renewal and presents a table for each system at the end of the section identifying the components that are subject to an AMR. The staff review of these tables generally found the component "tubing" on the table of components subject to an AMR. However, several systems did not list tubing as a component, even though some instrument lines originated from points of the system that were within the scope of license renewal. In the letter dated April 16, 1999, the staff requested that the applicant clarify the status of the instrumentation lines for the following systems:
On May 10, 1999, the applicant responded to the staff's RAI. The applicant stated that for three systems, reactor building spray, component cooling, and feedwater, stainless steel tubing is included within the scope of license renewal and was inadvertently omitted from Tables 2.5-2 and 3.5-2 of the LRA. For three systems, reactor building cooling system, auxiliary building ventilation system, and the SSF HVAC System, no tubing exists within the license renewal boundaries of the systems. For the Condenser Circulating Water (CCW) System, the applicant stated that this system does have instrumentation lines within the license renewal boundaries, but they do not perform any intended function and are, therefore, not subject to an AMR. Therefore, this tubing was not included on Table 2.5-9 for the CCW System. The staff reviewed the applicant's response and found it acceptable.
In Section 9.2.2.2.4 of the ONS UFSAR, the applicant described the design and operation of the Recirculation Cooling Water (RCW) System. One function of the RCW System is to remove decay heat from the stored fuel in the spent fuel pool by transferring the heat from the spent fuel pool coolers to the CCW System. In the UFSAR, the applicant also stated that the SFP Cooling System is designed to keep the pool bulk temperature below 150°F under a variety of postulated normal and upset conditions, and under 205°F when considering abnormally high heat loads and certain equipment failure. The UFSAR further stated that 205°F represents the actual operating limit, because calculations show that the seismic and structural integrity of the pool is not compromised below this temperature. In addition, Chapter 15 Section 11.2.1 of the UFSAR stated the assumptions for a fuel handling accident in the SFP, which include a fuel assembly gap pressure based on a bulk SFP coolant temperature of 150°F.
Since the RCW System is relied upon to supply cooling water to the SFP Cooling System coolers to maintain the bulk SFP coolant temperature below the SFP design limits and below assumptions for the fuel handling accident analysis described in Section 15.11.2.1 of the UFSAR, the staff requested that the applicant clarify the basis for excluding the RCW System from the scope of license renewal. This issue was identified as Open Item 2.2.3-1.
In letters dated October 15, and November 30, 1999, the applicant responded to Open Item 2.2.3-1. The applicant clarified the basis of the SFP Cooling System design and the reason for omitting the RCW System from the scope of license renewal. In its response, the applicant stated that the fuel handling accident analysis for ONS assumes that spent fuel pool cooling, and thus the RCW System, is not functional during or following such an event. The applicant stated that the results of the safety analysis for the fuel handling accident demonstrates that the consequences of such an accident are within the 10 CFR Part 100 guidelines. The normal operating temperature for the spent fuel storage pool established by the ONS operating procedures ensure that the results of a fuel handling accident analysis remain valid even if all forced cooling to the spent fuel pool is lost at the commencement of the accident. 10 CFR 54.4(a)(1) states that safety-related SSCs, which are those relied upon to remain functional during and following design-basis events to ensure the capability to mitigate or prevent the consequences of an event (such as a fuel handling accident), that could result in potential offsite exposure comparable to 10 CFR Part 100 guidelines shall be included within the scope. Since the applicant's analysis demonstrated that spent fuel pool cooling is not required to remain functional during or following a fuel handling accident or to prevent or mitigate the consequences that could result in potential offsite exposure comparable to 10 CFR Part 100 guidelines, the SSCs required to fulfill the function of decay heat removal from the spent fuel pool, including the RCW System, are not within the scope of license renewal. The staff reviewed the reasons for excluding the RCW System from the scope of license renewal and found the applicants justification acceptable. On the basis of this review and the staff's findings, Open Item 2.2.3-1 is closed.
The staff reviewed the information submitted by the applicant in the LRA, information in the ONS UFSAR, and additional information in the applicant's January 25, February 17, and May 10, 1999, responses to the NRC's December 2, 1998, and April 16, and October 15, 1999, letter and did not identify any systems or structures with intended functions that were not already evaluated in the LRA. Therefore, the staff has reasonable assurance that the applicant has appropriately identified the systems and structures within the scope of license renewal in accordance with 10 CFR 54.4.
2.2.3.1 Containment Structures
2.2.3.1.1 Concrete Components, Steel Components, and Post-Tensioning System
In Section 2.3, "Reactor Building (Containment) Structural Components," of Exhibit A of the LRA, the applicant identified the SCs that are within the scope of license renewal and which of those within-scope SCs are subject to an AMR.
Component (equipment and piping) supports for the SCs described below are covered separately in Section 2.7 of Exhibit A of the ONS LRA. Electrical components that support the operation of the systems are presented in Section 2.6 of Exhibit A of the LRA. The staff evaluated component supports and electrical components in Sections 2.2.3.6 and 2.2.3.7 of this SER. Although instrument lines are not individually highlighted as being within the scope of license renewal on the flow diagrams in OLRP-1002, instrumentation line components (e.g., tubing, valves) are within the scope if the lines are normally open to process flow, as stated in the rules for the identification of components within the scope of license renewal in OLRP-1002. The applicant included instrument line components with the system to which they are attached.
2.2.3.1.1.1 Summary of Technical Information in the Application
The reactor buildings are Class 1 structures which prevent uncontrolled release of radioactivity. The applicant has determined that Class 1 structures meet the intent of 10 CFR 54.4(a)(1) and are within the scope of license renewal. A part of the reactor building, the containment, includes the concrete containment structure, liner, and all penetrations. The containment has been divided into three groups according to material of construction and component-level function. These component groups are described in Section 2.3.2, "Concrete Components"; Section 2.3.3, "Steel Components"; and Section 2.3.4, "Post-Tensioning System." The three containment component groups within the scope of 10 CFR Part 54 and their intended functions are given in Table 2.3-2 of Exhibit A of the LRA.
The concrete component group consists of the cylinder wall, dome, floor, and foundation slab. The applicant identified the following intended functions for the concrete component group:
One additional intended function identified for the cylinder wall, was the need for the wall to provide a rated fire barrier to confine or retard a fire from spreading to or from adjacent areas of the plant.
The steel component group includes anchorages, embedments, attachments, electrical penetrations, emergency personnel hatch, equipment hatch, fuel transfer tubes, liner plate, mechanical penetrations, and personnel hatch. All the components of the steel component group have the intended function of providing an essentially leak-tight barrier to prevent uncontrolled release of radioactivity. For anchorages, embedments, and attachments, the applicant also identified the intended function of providing a structural and/or functional support to safety-related SSCs and non-safety-related SSCs where failure of the structural component could directly prevent satisfactory accomplishment of any of the required safety-related functions. Mechanical penetrations also provide structural and/or functional support to safety-related SSCs and this was identified as an intended function. Finally, the ability of the liner plate to provide a heat sink during design-basis accidents or station blackout was identified as an intended function.
The post-tensioning group comprises two component types, tendon anchorage and tendon wires. Providing structural and/or functional support to safety-related SSCs was identified as the intended function for the post-tensioning group. More specifically, this function involves imposing compressive forces on the concrete containment structure to resist the internal pressure resulting from a design-basis accident with no loss of structural integrity.
2.2.3.1.1.2 Staff Evaluation
The staff reviewed Section 2.3 of Exhibit A of the LRA to determine whether there is reasonable assurance that the applicant has identified the containment SCs subject to an AMR in accordance with the requirements of 10 CFR 54.21(a)(1).
2.2.3.1.1.2.1 Containment Structures, Systems, and Components Within the scope of License Renewal and Subject to an Aging Management Review
The staff reviewed Section 6.2.1, "Containment Functional Design," of the UFSAR and compared the description of the structures, systems, and other components in the UFSAR to the description in the application to determine if there were any additional portions of the system that the applicant should have identified as within the scope of license renewal. As described in Sections 2.3 and 2.7 of Exhibit A of the LRA, essentially all portions of the containment were determined to be within the scope of license renewal and subject to an AMR. The staff reviewed the few remaining components of the containment to verify that they do not perform any intended functions. The staff also reviewed Section 6.2.1 of the UFSAR to determine if there were any additional functions that were not identified as intended functions in the LRA. The staff found no omissions. Therefore, there is reasonable assurance that the applicant adequately identified all portions of the containment structures which fall within the scope of license renewal and are subject to an AMR in accordance with 10 CFR Part 54.
In RAI 2.3-8, the staff asked the applicant why the tendon gallery, which provides access to the bottom anchorages of the vertical tendons as part of the post-tensioning system, had not been included within the scope of license renewal under 10 CFR 54.4(a)(2). In response to the RAI the applicant stated that the function of the tendon access gallery is to provide access to the bottom of the vertical tendons so that they can be tested and that its failure would not prevent satisfactory accomplishment of any of the functions identified by 10 CFR 54.4(a)(1)(i), (ii) or (iii). The staff agrees that the tendon gallery itself is not within the scope of license renewal. However, operational experience, as documented in NUREG-1522, has shown that water infiltration and high humidity in the tendon gallery can be a significant aging effect on the vertical tendons that could potentially result in loss of the ability of the post-tensioning system to perform its intended function. This is reflected in Open Item 3.3.3.1-1.
In RAI 2.3-11, the staff asked the applicant why the function(s) of the containment sump was not identified as an intended function(s) of the containment. The applicant responded to the RAI by stating that the sumps were not included in Section 2.3 of Exhibit A of the LRA because they do not perform the function of providing an essentially leak-tight barrier to prevent uncontrolled release of radioactivity, which is the function of the containment building. However, the emergency and normal sumps are included in Section 2.7 with reactor building internal structures as components requiring an AMR.
2.2.3.1.1.3 Review Findings for Concrete Components, Steel Components, and Post-Tensioning System
As described above, the staff has reviewed the information provided in Section 2.3 of Exhibit A of the LRA and the additional information provided by the applicant in response to the staff's RAIs. The staff has reasonable assurance that the applicant has appropriately identified those portions of the containment, and the associated SCs thereof, that are within the scope of license renewal and subject to an AMR in accordance with the requirements of 10 CFR 54.4.
2.2.3.2 Reactor Coolant System
2.2.3.2.1 Reactor Coolant System
In Section 2.4, "Reactor Coolant System Mechanical Components and Class 1 Component Supports," of the LRA, the applicant described the SCs of the Reactor Coolant System (RCS) that are subject to an AMR for license renewal.
2.2.3.2.1.1 Summary of Technical Information in the Application
As described in the application, the following SCs of the RCS are within the scope of license renewal and are subject to an AMR: RCS piping (Class 1; non-Class 1 portions are addressed in Section 2.5 of the application), pressurizer, reactor vessel, reactor vessel internals, once-through steam generator, reactor coolant pumps, control rod drive motor tube housings, letdown coolers, Class 1 component supports, reactor coolant piping supports, pressurizer supports, reactor vessel support skirt, control rod drive service structure, once-through steam generator supports, and reactor coolant pump supports. The rest of this section lists the intended functions of these SCs according to 10 CFR 54.4(a) and briefly describes these SCs.
Reactor Coolant System Piping (Class 1)
Intended Function:
For the ONS, the following components are within the reactor coolant pressure boundary: reactor vessel, once-through steam generators (primary side), pressurizer, reactor coolant pump, main coolant piping and portions of systems attached to these components. The attached systems that contain Class 1 components include the Core Flood System, High-Pressure Injection System, Low-Pressure Injection System, and Chemical Addition System. In addition, vents, drains, and instrumentation lines contain Class 1 components. RCS piping includes piping (including fittings, branch connections, safe ends, and thermal sleeves), valve bodies (pressure retaining parts of RCS isolation/boundary valves), and bolted closures and connections.
Pressurizer
Intended Functions:
The pressurizer is a vertical cylindrical vessel with a bottom surge line penetration connected to the hot leg piping by the surge line piping. The pressurizer contains electric heaters in its lower section and a water spray nozzle in its upper section. Since all sources of heat in the RCS are interconnected by piping with no intervening isolation valves, relief protection is provided on the pressurizer. Overpressure protection consists of two code safety valves and one power- operated relief valve. Piping attached to the pressurizer is Class 1 up to and including the first isolation valve.
Reactor Vessel
Intended Functions:
The reactor vessel consists of the cylindrical vessel shell, lower vessel head, closure head, nozzles, interior attachments and all associated pressure-retaining bolting. Coolant enters the reactor through the inlet nozzles, passes down through the annulus between the thermal shield and vessel inside wall, reverses at the lower head, passes up through the core, turns around through the plenum assembly, and leaves the reactor vessel through the outlet nozzles.
The reactor vessel has two outlet nozzles, through which the coolant is transported to the steam generators, and four inlet nozzles, through which coolant enters the reactor vessel from the discharge of the reactor coolant pumps. Two smaller nozzles between the inlet nozzles serve as inlets for decay heat removal and emergency core cooling water injection. The reactor vessel is vented through the control rod drives. Instrumentation nozzles penetrate the lower vessel head.
Control rod drive mechanisms are attached to flanged nozzles, which penetrate the closure head, and are not within the scope of license renewal. However, the control rod drive motor tube housings are within the scope of license renewal and subject to an AMR.
Reactor Vessel Internals (RVI)
Intended functions:
The RVI consist of two structural subassemblies that are normally located within the reactor vessel. The RVI can be removed during refueling outages when necessary. The two subassemblies of the internals are the plenum assembly and the core support assembly. The RVI for the ONS are described in the B&WOG topical report, BAW-2248. The applicant states that it has reviewed the current design and operation of the ONS RVI, and has determined that they are bounded by the description in BAW-2248, with the exception of thermal shield and thermal shield upper restraint. The thermal shield and thermal shield upper restraint were omitted from the generic report; however, these items support an ONS RVI intended function and are subject to an AMR. The thermal shield surrounds the core barrel and is constructed of austenitic stainless steel. The thermal shield upper restraint is also constructed of austenitic stainless steel.
Once-Through Steam Generator (OTSG)
Intended Functions:
Each ONS unit has two OTSGs. Each is a vertical, straight-tube, once-through, counterflow, shell-and-tube heat exchanger with shell-side boiling. The steam generator consists of upper and lower hemispherical heads welded to tubesheets that are separated by a seven-course shell assembly. Over 15,000 straight Alloy 600 tubes are held in alignment by 15 tube support plates. Primary coolant from the reactor enters the steam generator through a single inlet nozzle in the top of the upper head. Coolant flows downward through the straight parallel tubes, is cooled by the secondary coolant on the shell side, and then exits through two outlet nozzles in the lower head. Secondary coolant enters through a ring of ports that penetrate the shell approximately midway up the shell assembly. The feedwater travels downward through an annulus between the lower baffle and the shell. Near the lower tubesheet the feedwater turns inward, and then flows upward around the tubes and through the tube support plates. As the feedwater absorbs heat from the primary coolant, it boils and then becomes superheated. The dry steam exits the steam generator through two steam outlet nozzles just above the feedwater inlet ports. The OTSG items that are subject to an AMR are the hemispherical heads, secondary shell, tubes, plugs, mechanical sleeves, tubesheets, primary nozzles, main and auxiliary feedwater nozzles, steam outlet nozzles, instrumentation nozzles, drain nozzles, all associated pressure retaining bolting, and integral attachments inspected in accordance with ASME Section XI, Subsections IWB and IWC.
Reactor Coolant Pumps (RCPs)
Intended Function:
The reactor coolant pumps provide the head required to transport the reactor coolant through the reactor core, piping, and steam generators. All four reactor coolant pumps of each ONS unit are required during normal operation. The four reactor coolant pumps installed on ONS Unit 1 are Westinghouse Model 93A, while those installed on ONS Units 2 and 3 are Sultzer-Bingham.
The reactor coolant pump items that are subject to an AMR are the casing, cover, and associated pressure-retaining bolting. The portion of the reactor coolant pump rotating element above the pump coupling, the electric motor, and the flywheel are not subject to an AMR in accordance with 10 CFR 54.21(a)(1).
The pump cover is a generic term used to describe the pressure-retaining closure of the pump casing. The cast austenitic stainless steel cover (stuffing box for Sultzer-Bingham pumps) serves as a housing for the mechanical seals, radial bearing, thermal barrier, and recirculating impeller for the Sultzer-Bingham pumps. The cover is clamped between the carbon steel driver mount (motor stand for Sultzer-Bingham pumps) and the stainless steel pump casing. The main flange serves as the cover for the Westinghouse design. The Westinghouse cover closure consists of the main flange, thermal barrier, and pump casing.
Each reactor coolan