| |
Search Options | |||
| Index | Site Map | FAQ | Facility Info | Reading Rm | New | Help | Glossary | Contact Us | |
|||
Safety Evaluation Report Related to the License Renewal of Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (NUREG-1705) |
[ Next | Top of file ]
Docket Nos. 50-317 and 50-318
Manuscript Completed: December 1999
Date Published: December 1999
D.L. Solorio, NRC Project Manager
Division of Regulatory Improvement Programs
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
[ Prev | Next | Top of file ]
This safety evaluation report (SER) documents the technical review of the Calvert Cliffs Nuclear Power Plant, Units 1 and 2 license renewal application by the U.S. Nuclear Regulatory Commission (NRC) staff. The Baltimore Gas and Electric Company requested renewal of the Class 104b operating licenses for the Calvert Cliffs units (license numbers DPR-53 and DPR-69) for a period of 20 years beyond the current expiration of midnight, July 31, 2014, for Unit 1 and midnight, August 13, 2016, for Unit 2. By letter dated April 8, 1998, the Baltimore Gas and Electric Company submitted the license renewal application for Calvert Cliffs in accordance with Part 54 of Title 10 of the Code of Federal Regulations.
The Calvert Cliffs nuclear station is located on the west shore of the Chesapeake Bay in Calvert County, Maryland, approximately 45 miles southeast of Washington, D.C., and 60 miles south of Baltimore, Maryland. Operation of the twin Combustion Engineering pressurized-water reactors results in an approximate net electrical output of 845 megawatts for each reactor.
This SER presents the results of the staff's review of information submitted in conjunction with the renewal application. In an earlier version of this safety evaluation report (SER) issued on March 21, 1999, the staff identified a number of open and confirmatory items. All of those items have been resolved, as discussed in this SER. On the basis of its evaluation of the application 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 Calvert Cliffs Nuclear Power Plant, Units 1 and 2.
[ Prev | Next | Top of file ]
This report describes the results of a review by the Nuclear Regulatory Commission (NRC) staff of an application to renew the licenses for the two units of the Calvert Cliffs nuclear power plant. Under the Atomic Energy Act, the NRC issues licenses for commercial power reactors to operate for up to 40 years. The Act 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 Part 54 focus on managing the effects of aging for passive structures and components, like buildings, tanks and pipes.
The NRC also established requirements for a license renewal environmental report in Part 51. Those requirements establish the scope of a review of environmental impacts, which is part of the NRC's responsibilities under the National Environmental Policy Act (NEPA). The results of that review are described in a separate NRC report.
In a letter dated April 8, 1998, the Baltimore Gas and Electric Company (BGE) filed an application to renew the licenses for their two-unit Calvert Cliffs plant. BGE requested a 20-year extension in the license term for both units. The existing licenses expire on midnight July 31, 2014 and August 13, 2016, respectively. If granted, the renewed licenses would extend to July 31, 2034 and August 13, 2036, respectively.
The Calvert Cliffs plant is located on the west shore of the Chesapeake Bay in Calvert County, Maryland. It is approximately 45 miles southeast of Washington, D.C., and 60 miles south of Baltimore, Maryland. Each unit is a Combustion Engineering pressurized water reactor that produces a net electric output of about 845 megawatts.
In accordance with Part 54, BGE submitted information in their renewal application that identifies all plant systems, structures, and components: (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. BGE'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 where BGE demonstrated that the existing programs provide adequate aging management. In other cases, BGE 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 about the length of time the plant will be operated are made and incorporated into design calculations for several of the plant's systems, structures, and components. 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 effect of aging on these structures, systems, and components will be adequately managed for the period of extended operation.
This report describes the results of the NRC staff's review of the BGE programs to manage aging effects. In this report, we conclude that BGE has demonstrated that aging effects applicable to the required scope of systems, structures and components 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 BGE 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 in the future. These procedure changes and future actions are summarized in a list included as Appendix E to this report. BGE will update their final safety analysis report, associated with the existing license, to include the changes to the licensing basis reflected in the Appendix E list, which we relied on to grant a renewed license.
During meetings to gather public comments about the environmental impacts of extending the Calvert Cliffs 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.
The conclusions in this report have been verified by inspections conducted by the NRC. 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 basis for the conclusions in this report are also reviewed by the NRC's Advisory Committee on Reactor Safeguards. They independently review the application, and submit their recommendation directly to the Commission. Their recommendation is included in the published version of this report.
In our recommendation for granting a renewed license for Calvert Cliffs, we have described the programs, maintenance activities, and inspection procedures that we rely on to conclude that there is reasonable assurance that actions have been or will be taken to manage effects of aging for a 20-year period of extended operation, such that the plant can continue to operate safely.
[ Prev | Next | Top of file ]
[ Prev | Next | Top of file ]
This document is a safety evaluation report (SER) on the application for license renewal for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, as filed by the applicant Baltimore Gas and Electric Company (BGE or Applicant). By a letter dated April 8, 1998, BGE submitted its application to the United States Nuclear Regulatory Commission (NRC) for renewal of the Calvert Cliffs operating licenses for an additional 20 years. This report was prepared by the NRC staff and summarizes the results of the staff's safety review of 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 Calvert Cliffs is David L. Solorio. Mr. Solorio may be contacted by calling 301-415-1973, or by writing to the License Renewal and Standardization Branch, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001.
In its April 8, 1998, submittal, BGE requested renewal of the Class 104b operating licenses for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (license numbers DPR-53 and DPR-69, respectively) for a period of 20 years beyond the current license expirations of midnight, July 31, 2014, and midnight, August 13, 2016, respectively. The nuclear station is located on the west shore of the Chesapeake Bay in Calvert County, Maryland, approximately 45 miles southeast of Washington, DC, and 60 miles south of Baltimore, Maryland. Operation of the twin Combustion Engineering pressurized-water reactors results in an approximate net electrical output of 845 megawatts for each reactor. Details concerning the plant and the site are contained in the Updated Final Safety Analysis Report (UFSAR) for Calvert Cliffs Nuclear Power Plant, Units 1 and 2.
The license renewal process proceeds along two tracks: a technical review of safety issues and an environmental review. The requirements for these reviews are stated in NRC regulations 10 CFR Parts 54 and 51, respectively. The safety review for the Calvert Cliffs license renewal is based on BGE's application for license renewal and on the licensee's answers to requests for additional information (RAIs) from the NRC staff. In meetings and docketed correspondence, BGE has also supplemented the answers that it has given to the RAIs. The license renewal application and all pertinent information and materials, including the UFSAR mentioned above, are available to the public for review at the NRC Public Document Room, 2120 L Street, NW., Washington, D.C. 20555-0001. In addition, the application and significant information and materials related to the renewal review are available on the NRC Web page at www.nrc.gov.
This SER summarizes the results of the staff's safety review of the Calvert Cliffs license renewal application 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 license renewal application was reviewed in accordance with the NRC regulations and the guidance provided 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 by the Advisory Committee on Reactor Safeguards (ACRS). The conclusions of this report are given in Chapter 6.
Appendix A is a chronology of NRC's principal correspondence related to the review of the application. Appendix B is a bibliography of the references used during the course of the review. Appendix C is a list of abbreviations used throughout the report. The NRC staff principal reviewers and its contractors for this project are listed in Appendix D.
Appendix E presents a summary listing of the programs, maintenance activities and inspection procedures that formed a significant basis for the staff's conclusion. As such, this list represents those commitments that warrant regulatory control. BGE will incorporate appropriate changes to the next update of the final safety analysis report (FSAR), following the issuance of the renewed license. The FSAR will be updated for each item in Appendix E in accordance with the guidance for 10 CFR Section 50.71(e). Since future changes to the FSAR will be made in accordance with 10 CFR Section 50.59, these programs, maintenance activities and inspection procedures will be adequately controlled. Until the FSAR update is complete, a license condition requires that any changes to the items on the list be made in accordance with Section 50.59.
The listing in Appendix E also identifies future actions. Throughout this safety evaluation report, the staff has described various schedules for future actions. The staff has determined that none of the future actions are required prior to the end of the current license term in order to effectively manage aging. Therefore, as long as they are completed by the end of the current license term, licensee can make changes to such schedules without prior NRC approval. However, all of the future actions must be completed before the plant enters the period of extended operation, except for the volumetric inspections of the control element drive mechanisms in Unit 1 which will be completed by 2029 as described in Section 3.2.3.2.1.C (6) of the SER. Accordingly, the renewed license also includes a condition that all of the future actions must be completed by the end of the existing license term.
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 considerations related to renewing the license for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2. The draft and final plant-specific supplements to the GEIS were issued separate from this report.
[ Prev | Next | Top of file ]
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, and 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 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 Part 54 established a regulatory process that is simpler, more stable, and more predictable than the previous license renewal rule. In particular, Part 54 was clarified to focus on managing the adverse effects of aging rather than on identification of all aging mechanisms. The rule changes were intended to ensure that important systems, structures, and components 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 to similarly focus the scope of the review of environmental impacts of license renewal, under 10 CFR Part 51, which is part of the 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 systems, structures, and components 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 systems, structures, and components (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 systems, structures, and components 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 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 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 by 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. Section 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 about the length of time the plant will be operated are made and incorporated into design calculations for several of the plant's systems, structures, and components. Under 10 CFR 54.21(c)(1), 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 effect of aging on these structures, systems, and components 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 for the safety review, which was made available in the Public Document Room in September 1997. The draft regulatory guide will be used, along with the draft standard review plan, 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 standard review plan and clarify regulatory guidance.
The environmental protection regulations, 10 CFR Part 51, were revised 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, in which the staff 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 CFR51.53(c)(3)(i), an applicant for license renewal may incorporate these generic findings in an environmental report and address only those environmental impacts that are required to be evaluated on a plant-by-plant basis.
The NRC performs plant-specific reviews of the remaining environmental impacts of license renewal (those identified as Category 2 issues in 10 CFR Part 51, Subpart A, Appendix B) as well as any new and significant information, in accordance with NEPA and the requirements of 10 CFR Part 51. A public meeting was held on July 9, 1998, near Calvert Cliffs nuclear power plant as part of the scoping process to identify environmental issues specific to the plant. The result of the environmental review is an NRC preliminary recommendation with respect to the license renewal action. This is known as a draft plant-specific supplement to the GEIS, which is published for comment and discussed at a separate public meeting. After consideration of comments on the draft, NRC prepares and publishes a final plant-specific supplement to the GEIS.
Two public scoping meetings were held on July 9, 1998 to identify environmental issues specific to the plant. On February 24, 1999, the staff issued the Draft Supplement 1 to the GEIS, regarding the results of the staff's environmental review of Calvert Cliffs. During the 75-day comment period that followed, two public meetings were held on April 6, 1999, in 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 October 5, 1999, the staff issued the Final Supplement 1 to the GEIS on Calvert Cliffs, in which it presents its final environmental analysis that considers and weighs the environmental effects of the license renewal, the environmental impacts of alternatives to 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 BGE; (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 1 to NUREG-1437 that the Commission determine that the adverse environmental impacts of license renewal for Calvert Cliffs Nuclear Power Plant Units 1 and Unit 2 are not so great that preserving the option of license renewal for energy planning decisionmakers would be unreasonable.
[ Prev | Next | Top of file ]
The requirements for the renewal of operating licenses for nuclear power plants are described in 10 CFR Part 54. The staff performed its technical review of the Calvert Cliffs application for license renewal in accordance with Commission guidance and the requirements of 10 CFR Sections 54.19, 54.21, 54.22, 54.23, and 54.25. The standards for issuance of a renewed 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 provide general information. Baltimore Gas and Electric provided this general information in Attachment 1 to its April 8, 1998, submittal letter regarding the application for renewed operating licenses for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2. The staff finds that Calvert Cliffs has provided the information required by 10 CFR 54.19(a) in Attachment 1 of the April 8, 1998, submittal letter.
In 10 CFR 54.19(b), the Commission requires that license renewal applications 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." BGE states the following in its renewal application regarding this issue:
The current indemnity agreement (B-70) for licenses DPR-53 and DPR-69 does not contain a specific expiration term. Expiration is expressed in terms of the time of the expiration of the licenses specified. Therefore, conforming changes to account for the expiration term of the proposed renewed licenses are unnecessary.
The staff notes that the current indemnity agreement for Calvert Cliffs 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 lists two license numbers. By maintaining the license numbers on issuance of the renewed license, there is no need to make conforming changes to the indemnity agreement. Therefore, the requirements of 10 CFR54.19(b) have been met.
In 10 CFR 54.21, the Commission requires that each application for a renewed license for a nuclear facility shall include an integrated plant assessment (IPA), current licensing basis (CLB) changes during NRC review of the application, an evaluation of time-limited aging analyses (TLAAs) and a final safety analysis report (FSAR) supplement. In 10 CFR 54. 22, the Commission states requirements regarding technical specifications. 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 provided by the draft standard review plan entitled "Review of License Renewal Applications for Nuclear Power Plants," which was published in September 1997. The staff's evaluation of the license renewal application in accordance with 10 CFR 54.21 and 54.22 are contained in Chapters 2, 3, and 4 of this report.
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 1, that state the considerations related to renewing the license for the Calvert Cliffs Nuclear Power Plant, Units 1 and 2.
The report by the Advisory Committee on Reactor Safeguards required by 10 CFR 54.25 is included in Chapter 5 of this SER. The finding required by 10 CFR 54.29 is contained in Chapter 6 of this report.
[ Prev | Next | Top of file ]
As a result of its initial review of the license renewal application for Calvert Cliffs, including the additional information provided to the NRC, the staff identified a number of open issues and confirmatory items when this report was issued in March 1999. This report has been revised to include a description, in each applicable section, of the manner by which those matters have been resolved.
[ Prev | Next | Top of file ]
[ Prev | Next | Top of file ]
Applicants for license renewal are required by the license renewal rule to perform, among other things, an integrated plant assessment (IPA). The first two steps of the IPA, 10 CFR 54.21(a)(1) and 10 CFR 54.21(a)(2), require the applicant to identify and list, from those systems, structures, and components (SSCs) within the scope of the license renewal rule, those structures and components that are subject to an aging management review and to describe and justify the methods used to determine those structures and components subject to review. SSCs within the scope of the license renewal rule are those meeting the criteria in 10 CFR 54.4. Structures and components subject to an aging management review are those that meet the criteria of 10 CFR 54.21(a)(1)(i) and (ii).
In a letter dated August 18, 1995, BGE (the applicant) submitted its "Integrated Plant Assessment Methodology," which was subsequently amended to incorporate changes required by the staff. The amendment to the IPA was submitted in a BGE letter dated January 11, 1996. The staff reviewed this methodology and found it acceptable as documented in a Final Safety Evaluation (FSE) dated April 4, 1996. The BGE license renewal application (LRA) dated April 8, 1998, contains the IPA methodology, technically unchanged from that previously submitted in Attachment 1, Appendix A, Section 2. The staff concluded in its FSE that:
The BGE methodology sufficiently describes and justifies an acceptable process for identifying structures and components at Calvert Cliffs, Units 1 and 2, that are subject to an aging management review for license renewal and therefore would meet the requirement of 54.21(a)(2). In addition, this process, if implemented, provides reasonable assurance that all structures and components subject to an aging management review pursuant to 10 CFR 54.21(a)(1) will be identified.
The staff's evaluation of the implementation of the process for identifying SSCs that are subject to an aging management review pursuant to 10 CFR 54.21(a)(1) is contained in Section 2.2 of this safety evaluation report (SER).
[ Prev | Next | Top of file ]
In Sections 3 through 6 of Appendix A, "Technical Information," to the LRA, BGE (the applicant) described the structures and components that are subject to an aging management review (AMR) for license renewal. The staff reviewed these sections of the application to determine if there is reasonable assurance that the applicant has identified and listed those structures and components subject to an AMR to meet the requirements stated in 10 CFR 54.21(a)(1).
The staff reviewed Sections 3 through 6 of Appendix A to the LRA to determine if there is reasonable assurance that the applicant has appropriately identified and listed those structures and components subject to an AMR to meet the requirements stated in 10 CFR 54.21(a)(1). The statements of consideration (SOC) for the license renewal rule (60 FR 22478) indicate that an applicant has the flexibility to determine the set of structures and components for which an AMR is performed, provided that this set encompasses the structures and components 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 staff SER did not result in the omission of structures and components subject to an AMR in accordance with 10 CFR 54.21(a)(1). The staff performed the following two-step evaluation:
| (1) | The first step was to determine whether the applicant has properly identified the systems, structures, and components (SSCs) within the scope of license renewal, pursuant to 10 CFR 54.4. As described in more detail below, the staff reviewed selected structures and components that the applicant did not identify as within the scope of license renewal to verify that they do not have any intended functions. |
| (2) | The second step was to determine whether the applicant has properly identified the structures and components (S&Cs) subject to an AMR from among those identified in the first step. As described in more detail below, the staff reviewed selected S&Cs that the applicant identified as within the scope of license renewal to verify that the applicant has identified these S&Cs as subject to an AMR if they perform intended functions without moving parts or without a change in configuration or properties and are not subject to replacement on the basis of a qualified life or specified time period. To determine whether the applicant identified the S&Cs subject to an AMR, the staff did not review S&Cs that the applicant had identified as subject to an AMR because it is an applicant's option to include more S&Cs than those required by 10 CFR 54.21(a)(1). |
The staff used the Calvert Cliffs 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 structures, systems, and components of the facility, with emphasis upon performance requirements, the bases, with technical justification therefor, 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 systems, SSCs, and their functions.
The applicant presented its methodology (i.e., the integrated plant assessment (IPA)) to identify the systems, structures, and components (SSCs) within the scope of license renewal in Section 2.0 of Appendix A to 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 in a letter to the applicant dated April 4, 1996, the staff concluded that the methodology was acceptable for meeting the requirements of 10 CFR 54.21(a)(2) and, if implemented, offered reasonable assurance that all structures and components subject to an aging management review (AMR), as required by 10 CFR 54.21(a)(1), would be identified. Additionally, the letter stated that the staff concluded that the methodology provides processes for demonstrating that the effects of aging would be adequately managed pursuant to 10 CFR 54.21(a)(3) and for evaluating time-limited aging analyses pursuant to 10 CFR 54.21(c) that are conceptually sound and consistent with the intent of the license renewal rule.
To ensure that the IPA methodology described in Section 2.0 of Appendix A to the LRA properly implemented and identified the systems and structures within the scope of license renewal, the staff performed the following additional review. The staff compared the list of systems and structures at the Calvert Cliffs Nuclear Power Plant (CCNPP) listed in Table 3-1 in Section 2.0 of Appendix A to the LRA, to a list of the 66 systems and structures identified by the applicant as conforming to the scoping requirements of 10 CFR 54.4. The staff identified those systems and structures not included within the scope of license renewal and reviewed the information contained in the UFSAR for a sample of these systems and structures to determine whether they performed any intended function defined by 10 CFR 54.4, and thus would be required to be included within the scope of license renewal. The staff found no omissions. However, to ensure the applicant did not omit any system or structures with intended functions, by letter dated August 27, 1998, the staff requested additional information about eight systems and structures outside the scope of license renewal. In response to the staff's request for additional information, on November 2, 1998, the applicant submitted additional information about the five systems and three structures. For each system and structure, the applicant submitted a general description, listed the specific intended functions (active and passive), and identified the portion of the LRA in which the system's components were reviewed (if the system or structure performed an intended function). For example, the staff requested additional information about the reactor protective system. In its response, the applicant identified the three passive intended functions performed by this system and added that the components within the scope of license renewal that performed this intended function were evaluated in either Section 6.2, "Electrical Commodities"; Section 5.9, "Feedwater System"; or Section 6.1, "Cable Commodities."
The staff reviewed the information submitted by the applicant in the LRA and additional information submitted in response to the NRC's August 27, 1998, memorandum, and did not find any systems or structures with intended functions that were not already evaluated in the LRA. Therefore, the staff has reasonable assurance that the applicant had appropriately identified the systems and structures within the scope of license renewal in accordance with 10 CFR 54.4.
2.2.3.1 Component Supports Commodity Group
In Section 3.1, "Component Supports," of Appendix A to the LRA, the applicant described the systems with component supports at CCNPP that are within the scope for license renewal, and identified which of those structures and components are subject to an AMR.
2.2.3.1.1 Summary of Technical Information in the Application
As described in the LRA, component supports are associated with almost every plant system. A component support is the connection between a system, or a component within a system, and a plant structural member. Because component supports perform the same basic function regardless of the system, the applicant reviewed these components as a commodity group.
The applicant prepared a generic list of component supports by reviewing industry and plant-specific information, including the Seismic Qualification Utility Group guidance, American Society of Mechanical Engineers, Section XI, component support inspection documentation, and the CCNPP system level scoping results for license renewal. The applicant identified all component support types that provide support to plant components that are within the scope of license renewal and listed them as being within the scope of license renewal. The applicant identified 48 systems within the scope of license renewal that contained supports within this commodity group evaluation.
The applicant grouped the total population of component supports into four categories. The categories include supports for both the distributive portions of systems (e.g., piping and cable raceways) and for system equipment. The categories are defined by the components they support: piping; cable raceways; heating, ventilating, and air conditioning ducting; and equipment. These four categories are further separated into 19 sub-categories based on similarities of physical characteristics, loading conditions, and environment.
The applicant identified the following intended functions for the component supports within the scope of license renewal:
The applicant identified the following component supports within the scope of license renewal that are evaluated elsewhere in Appendix A to the LRA:
The applicant noted that all of the intended functions listed above are passive because they accomplish their function without moving parts or a change in configuration or property. The applicant therefore concluded that all component supports within the scope of license renewal are also subject to an AMR.
On the basis of the intended functions listed above, the applicant identified the following 19 component support types from the component support groups within the scope of license renewal as being subject to an AMR:
| COMMODITY SUPPORT GROUPS AND TYPES | |
|---|---|
| Piping Supports | Spring hangers, constant load, snubber supports--OC |
| Spring hangers, constant load, snubber supports--IC | |
| Piping frames and stanchions--OC | |
| Piping frames and stanchions--IC | |
| Cable Raceway Supports | Trapeze, cantilever, other supporting styles--OC |
| Piping frames and stanchions--IC | |
| HVAC Ducting Supports | HVAC ducting supports--OC |
| HVAC ducting supports--IC | |
| Equipment Supports | Elastomer vibration isolators--OC |
| Electrical cabinet anchorage--OC | |
| Electrical cabinet anchorage--IC | |
| Equipment frames and stanchions--OC | |
| Equipment frames and stanchions--IC | |
| Frames and saddles--OC | |
| Frames and saddles--IC | |
| Metal spring isolators and fixed bases--OC | |
| Metal spring isolators and fixed bases--IC | |
| Loss-of-coolant accident restraints--IC | |
| Ring foundations for flat-bottomed vertical tanks--OC | |
OC - Outside Containment, IC - Inside Containment
2.2.3.1.2 Staff Evaluation
The staff reviewed Section 3.1 of Appendix A to the LRA to determine whether there is reasonable assurance that the applicant has appropriately identified the component supports within the scope of license renewal in accordance with 10 CFR 54.4 and subject to an AMR in accordance with the requirements of 10 CFR 54.21(a)(1). After completing the initial review, by letter dated September 7, 1998, the staff issued a request for additional information (RAI) regarding component supports, and by letter dated November 19, 1998, the applicant responded to the RAI.
2.2.3.1.2.1 Component Supports Within the Scope of License Renewal
In the first step of its evaluation, the staff reviewed the information submitted by the applicant in the LRA to identify if there were systems or portions of systems with component supports that the applicant failed to identify as within the scope of license renewal that should have been so identified. The applicant stated in the LRA that all component support types that provide support to plant components that are within the scope of license renewal are identified and these component support types are listed as being within the scope of license renewal. The staff compared Table 3.1-1, which is found in Section 3.1 of Appendix A to the LRA, with Table 3-1, which is found in Section 2.0 of Appendix A to the LRA, to determine if the applicant omitted any component supports when compiling its list of such systems within the scope of license renewal. The staff also sampled selected systems not listed in Table 3.1-1 to verify that they do not have any intended functions as defined in Section 3.1 of Appendix A to the LRA.
To help ensure that all systems with component supports within the scope of license renewal were listed in Table 3.1-1, the staff requested more detailed information from the applicant. In NRC Question Nos. 3.1.1 and 3.1.8, the staff noted seven systems in Table 3-1 of Section 2.0 of Appendix A to the LRA that were within the scope of license renewal but that did not appear in Table 3.1-1 of Section 3.1. The applicant responded that two of the systems were within the scope of license renewal, but contained no component supports; one was a portion of a system already listed in Table 3.1-1 (SG blowdown system is part of the MS system); three systems were evaluated in other commodity or system reports (e.g., the containment isolation group's individual containment penetrations are evaluated in each individual system's section); and one system was determined to be outside the scope of license renewal and, therefore, its component supports were outside the scope. One system, diesel generator building HVAC system, was inadvertently omitted from Table 3.1-1. The applicant corrected this error in its November 19, 1998, response to the staff's RAI, by adding the diesel generator building HVAC component supports to Table 3.1-1.
In NRC Question No. 3.1.4, the staff requested clarification on whether steel structural frames used for the support of piping systems were treated as component supports or as structural components. In its response, the applicant stated that the piping support frames were considered component supports and were discussed in Section 3.1 of Appendix A to the LRA. Information regarding the boundary of commodity supports was requested in NRC Question No. 3.1.6, specifically, were fasteners included, and if fasteners have welded connections, are they included within the scope of the components commodity report. The applicant clarified in its response that fasteners and attachments associated with the component side of the component support are evaluated in the component supports commodity group. Fasteners on the structure side of the component support are evaluated in both the component support commodity evaluation and in the evaluation for the specific structure. Welds and fasteners were not identified specifically, rather, they were considered part of the support.
As described above, the staff has reviewed the information in Section 3.1 of Appendix A to the LRA and the additional information submitted by the applicant in response to the staff's RAIs. On the basis of that review, the staff finds that there is reasonable assurance that the applicant has appropriately identified the component supports within the scope of license renewal in accordance with the requirements of 10 CFR 54.4.
2.2.3.1.2.2 Component Supports Subject to an Aging Management Review
In Table 3.1-1 of Appendix A to the LRA, the applicant identified systems and their associated component supports within the scope of license renewal. In Section 3.1.1.1 of Appendix A to the LRA, the applicant stated that because these component supports performed their intended function without moving parts or without a change in configuration or properties, they have passive intended functions. Therefore, all component supports (except for snubbers, which were excluded as "active" equipment by 10 CFR 54.21(a)(1)(i)), are within the scope of license renewal. The applicant further clarified that the snubber subcomponents that mount the snubber to the pipe or component and to the structural component are referred to as snubber supports, and are included within the scope of license renewal and are subject to an AMR. Table 3.1-2 of Appendix A to the LRA summarizes all the component support types requiring an AMR. The staff agrees with the applicant's inclusion of all the component support types listed in Table 3.1-2 as requiring an AMR.
The staff reviewed the information in Section 3.1 of Appendix A to the LRA and has determined that there is reasonable assurance that the applicant has appropriately identified the component supports subject to an AMR to meet the requirements stated in 10 CFR 54.21(a)(1).
2.2.3.2 Piping Segments That Provide Structural Support
In Section 3.1A, "Piping Segments That Provide Structural Support," of Appendix A to the LRA, the applicant described the piping segments that provide structural support and that are within the scope for license renewal and identified which of those piping segments are subject to an AMR.
2.2.3.2.1 Summary of Technical Information in the Application
Systems that have safety-related/non-safety-related (SR/NSR) boundaries or changes in piping classification have a boundary valve at the functional transition point. The structural integrity of the boundary valve, which functions as the system pressure boundary, must not be compromised. To ensure proper seismic structural support if the valve itself is not anchored, the system's structural boundary must be extended beyond the boundary valve to the first seismic anchor (or equivalent) and must include the pipe segment connecting the boundary valve to the pipe support. These components together act as a single support system, ensuring the integrity of the SR/NSR functional boundary under all design-basis conditions.
Providing structural support under all current licensing-basis design loading conditions for safety-related components (within the scope of license renewal) is the only intended function identified by the applicant for these piping segments. Because the intended function is performed without moving parts or a change in configuration or properties, it is a passive intended function and, therefore, piping segments that provide such support are subject to an AMR.
All fluid systems containing safety-related piping are within the scope of license renewal. These systems have the potential for having SR/NSR functional boundaries where piping segments beyond the functional boundary would be credited for structural support of the boundary. The applicant reviewed all of the fluid systems at CCNPP and identified those systems with safety-related piping in Table 3.1A-1 of Appendix A to the LRA. A total of 25 systems were identified as having the potential for SR/NSR functional boundaries with seismic boundaries extending beyond them for structural support.
2.2.3.2.2 Staff Evaluation
The staff reviewed Section 3.1A of Appendix A to the LRA to determine whether there is reasonable assurance that the applicant has appropriately identified the piping segments providing structural support within the scope of license renewal in accordance with 10 CFR 54.4 and subject to an AMR in accordance with the requirements of 10 CFR 54.21(a)(1).
2.2.3.2.2.1 Piping Segments That Provide Structural Support Within the Scope of License Renewal
To determine which piping segments are credited with providing structural support for boundary valves and isolation points at SR/NSR boundaries, the staff performed the following reviews. The staff compared Table 3.1A-1 in Section 3.1 of Appendix A to the LRA and Table 3-1 in Section 2.0 of Appendix A to the LRA to determine if the applicant omitted any safety-related fluid systems when compiling its list of systems to evaluate for functional boundaries. The applicant considers all piping segments beyond the SR/NSR functional boundary that perform the intended function of providing structural support to the safety-related piping and boundary isolation valve or isolation point as being within the scope of license renewal. The staff also reviewed the UFSAR to determine if there were CCNPP fluid systems that might perform safety-related functions or other intended functions as described in 10 CFR 54.4 that were not identified in Table 3.1A-1. The staff sampled CCNPP fluid systems not included in Table 3.1A-1 to determine if the applicant had omitted any systems having the potential for safety-related or non-safety-related functional boundaries. No omissions were identified.
Safety-related systems have the potential for SR/NSR functional boundaries where non-safety-related piping segments may provide structural support beyond the functional boundary. The LRA identified the safety-related fluid systems that have the potential for SR/NSR functional boundaries with structural boundaries extending beyond the functional boundaries within the scope of license renewal. As described above, the staff reviewed the information in Section 3.1A of Appendix A to the LRA and concluded that there is reasonable assurance that the applicant has appropriately identified the piping segments providing structural support to safety-related piping and boundary valves within the scope of license renewal in accordance with the requirements of 10 CFR 54.4.
2.2.3.2.2.2 Piping Segments That Provide Structural Supports Subject to an Aging Management Review
In Table 3.1-1 of Appendix A to the LRA, the applicant identified systems within the scope of license renewal with the potential for containing piping segments beyond SR/NSR boundaries that provide structural support to the safety-related piping and boundary isolation valve or isolation point. In Section 3.1.A.1.1 of Appendix A to the LRA, the applicant stated that because these portions of piping segments performed their intended function without moving parts or without a change in configuration or properties, they have passive intended functions. Therefore, all of these piping segments are included within the scope of license renewal and are subject to an AMR. The staff agrees with the applicant's inclusion of all these piping segments as requiring an AMR.
The staff has reviewed the information in Section 3.1A of Appendix A to the LRA. On the basis of the staff's review, the staff finds that there is reasonable assurance that the applicant has appropriately identified the piping segments that provide structural supports subject to an AMR in accordance with the requirements of 10 CFR 54.21(a)(1).
2.2.3.3 Fuel Handling Equipment and Other Heavy Load Handling Cranes
In Section 3.2, "Fuel Handling Equipment (FHE) and Other Heavy Load Handling Cranes (HLHCs)," of Appendix A to the LRA, the applicant described structures and components of the FHE and HLHCs that are within the scope of license renewal (10 CFR 54.4). The applicant also identified which of those within-scope structures and components are subject to an AMR in accordance with 10 CFR 54.21(a)(1)(i) and (ii). By a letter to the NRC dated February 4, 1999, the applicant supplemented the scope of Section 3.2 by identifying additional structures and components that are within the scope of license renewal and subject to an AMR. In addition, the staff issued RAIs by letter dated August 26, 1998, regarding the FHE and HLHC commodity report. By letter dated November 4, 1998, the applicant responded to the staff's RAIs.
The staff reviewed Section 3.2, of Appendix A to the LRA, against the requirements of 10 CFR 54.4 (a)(1), (2), and (3) and 10 CFR 54.21(a)(1)(i) and (ii). More specifically, the staff focused its review on determining whether there is reasonable assurance that the applicant identified and listed (1) FHE and HLHC structures and components that are within the scope of license renewal and (2) FHE and HLHC structures and components that are subject to an AMR in accordance with the requirements stated in 10 CFR 54.21(a)(1).
2.2.3.3.1 Summary of Technical Information in the Application
The applicant stated that the structures and components of the FHE and HLHCs are common to many systems. Therefore, the applicant's evaluation is presented in Section 3.2 of Appendix A to the LRA as a separate commodity report on all the FHE and HLHC structures and components within the plant. Some of the FHE and HLHC structural type components, as discussed later in this section of the SER, are identified in Section 3.2 but are evaluated in the individual system sections or buildings in which they are housed.
The FHE and HLHC commodity report addresses (1) all structures and components involved in fuel handling and transfer and (2) cranes that routinely lift heavy loads over safety-related equipment. The applicant identified seven systems with structures and components that define the FHE and HLHC that are within scope for license renewal: (1) spent fuel storage (spent fuel pool), (2) refueling pool, (3) new fuel storage and elevator, (4) spent fuel cask washing pit, (5) fuel transfer tube, (6) fuel handling system, and (7) cranes. These major systems are described as follows:
| (1) | Spent Fuel Storage System: The CCNPP Units 1 and 2 spent fuel storage system (SFSS), or spent fuel pool (SFP), is located in the auxiliary building and consists of the SFP, the spent fuel shipping cask pit (within the SFP), the spent fuel shipping cask support platform, the SFP work platform, and SFP storage racks. |
| |
| |
| |
| |
| |
| (2) | Refueling Pool: CCNPP's refueling pool is constructed of reinforced concrete and lined with SS. It is located around the upper portion of the reactor vessel and filled with water from the refueling water storage tank by the SFP cooling pumps. The refueling pool is connected to the SFP by the fuel transfer tube, the safety injection system, and the spent fuel pool cooling system. |
| (3) | New Fuel Storage System and Elevator: The new fuel storage system consists of the new fuel dry storage racks and the new fuel inspection machine (new fuel storage inspection platform). It does not include the new fuel elevator which is part of the fuel handling system discussed under item 6 below. New fuel is removed from its shipping cask using the spent fuel cask handling crane and transferred to the storage racks. Each rack provides storage for 144 fuel assemblies (two-thirds of a core). New fuel is stored in the SFP as space allows. The new fuel inspection machine is located near the new fuel storage area. The new fuel inspection machine is designed to automatically check the straightness and sectional size of a fuel bundle through its full length. |
| (4) | Spent Fuel Cask Washing Pit: The spent fuel cask washing pit is constructed of reinforced concrete lined with SS and provides for storage and decontamination of spent fuel transfer/shipping casks. (This component is evaluated in Section 3.3E of Appendix A to the LRA.) |
| (5) | Fuel Transfer Tube: The fuel transfer tube connects the refueling pool with the SFP and accommodates the transfer of fuel between the two areas. (This component is evaluated in Section 3.3A of Appendix A to the LRA.) |
| (6) | Fuel Handling System: The fuel handling system contains those components used to move fuel from the time new fuel is received until the spent fuel is stored in the SFP. The system includes (a) the new fuel elevator, (b) the spent fuel handling machine, (c) fuel upending machines, (d) the transfer carriage, (e) the reactor refueling machine, and (f) the spent fuel inspection elevator. These components are described as follows: |
| |
| |
| |
| |
| |
| |
| (7) | Cranes: The crane system is described as all cranes, monorails, and hoisting and jib equipment at CCNPP. The applicant stated that there are approximately 85 cranes in the plant and grouped them into three types: overhead gantry cranes, monorail systems and underhung cranes, and overhead hoists. The applicant further grouped the components of the cranes into mechanical components and electrical components. The mechanical components include overhead monorail systems, cranes, monorail tracks, carriers or trolleys, motor-driven electric hoist carriers, gears, hoists, hooks, bridges, and lift-drop sections. Electrical components include motors, connectors, contacts, electric lift and drop sections, motor starters, and control panels. The applicant also identified the specially designed structural load handling devices such as the lifting rig for the reactor vessel cooling shroud and the reactor vessel head (reactor vessel internals system) as structural components in the crane system. |
As noted above, two of the systems identified as within scope for license renewal are addressed in other sections of Appendix A to the LRA.
In the LRA, the applicant identified the following intended functions for the above noted structures and components in the FHE and HLHC based on the requirements of 10 CFR 54.4(a)(1) and (2):
The applicant also determined that there are no intended functions of the FHE and HLHC based on the requirements of 10 CFR 54.4(a)(3).
On the basis of its evaluation of the structures and components that provide the intended functions noted above, the applicant identified a total of 57 structural components/ subcomponents that are within the 5 systems and/or structures and components that constitute the FHE and HLHC and are within scope for license renewal and subject to an AMR.
As discussed in the LRA and the UFSAR, the FHE and HLHC structural components are designated as safety-related and are designed to meet seismic Category I criteria because they must remain functional before, during, and after a safe-shutdown earthquake. Therefore, most of FHE and HLHC structural components perform the first and second intended functions noted above. For example, the SFP is designed to maintain structural integrity during a seismic event in order to support spent fuel in the SFP. Also, the SFP storage racks are designed to withstand all anticipated loadings and are separated in such a manner as to preclude a reduction in separation space under either operating-basis or safe-shutdown earthquake.
In addition, the applicant cited five major cranes in the crane system that handle heavy loads that are functionally not safety-related, but are considered safety-related because they are used to handle heavy loads in the vicinity of the reactor vessel, near spent fuel in the SFP, or in areas in which, if a load is dropped, could damage safe-shutdown or decay-heat-removal equipment. These cranes are the polar crane, the intake structure semi-gantry crane, the transfer jib machine crane, the containment purge exhaust monorail hoist, and the spent fuel cask handling crane (SFCHC).
These cranes are categorized as seismic Category I/II and satisfy the intended functions as noted above. The SFCHC crane (auxiliary building crane) is also designed in accordance with the single-failure-proof criteria in NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power Plants," and NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."
In Table 3.2-1 of Appendix A to the LRA, the applicant listed 48 of the 57 components and subcomponents that are identified for an AMR. The remaining 9 structures and components are structural-type components that are addressed in Section 3.3 of Appendix A to the LRA where they are treated for their intended functions as part of the buildings in which they are housed. Those 9 components are (1) polar crane girders, (2) spent fuel cask handling crane rail/support girders, (3) refueling pool reinforced concrete, (4) refueling pool SS liner, (5) fuel transfer tube SS liner, (6) spent fuel pool reinforced concrete, (7) spent fuel pool SS liner, (8) spent fuel pool storage racks, and (9) new fuel storage racks.
2.2.3.3.2 Staff Evaluation
The staff reviewed Section 3.2 of Appendix A to the LRA to determine whether there is reasonable assurance that the applicant has appropriately identified the FHE and HLHC components and supporting structures that are within the scope of license renewal in accordance with 10 CFR 54.4 and subject to an AMR in accordance with the requirements of 10 CFR 54.21(a)(1).
2.2.3.3.2.1 Fuel Handling Equipment and Other Heavy Load Handling Cranes Within the Scope of License Renewal
The staff reviewed Section 9.7, "Fuel and Reactor Component Handling Equipment," of the UFSAR to determine if there were any additional portions of the structure and other components that the applicant should have identified as within the scope of license renewal. The staff also reviewed Section 9.7 of the UFSAR for any safety-related functions that were not identified as intended functions in the LRA to verify that no structure or component having an intended function was omitted from the scope of the rule.
The staff has reviewed the information presented in Section 3.2 of Appendix A to the LRA and Section 9.7 of the UFSAR. Table 3.2-1 of Appendix A to the LRA shows that all of the FHE and HLHC structures and components that comprise the 48 structural component types within the scope of license renewal require an AMR. Upon completing the initial review, the staff issued RAIs by letter dated August 26, 1998, regarding the FHE and HLHC commodity report. By letter dated November 4, 1998, the applicant responded to the staff's RAIs. As documented by a letter from BGE to NRC, dated February 4, 1999, an additional component type, the containment purge exhaust monorail, was added to the list of components that are within the scope of license renewal and subject to an AMR. In addition, the HLHC carbon steel chain hoist for the containment purge exhaust monorail is identified as a subcomponent that is within the scope of license renewal and subject to an AMR. The staff agrees that this non-safety-related component does perform the intended functions as defined in 10 CFR 54.4(a)(1), (2), and (3), and is within the scope of license renewal. On the bases discussed above, the staff finds that there is reasonable assurance that the applicant has appropriately identified the portions of the FHE and HLHC and the associated structures and components thereof that are within the scope of license renewal in accordance with the requirements of 10 CFR 54.4.
2.2.3.3.2.2 Fuel Handling Equipment and Other Heavy Load Handling Cranes Subject to an Aging Management Review
In accordance with the license renewal rule, the following structures and components are subject to an AMR: (1) those that perform an intended function without moving parts or without change in configuration or properties, and (2) those that are not subject to periodic replacement based on a qualified life or specified time period.
The applicant's process determined that some structural devices, such as drums, hydraulic cylinders, and wheels, perform their intended function(s) while in motion. Such devices were considered to be active subcomponents and were eliminated from an AMR. It was assumed that no structural components or subcomponents in the fuel handling equipment (FHE) and heavy load handling cranes (HLHCs) were replaced on the basis of time or qualified life.
On the basis of the results of the process described above, the portion of the FHE and HLHCs that is within the scope of license renewal and subject to an AMR includes 57 structural components and their supports.
The following FHE and HLHC components are addressed for their structural intended function(s) as parts of the building in which they are housed in Section 3.3 of Appendix A to the LRA, and are, therefore, not reviewed in this section:
The remaining 48 components, listed in Table 3.2-1 in Appendix A to the LRA are subject to an AMR and are evaluated within this section. The staff reviewed the information submitted by the applicant and verified that the grouping was correct. Therefore, the staff finds that there is reasonable assurance that the applicant has appropriately identified the structures and components subject to an AMR for the FHE and HLHC's in accordance with the requirements of 10 CFR 54.21(a)(1).
2.2.3.4 Primary Containment Structure
In Section 3.3A, "Primary Containment Structure," of Appendix A to the LRA, the applicant describes portions of the primary containment and the components therein that are within the scope of license renewal, and identified which of those within-scope components are subject to an AMR.
2.2.3.4.1 Summary of Technical Information in the Application
As described in Appendix A to the LRA, the primary containment is designed to withstand an internal pressure of 50 psig with a coincident concrete surface temperature of 276º F, and to limit leakage to no more than 0.20 percent by weight per day at the design temperature and pressure. The containment structure is designated a seismic Category I structure and is designed for all loading combinations described in Section 5A.3 of the UFSAR. The primary containment consists of two categories of components -- the containment structure and the containment system. The containment structure embraces the majority of structural components, such as beams, columns, walls, and liners. The containment system covers penetrations, hatches, air locks, and associated instrumentation.
In Appendix A to the LRA, the applicant identified the following intended functions for the primary containment in accordance with 10 CFR 54.4(a)(1) and 54.4(a)(2):
The applicant also determined that the following were intended functions of the primary containment according to the requirements of 10 CFR 54.4(a)(3):
On the basis of the intended functions stated above, the applicant identified a total of 37 structural component types as being within the scope of license renewal. These structural component types were further combined into the following 4 structural component categories on the basis of their design and materials: (1) concrete, (2) structural steel, (3) architectural, and (4) unique (e.g., post-tensioning system, basemat and containment liner, permanent cavity seal ring, trisodium phosphate baskets, and emergency sump cover and screen). The applicant identified all 37 structural component types as subject to an AMR. The applicant identified the following 3 component types for the containment system: (1) air locks and equipment hatch, (2) containment penetrations, and (3) limit switches. Of these 3 component types, the applicant identified 2 as subject to an AMR.
The applicant also indicated that some components in the containment system that are common to many systems have been included in the separate commodity reports that address those components for the entire plant. Therefore, the following components were not included in the individual system sections:
2.2.3.4.2 Staff Evaluation
The staff reviewed Section 3.3A of Appendix A to the LRA to determine whether there is reasonable assurance that the applicant has appropriately identified the primary containment components and supporting structures within the scope of license renewal in accordance with 10 CFR 54.4 and subject to an AMR in accordance with the requirements of 10 CFR 54.21(a)(1).
2.2.3.4.2.1 Systems, Structures, and Components Within the Scope of License Renewal
The staff reviewed Section 5.1, "Containment Structure," of the UFSAR and compared the description of the structures and components in the UFSAR to the description in the application to determine if there were any portions of the structure, and other components, that the applicant should have identified as within the scope of license renewal. The staff also reviewed Section 5.1 to determine if there are any safety-related functions that were not identified as intended functions in the LRA to determine if there are any structures or components with intended functions that might have been omitted from the scope of license renewal. On the basis of its review, the staff found that the applicant did not omit anything.
Table 3.3A-1 of Appendix A to the LRA shows that all of the containment structure components that comprise the 37 structural component types within the scope of license renewal also require an AMR. As mentioned in Section 2.2.3.17.2.1 of this SER, the containment sump, trisodium phosphate baskets, and the emergency sump cover and screens were adequately identified in Table 3.3A-1 as requiring an AMR. Only one of the three component types within the scope of license renewal for the containment system did not require an AMR. The component type, limit switches, was found to only support the active function of providing closure of the containment air lock and access/egress hatches during a station blackout. In performing their functions, limit switches change configuration; therefore, the limit switches do not require an AMR. The remaining component types requiring an AMR are shown in Table 3.3A-2 of Appendix A to the LRA. On the basis of the components identified in the tables referenced above and the supporting information in Section 5.1 of the USAR, the staff concludes that those portions of the primary containment structure that are not identified as within the scope of license renewal do not perform any intended functions.
As noted above, the staff has reviewed the information in Section 3.3A of Appendix A to the LRA and Section 5.1 of the USAR. On the basis of that review, the staff finds that there is reasonable assurance that the applicant has appropriately identified the portions of the primary containment and the associated structures and components thereof that are within the scope of license renewal in accordance with the requirements of 10 CFR 54.4.
2.2.3.4.2.2 Primary Containment Structures Subject to an Aging Management Review
Of 45 component types within the scope of the license renewal rule, 37 are structural component types and are identified in Table 3.3A-1. The remaining 8 are system component types, 7 of which are identified in Table 3.3A-2, and the eighth is a limit switch. The staff reviewed the component types that are electrical/instrumentation components to verify that the applicant did not miss any electrical/instrumentation components that should be subject to an AMR. The applicant classified the limit switch as having only an active function and, therefore, not requiring an AMR. Electrical control/power cabling is evaluated in Section 2.2.3.32, "Cables," of this SER. One electrical/instrumentation component, electrical penetrations, evaluated in this section was classified as subject to an AMR. The staff agrees with this BGE determination covering electrical/instrumentation components, which is consistent with 10 CFR 54.21(a)(1).
Some components in the containment system are common to many other plant systems (e.g., structural supports for piping, cables, electrical control, and power cabling) and have been discussed by the applicant in separate sections of the LRA that address those components as commodities for the entire plant.
On the basis of the applicant's integrated plant assessment (IPA) methodology provided in Appendix A to the LRA and provisions of 10 CFR 54.21(a)(1), the applicant identified 44 component types for the containment structure and component system as components subject to an AMR, and listed these component types in Tables 3.3A-1 (37 structural type components) and 3.3A-2 (7 system type components) of Appendix A to the LRA.
The staff focused its evaluation of the applicant's approach for defining the applicability of an AMR for the containment structure and containment system on the issue of whether the requirements and intent of 10 CFR 54.4 and 54.21(a)(1) are fully complied with. The staff reviewed each of the 44 component types noted above for the containment structure and containment system to verify that these items are part of the containment structure and the containment system. The staff further verified that the applicant had not omitted any items from an AMR that are part of the containment structure and containment system, and that perform an intended function without moving parts or without a change in configuration or properties and are not subject to replacement based on a qualified life or specified time period. The staff also reviewed the manner in which the applicant handled some components in the containment system that are common to many other plant systems and have been reviewed by the applicant in separate sections of the LRA, that address those components as commodities for the entire plant. On the basis of the review described above, the staff concludes that the applicant has implemented an adequate procedure for defining structural and system component types for the CCNPP containment structure and the containment system that are subject to an AMR, because the applicant's approach included 100 percent of the structural and system component types that constitute the CCNPP containment structure and the containment system.
Table 3.3A-1, "Containment Structure Component Types Requiring an AMR," in Appendix A to the LRA designates the containment structural components subject to an AMR. The containment tendon gallery protects the bottom anchorages of the vertical tendons, and gives access to the tendon anchorages for inservice inspection activities. The tendon gallery is categorized as a non-safety-related element of the containment structures. BGE indicated that the tendon gallery is not relied upon for containment integrity in the seismic analyses or design-basis events. Documentation of this basis for excluding the tendon gallery from the scope of the structural elements subject to an AMR was identified as Confirmatory Item 2.2.3.4.2.2-1 in the previous SER.
In its July 2, 1999, response to this issue, the applicant committed to perform the aging management of tendon anchorages through procedures STP-M-663-1and -2, (Containment Tendon Surveillance Tests) and MN-1-319 (Structures and Systems Walkdown). The staff concludes that these procedures provide adequate monitoring of the condition of the tendon anchorages and allow detection of anchorage degradation in sufficient time to correct the degradation before the intended function is compromised. On this basis, the staff concludes that the applicant has provided an acceptable basis to exclude the tendon galleries from an AMR and considers Confirmatory Item 2.2.3.4.2.2-1 closed. The staff notes that managing the condition and environment in the tendon galleries (e.g., moisture and humidity) may be a prudent way to manage the degradation (i.e., corrosion) of bearing plates and other vertical tendon anchorage components in the tendon galleries.
The staff finds that there is reasonable assurance that the applicant has appropriately identified the structural and system component types for the primary containment structure that are subject to an AMR pursuant to 10 CFR 54.21(a)(1).
2.2.3.5 Turbine Building Structure
In Section 3.3B, "Turbine Building Structure," of Appendix A to the LRA, the applicant described the turbine building and noted the components that are within the scope of license renewal, and identified which of those components are subject to an AMR.
2.2.3.5.1 Summary of Technical Information in the Application
As described in the LRA, the turbine building is within the scope of license renewal because its structural components perform one or more of the following generic functions:
In Section 3.3B.1 of Appendix A to the LRA, the applicant described the turbine building, including the conceptual boundaries, and listed the intended functions performed by its structural components. The applicant then identified the structural component types within the scope of license renewal. Finally, the components subject to an AMR were identified and dispositioned in accordance with the integrated plant assessment methodology described in Section 2.0 of Appendix A to the LRA.
The turbine building for the CCNPP is common to both units and is oriented parallel to the Chesapeake Bay shoreline between the North Service Building and the auxiliary building. It is a steel structure with metal siding supported on reinforced-concrete foundations. The turbine building is a seismic Category II structure. The conceptual boundary of the turbine building includes the AFW pump rooms and portions of the electrical ductbanks that are seismic Category I structures. Since the seismic Category I structures are enclosed within the turbine building that serves such intended functions as providing support and shelter to safety-related equipment, the turbine building and its enclosures are within the scope of license renewal.
The electrical ductbanks that run under the turbine building are connected between the AFW pump rooms and the intake structure. These ductbanks are seismic Category I reinforced-concrete structures that encase the safety-related electrical conduits. The siding on the turbine building wall is not safety-related, but the siding clips that hold the siding in place are safety- related. The siding clips are designed to fail when a differential pressure across the siding reaches a pre-determined pressure, which allows the siding to blow off for venting blowdown pressure following an accident and protects vital equipment and structures within the turbine building. The wall at the end of the main steam pipe tunnel that separates the turbine building and the auxiliary building is designed to fail at 0.5 psi to release pressure if a main steam line breaks near the main steam pipe tunnel. The wall is also designed to fail at a hydraulic pressure of 3 feet of water from a main feedwater line rupture in the main steam piping area.
The applicant identified that the turbine building and the AFW pump rooms are within the scope of license renewal according to 10 CFR 54.4(a). Six of the seven generic structural functions (except for the pressure boundary for fission products) listed in Table 3.3B-1 of Appendix A to the LRA are the intended functions for the turbine building and the AFW pump rooms. As described in the IPA, the applicant developed a generic list of component types for use during the structural component scope task. On the basis of this generic list, the applicant determined 24 structural component types for the turbine building (as listed in Table 3.3B-2 of Appendix A to the LRA) that identify such structural components as walls, slabs, and equipment pads, which do not have unique equipment identifiers in the site equipment database. These structural component types were combined into the following four structural categories on the basis of their design and material:
The structural component types identified for the turbine building contribute at least one of the structural intended functions discussed in the LRA. For example, the electrical ductbanks that run under the turbine building have been identified as structural components under the category of concrete components and are included in the turbine building conceptual boundary because they are seismic Category I. The turbine building siding clips and retainer clips are identified as structural components under the category of architectural components because they are safety related. These structural components that fall within the scope of license renewal are functionally passive and are not subject to periodic replacement. All the structural components listed in Table 3.3B-2 of Appendix A to the LRA are subject to an AMR and are evaluated in this section.
Component supports that are connected to structural components in the turbine building are evaluated in Section 3.1 of Appendix A to the LRA under the component support commodity evaluation. A component support is defined as the connection between a system (or component within a system) and a plant structural member. Component supports interface with the component they support in the applicable systems and interface with the structural component to which they are attached. For example, a fixed base that supports a pump is considered a component support since it connects the concrete equipment pad to the pump. The pump itself would be included and evaluated within the associated system in Appendix A to the LRA. The fixed base would be included within the component support commodity evaluation, and the concrete equipment pad would be included within the evaluation for the associated structure. If anchor bolts are used at the interface with the structural member, there is overlap between the component support commodity evaluation and the evaluation for the structural component. Evaluations for structural components considered the effects of aging caused by the surrounding environment; the component support commodity evaluation considered the effects of aging caused by the supported equipment (thermal expansion, rotating equipment, etc.) as well as by the surrounding environment. Supports for structural components such as platform hangers are not "component supports" in this sense because any support for a structural component is itself a structural component (i.e., is included in the scope of the associated structure). All the component supports in the turbine building are evaluated in Section 3.1 of Appendix A to the LRA.
2.2.3.5.2 Staff Evaluation
The staff reviewed Section 3.3B of Appendix A to the LRA to determine whether there is reasonable assurance that the applicant has appropriately identified the turbine building structural components that are within the scope of license renewal in accordance with 10 CFR 54.4 and subject to an AMR in accordance with the requirements of 10 CFR 54.21(a)(1).
2.2.3.5.2.1 Systems, Structures, and Components Within the Scope of License Renewal
As part of the first-step evaluation (i.e., to determine whether the applicant has properly identified the systems, structures, and components within the scope of license renewal), the staff reviewed portions of the UFSAR, including the layout drawings for the turbine building, the AFW pump rooms, and the ductbanks, and compared them with the structural components listed in Table 3.3B-2 and shown in Figure 3.3B-1 in Appendix A to the LRA to determine if there were any portions of the structures and associated components that the applicant did not identify as within the scope of license renewal. The staff also reviewed the UFSAR to determine if there were any safety-related system functions that were not identified as intended functions in the LRA to determine if there were any structural components having intended functions that might have been omitted from consideration within the scope of license renewal. Although the staff found no omissions, the staff questioned why the turbine building roof trusses were described in the Structural Description portion of Section 3.3B, but not included in Table 3.3B-2, "Structural Component Types Requiring AMR for the Turbine Building."
During a site visit to the CCNPP on February 18, 1999 (summarized in an NRC letter dated March 19, 1999), the staff asked the applicant why the roof trusses were not subject to an AMR. The applicant stated that the roof trusses are not within the scope of license renewal. The applicant explained that the roof trusses are seismic Category II structures, but their failure during an abnormal (e.g. seismic) event could not affect the operability of any safety-related equipment in the turbine building. Therefore, the roof trusses do not meet the scoping criteria of 10 CFR 54.4. The staff reviewed the information and agreed that the roof trusses are not within scope.
On the basis of this review, the staff finds that there is reasonable assurance that the applicant has appropriately identified the structural components of the turbine building and the AFW pump rooms that are within the scope of license renewal in accordance with the requirements of 10 CFR 54.4.
2.2.3.5.2.2 Turbine Building Structure Subject to an Aging Management Review
The staff determined whether the applicant has properly identified the structural component types of the turbine building subject to an AMR from among all of the structural component types in the turbine building. The applicant identified 24 structural component types under 4 structural component categories for the turbine building in Table 3.3B-2 in Section 3.3B of Appendix A to the LRA. In the "concrete" category, the structural components are walls, ground floor slabs and equipment pads, elevated floor slabs, cast-in-place anchors/embedments, ductbanks, grout, fluid-retaining walls and slabs, and post-installed anchors. In the "structural steel" category, the structural components are beams, baseplates, floor framing, platform hangers, decking, jet impingement barriers, floor grating, and stairs and ladders. In the "architectural components" category, the structural components are building siding clips, retainer clips, fire doors, jambs, hardware, and caulking and sealants. In the "unique components" category, the structural components are watertight doors, pipe whip restraints, and pipe encapsulations. The staff reviewed the list of 24 structural component types within the scope of license renewal and determined that they perform their intended functions without moving parts or changes in configuration, and are not replaced on a periodic basis.
Based on this review, the staff finds that since all 24 structural component types within the scope of license renewal are subject to an AMR, there is reasonable assurance that the applicant has identified the structural components subject to an AMR in accordance with 10 CFR 54.21(a)(1).
2.2.3.6 Intake Structure
In Section 3.3C, "Intake Structure," of Appendix A to the LRA, the applicant described the technical information related to the intake structure at the plant site. The staff reviewed this section of the application to determine if there is reasonable assurance that the applicant has identified and listed those structures and components of the intake structure that are subject to an AMR to meet the requirements stated in 10 CFR 54.21(a)(1).
2.2.3.6.1 Summary of Technical Information in the Application
As described in the LRA, the intake structure is situated to the east of the main plant between the North Service Building and the Chesapeake Bay shoreline. The structure houses 12 circulating water pumps that supply water from the Chesapeake Bay to the condensers, and 6 saltwater pumps that provide cooling water to various plant equipment. Trash racks and traveling screens protect the condensers from foreign bodies present in the bay water. A gantry crane, having a