ML040200983
| ML040200983 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 01/20/2004 |
| From: | Mitra S Office of Nuclear Reactor Regulation |
| To: | Moyer J Carolina Power & Light Co |
| Mitra, SK, NRR/DRIP/RLEP, 415-2783 | |
| Shared Package | |
| ML040200981 | List: |
| References | |
| Download: ML040200983 (29) | |
Text
1-1 1 Introduction and General Discussion 1.1 Introduction This document is a safety evaluation report (SER) on the application for license renewal for the H.B. Robinson Steam Electric Plant, Unit No. 2 (RNP), as filed by the Carolina Power & Light Company (CP&L or the applicant). By letter dated June 14, 2002, CP&L submitted its application to the U.S. Nuclear Regulatory Commission (NRC or the Agency) for renewal of the RNP operating license for an additional 20 years. The NRC staff (the staff) prepared this report which summarizes the results of its safety review of the renewal application for compliance with the requirements of Title 10, Part 54 of the Code of Federal Regulations (10 CFR Part 54),
Requirements for Renewal of Operating Licenses for Nuclear Power Plants. The NRC license renewal project manager for the RNP license renewal review is Mr. Sikhindra K. Mitra. Mr.
Mitra may be contacted by calling 301-415-2783, or by writing to the License Renewal and Environmental Impacts Program, U.S. Nuclear Regulatory Commission, Mail Stop O-11F1, Washington, D.C. 20555-0001.
In its June 14, 2002, submittal letter, CP&L requested renewal of the operating license issued under Section 104b of the Atomic Energy Act of 1954, as amended, for RNP (License Number DPR-23) for a period of 20 years beyond the current license expiration date of July 31, 2010.
RNP is adjacent to Unit 1 of the H.B. Robinson Steam Electric Plant, a coal-fired steam power plant. RNP is located on Lake Robinson, a man-made lake in Darlington and Chesterfield Counties, South Carolina. RNP is a pressurized light-water-moderated and cooled system.
The nuclear power plant incorporates a three-loop closed-cycle, pressurized water, nuclear steam supply system (NSSS) designed by Westinghouse Electric Corporation and licensed to generate 2339 Mw-thermal, or approximately 769 Mw-electric. Details concerning the plant and the site are found in the Updated Final Safety Analysis Report (UFSAR) for RNP.
The license renewal process proceeds along two tracksa 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 RNP license renewal is based on the applicants license renewal application (LRA), RNP UFSAR and on the answers to requests for additional information (RAIs) from the staff. In meetings and docketed correspondence, the applicant has also supplemented its LRA and answers to the RAIs. The LRA and all pertinent information and materials, including the UFSAR mentioned above, are available to the public for review at the NRC Public Document Room, 11555 Rockville Pike, Room 1-F21, Rockville, MD 20852-2738 (301-415-4737/800-3974209). Material related to the LRA is also available through the NRCs website, at www.nrc.gov.
This SER summarizes the results of the staffs safety review of the RNP LRA and delineates the scope of the technical details considered in evaluating the safety aspects of RNPs proposed operation for an additional 20 years beyond the term of the current operating license.
The LRA was reviewed in accordance with the NRC regulations and the guidance provided in NUREG-1800, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, dated July 2001 (SRP-LR).
Sections 2 through 4 of the SER address the staffs review and evaluation of license renewal issues that have been considered during the review of the application. Section 5 is reserved for
1-2 the report of the Advisory Committee on Reactor Safeguards (ACRS). The SER conclusions are in Section 6 of this document.
Appendix A of this SER is a table that identifies the applicants commitments associated with the renewal of the operating license. Appendix B is a chronology of the NRCs and the applicants principal correspondence related to the review of the applications. Appendix C is a list of the NRC staff's principal reviewers and its contractors for this project. Appendix D is a list of the major references used in support of this SER.
In accordance with 10 CFR Part 51, the staff prepared a draft for comment on the plant-specific supplement to the Generic Environmental Impact Statement (GEIS) that discusses the environmental considerations related to renewing the license for RNP. NUREG-1437, Supplement 13, the plant-specific draft supplement to the GEIS, was issued on May 5, 2003.
The final supplement to the GEIS was issued in December 2003.
1.2 License Renewal Background Pursuant to the Atomic Energy Act of 1954, as amended, and NRC regulations, licenses for the operation of commercial power reactors 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, rather than technical limitations. However, some plant equipment may have been designed on the basis of an expected 40-year service life.
In 1982, the NRC anticipated interest in license renewal and held a workshop on the aging of nuclear power plants. This workshop led the NRC to establish a comprehensive program for nuclear plant aging research (NPAR). As a result of this research, a technical review group concluded that many aging phenomena are readily manageable and do not involve technical issues that would preclude extending the life of nuclear power plants. In 1986, the NRC published a request for comments regarding a policy statement on major policy, technical, and procedural issues related to license renewal for nuclear power plants.
In 1991, the NRC published a license renewal rule in 10 CFR Part 54 (the Rule). The NRC participated in an industry-sponsored demonstration program to apply the Rule to a pilot plant and to 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 10 CFR Part 54 established a regulatory process that is simpler, more stable, and more predictable than the previous license renewal rule. In particular, 10 CFR Part 54 was amended to focus on managing the adverse effects of aging rather than on identifying age-related degradation unique to license renewal. The rule changes were intended to ensure that important systems, structures, and components (SSCs) will continue to perform their intended functions 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 (SCs).
1-3 In parallel with these efforts, the NRC pursued a separate rulemaking effort, 10 CFR Part 51, to focus the scope of the review of the environmental impacts of license renewal, in fulfillment of the NRCs responsibilities under the National Environmental Policy Act of 1969 (NEPA).
1.2.1 Safety Review License renewal requirements for power reactors are based on two 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 system, structures, and components during the period of extended operation and a few other safety issues.
(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, 10 CFR 54.4 defines the scope of license renewal as including those plant SSCs (a) that are safety related, (b) nonsafety related whose failure could affect safety-related functions, and (c) that are relied on to demonstrate compliance with the Commissions regulations for fire protection (FP), environmental qualification (EQ), pressurized thermal shock (PTS), anticipated transients without scram (ATWS), and station blackout (SBO).
Pursuant to 10 CFR 54.21(a), an applicant for a renewed license must review all SSCs within the scope of the Rule to identify SCs that are subject to an aging management review (AMR).
SCs subject to an AMR are those that perform an intended function without moving parts or without a change in configuration or properties, and that are not subject to replacement based on a qualified life or a specified time period. As required by 10 CFR 54.21(a), an applicant for a renewed license must demonstrate that the effects of aging will be managed in such a way that the intended functions of the SCs within the scope of license renewal will be maintained, consistent with the current licensing basis (CLB), 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 effects of aging on 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 plant design and licensing basis, are required throughout the period of extended operation. Section 54.21(d) of the Rule requires that a supplement to the final safety analysis report (FSAR) contain a summary description of the programs and activities for managing the effects of aging be submitted by the applicant.
Another requirement for license renewal is the identification and updating of time-limited aging analyses (TLAAs). During the design phase for a plant, certain assumptions are made about the initial operating term of the plant, and these assumptions are incorporated into design calculations for some of the plants 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 projected to the end of the period of extended operation, or the applicant must demonstrate that the effects of aging of these SSCs will be adequately managed for the period of extended operation.
1-4 In 2001, the NRC developed and issued Regulatory Guide (RG) 1.188, Standard Format and Content for Applications to Renew Nuclear Power Plant Operating Licenses. This guide endorses an implementation guideline prepared by the Nuclear Energy Institute (NEI) as an acceptable method of implementing the license renewal rule. The NEI guideline, issued in March 2001, is NEI 95-10, Revision 3, Industry Guideline for Implementing the Requirements of 10 CFR Part 54The License Renewal Rule. The NRC also prepared the SRP-LR which, along with the RG 1.188, was used to review this application.
CP&L utilizes the process defined in NUREG-1801, Generic Aging Lessons Learned (GALL)
Report, dated July 2001. The purpose of GALL is to provide the staff with a summary of staff-approved aging management programs (AMPs) for the aging of most structures and components that are subject to an AMR. If an applicant commits to implementing these staff-approved AMPs, the time, effort, and resources used to review an applicants LRA will be greatly reduced, thereby improving the efficiency and effectiveness of the license renewal review process. The GALL Report summarizes the aging management evaluations, programs, and activities credited for managing aging for most of the structures and components used throughout the industry, and serves as a reference for both applicants and staff reviewers to quickly identify those aging management programs and activities that the staff has determined will provide adequate aging management during the period of extended operation.
1.2.2 Environmental Review The environmental protection regulation, 10 CFR Part 51, was revised in December 1996 to facilitate the environmental review for license renewal. The staff prepared a GEIS 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 into 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 staff performed a plant-specific review of the environmental impacts of license renewal, including whether new and significant information existed that was not considered in the GEIS. As part of the NRC environmental scoping process, a public meeting was held near RNP on September 25, 2002, in Hartsville, SC, 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 the NRCs draft plant-specific supplement to the GEIS for RNP, which was issued by the NRC in May 2003. After considering comments on the draft, the NRC prepared NUREG-1437, Supplement 13, Generic Environmental Impact Statement for License Renewal of Nuclear Power Plants, which was published in December 2003.
1-5 1.3 Principal Review Matters 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 RNP LRAs in accordance with Commission guidance and the requirements of 10 CFR Part 54. The standards for renewing a license are contained in 10 CFR 54.29. This SER describes the results of the staffs safety review.
In 10 CFR 54.19(a), the Commission requires a license renewal applicant to submit general information. The applicant provided this general information in Section 1 to its letter of June 14, 2002, forwarding its applications for renewed operating licenses for H.B. Robinson Steam Electric Plant Unit 2. The staff reviewed Section 1 and found that the applicant submitted the information required by 10 CFR 54.19(a).
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. The applicant states the following in its LRA regarding this issue:
The current indemnity agreement for H.B. Robinson Steam Electric Plant Unit 2 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 Agreement to the indemnity agreement, as revised by Amendment No. 1, lists H.B. Robinson Operating License DPR-23. CP&L requests that conforming changes be made to the indemnity agreement, and/or the Attachment to that agreement, specifying the extension of agreement until the expiration date of the renewed H.B.
Robinson operating license as sought in this application. In addition, should the license number be changed upon issuance of the renewed license, CP&L requests that conforming changes be made to the Attachment and any other sections of the indemnity agreement as appropriate.
The staff intends to maintain the license type and number upon 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 renewed license for a nuclear facility must contain (a) an IPA, (b) a description of CLB changes during staff review of the application, (c) an evaluation of TLAAs, and (d) an FSAR Supplement. Sections 2, 3 and 4 of the LRA address the license renewal requirements of 10 CFR 54.21(a), (c), and (d),
respectively.
In 10 CFR 54.21(b), the Commission requires that each year following submittal of the application, and at least 3 months before the scheduled completion of the staffs review, an amendment to the renewal application must be submitted that identifies any change to the CLB of the facility that materially affects the contents of the license renewal application, including the FSAR Supplement. This information was provided by letter dated June 25, 2003. Therefore, the requirements of 10 CFR 54.21(b) have been met.
In 10 CFR 54.22, the Commission lists requirements regarding technical specifications. In Appendix D of the LRA, the applicant stated that no changes to the RNP technical specifications are necessary. This adequately addresses the requirements of 10 CFR 54.22.
1-6 The staff evaluated the technical information required by 10 CFR 54.21 and 10 CFR 54.22 in accordance with the NRCs regulations and the guidance provided by the SRP-LR. The staffs evaluation of the LRA, in accordance with 10 CFR 54.21 and 10 CFR 54.22, is contained in Sections 2, 3, and 4 of this report.
The staffs evaluation of the environmental information required by 10 CFR 54.23 is included in the draft, and the final plant-specific supplements to the GEIS state the considerations related to renewing the license for RNP. When the report of the ACRS, required by 10 CFR 54.25, is issued, it will be incorporated into Section 5 of this SER. The findings required by 10 CFR 54.29 are included as Section 6 of this report.
1.3.1 Westinghouse Topical Reports In the LRA the applicant referenced certain Westinghouse Commercial Atomic Power (WCAP) reports. In accordance with 10 CFR 54.17(e), the applicant referenced the following WCAP reports in the LRA:
WCAP-10322, Revision No. 1, Stress Report of 312 Standard Reactor Core Support Structures and Internal Structures Structural and Fatigue Analysis, October 1984 WCAP-12962, Supplement 1, Structural Evaluation of the H.B. Robinson Unit 2 and Shearon Harris Pressurizer Surge Lines, Considering the Effects of Thermal Stratification, October 1995 WCAP-13587, Revision No. 1, Reactor Vessel Upper Shelf Energy Bounding Evaluation for Westinghouse Pressurized Water Reactors, September 1993 WCAP-14209, Evaluation of the Effects of Insurge/Outsurge Transients of the Integrity of the Pressurizer at H.B. Robinson Unit 2, October 28, 1994 WCAP-15338, A Review of Cracking Associated with Weld Deposited Cladding in Operating PWR Plants, March 2000 WCAP-15363, Revision No. 1, A Demonstration of Applicability of ASME Code Case N-481 to the Primary Loop Pump Casings of H.B. Robinson Unit 2 for the License Renewal Program, July 2002 WCAP-15628, Technical Justification for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for the H.B. Robinson Unit 2 Nuclear Power Plant for the License Renewal Program, July 2001 The applicant states that in support of license renewal, a new report, WCAP-15363, Revision No. 1, was prepared. WCAP-15363, Revision No. 1, supercedes WCAP-15363, Revision 0, and includes an evaluation of the plant-specific pump casing material properties.
The safety evaluations of the topical reports are intended to be stand alone documents. An applicant that incorporates the topical reports by reference into an LRA must ensure that the conditions of approval stated in the safety evaluations are met. The staff's evaluation of the
1-7 applicants incorporation of the topical reports into the application is documented in Section 3 of this SER.
1.4 Interim Staff Guidance The license renewal program is a living program. The NRC staff, industry, and other interested stakeholders gain experience and develop lessons learned with each renewed license. The lessons learned address the NRCs performance goals of maintaining safety, improving effectiveness and efficiency, reducing regulatory burden, and increasing public confidence.
The lessons learned are captured in interim staff guidance (ISG) for use by the staff and interested stakeholders until the improved license renewal guidance documents are revised.
The current set of relevant ISGs that have been issued by the staff and the SER sections where the issues are addressed are provided below.
Purpose SER Section Station Blackout (SBO) Scoping (ISG-02)
The license renewal rule 10 CFR 54.4(a)(3) includes 10 CFR 50.63(a)(1)-SBO.
The SBO rule requires that a plant must withstand and recover from an SBO event. This includes recovery of offsite power.
The offsite power system should be included within the scope of license renewal.
2.5.4 3.6.2.4.3 3.6.2.4.4 3.6.2.4.5 Concrete Aging Management Program (ISG-03)
Lessons learned from the GALL Demonstration Project indicated that GALL is not clear whether concrete needs any AMPs.
3.5.2.4.1
1-8 Fire Protection System Piping (ISG-04)
To clarify staff position for wall thinning of FP piping system in GALL AMPs (XI.M26 and XI.M27).
New position is that there is no need to disassemble FP piping, as oxygen can be introduced in the FP piping which can accelerate corrosion. Instead, use nonintrusive method such as volumetric inspection.
Field service testing of sprinkler heads should be performed at 50 years and every 10 years after initial field service testing.
Eliminated Halon/carbon dioxide system inspections for charging pressure, valve line ups, and automatic mode of operation tests from GALL, as the staff considers these test verifications to be operational activities.
2.3.3.15 3.3.2.3.3.2
1-9 Identification and Treatment of Electrical Fuse Holder (ISG-05)
To include fuse holder AMR and AMP (i.e., same as terminal blocks and other electrical connections).
The position includes only fuse holders that are not inside the enclosure of active components (e.g., inside of switchgears and inverters).
Operating experience finds that metallic clamps (spring-loaded clips) have a history of age-related failures from aging stressors such as vibration, thermal cycling, mechanical stress, corrosion, and chemical contamination.
The staff finds that visual inspection of fuse clips is not sufficient to detect the aging effects from fatigue, mechanical stress and vibration.
3.6.2.3.1 1.5 Summary of Open Items As a result of its review of the LRA for RNP, including additional information submitted to the NRC through April 28, 2003, the staff identified the following issues that remained open at the time this report was prepared. An issue was open if the applicant had not presented a sufficient basis for resolution. Each open item has been assigned a unique identifying number. The items identified in this section have been properly closed by the technical staff.
Open Item 2.3.1.6-1 (steam generator feedrings)
The staff believes that the steam generator (SG) feedrings should be included in the scope of license renewal (Open Item 2.3.1.6-1). Since this component is completely enclosed by safety-related, pressure-boundary components, it is important to show that failures of this component could not impede certain safety-related functions of the components in which it is contained (10 CFR 54.4(a)(2)).
The possibility that loose parts might be generated and that they might prevent the accomplishment of certain safety functions of the steam generator is not, by itself, sufficient to require that the feedring be included in the scope for license renewal. There must be some basis in operating experience. The NEI guidelines indicated that the hypothetical failure (the loose part scenario) need not be considered, if it has not been previously experienced.
1-10 In response to a staff request for further information in RAI 2.3.1.6-1, RNP surveyed operating history experience compiled by the World Association of Nuclear Operators (WANO) and the Institute of Nuclear Power Operations (INPO) and found that there were no recorded instances of this type of failure. They did find, however, instances wherein J-tubes were replaced, due to corrosion problems, and an instance wherein there was direct leakage for the feedring. These can be considered to be preconditions to the loose part scenario. Therefore, the staff believes that the feedring should be within the scope of license renewal.
In a letter dated September 16, 2003 (ADAMS accession no. ML032650884), the applicant agreed to include the steam generator feedrings in the scope of the license renewal application.
The steam generator feedrings and their associated aging management program are discussed in Section 3.1.2.2.14 of this report. The staff reviewed the steam generator feedrings and their associated components that were subject to an AMR and found that the applicant has adequately included components of the steam generator feedrings, as required by 10 CFR 54.21(a)(1). Therefore, Open Item 2.3.3.6-1 is closed.
Open Item 2.3.3.8-1 (exclusion of deepwell pumps, piping, and valves from an AMR)
The staff requested the applicant to provide adequate justification for the exclusion of the deepwell pumps and associated piping from an AMR. The staff found that the applicant has not adequately justified the referred exclusion. The context of Section 10.4.8 of the UFSAR does not link dam failure to any particular set of initiating events, and seismic events and age-related degradation do not encompass all credible causes of dam failure. Dam failure results in loss of the ultimate heat sink and loss of the normal backup supply of feedwater from the service water system through the auxiliary feedwater system. Following dam failure and depletion of the condensate storage tank inventory, failure of the deepwell pumps would cause failure of the safety-related auxiliary feedwater system and prevent the residual heat removal (RHR) necessary to maintain a safe shutdown condition. Therefore, the deepwell pumps and associated piping are within the scope of license renewal (LR) in accordance with 10 CFR 54.4(a)(2). Therefore, the staff found that the applicant has not adequately justified excluding the deepwell pumps and associated piping and valves from an AMR, and this issue remains as Open Item 2.3.3.8-1.
By letter dated September 16, 2003, the applicant agreed to include, within the scope of license renewal, the three deepwell pumps and associated piping required to provide a backup source of water for the auxiliary feedwater system. The staff found that the applicant adequately identified components of the deepwell pumps and associated piping within the scope of license renewal, as required by 10 CFR 54.4(a)(2).
The applicant completed an AMR of the deepwell pumps and associated piping, which resulted in the identification of material/environment combinations not previously identified in the LRA for the primary and demineralized water makeup system. The applicant presented the results of the revised aging management evaluations in an update to LRA Table 2.3-14. The staff reviewed the components that were subject to an AMR and found that the applicant has adequately included components of the deepwell pumps and associated piping, as required by 10 CFR 54.21(a)(1). Therefore, Open Item 2.3.3.8-1 is closed.
1.6 Summary of Confirmatory Items
1-11 Confirmatory Items are items for which the staff and the applicant have reached a satisfactory resolution, but the resolution has not yet been formally submitted to the staff.
As a result of its review of the LRA for RNP, including additional information submitted to the NRC through April 28, 2003, the staff identified the following issues that remained confirmatory at the time this report was prepared.
Confirmatory Item 2.3.1.3-1 (pressurizer spray head)
The staff believed that the pressurizer spray head should be included in the scope of license renewal (RAI 2.3.1.3-1). Since this component is completely enclosed by safety-related, pressure-boundary components, it is important to show that its failure could not impede certain safety-related functions of the components in which they are contained (10 CFR 54.4(a)(2)).
The possibility of a failure in the pressurizer spray head, affecting the functioning of the PORVs or pressurizer safety valves, was noted. The applicant surveyed operating experience and concluded that such a failure had not occurred anywhere. The applicant provided supplemental information in support of a revised response to RAI 2.3.1.3-1. Pending the applicants formal submittal of this information and the NRC staffs review of the acceptability of the supplemental information, RAI 2.3.1.3-1 will be considered to be Confirmatory Item 2.3.1.3-1.
After reviewing the applicants response, the staff concluded that it was not necessary to include the pressurizer spray head in the license renewal scope to meet the requirements of either 10 CFR 54.4(a)(1) or 10 CFR 54.4(a)(3). Furthermore, the possibility of a failure in the pressurizer spray head, affecting the functioning of the PORVs or pressurizer safety valves, was postulated and considered under the terms of 10 CFR 54.4(a)(2). In accordance with the NEI guidelines in NEI 95-10, Revision 3, the staff requested CP&L to provide information to show that its hypothetical failure has not been experienced at RNP or at other plants. The applicant surveyed plant-specific and industry-wide operating experience and found that there were no known occurrences of the postulated failure scenario. Therefore, the staff concludes that inclusion of the pressurizer spray head in the license renewal scope is not required by 10 CFR 54.4(a)(2).
Confirmatory Item 2.3.2.5-1 (hydrogen recombiners and supporting components)
The staff considered the applicant's responses to RAIs 2.3.2.5-1, 2.3.2.5-2, and 2.3.2.5-3 to be unacceptable because they are incomplete. Although the responses provided sufficient information to demonstrate that 10 CFR 54.4(a)(1) and (a)(3) did not apply to the hydrogen recombiners and supporting components, they did not adequately demonstrate that these components were not within the scope of license renewal in accordance with 10 CFR 54.4(a)(2). Specifically, although ample time is available to effect hydrogen control, 10 CFR 54.4 does not explicitly permit components required for accident mitigation to be excluded from the scope of license renewal on that basis. In addition, although the response states that sufficient time exists to ensure that all components of the recombiner system are operable before its operation is required, UFSAR Section 6.2.5.2.2 indicates that the majority of the lines associated with this system cannot be repaired due to the high radiation rates present
1-12 during post accident conditions. As described further in Section 2.3.2.5.2 of this SER, the applicant has transmitted a revised draft response to these RAIs that would bring within scope the components of the hydrogen recombiner system that are necessary to fulfill the hydrogen control intended function. Pending the applicants formal submittal of this information and the NRC staffs review of the acceptability of the aging management results for the components that would be added within scope, RAIs 2.3.2.5-1, 2.3.2.5-2, and 2.3.2.5-3 are considered to be Confirmatory Item 2.3.2.5-1.
By a letter dated September 16, 2003, the applicant transmitted a revised response to these items that would bring within scope the components of the hydrogen recombiner system that are necessary to fulfill the hydrogen control intended function. Specifically, in addition to the components necessary for containment isolation, the response brings within scope the hydrogen recombiner, permanently installed piping, and temporary flexible piping associated with the post-accident hydrogen system pressure boundary, as well as the passive pressure boundary components of the associated nitrogen system that actuates the containment isolation valves which would permit the flow of containment atmosphere to and from the hydrogen recombiner. Based on the applicants decision to bring those components within scope of license renewal the staff finds the applicants responses to RAIs 2.3.2.5-1, 2.3.2.5-2, and 2.3.2.5-3 acceptable, and Confirmatory Item 2.3.2.5-3 is closed.
Confirmatory Item 2.3.3.9-1 (issued with regard to the exclusion from an AMR of the refueling water purification pump, piping, and valves necessary for spent fuel pool (SFP) makeup from the refueling water storage tank)
In discussions regarding the provision of makeup water to the spent fuel pool following loss of cooling, the applicant agreed to include components along the flow path from the refueling water storage tank (RWST) to the spent fuel pool within the scope of license renewal. The applicant indicated that a revised drawing highlighting the additional components added to the scope of license renewal and a revised list of components (including the purification pump casing, demineralizer vessel, and filter housing) that are subject to an AMR and the associated AMP would be transmitted by letter. This is Confirmatory item 2.3.3.9-1.
By letter dated August 14, 2003, the applicant formally agreed to include the SFP makeup path from the RWST to the SFP within the scope of license renewal, and described the specific boundaries of the components within the scope of license renewal. As a result of the expansion of the evaluation boundary, the applicant revised LRA Table 2.3-15 to include the SFP cooling demineralizer, SFP filter, and refueling water purification pump. The remainder of the piping components fell within existing commodity groups in LRA Table 2.3-15. The staff found that the formal description of the components subject to an AMR was consistent with the previous communication. Therefore, Confirmatory Item 2.3.3.9-1 has been resolved.
Confirmatory Item 3.0.3.2.2-1 (commitment inspections for the steam generator upper shell-to-transition cone weld)
The purpose of this item is to confirm that CP&L will commit to performing augmented inspections of the steam generator upper shell-to-transition cone weld during the two 10-year inservice inspection (ISI) intervals for the extended period of operation for RNP.
1-13 In a letter dated September 16, 2003, the applicant provided the following response to Confirmatory Item 3.0.3.2.2-1:
RNP will continue to perform examinations of the steam generator transition girth welds as required by ASME Section XI during the period of extended operation.
The applicants response to Confirmatory Item 3.0.3.2.2-1 confirms that the applicant will continue to perform the required ultrasonic examinations of the steam generator shell-to-transition cone girth welds during the two 10-year ISI intervals that are scheduled for the extended period of operation. This resolves Confirmatory Item 3.0.3.2.2-1 and Confirmatory Item 3.0.3.2.2-1 is closed.
Confirmatory Item 3.1.2.1-1, Parts 1 and 2 (issued with regard to the staffs assessment of AMR Item No. 22 of LRA Table 3.1-1, as evaluated in Section 3.1.2.1 of the SER)
The staff seeks confirmation as to whether or not there is any plant-specific or generic industry experience that supports the conclusion that crack initiation and growth due to stress corrosion cracking (SCC) is an applicable aging effect for carbon steel bolting materials in the reactor coolant system (RCS). If industry experience does support that crack initiation and growth due to SCC is an applicable aging effect for carbon steel bolting, the applicant should propose an AMP to manage this effect. This is Confirmatory Item 3.1.2.1-1, Part 1.
The applicants response to RAI 3.1.2.1-3 states that stress relaxation is not applicable to valve closure bolting in the reactor coolant pressure boundary (RCPB) (i.e., RCPB valve bolting) and other closure bolting in high pressure and high temperature systems. However, the applicants discussion for AMR 22 to LRA Table 3.1-1 states that the Bolting Integrity Program is applicable to all RCPB bolting except reactor vessel studs for which the Reactor Head Closure Studs Program applies, and that the Bolting Integrity Program relies on the ASME Section XI, Subsection IWB, IWC, and IWD Program to assure that aging effects associated with wear and stress relaxation are managed for RCS Class 1 closure bolting and for Class 2 bolting greater than 2 inches in diameter. The applicants discussion of AMR 22 in LRA Table 3.1-1 did not indicate that the applicant was exempting stress relaxation as an applicable aging effect for the RCPB valve bolting or other closure bolting in high pressure and high temperature systems. Therefore, the staff concludes that the applicants response to RAI 3.1.2.1-3, as it pertains to the management of stress relaxation in the RCPB valve bolting or other closure bolting in high pressure and high temperature systems, contradicts the applicants discussion of AMR 22 in LRA Table 3.1-1. The staff requests confirmation that, other than SCC, the aging effects identified in AMR 22 to LRA Table 3.1-1 are still applicable to the RCS bolting within the scope of the commodity group, other than the steam generator primary and secondary manway and handhole bolting. The applicant must explain the contradiction in the RAI response and the information in AMR 22 of LRA Table 3.1-1. This is Confirmatory Item 3.1.2.1-1, Part 2.
In a letter dated September 16, 2003, the applicant provided the following response to Confirmatory Item 3.1.2.1-1, Part 1:
The RNP Aging Management Review (AMR) has not identified plant-specific or generic industry experience which supports a conclusion that crack initiation and growth due to Stress Corrosion Cracking (SCC) is an applicable aging effect for carbon steel or low-
1-14 alloy steel bolting materials in the reactor coolant system (RCS). This is supported by operating experience and existing data which indicate that SCC failure should not be a significant issue for closure bolting within the RCS.
The applicants response to Confirmatory Item 3.1.2.1-1, Part 1, confirms that there has not yet been any RNP-specific or generic operating experience to support the conclusion that SCC-induced cracking is an aging issue for carbon steel bolting materials in ASME Class 1 systems.
The staff therefore concludes that SCC-induced cracking is not an aging effect requiring aging management for ASME Class 1 carbon steel bolting made from carbon steel materials. The staff therefore considers Confirmatory Item 3.1.2.1-1, Part 1, to be resolved, and Confirmatory Item 3.1.2.1-1, Part 1, is closed.
In the applicants response to Confirmatory Item 3.1.2.1-1, Part 2, dated September 16, 2003, the applicant explained their response to RAI 3.1.2.1-3 and confirmed that loss of preload due to stress relaxation is an applicable aging effect requiring aging management for the RCS bolting materials within the scope of AMR 22 in LRA Table 3.1-1. The staff therefore considers Confirmatory Item 3.1.2.1-1, Part 2, to be resolved, and Confirmatory Item 3.1.2.1-1, Part 2 is closed.
Confirmatory Item 3.1.2.1-1, Part 3 (issued with regard to the staff's assessment of AMR Item No. 22 of LRA Table 3.1-1, as evaluated in Section 3.1.2.1 of the SER)
In its response to RAI 3.1.2.1-3, the applicant stated that it recognizes that stress relaxation can occur in the SG manway and handhole bolting, at least for the bolting on the secondary side of the SGs, and stated that it has a bolting and torque program to determine the closure and torque requirements for RCS closure bolting. However, in its response to RAI 3.1.2.1-3, the applicant did not identify loss of preload as an aging effect and did not identify an AMP to manage the aging effect associated with SG bolting. GALL IV.D.1.1.7 identifies that loss of pre load due to stress relaxation is an aging effect for the steam generator secondary manway and handhole bolting, and GALL XI.M18, "Bolting Integrity," is the AMP to manage this aging effect.
According to 10 CFR 54.21(1), license renewal applicants must perform AMRs and identify all applicable aging effects for passive components within the scope of license renewal. The SG primary and secondary manway and handhole bolts are passive components within the scope of license renewal. The applicant has stated that stress relaxation is an applicable aging effect for the SG secondary manway and handhole bolting; therefore, the applicant is required by 10 CFR 54.21(a)(3) to propose an AMP to manage the aging effect. The staff also requests the applicant to provide technical justification as to why loss of preload stress relaxation does not have to be managed for the primary SG manway bolts in the manner required for the management of the SG secondary side bolting. In subsequent discussions with the NRC staff to resolve this issue, the applicant stated that the RNP Bolting Integrity Program in LRA Section B.3.4 will be applied to the pressure retaining bolting for the primary and secondary side of the steam generators because the RNP Bolting Integrity Program can be relied upon to prevent the loss of preload and that the RNP Bolting Integrity Program will not take exception to the Scope of Program in GALL XI.M18, Bolting Integrity. The staff evaluates the RNP Bolting Integrity Program in Section 3.0.3 of this SER. The staff finds the applicants resolution of the issue acceptable because the applicant credits its Bolting Integrity Program to manage loss of preload due to stress relaxation in the SG primary and secondary manway and handhole bolts.
However, the applicant needs to submit its resolution under oath and affirmation; therefore, this is Confirmatory Item 3.1.2.1-1, Part 3.
1-15 In its response to Confirmatory Item 3.1.2.1-1, Part 3, dated September 16, 2003, the applicant stated that the RNP Bolting Integrity Program is applied to pressure retaining bolting for the primary and secondary side of the steam generator. The applicant modified the Bolting Integrity Program to include the aging management of the SG primary and secondary bolting.
As specified in LRA section B.3.4, Bolting Integrity Program, loss of preload due to stress relaxation is one of the aging effects that will be managed. The staffs evaluation of the applicants Bolting Integrity Program is discussed in Section 3.0.3.6 of this SER. The staff concludes that Confirmatory Item 3.1.2.1-1, Part 3, is closed because the applicants Bolting Integrity Program will adequately manage the aging effect of loss of preload due to stress relaxation in the steam generator primary and secondary side bolting.
Confirmatory Item 3.1.2.1-2 (issued with regard to the staffs assessment of AMR Item No. 26 of LRA Table 3.1-1, as evaluated in Section 3.1.2.1 of the SER)
In order to provide reasonable assurance that general corrosion is not an applicable aging effect for the Class 1 carbon steel or low-alloy steel components in containment air or indoor air environments, the staff seeks confirmation that the Class 1 carbon steel or lower alloy steel components operate at temperatures that are equivalent to or hotter than the ambient temperature for the surrounding containment air or indoor air environments. This is Confirmatory Item 3.1.2.1-2.
The applicant provided the following response to Confirmatory Item 3.1.2.1-2 in a letter dated September 16, 2003:
RNP confirms that Class 1 carbon steel or low alloy steel components operate at temperatures that are equivalent to or hotter than the ambient temperature for the surrounding containment air or indoor air environments.
The applicants response to Confirmatory Item 3.1.2.1-2 confirms that the Class 1 carbon steel or low-alloy steel components in the RCS operate at temperatures equivalent to or hotter than the ambient temperatures for their external atmospheric environments (i.e., the containment air or indoor air environments). Based on the applicants response, the staff concludes that precipitation on the components therefore will not be a concern for the extended period of operation for RNP and that general corrosion induced by precipitation on the Class 1 carbon steel or low-alloy steel components is not an aging effect requiring aging management during the extended period of operation for RNP. Confirmatory Item 3.1.2.1-2 is therefore resolved, and Confirmatory Item 3.1.2.1-2 is closed.
Confirmatory Item 3.1.2.1-3, Parts 1 and 2 (issued with regard to the staffs assessment of AMR Item No. 31 of LRA Table 3.1-1, as evaluated in Section 3.1.2.1 of the SER)
The staff seeks confirmation that the reactor vessel (RV) thermal shield is adjacent to the fuel zone region of the RV, receives a neutron fluence greater than 1x1017 n/cm2, is within the scope of the commodity group in AMR 31 to LRA Table 3.1-1, and will be managed by the Pressurized Water Reactor Internal Program. This is Confirmatory Item 3.1.2.1-3, Part 1.
The staff seeks confirmation whether or not the RV internal lower support and lower support plate columns are fabricated from cast austenitic stainless steel (CASS) materials and are
1-16 within the scope of AMR Item 8 of LRA Table 3.1-1, AMR Item 33 of LRA Table 3.1-1, and AMR Item 14 of LRA Table 3.1-2. This is Confirmatory Item 3.1.2.1-3, Part 2.
The applicants response to RAI 3.1.2.1-9, Part 1, as amended by the applicants response to Confirmatory Item 3.1.2.1-3, Part 1, also provides an acceptable basis for omitting the RNP thermal shield from the scope of AMR Item 31 of LRA Table 3.1-1, because the applicant has committed to continued participation in the EPRI-MRPs activities for investigating the aging effects that are applicable to the pressurized-water reactor (PWR) internals of PWR-designed light-water reactors, and to use its participation in the activities as the basis for developing its inspection plan for the PWR Vessel Internals Program. This will include industry initiatives to study the aging effects that are applicable to the thermal shields of PWR-designed light-water reactors and to determine whether nondestructive inspections are warranted for the thermal shields and, if warranted, which inspection methods are most appropriate for the examinations.
The applicant has also committed to submitting its inspection plan for the PWR Vessel Internals Program to the staff for review and approval 24 months prior to its implementation. These commitments are given in Commitment No. 33 of Attachment II of CP&L Serial Letter No. RNP-RA/03-0031, dated April 28, 2003. The staff considers that this commitment will permit the staff an opportunity to determine and resolve with the applicant whether additional inspections are warranted for the RNP RV internals, including the thermal shield. The staff therefore considers RAI 3.1.2.1-9, Part 1, and Confirmatory Item 3.1.2.1-3, Part 1, to be resolved and RAI 3.1.2.1-9, Part 1, and Confirmatory Item 3.1.2.1-3, Part 1, are closed.
The applicant provided its response to Confirmatory Item 3.1.2.1-3, Part 2, in a letter dated September 16, 2003. In this response, the applicant clarified that only the upper support tube base, lower support plate columns, and bottom-mounted instrumentation column cruciform are fabricated from CASS. The applicant clarified that the lower support column forging is fabricated from austenitic stainless steel and that the AMRs for this forging are given in AMR Items 8 and 33 of LRA Table 3.1-1. The applicant confirmed that the lower support forging is not within the scope of AMR Item 14 of LRA Table 3.1-2 because the component is not fabricated from CASS. Since the applicant has provided the clarifications requested by the staff relative to the CASS RV internal components, the staff considers Confirmatory Item 3.1.2.1-3, Part 2, to be resolved, and Confirmatory Item 3.1.2.1-3, Part 2 is closed.
Confirmatory Item 3.1.2.2.4-1 (issued with regard to the staffs assessment of AMR Item No. 6 of LRA Table 3.1-1, as evaluated in Section 3.1.2.2.4 of the SER)
The staff is concerned that the AMPs credited by the applicant for managing crack initiation and growth of small bore Class 1 piping may be used as a precedent for relieving the applicant of performing the required ASME ISI examinations for the small bore Class 1 piping welds during the period of extended operation for RNP. Therefore, the staff seeks confirmation that the applicant will continue to perform the ISI examinations of the small bore Class 1 piping that are required by Section XI of the ASME Boiler and Pressure Vessel Code during the period of extended operation for RNP.
In its response to Confirmatory Item 3.1.2.2.4-1, dated August 14, 2003, the applicant confirmed that it would continue to conduct all applicable ISI inspections of the Class 1 small bore piping required by Section XI of the ASME Boiler and Pressure Vessel Code, unless relief is requested from and granted by the staff under applicable provisions in 10 CFR 50.55a.
1-17 Since the applicant response indicates that the applicant will continue to meet the inspection requirements for Class 1 small bore pipe, as required by 10 CFR 50.55a and Section XI of the ASME Boiler and Pressure Vessel Code, during the period of extended operation for RNP, the applicants response to Confirmatory Item 3.1.2.2.4-1 is acceptable. Confirmatory Item 3.1.2.2.4-1 is resolved.
Confirmatory Item 3.1.2.2.7-1 (issued with regard to the staffs assessment of AMR Item No. 9 of LRA Table 3.1-1, as evaluated in Section 3.1.2.2.7 of the SER)
The staff seeks confirmation that the welds used to join the SG instrumentation nozzles to the SG shells were fabricated using Alloy 600 weld material (i.e., Alloy 82/182 filler metals). If Alloy 600 weld materials are utilized, the applicant should discuss whether the welds are within the scope of and managed by the Nickel-Alloy Nozzles and Penetrations Program. This is Confirmatory Item 3.1.2.2.7-1.
In its response to Confirmatory Item 3.1.2.2.7-1, dated September 16, 2003, the applicant stated that the welds joining the carbon steel steam generator shell to the carbon steel instrumentation nozzles are not fabricated from Alloy 600 weld material. The staff finds that the Nickel-Alloy Nozzles and Penetrations Program would not be an appropriate AMP to manage the aging effects of the instrumentation nozzle welds because Alloy 600 materials (i.e., Alloy 82/182 filler metals) are not used in the welds. However, the steam generator instrumentation nozzles and associated welds are being managed by other applicable AMPs as discussed above. The staff concludes that Confirmatory Item 3.1.2.2.7-1 is closed because the applicant has clarified that the welds joining the carbon steel steam generator shell to the carbon steel instrumentation nozzles are not made of Alloy 600 materials.
Confirmatory Item 3.1.2.4.4.3-1 (issued with regard to the staffs assessment of AMR Item No. 10 to LRA Table 3.1-2, as evaluated in Section 3.1.2.4.4.3 of the SER)
The staff seeks confirmation that CP&L is crediting the Nickel-Alloy Nozzles and Penetrations Program as an additional AMP for managing primary water stress corrosion cracking (PWSCC) in the RNP bottom head instrumentation tube nozzles. This is Confirmatory Item 3.1.2.4.4.3-1.
The applicant provided the following response to Confirmatory Item 3.1.2.4.4.3-1 by letter dated September 16, 2003. RNP will maintain its involvement in industry initiatives and will implement any actions, unless impracticable, that are agreed upon between the NRC and the nuclear power industry to monitor for, detect, evaluate, and correct cracking in the VHP nozzles, specifically as the actions relate to ensuring the integrity of VHP nozzles in the RNP upper reactor vessel head during the extended period of operation. RNP also agreed to submit, for review and approval, its inspection plan for the Nickel-Alloy Nozzles and Penetrations Program, as it will be implemented from participation in industry initiatives prior to July 31, 2009.
Based on the applicants response to Confirmatory Item 3.1.2.4.4.3-1, the applicants commitment to Commitment # 31 to attach the CP&Ls serial letter No. RNP-RA/03-0031 and the clarification provided in the applicants response to RAI 3.1.2.4.4-1 and B.4.1-1, the staff
1-18 concludes that the applicant has provided an acceptable method of determining which inspection methods will be necessary for the RNP bottom head instrumentation tube nozzles during the extended period of operation for RNP, as determined from the industrys initiatives on managing degradation of nickel-based alloy components and welds, the state of pertinent industry operating experience (OE) on degradation of PWR bottom head instrumentation tube nozzles (including that for STP), and the staffs resolution of this OE with licensed utilities in the industry. Confirmatory Item 3.1.2.4.4.3-1 is resolved.
Confirmatory Item 3.1.2.4.5.2-1 (issued with regard to the staffs assessment of AMR Item No. 9 to LRA Table 3.1-2, as evaluated in Section 3.1.2.4.5.2 of the SER)
The staff seeks confirmation that CP&L is crediting the Nickel-Alloy Nozzles and Penetrations Program as an additional AMP for managing PWSCC in the RV core support pads. This is Confirmatory Item 3.1.2.4.5.2-1.
The applicant provided the following response to Confirmatory Item 3.1.2.4.5.2-1 by letter dated September 16, 2003. RNP will maintain its involvement in industry initiatives and will implement any actions, unless impracticable, that are agreed upon between the NRC and the nuclear power industry to monitor for, detect, evaluate, and correct cracking in the VHP nozzles, specifically as the actions relate to ensuring the integrity of VHP nozzles in the RNP upper reactor vessel head during the extended period of operation. RNP also agreed to submit, for review and approval, its inspection plan for the Nickel-Alloy Nozzles and Penetrations Program, as it will be implemented from participation in industry initiatives prior to July 31, 2009.
Based on the applicants response to Confirmatory Item 3.1.2.4.5.2-1, the applicants commitment to Commitment # 31 to attach the CP&Ls serial letter No. RNP-RA/03-0031 and the clarification provided in the applicants response to RAI 3.1.2.4.4-1 and B.4.1-1, the staff concludes that the applicant has provided an acceptable method of determining which inspection methods will be necessary, if any, for the RNP RV core support pads during the extended period of operation for RNP, as determined from the applicants commitment to maintain its continued participation in the industrys initiatives on nickel-based alloy components and welds and its commitment to submit the Nickel-Alloy Nozzles and Penetrations Program to the staff for review and approval. Confirmatory Item 3.1.2.4.5.2-1 is resolved.
Confirmatory Item 3.1.2.4.5.5-1 (nickel-based alloy in-core flux thimble tubes)
The staff seeks confirmation that the scope of AMR Item 16 of LRA Table 3.1-2 is for nickel-based alloy in-core flux thimble tubes and not for the retractable in-core flux thimbles. An inspection-based program should be used in conjunction with the Water Chemistry Program to manage SCC in these components. Therefore, the staff also seeks confirmation that the applicant will credit both the PWR Vessel Internals Program and the Water Chemistry Program to manage SCC (including PWSCC and/or irradiation-assisted stress-corrosion cracking (IASCC)) in the nickel-based alloy in-core flux thimble tubes. This is Confirmatory Item 3.1.2.4.5.5-1.
In response to this confirmatory item, the applicant revised Commitment No. 31 on the Nickel-Alloy Nozzles and Penetrations Program. This revision was submitted to the NRC by the CP&L Serial Letter No. RNP-RA/03-0154, dated December 10, 2003. This version of the commitment
1-19 included a commitment to: (1) participate in the MRPs industry initiatives on cracking of nickel-based alloy components, (2) implement those recommendations that result for the MRPs studies on these matters and are acceptable to the NRC, and (3) to submit the inspection plan for the Nickel-Alloy Nozzles and Penetrations Program for NRC review and approval by July 31, 2009. The commitment to submit the Nickel-Alloy Nozzles and Penetrations Program for staff review and approval will provide a sufficient opportunity to determine whether cracking is an issue for the Alloy 600 thimble outer sheaths that are exposed to the reactor coolant and to discuss with the applicant whether inspections of the components will be needed during the extended period of operation for RNP. The staff therefore concludes that this is an acceptable process for managing cracking that may potentially occur in the thimble outer sheaths. Based on this assessment, the staff concludes that the applicant has proposed an acceptable basis for managing cracking in the flux thimbles at RNP and that AMR 16 of LRA Table 3.1-2 is acceptable. Therefore, Confirmatory Item 3.1.2.4.5.5-1 is resolved.
Confirmatory Item 3.3.2.3.3-1 (confirmation that the diesel-and motor-driven fire pumps are overhauled on a 10-year cycle, and this overhaul includes inspection of the bowls)
During the AMR inspection (June 9-13, 2003), the staff reviewed the applicants replacement frequency for fire pump casings for the Fire Protection Program (see LRA Table 3.3-2, Item 30).
The audit noted that there is an error in the application and the fire pumps do not have casings, rather the vertical shaft pumps used at RNP use bowls for the pressure boundary function.
Furthermore, the inspection indicated that these bowls are not replaced on a 10 year cycle, rather the pumps are overhauled on a 10-year cycle. Overhaul does not specifically require replacement of the bowls. The applicant explained during a phone call on June 12, 2003, that the frequency of the overhaul of the fire pumps is consistent with OE and that the current Preventive Maintenance Program is effective at ensuring the pumps remain operable during a 10-year service between overhauls. A Confirmatory Item 3.3.2.3.3-1 will be included for the applicant to confirm that the diesel-and motor-driven fire pumps are overhauled on a 10-year cycle and this overhaul includes inspection of the bowls (i.e., the pressure retaining portion of the pump), and the bowls may or may not be replaced based upon their condition.
In its response dated September 16, 2003, the applicant included a revision of LRA Table 3.3-2, Item 30. This revision corrected the language to reference bowls rather than casings. The same letter also corrected the discussion to state that the diesel-and motor-driven fire pumps are overhauled on a 10-year cycle, and this overhaul includes inspection of the bowls. This is a change from the previous language which stated that the bowls are replaced on a 10-year frequency. The applicant has determined that based on OE this frequency is adequate to manage aging-related degradation. The staff found the applicants response to be acceptable, and Confirmatory Item 3.3.2.3.3-1 is considered to be closed.
Confirmatory Item 3.3.2.4.7-1 (AMP of radioactive equipment drains)
This confirmatory item relates to radioactive equipment drain system (REDS). In RAI 2.3.3.7-2, the staff requested the applicant to clarify which portions of this system are included within the scope of license renewal and subjected to an AMR. In its response dated April 28, 2003, the applicant described the portions of the REDs that are within the scope of license renewal and identified the aging effect of loss of material due to crevice corrosion, pitting corrosion, and microbiologically induced corrosion (MIC). In its response to RAI 2.3.3.7-2, the applicant stated
1-20 that the identified aging effects do not affect the intended function of the REDS and, therefore, do not require management for the period of extended operation. Based on the information provided in the LRA and the additional information included in the applicants response to RAI 2.3.3.7-2, the staff requested the applicant to provide additional information to support its conclusion that the identified aging effects do not affect the intended function of the REDS and, therefore, do not require management for the period of extended operation. On June 17, 2003, in a telephone conference, the staff discussed the issue further with the applicant. Subsequent to the telephone conference, by an electronic correspondence dated June 19, 2003, the applicant provided information to support its conclusion on the aging management of REDS.
This explanation has been discussed in Section 3.3.2.4.7.2 of this SER. The staff finds that the applicant has provided adequate information to justify that no AMP is required to manage the aging effects of the REDS because the applicant has demonstrated that leaking and blockage of the REDS are unlikely, the potential flow blockage will be identified and corrected timely by the applicants routine inspection and other activities, and leakage of the REDS would not adversely impact the performance of the SSCs. However, the applicant was requested to clarify the applicable aging effects for these REDs components and to incorporate the supporting explanation as discussed above into its response to RAI 2.3.3.7-2. This is Confirmatory Item 3.3.2.4.7-1.
By letter dated August 14, 2003, the applicant provided the requested information. Based on its review of the information provided in the LRA and the additional information provided in the applicants response to RAI 3.3.2.4.7-1, the staff concurs with the applicants conclusion that no AMP is required to manage the aging effects of the REDS and that there is reasonable assurance that the intended functions of the REDS will remain. Therefore, Confirmatory Item 3.3.2.4.7-1 is resolved.
Confirmatory Item 3.3.2.4.17-1 (aging effects for the components in the dedicated shutdown diesel generator)
This confirmatory item relates to the aging effects for the materials and environments associated with the components in the dedicated shutdown diesel generator. In RAI 3.3.17-1, the staff requested the applicant to provide a detailed discussion on the AMR performed for the stainless steel valves, piping, tubing, and fittings listed in Table 3.3-2, row numbers 12, 13, and 23, and explain why the AMR results are different among them. In its response, the applicant stated that the air and gas environments in row numbers 12 and 13 include the potential for wetting of stainless steel by untreated water, which is the genesis of the potential aging effects.
A detailed explanation of the response has been included in Section 3.3.2.4.17 of this SER.
The staff found the referenced explanation appropriate. However, the applicant is requested to provide the above information under oath and affirmation, and this remains as Confirmatory Item 3.3.2.4.17-1.
By letter dated August 14, 2003, the applicant provided the requested information. On the basis of its review of the information provided in the LRA and the additional information included in the applicants response to RAIs 3.3-3 and 3.3-5, the staff finds that the aging effects that result from contact of the dedicated shutdown (DS) diesel generator (DG) SSCs to the environments described in Tables 2.3-23, 3.3-1, and 3.3-2 are consistent with industry experience for these combinations of materials and environments. Therefore, the staff finds the
1-21 applicant has identified the appropriate aging effects for the materials and environments associated with the components in the DS DG. Confirmatory Item 3.3.2.4.17-1 is resolved.
Confirmatory Item 3.3.2.4.19-1 (aging effects for the components in the fuel oil system)
This confirmatory item relates to the aging effects for the materials and environments associated with the components in the fuel oil system. In RAI 3.3.17-1, the staff requested the applicant to provide a detailed discussion of the AMR performed for the stainless steel valves, piping, tubing, and fittings listed in Table 3.3-2, row numbers 12, 13, and 23, and explain why the AMR results are different among them. The air and gas environments in row numbers 12 and 13 include the potential for wetting of stainless steel by untreated water, which is the genesis of the potential aging effects. In row number 23, the environment is considered a reasonably dry environment which results in no potential aging effects for stainless steel. For the fuel oil system, it has a stainless steel valve and instrumentation tubing, valves, and fittings that are conservatively modeled in a wetted outdoors environment. The fuel oil tank level instrumentation is located outdoors and has components that are near the ground. A detailed explanation of the response has been included in Section 3.3.3.4.19 of this SER. The staff found the referenced explanation appropriate. However, the applicant is requested to provide the above information under oath and affirmation, and this remains as Confirmatory Item 3.3.24.19-1.
The applicant has provided additional information related to the aging effects of the external surfaces of the SS components/environments combination in the response to Confirmatory Item 3.3.2.4.19-1, in letter RNP-RA/03-0094, dated August 14, 2003. The staff found the applicants response to be acceptable. Therefore, Confirmatory Item 3.3.2.4.19-1 is resolved.
Confirmatory Item 3.5-1 (AMP for below-grade reinforced concrete)
In RAI 3.5.1-3, the staff requested the applicant to provide available RNP ground-water chemistry test results including chlorides, sulphate, and pH values and discuss the proposed AMP, as well as past inspection results of below-grade concrete at RNP, since the below-grade reinforced concrete at RNP is exposed to an aggressive environment (low pH). In RAI 3.5.1-9 the staff stated that it is unclear how the inspection for below-grade containment concrete will be performed by the ASME Section XI, Subsection IWL Program and requested that additional information, such as the locations, depth, and frequency of soil excavation, related to the AMR of below-grade containment concrete be provided. The applicant responded to both RAIs offering commitments that adequately address the staff concerns regarding the aging management of below-grade in-scope concrete structural components at RNP. Because of the slightly acidic RNP ground-water environment, the applicant conservatively assumed existence of an aggressive chemical environment and proposed the plant-specific AMPs (an enhanced ASME,Section XI, Subsection IWL Program for containment and an enhanced Structures Monitoring Program for other Category 1 structures) described in Section 3.5.2.2.1.1 of this SER to manage the aging effects of below-grade concrete. The staff finds RAIs 3.5.1-3 and 3.5.1-9 are fully resolved, pending satisfactory resolution of Confirmatory Item 3.5-1.
By letter dated August 14, 2003 (RNP Serial RNP-RA/03-0094), the applicant responded to a number of confirmatory items identified by the staff. The staff reviewed the revised contents of Items 25, 26, and 27 of Attachment II (Revised License Renewal Commitments). The staff also
1-22 reviewed the specific response to Confirmatory Item 3.5-1 provided in Attachment III (Response to License Renewal Confirmatory Items) in the same letter. Based on these reviews, the staff finds that the applicant has provided adequate information, and Confirmatory Item 3.5-1 is closed.
Confirmatory Item 3.6.2.3.1.2-1 (non-EQ insulated cables and connections program)
In LRA Section B.4.6, Non-EQ Insulated Cables and Connections Program, the applicant described its AMP to manage aging in non-EQ insulated cables and connections. The LRA stated that this AMP is consistent with GALL AMPs XI.E1, Electrical Cables and Connections Not Subject to 10 CFR 50.49 Environmental Qualification Requirements, with no deviations. In response to the staffs concern (RAI B.4.6-2) about excluding non-polyvinyl chloride (PVC) cables inside and outside containment in an adverse localized environment from the sample, the applicant in a letter dated June 13, 2003, stated that the scope of this program includes plant cables and connections of various insulation material types (not just PVC) that may be located in an adverse, localized environment. On the basis of its review, the staff finds that its concern is not resolved. In subsequent discussions with the NRC staff to resolve this issue, the applicant stated that the statement in LRA Section B.4.6 regarding The sample locations will consider the location of PVC cables inside and outside containment as well as any known adverse localized environments, (PVC was determined to be the limiting insulation material) will be modified by The sample locations will consider the location of cables and connections inside and outside containment as well as any known adverse localized environments." The staff finds that the applicants resolution of this issue is acceptable because the sample will consider all insulation material types used inside and outside containment as well as any known adverse localized environments. However, the applicant needs to submit its resolution under oath and affirmation; therefore, this is Confirmatory Item 3.6.2.3.1.2-1 In its response dated September 16, 2003, the applicant revised the LRA Section B.4.6 to read, The sample locations will consider the location of cables and connections inside and outside containment as well as any known adverse localized environments. The staff found the applicants response to be acceptable, and Confirmatory Item 3.6.2.3.1.2-1 is considered to be closed.
Confirmatory Item 3.6.2.3.2.2-1 (AMP for non-EQ electrical cables used in instrumentation circuits (B.4.7))
For the GALL attribute, Operating Experience, the applicant stated that changes in instrument calibration data can be caused by degradation of the circuit cable and are a possible indication of potential cable degradation. The staff finds that the applicant did not address the operating experience in the formal response. In subsequent discussions with the NRC staff to resolve this issue, the applicant stated that this element will be revised to address the operating experience as follows. Industry operating experience indicates that changes in instrument calibration data can be caused by degradation of the circuit cable and are a possible indication of potential cable degradation. This program is for the non-EQ portions of the high range radiation monitoring cabling systems. These cabling systems are located in non harsh environments and none have experienced age-related degradation. The staff finds that the applicants resolution of the open item is acceptable because the applicant adequately
1-23 addressed the operating experience. However, the applicant needs to submit its resolution under oath and affirmation; therefore, this is Confirmatory Item 3.6.2.3.2.2-1.
In its response dated September 16, 2003, the applicant revised the operating experience to include the following, Industry operating experience indicates that changes in instrument calibration data can be caused by degradation of the circuit cable and are a possible indication of potential cable degradation. This program is for the non-EQ portions of the high range radiation monitoring cabling systems. These cabling systems are located in non harsh environments and none have experienced age related degradation. The staff found the applicants response to be acceptable, and Confirmatory Item 3.6.2.3.2.2-1 is considered to be closed.
Confirmatory Item 3.6.2.3.2.2-2 (AMP for neutron flux instrumentation (B.4.8))
To detect aging effects, the cables used in neutron flux instrumentation circuits will be tested at least once every 10 years. Testing may include insulation resistance tests, TDR tests, I/V testing, or other testing judged to be effective in determining cable insulation condition.
Following issuance of a renewed operating license for RNP, the initial test will be completed before the end of the initial 40-year license term for Unit 2 (July 31, 2010). The staff finds that this testing is acceptable because the testing will determine cable insulation resistance (potential degradation); however, the staff is concerned about the 10-year testing frequency. In subsequent discussions with the NRC staff to resolve this issue, the applicant stated that a review of site operating experience found no age-related failures for neutron monitoring cables or connectors. The only industry operating experience identified for these cables was Westinghouse Technical Bulletin 86-01. This bulletin identified industry concerns with cables used for the source range detector regarding cable degradation due to high operating voltage, radiation, heat, and moisture. Both the source range and intermediate range detector cables inside containment were replaced in 1991 as a result of that bulletin. These cables had operated for 20 years without failure prior to being replaced. The replacement cables were manufactured to Class 1E standards and have remained functional during the last 12 years.
The power range cables are the original installed cables and are the same cable type (Amphenol/Essex 21-529) that was originally used in the source range and intermediate range circuits. They have operated for over 32 years without failure, which demonstrates their ability to operate over long periods without a loss of intended function.
In addition, the licensee stated that initial testing of all in-scope neutron monitoring cables will be performed prior to the end of the current license term. This testing will provide a positive means of detecting any significant aging that has occurred since the cables were installed, which in the case of the power range cables will be after 3340 years of operation. Given the operating experience of these cables and the gradual nature of cable insulation aging, the 10-year testing frequency subsequent to the initial testing provides reasonable assurance that the cables will continue to perform their intended function. The staff finds that the applicants resolution of the issue is acceptable because the cable insulation degradation is a slow process and RNP operating experience did not identify any cable insulation degradation. Additionally, this 10-year frequency is consistent with NUREG-1801 cable aging management programs frequency. However, the applicant needs to submit its resolution under oath and affirmation; therefore, this is Confirmatory Item 3.6.2.3.2.2-2.
1-24 In response to the above confirmatory item, the applicant, in a letter dated September 16, 2003, stated the following:
A review of site operating experience found no age related failures for neutron monitoring cables or connectors. The only industry operating experience identified for these cables was Westinghouse Technical Bulletin 86-01. This Bulletin identified industry concerns with cables used for the source range detector regarding cable degradation due to high operating voltage, radiation, heat, and moisture. Both the source range and intermediate range detector cables inside containment were replaced in 1991 as a result of that bulletin. These cables had operated for 20 years without failure prior to being replaced. The replacement cables were manufactured to Class 1E standards and have remained functional during the last twelve years. The power range cables are the original installed cables and are the same cable type (Amphenol/Essex 21-529) that was originally used in the source range and intermediate range circuits. They have operated for over 32 years without failure, which demonstrates their ability to operate over long periods without a loss of intended function.
In addition, the licensee stated that the following:
Initial testing of all in-scope neutron monitoring cables will be performed prior to the end of the current license term. This testing will provide a positive means of detecting any significant aging that has occurred since the cables were installed, which in the case of the power range cables will be after 3340 years of operation. Given the operating experience of these cables and the gradual nature of cable insulation aging, the 10 year testing frequency subsequent to the initial testing provides reasonable assurance that the cables will continue to perform their intended function.
In addition, the applicant modified the operating experience element as described in Section 3.6.2.3.2. The staff found the applicants response to be acceptable, and on such basis Confirmatory Item 3.6.2.3.2.2-2 is considered to be closed.
Confirmatory Item 4.2.3-1 (update of UFSAR Supplement in accordance with the reference temperature (RT)PTS and upper-shelf energy (USE) values listed in WCAP-15828)
The staff requests confirmation that, at the next update of the UFSAR Supplement for RNP, the applicant will update Sections A.3.2.1.1 and A.3.2.1.2 of Appendix A in the LRA to reference the applicability of PTS and USE analyses in WCAP-15828, Revision 0, to the 60-year PTS and USE assessments for the RNP RV beltline materials and will update the corresponding UFSAR Supplement summary descriptions to reference the RTPTS and USE values listed in the report for the limiting PTS and USE materials in the beltline of the reactor vessel.
In its response to Confirmatory Item 4.2.3-1 dated September 16, 2003, the applicant stated that it would amend the FSAR Supplement summary descriptions for the TLAAs on PTS and USE, as given in Sections A.3.2.1 and A.3.2.2, respectively. This proposed amendment has been included in Section 4.2.3 of this SER. The applicants amended FSAR Supplement summary descriptions for the TLAAs on PTS and USE accomplish the following objectives (1) the amendments provide a sound basis why the TLAAs for PTS and USE, as given in Sections A.3.2.1 and A.3.2.2 of the LRA, comply with the requirements in 10 CFR 50.61 for PTS and in 10 CFR Part 50, Appendix G, for USE through the expiration of the extended period of operation for RNP, and (2) the amendments provide a reference to the extended period of
1-25 operation licensing basis documents containing the TLAAs for PTS and USE. Since the FSAR Supplement summary descriptions demonstrate while the TLAAs are acceptable and reference the applicable licensing basis documents, the staff therefore concludes that the applicants FSAR Supplement summary descriptions for the TLAAs on PTS and USE, as given in Sections A.3.2.1 and A.3.2.2 of the LRA, and amended by the applicants response to Confirmatory Item 4.2.3-1, are acceptable. Confirmatory Item 4.2.3-1 is resolved.
Confirmatory Item 4.3.2-1 (auxiliary feedwater fatigue analysis)
In RAI 4.3-7, the staff requested the applicant to provide (1) calculated cumulative utilization factors (CUFs) of the six replacement branch connections, (2) confirmation that no other nonstandard components were used or provide justification of the acceptability for use in safety systems at RNP, and (3) description of the AMPs that will be used to provide assurance that the CUFs for these connections will not exceed the limit of 1.0 for the period of extended operation.
In its response by a letter dated June 13, 2003, the applicant stated that there are three 4" to 16" auxiliary feedwater-to-feedwater connections downstream of the motor-driven and the steam-driven AFW pump. The three connections downstream from the steam-driven pumps could not be qualified for the full 40-year design transient set, so a reduced number of design transients was postulated. This resulted in a CUF value of 0.99 for 40-year life. Based upon projections of actual transients to date, the qualified number of transients is not expected to be reached until approximately year 50. The applicant indicated that the number of transients used in the analysis will be tracked by the Fatigue Monitoring Program. The applicant further indicated that the components will be either reanalyzed or replaced prior to exceeding the number of transients tracked by the Fatigue Monitoring Program. The staff finds that the applicants proposed options provide acceptable plant-specific approaches to address fatigue of the connections between the auxiliary and main feedwater lines for the period of extended operation in accordance with 10 CFR 54.21(c)(1). However, in accordance with 10 CFR 54.21(d), these options need to be included in the UFSAR Supplement (Confirmatory Item 4.3.2-1).
By letter dated September 16, 2003, the applicant provided a modification to UFSAR Supplement Section A.3.2.2.1, which includes the proposed options to address fatigue of the connections between the auxiliary and main feedwater lines for the period of extended operation. The staff finds the modification to UFSAR Supplement Section A.3.2.2.1 acceptable.
Confirmatory Item 4.3.2-1 is closed.
Confirmatory Item 4.3.2-2 (aging management of surge line for period of extended operation)
In RAI 4.3-10, the staff requested that the applicant provide additional clarification regarding aging management of the surge line during the period of extended operation. The applicants June 13, 2003, response indicated that fatigue of the surge line will be managed using one or more options. Options include further refinement of the fatigue analyses to maintain the environmentally assisted fatigue (EAF)-adjusted CUF below 1.0, repair of the affected locations, replacement of the affected components, or management of the effects of fatigue through the use of an augmented ISI program reviewed and approved by the NRC.
The applicant commits to provide the NRC with the details of the inspection program prior to the period of extended operation if the last option is selected. As indicated by the applicant, the
1-26 use of an inspection program to manage fatigue will require prior staff review and approval.
The applicant indicated that LRA Section A.3.2.2.2 would be revised to include the applicants proposed options for managing the surge line fatigue. The staff finds the applicants proposed options provide acceptable plant-specific approaches to address EAF of the RNP pressurizer surge line for the period of extended operation in accordance with 10 CFR 54.21(c)(1).
Revision of the UFSAR Supplement is Confirmatory Item 4.3.2-2.
By letter dated September 16, 2003, the applicant provided a modification to UFSAR Supplement Section A.3.2.2.1, which includes the proposed options to address fatigue of the surge line for the period of extended operation. The staff finds the modification to UFSAR Supplement Section A.3.2.2.1 acceptable. Confirmatory Item 4.3.2-2 is closed.
Confirmatory Item 4.6.3-1 (elimination of containment penetration coolers)
This confirmatory item relates to RAI 4.6.3-2. The staff requested the applicant to describe how the analysis was performed and submit the analysis results of concrete properties at the end of 252 cycles. The staff requested the applicant to clarify whether the conclusion of 252 cycles was obtained from its operating experience. During a teleconference call on June 10, 2003, the applicant stated it had found an analysis result indicating that the temperature in concrete around the containment penetration would always remain below 200 F. Therefore, the applicant is withdrawing this TLAA item and will submit a new writeup to indicate the withdrawal.
Since the applicants analysis results indicate that the concrete temperature around the containment penetration will always remain below 200 F with the elimination of containment penetration coolers, the applicant informed the staff in the teleconference that it had withdrawn this TLAA issue and would submit its new writeup accordingly (Confirmatory Item 4.6.3-1). The staff finds the applicants approach acceptable.
The staff agreed with the applicants approach of withdrawing this TLAA issue because its analysis results indicate that there is no need for the TLAA. The applicant submitted a letter dated August 14, 2003, to withdraw this TLAA item from the LRA. Therefore, Confirmatory Item 4.6.3-1 is closed.
Confirmatory Item 4.6.4-1 (issued with regard to the staffs assessment of LRA Section B.4.6.4, Aging of Boraflex, as evaluated in Section 4.6.4.2 of the SER)
By letter dated May 28, 2003, the applicant submitted for staff review a license amendment to change the technical specifications regarding removal of Boraflex monitoring procedures. The staff will need confirmation that the license amendment to remove the requirements to credit the Boraflex panels from the RNP technical specification has been approved and that the Boraflex panels will no longer be needed to maintain the effective neutron multiplication factor (Keff) for the geometry of the spent fuel rods stored in the spent fuel pool within acceptable levels. As part of this confirmatory item, the staff will need the applicant to provide a reference regarding the staffs safety evaluation to CP&L approving the license amendment for the Boraflex panels. This confirmatory item also requires the applicants statement that it will not be necessary to include a summary description of the Boraflex TLAA in the UFSAR Supplement of the application (i.e., in Appendix A of the LRA). This is Confirmatory Item 4.6.4-1.
1-27 By letter dated December 22, 2003, License Amendment No. 198, the staff approved the applicants request to eliminate the need to credit the Boraflex neutron absorbing material for reactivity control in the spent fuel storage pool. In place of the Boraflex material (i.e., panels),
the staff approved the applicants request to take credit for a combination of soluble boron and controlled fuel loading patterns in the spent fuel pool to maintain the required subcriticality margins in the spent fuel storage pool. On the basis of the final issuance of License Amendment No. 198, the staff finds that Confirmatory Item 4.6.4-1 is closed. In addition, the applicant may eliminate its Commitment No. 47 and eliminate any discussion in the RNP UFSAR regarding the Boraflex TLAA or the Boraflex monitoring program.
Confirmatory Item B.3.11-1 (issued with regard to the staffs assessment of LRA Section B.3.11, Reactor Vessel Surveillance Program, as evaluated in Section 3.1.2.3.4 of the SER)
The withdrawal schedule in WCAP-15805 indicates that the in-vessel location for Capsule U was moved sometime within the current life of the plant. Therefore, in a meeting with the applicant on May 21, 2003, the staff requested additional clarifying information regarding the elapsed time when Capsule U was moved in the vessel, what the lead factors were for Capsule U at the different in-vessel locations, and what CP&Ls basis was for determining that the projected fluence for Capsule U at its projected time of withdrawal would be indicative of the fluence for the RV shell at 50 effective full-power years (EFPY) (i.e., at the EFPY projected for the end of the extended period of operation for RNP). During the meeting of May 21, 2003, the applicant informed the staff that it would provide the additional information requested by the staff. The applicant submitted the requested information in an E-mail to the staff dated June 9, 2003. The applicant must formally submit the information in the E-mail of June 9, 2003, into the docket for RNP (i.e., into Docket No. 50-261) under Oath and Affirmation. This is Confirmatory Item B.3.11-1.
In its response to Confirmatory Item B.3.11-1, the applicant submitted the information provided in the email of June 9, 2003, for incorporation into the docket for RNP (i.e, Docket No. 50-261) under oath and affirmation. Since the requested information in the email of June 9, 2003, has been incorporated into the docket for RNP and since the information indicates the RV surveillance capsule withdrawal schedule is acceptable for the period of extended operation for RNP, the staff concludes that the applicants response to Confirmatory Item B.3.11-1 is acceptable. Confirmatory Item B.3.11-1 is resolved.
Confirmatory Item B.4.1-1 (issued with regard to the staffs assessment of LRA Section B.4.1, Nickel-Alloy Nozzles and Penetrations Program, as evaluated in Section 3.1.2.3.6 of the LRA)
The first paragraph in the UFSAR Supplement summary description for the Nickel-Alloy Nozzles and Penetrations Program is not up to date and needs to be amended to reflect that the applicants inspection program for the RNP vessel head penetration (VHP) nozzles is based on the requirements in NRC Order No. EA-03-009 (February 11, 2003) and the applicants response to the order dated March 3, 2003. The applicant must confirm that the UFSAR Supplement summary description for the Nickel-Alloy Nozzles and Penetrations Program (as given in Section A.3.1.28 of Appendix A to the LRA) will be amended to reflect the augmented requirements in NRC Order No. EA-03-009 for the RNP upper reactor vessel head and its VHP nozzles. This is Confirmatory Item B.4.1-1.
1-28 The applicant provided its response to Confirmatory Item B.4.1-1 by letter dated September 16, 2003. In this response, the applicant confirmed that the scope of the FSAR Supplement summary description for the Nickel-Alloy Nozzles and Penetrations Program will be amended to include the augmented requirements in NRC Order EA-03-009, as they apply to augmented inspections of the RNP reactor vessel head and VHP nozzles. Since the response confirms that the FSAR Supplement summary description for the AMP will be amended to reflect the applicability of the requirements in NRC Order EA-03-009, the staff concludes that the applicants response to Confirmatory Item B.4.1-1 is acceptable and Confirmatory Item B.4.1-1 is resolved.
Confirmatory Item B.4.2-1 (issued with regard to the staffs assessment of LRA Section B.4.2, Thermal Aging of Cast Austenitic Stainless Steel Program, as evaluated in Section 3.1.2.3.7 of the SER)
The staff seeks confirmation that, although a leak before break flaw tolerance evaluation has been performed for the extended period of operation for RNP (as given in WCAP-15628), the applicant will continue to perform those ISI examinations for the primary coolant loop piping, valve, and pump casings that are required by Table IWB-2500-1 of Section XI to the ASME Boiler and Pressure Vessel Code, unless relief has been granted by the NRC under applicable provisions in 10 CFR 50.55a from meeting the staffs ISI requirements of 10 CFR 50.55a(g)(4).
If relief has been granted from any of the required ISI examinations for the primary coolant loop piping, valve, or pump casings, the staff seeks confirmation of the applicable NRC staff safety evaluation granting this relief and the specific ISI examination requirements for which relief has been granted. The staff also seeks confirmation that the UFSAR Supplement summary description will be amended to reflect the information in the applicants response to this confirmatory item. This is Confirmatory Item B.4.2-1.
In its response to Confirmatory Item B.4.2-1, dated August 14, 2003, the applicant confirmed that the UFSAR Supplement summary description for the CASS Program will be amended to indicate that the applicant will continue to perform the inservice inspections of the ASME Class 1 primary loop piping, valve bodies, and pump casings, as required by 10 CFR 50.55a and Section XI of the ASME Boiler and Pressure Vessel Code, unless relief has been requested and granted by the NRC under applicable provisions in 10 CFR 50.55a. The applicant also confirmed that the summary description for the CASS program will also be amended to indicate that the NRC did approve some specific relief requests (i.e., in NRC safety evaluation dated September 26, 2002) on some of the specific ISI requirements for the ASME Class 1 primary loop piping, valve bodies, and pump casings for the fourth 10-year ISI interval for RNP.
The staff reviewed the information in the safety evaluation of September 26, 2002, and confirmed that the reliefs granted would not impact the acceptability of the program attributes for the CASS Program. Since the applicants response to Confirmatory Item B.4.2-1 indicates that the UFSAR Supplement summary description will be modified to demonstrate continued compliance with the requirements of 10 CFR 50.55a and Section XI of the ASME Boiler and Pressure Vessel Code, the staff concludes that the UFSAR Supplement summary description for the CASS Program is acceptable. Confirmatory Item B.4.2-1 is resolved.
1-29 Confirmatory Item B.4.3-1 (issued with regard to the staffs assessment of LRA Section B.4.3, PWR Vessel Internals Program, as evaluated in Section 3.1.2.3.8 of the SER)
The staff will confirm that the applicant has incorporated the commitment regarding the Nickel-Alloy Nozzles and Penetrations Program into the UFSAR Supplement summary description of Section A.3.1.30 of Appendix A to the LRA when the applicant revises its UFSAR Supplement for this AMP. This is Confirmatory Item B.4.3-1.
In its response to Confirmatory Item B.4.3-1, the applicant provided the staff with an updated version of Commitment No. 33 in RNP Serial Letter RNP-RA/03-0031, dated April 28, 2003, which included a commitment to submit the inspection plan for the PWR Vessel Internal Program for NRC review and approval. In the response to Confirmatory Item B.4.3-1, the applicant also confirmed that it would amend to UFSAR Supplement summary description for the PWR Vessel Internals Program, as given in Section A.3.1.30 of Appendix A to the LRA, to incorporate a statement that reflects that the PWR Vessel Internal Program will be submitted to the staff for review and approval 24 months prior to implementation. Since the applicants response reflects the commitment in Commitment No. 33 for submittal of the AMP for staff review and approval, the staff concludes that the applicants response to Confirmatory Item B.4.3-1 is acceptable and Confirmatory Item B.4.3-1 is resolved.
1.7 Summary of Proposed License Conditions As a result of the staffs review of the RNP application for license renewal, including the additional information and clarifications submitted subsequently, the staff identified two proposed license conditions. The first license condition requires the applicant to include the UFSAR Supplement in the next UFSAR update required by 10 CFR 50.71(e) following issuance of the renewed license. The second license condition requires that the future inspection activities identified in the UFSAR Supplement be completed prior to the period of extended operation.