ML15112A699
| ML15112A699 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/24/1998 |
| From: | Hoffman S NRC (Affiliation Not Assigned) |
| To: | Mccollum W DUKE POWER CO. |
| References | |
| NUDOCS 9811300153 | |
| Download: ML15112A699 (6) | |
Text
November 24, 1998 Mr. William R. McCollum, Jr.
Vice President, Oconee Nuclear Site Duke Energy Corporation P. 0. Box 1439 Seneca, SC 29679
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE OCONEE NUCLEAR STATION, UNITS 1, 2, AND 3, LICENSE RENEWAL APPLICATION
Dear Mr. McCollum:
By letter dated July 6, 1998, Duke Energy Corporation (Duke) submitted for the Nuclear Regulatory Commission's (NRC's) review an application pursuant to 10 CFR Part 54, to renew the operating licenses for the Oconee Nuclear Station (Oconee), Units 1, 2, and 3. Exhibit A to the application is the Oconee Nuclear Station License Renewal Technical Information Report (OLRP-1001), which contains the technical information required by 10 CFR Part 54. The NRC staff is reviewing the information contained in OLRP-1001 and has identified, in the enclosure, areas where additional information is needed to complete its review. Specifically, the enclosed questions are from the Mechanical Engineering Branch regarding OLRP-1001 Sections 1.5.5 and 5.4.1.
Please provide a schedule by letter, electronic mail, or telephonically for the submittal of your responses within 30 days of the receipt of this letter. Additionally, the staff would be willing to meet with Duke prior to the submittal of the responses to provide clarifications of the staffs requests for additional information.
Sincerely, Stephen T. Hoffman, Senior Project Manager License Renewal Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation and 50-287
Enclosure:
Request for Additional Information cc w/encl: See next page DISTRIBUTION: See next page DOCUMENTNAME:G:\\SEBROSKY\\RAl17.WPD OFFICE LA PM:PDLR PDLR:D NAME
_Berry Soffma CIGrimes DATE 11//98 11/ 23198 11/20098 OFFICIAL RECORD COPY 9811300153 981124 PDR ADOCK 05000269 ui UL:
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Oconee Nuclear Station (cense Renewal) cc:
Paul R. Newton, Esquire Duke Energy Corporation Mr. J. E. Burchfield 422 South Church Street Compliance Manager Mail Stop PB-05E Duke Energy Corporation Charlotte, North Carolina 28201-1006 Oconee Nuclear Site P. 0. Box 1439 J. Michael McGarry, Ill, Esquire Seneca, South Carolina 29679 Anne W. Cottingham, Esquire Winston and Strawn Ms. Karen E. Long 1400 L Street, NW.
Assistant Attorney General Washington, DC 20005 North Carolina Department of Justice P. 0. Box 629 Mr. Rick N. Edwards Raleigh, North Carolina 27602 Framatome Technologies Suite 525 L. A. Keller 1700 Rockville Pike Manager - Nuclear Regulatory Licensing Rockville, Maryland 20852-1631 Duke Energy Corporation 526 South Church Street Manager, LIS Charlotte, North Carolina 28201-1006 NUS Corporation 2650 McCormick Drive, 3rd Floor Mr. Richard M. Fry, Director Clearwater, Florida 34619-1035 Division of Radiation Protection North Carolina Department of Senior Resident Inspector Environment, Health, and U. S. Nuclear Regulatory Commission Natural Resources 7812B Rochester Highway 3825 Barrett Drive Seneca, South Carolina 29672 Raleigh, North Carolina 27609-7721 Regional Administrator, Region II Gregory D. Robison U. S. Nuclear Regulatory Commission Duke Energy Corporation Atlanta Federal Center Mail Stop EC-12R 61 Forsyth Street, SW, Suite 23T85 P. 0. Box 1006 Atlanta, Georgia 30303 Charlotte, North Carolina 28201-1006 Virgil R. Autry, Director Robert L. Gill, Jr.
Division of Radioactive Waste Management Duke Energy Corporation Bureau of Land and Waste Management Mail Stop EC-12R Department of Health and P. 0. Box 1006 Environmental Control Charlotte, North Carolina 28201-1006 2600 Bull Street RLGILL@DUKE-ENERGY.COM Columbia, South Carolina 29201-1708 Douglas J. Walters County Supervisor of Oconee County Nuclear Energy Institute Walhalla, South Carolina 29621 1776 I Street, NW Suite 400 Washington, DC 20006-3708 Chattooga River Watershed Coalition DJW@NEI.ORG P. 0. Box 2006 Clayton, GA 30525
DISTRIBUTION: Hard copy DEk-et-Eil0 PUBLIC PDLR RF M. EI-Zeftawy ACRS T2E26 F. Miraglia J. Fair J. Roe D. Matthews C. Grimes T. Essig G. Lainas J. Strosnider G. Bagchi H. Brammer T. Hiltz G. Holahan S. Newberry C. Gratton L. Spessard R. Correia R. Latta J. Peralta J. Moore R. Weisman M. Zobler E. Hackett A. Murphy T. Martin D. Martin W. McDowell S. Droggitis PDLR Staff M. Banic G. Hornseth H. Berkow D. LaBarge L. Plisco C. Ogle R. Trojanowski M. Scott C. Julian R. Architzel J. Wilson R. Wessman E. Sullivan R. Gill, Duke D. Walters, NEI
REQUEST FOR ADDITIONAL INFORMATION OCONEE NUCLEAR STATION, UNITS 1. 2. AND 3 LICENSE RENEWAL APPLICATION, EXHIBIT A OLRP-1001 Section No.
1.5.5 Generic Safety Issue 190 - Fatigue Evaluation of Metal Components for 60-Year Plant Life 1.5.5-1 Section 1.5.5.3 of the license renewal application indicates that additional confirmatory research is ongoing at Oconee in support of the generic resolution of issues associated with Generic Safety Issue (GSI) 190, "Fatigue Evaluation of Metal Components for 60-year Plant Life." The application further indicates that the Oconee study will apply the methodology developed in EPRI Report TR-1 05759, "An Environmental Factor Approach to Account for Reactor Water Effects in Light Water Reactor Pressure Vessel and Piping Fatigue Evaluations." Application of this methodology was discussed during a March 19, 1998, meeting between the industry and the staff. A letter from Christopher Grimes to NEI dated November 2, 1998, titled "Request for Additional Information on the Industry's Evaluation of Fatigue Effects for License Renewal", summarizes the staffs technical concerns regarding the methodology in EPRI Report TR-1 05759. Upon resolution of these concerns and when a final determination regarding GSI-190 has been made, you will be expected to address any particular action that may arise as a result of such determination.
Since the conclusion regarding GSI-1 90 in Section 1.5.5.4 of your application for license renewal relies on the conclusions from the referenced EPRI report, discuss how Oconee meets the relevant portion of Section 54.29 of the license renewal rule as discussed in the statement of considerations (SOC) (60 FR 22484, May 1995) in the absence of the staffs endorsement of EPRI Report TR-1 05759. Although the staff expects timely resolution of GSI-190, your response should address the situation in which GSI-190 is not resolved prior to the current license term.
Consistent with the SOC, it is expected that Duke will "submit a technical rationale which demonstrates that the CLB [current licensing basics] will be maintained until some later point in time in the period of extended operation, at which time one or more reasonable options (e.g., replacement, analytical evaluation, or a surveillance/maintenance program) would be available to adequately manage the effects of aging...and briefly describe options that are technically feasible during the period of extended operation to manage the effects of aging...."
5.4.1 Reactor Coolant System Piping and Components 5.4.1-1 Section 5.4 of the license renewal application indicates that B&W Owners Group Report BAW-2243A, Demonstration of the Management of Aging Effects for the Reactor Coolant System Piping, June 1996, identified leak-before-break and high energy line break postulation based on fatigue cumulative usage factor (CUF>0.1) as generically applicable time-limited aging analyses (TLAAs). However, the application indicates that, "the review conducted of Oconee documentation Enclosure
2 determined that neither the leak-before-break analyses nor the cumulative usage factor (CUF>0.1) analyses are time-limited aging analyses for Oconee." Provide the bases for this conclusion. Describe the documentation that was reviewed. Include a discussion of the applicability of the definition of TLAA in 10 CFR 54.3 to leak before-break and high energy line break postulation at Oconee.
5.4.1-2 Section 5.4.1.1.2 of the license renewal application identifies locations within the B&W scope of supply that require further evaluation for thermal fatigue. The locations that require further evaluation include the reactor vessel studs for all three units, the pressurizer spray line for Unit 3, and the Emergency Feedwater System nozzle for Unit 3. Describe the planned evaluation of these components. Provide a schedule for the completion of this evaluation. Discuss your compliance with the requirements in 10 CFR 54.21(c)(1) for these items.
5.4.1-3 Section 5.4.1.1.3 of the license renewal application indicates that the reactor coolant loop attached piping was originally analyzed to USAS B31.7, Class II standards.
The application further indicates that the fatigue evaluation of this piping to Class I standards is currently underway. Provide the schedule for the completion of these analyses. Discuss your compliance with the requirements in 10 CFR 54.21(c)(1) for these items.
5.4.1-4 Section 5.4.1.1.5 of the license renewal application addresses NRC Bulletin 88-08, "Thermal Stresses in Piping Connected to Reactor Coolant Systems." The application indicates that a supplemental response to the bulletin will be provided by July 1, 2000. Discuss your compliance with the requirements in 10 CFR 54.21 (c)(1) considering the ongoing effort regarding NRC Bulletin 88-08.
5.4.1-5 Section 5.4.1.3 of the license renewal application describes the Thermal Fatigue Management Program. The application indicates that the program, "tracks actual plant thermal cycles for those components that contain design features that have explicit design basis transient cycle assumptions in order to assure the continued validity of the component design basis." Provide a summary of the Thermal Fatigue Management Program that addresses the elements listed below. The summary should also include a discussion of the bases for each of these elements.
- a. Scope of the program that includes the specific structures and components subject to fatigue monitoring including the location monitored for each structure or component;
- b. Preventive actions that will be used to mitigate or prevent fatigue degradation;
- c. Parameter(s) to be monitored and the monitoring device(s) at each location monitored by the program;
- d. Assurance that detection of fatigue degradation will occur before loss of the structure or component intended functions;
- e. Program monitoring, trending, inspection technique, testing frequency, and sample size to ensure structure and component intended functions;
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- f. The method used to compare the monitored data to the fatigue analysis of record;
- g. Acceptance criteria to ensure structures and components can preform intended functions; and
- h. Operating experience from similar programs or inspection techniques used by Duke or the industry.