ML15112A684
| ML15112A684 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/20/1998 |
| From: | Joseph Sebrosky NRC (Affiliation Not Assigned) |
| To: | Mccollum W DUKE POWER CO. |
| References | |
| NUDOCS 9811250094 | |
| Download: ML15112A684 (7) | |
Text
November 20, 199 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 U.S. 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 Materials and Chemical Engineering Branch regarding Sections 3.4.5, 4.10, 4.3.1, 4.24, and 5.4.2 of OLRP-1001.
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 staff's requests for additional information.
Sincerely, original signed by:
Joseph M. Sebrosky, Project Manager License Renewal Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270, and 50-287
Enclosure:
Request for Additional Information cc w/encl: See next page DISTRIBUTION: See next page DOCUMENT NAME:G:\\SEBROSKY\\RAI15.WPD OFFICE LA PM:PDLR PDLR:D NAME LBerry JSebrosky ClGrimes DATE 11/1 198 11/198 1 1/ZO/98 OFFICIAL RECORD COPY
~9811250094 981120 PDR ADOCK 05000269 P
DISTRIBUTION: Hard copy Docket File PUBLIC PDLR RF M. EI-Zeftawy ACRS T2E26 E-mail; F. Miraglia B. Elliot J. Roe H. F. Conrad D. Matthews J. Medoff 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
Oconee Nuclear Station (License 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
REQUEST FOR ADDITIONAL INFORMATION OCONEE NUCLEAR STATION, UNITS 1. 2. AND 3 LICENSE RENEWAL APPLICATION, EXHIBIT A OLRP-1001 Section No.
3.4.5 Reactor Vessel 3.4.5-1 The following general issue with respect to plant aging needs to be addressed:
Based on its evaluation of operating experience, the NRC has determined that potential aging effect mechanisms in components of pressurized water reactor vessels are as indicated in the Table 3.1-3 of the Draft Standard Review Plan for License Renewal. Table 3.1-3 identifies components that are considered part of the reactor pressure vessel (RPV) and identifies the associated aging effects for the components. Identify the equivalent components in the Oconee reactor pressure vessels and identify the aging effects (identified as significant or unresolved in Table 3.1-3) applicable to these components and where they are addressed in the application. For those aging effects that are not addressed explain why they are not applicable.
Note: Questions 3.4.5-2 through 3.4.5-8 discuss how the Oconee license renewal application relates to BAW-2251. There are aspects of the questions that involve sections 3.4.5, 4.24, and 5.4.2 of Oconee's license renewal application. The questions have all been placed in this section for convenience.
3.4.5-2 The following are action items to be addressed by a plant-specific license renewal application when incorporating by reference the Babcock & Wilcox Owners Group (B&WOG) topical report, BAW-2251. Provide the following:
a) The license renewal applicant is to verify that its plant is bounded by the topical report. Further, the renewal applicant is to commit to programs described as necessary in the topical report to manage the effects of aging during the period of extended operation on the functionality of the reactor vessel components.
Duke Energy, the applicant for license renewal will be responsible for verifying that any such commitments are subject to appropriate regulatory control. As such, identify any deviations from the aging management programs described in Topical Report BAW-2251. Evaluate any deviations on a plant-specific basis in accordance with 10 CFR 54.21(a)(3) and (c)(1).
b) B&WOG has determined that the lower control rod drive mechanism (CRDM) service support structure, including the weld that connects the lower CRDM service support skirt to the reactor vessel closure head, and the reactor vessel support skirt, including the weld that connects the reactor vessel support skirt to the transition forging, are subject to an aging management review for license renewal. However, the B&WOG has decided to exclude them from the scope of the Topical Report BAW-2251. Identify which aging effects are applicable to these components and describe your aging management program for these components in the license renewal application.
Enclosure
2 Note: Additional plant-specific open Items that need to be addressed relative to the contents of the license renewal application and Topical Report BAW-2251 are discussed in question 3.4.5-3 through 3.4.5-8.
3.4.5-3 Intended Function of Reactor Vessel Components Identify whether the intended function of the reactor vessel internals is to maintain the capability to shut down the reactor and maintain it in a safe-shutdown condition.
3.4.5-4 Flow Stabilizers Subject to Aging Management Review The staff has concerns about whether the flow stabilizers should be excluded from an aging management review for license renewal. Although the flow stabilizers themselves do not have safety-related functions, they were installed to address flow induced vibration (FIV) problems experienced during hot functional testing. Thus, cracking of the attachment weld may cause the reactor vessel shell to crack thereby affecting its intended functions. Indicate if an aging management program is provided to manage the aging effects on the flow stabilizers. If so, provide the details of such a program; if not justify why such a program is not needed to ensure the integrity of these stabilizers over the extended life for the units.
3.4.5-5 Wear of Core Guide Lugs The staff considers loss of material due to mechanical wear of the core guide lugs a potential applicable aging effect that should be managed for license renewal. This potential aging effect is discussed in Section 3.1 of the working draft standard review plan for license renewal. Indicate if an aging management program is provided to manage the aging effects on the lugs. If so, provide the details of such a program; if not justify why such a program is not needed to ensure the integrity of the lugs over the extended life for the units.
3.4.5-6 Underclad Cracking Cracking has been detected under the austenitic stainless steel weld cladding in reactor vessel forgings. When cracks are detected, the licensee performs a time limited aging analysis (TLAA) to evaluate the integrity of the reactor vessel.
However, the staff considers the potential for underclad cracks to grow during plant operation an applicable aging effect to be managed for license renewal. Indicate if an aging management program is provided to manage the aging effects on the stainless steel cladding in the forgings. If so, provide the details of such a program.
If not, justify why such a program is not needed to ensure the integrity of reactor vessel forgings.
3.4.5-7 Reactor Vessel Materials Surveillance Program To ensure that the results of fracture toughness tests remain valid during the extended license period, describe the operating limitations necessary for ensuring that each plants' operating conditions (temperature and neutron fluence) do not invalidate the results of fracture toughness tests conducted on surveillance capsules
3 removed from the Oconee reactor pressure vessels during the original 40-year license periods for the plants.
3.4.5-8 Additional Limitations on Pressure-temperature (P-T) Limits and Reactor Coolant Pump Seal Limits Based on the projected P-T limits at the end of the extended license period and other plant operating limits (e.g., limits of pump seal pressure), identify whether the.
operating windows for the Oconee units will be sufficient to start up and shut down the units at the end of the extended license period. If the operating windows are insufficient, provide aging management programs to increase the operating windows or reduce the amount of neutron embrittlement to the Oconee RPVs.
4.3.1 Alloy 600 Aging Management Program Note: Question 4.3.1-1 and 4.10-1 were originally grouped together.
4.3.1-1 In regard the content of Section 4.3.1, "Alloy 600 Aging Management Program" (henceforth the Alloy 600 AMP) to the License Renewal Application:
- a. The section states that the Alloy 600 AMP will be used to identify and inspect the four most susceptible locations within the Oconee reactor coolant systems (RCS). Clarify whether the scope of the proposed inspections of the four most susceptible locations will be on different components within the RCS or on redundant ("sister") components in the RCS.
- b. Clarify whether the aging management program (Section 4.10 of the License Renewal Application) for the Oconee Alloy 600 vessel head penetration (VHP) nozzles and associated Alloy 82/182 partial penetration welds is a separate program from the Alloy 600 AMP and if it will be implemented in addition to the Alloy 600 AMP.
4.10 Control Rod Drive Mechanism Nozzle and Other Vessel Closure Penetrations Inspection Program 4.10-1 Regarding the content of Section 4.10, "Control Rod Drive Mechanism Nozzle and Other Vessel Closure Penetrations Inspection Program:"
In Section 4.10 to the License Renewal Application Duke indicated, in part, that the existing regulatory basis for the aging management program for Alloy 600 VHP nozzles is provided in the Duke response to Generic Letter 97-01, "Degradation of Control Rod Drive Mechanism Nozzle and Other Vessel Closure Head Penetrations." In its response to GL 97-01, Duke indicated that it was a participant in the joint Babcock and Wilcox Owners Group (BWOG)/Nuclear Energy Institute (NEI) integrated program for assessing primary water stress corrosion cracking (PWSCC) in VHP nozzles to B&W designed VHP nozzles, and that this program was contained in BWOG Topical Report BAW-2301. On May 14, 1998, the NEI submitted an integrated "industry Histogram for Reactor Vessel Head Penetration" on behalf of PWR licensees participating in NEI's integrated assessment program for control rod drive mechanism (CRDM) penetration nozzles and other VHP nozzles in
4 domestic PWR designs. The histogram ranked the CRDM penetration nozzles in "less than 5 year," "5 to 15 year," and "beyond 15 year" probabilities of failure categories. The CRDM penetration nozzles of Oconee have been designated as falling into the "less than 5 year" category and inspections of the Oconee Unit 2 CRDM penetration nozzles have been scheduled to be reinspected for a second time in the year 1999. F-owever, the current integrated program and susceptibility assessment for the PWR industry is based on a 40-year (normal life) time frame.
Provide the following information with respect to how the license renewal term for the Oconee units relates to the industry's integrated program for assessing domestic PWR VHPs:
- a. Indicate whether Duke is committed to extending its participation in the BWOG integrated aging management program for VHP nozzles during the license renewal term for the Oconee units.
If Duke is committed to extending its participation in the integrated program to the license renewal term, indicate how the integrated program will be used as the basis for proposing any further inspections of the VHP nozzles at Oconee Units 1, 2, and 3 during the extended license terms for the facilities.
ii.
If Duke is not committed to extending its participation in the integrated program to the license renewal term, describe what the basis (in addition to the inspections of the Oconee Unit 2 VHP nozzles in 1999) will be for assessing the potential for primary water stress corrosion cracking to exist in the Oconee VHP nozzles and for proposing any further inspections of the Oconee VHP nozzles during the extended license terms for the facilities.