ML15112A691
| ML15112A691 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/19/1998 |
| From: | Joseph Sebrosky NRC (Affiliation Not Assigned) |
| To: | Mccollum W DUKE POWER CO. |
| References | |
| NUDOCS 9811250136 | |
| Download: ML15112A691 (7) | |
Text
November 19, 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 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-1 001 and has identified, in the enclosure, areas where additional information is needed to complete its review. Specifically, the enclosed questions are from the Civil Engineering and Geosciences Branch and Mechanical Engineering Branch regarding OLRP-1001 Sections 2.3, 3.3, 4.8, 5.3.1, and 5.3.2.
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, 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\\RAIl6.WPD OFFICE 1, PM:PDLR PDLR:D NAME LBerr JSebrosky ClGrime DATE 1
198 11/ 1/98 11/
/98 OFFICIAL RECORD COPY 9811250136 981119 PDR ADOCK 050002689 P
DISTRIBUTION: Hard copy Docket File PUBLIC PDLR RF M. EI-Zeftawy ACRS T2E26 F. Miraglia H. Ashar J. Roe M. Hartzman 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
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.
2.3 Reactor Building (Containment) Structural Components Note: Previous questions in Sections 2.3 and 3.3 were sent to Duke in a November 14, 1997, letter. The numbers below are a continuation of the numbering system that was established in the November 14, 1997, letter.
2.3-8 (Question refers to Section 2.3.2 of OLRP-1001) The tendon gallery provides access to the bottom anchorages of the vertical tendons of containment's post-tensioning system. It also protects the anchorages from the direct influence of soil and ground water, which they would be subjected to if there were not a tendon gallery. It serves intended function 3 as defined in Table 2.3-1 of Duke Energy Corporation Report OLRP-1001, June 1998. Provide justification as to why it should not be within the scope of license renewal as per 10 CFR 54.4(a)(2).
In addition, please provide a discussion regarding what aging effects of water infiltration would be in the tendon gallery on the tendon anchorage system (e.g.,
tendon end caps, tendons, and basemat concrete) to ensure that the intended function of the tendon anchorage system is maintained during the period of extended operation.
2.3-9 to the Duke Power's letter of August 12, 1998, indicates that the miscellaneous attachment welds to the liner are added to Sections 2.3.3.2 and 2.7.7.
A review of Section 2.3.3.2 indicates that these welds are not considered within the boundary of containment, but they are addressed in Section 2.7.7 (Reactor Building Internal Structural components). Section 2.7.7 does not specifically address attachment welds to the liner. Provide justification as to why these and other attachments to the liner, which could influence the leak-tight integrity of the liner, should not be considered as part of the containment pressure boundary.
2.3-10 (Question refers to Section 2.3.3 of OLRP-1001) With the confirmation from Duke that there are no bellows in Oconee containment penetrations, the staff had resolved RAI 2.3-7 (Draft Technical Evaluation Report for OLRP-1001, Brookhaven National Laboratory, July 1998). However, in Attachment 2 to Duke's Letter to NRC on containment review for renewed operating license, August 12, 1998, in response to RAI 2.3-7, Duke states, "No revisions to OLRP-1001 are required at this time."
Explain the phrase "at this time."
3.3 Aging Effects for Reactor Building (Containment) Structural Components 3.3-19 (Question refers to Section 3.3.1 of OLRP-1 001) In view of the fact that expansion joints, caulking, and sealants associated with containment structure integrity (e.g.
moisture barrier at the junction of liner plate and basemat fill concrete) are not Enclosure
2 subjected to replacement based on qualified life or specified time period, explain why they should not be considered for aging management review [see 10 CFR 54.21(a)(1)(ii)] and addressed in this Section.
3.3-20 Section 3.3.2.3 indicates that the degradation mechanisms applicable to cracking of concrete in the Oconee containments are as per Ref. 3.3-5 (ACI-201) of OLRP 1001. However, the Oconee containments are subjected to significant compression and splitting tensile stresses at the locations of post-tensioning anchorages (buttresses, top and side of ring-girders, bottom of the basemat in tendon galleries).
Environmental effects such as freeze-thaw and temperature, and material characteristics such as alkali-silica reaction can accentuate the mechanically induced cracking. Explain why such synergistic effects should not be considered as contributing to cracking in Oconee containments and addressed as such as part of the aging management programs for the containments..
3.3-21 In response to RAI 3.3-1 (Attachment 2 to Duke's letter to NRC on containment review for renewed operating license, August 12, 1998) you state that Section 3.3.2.7 has been revised to address inaccessible areas of containment concrete components. Your OLRP-1001, Section 3.3.2.6 indicates one sentence related to symptomatic evidence of cracking and leaching. Provide an explanation for why the below ground areas of concrete surfaces and embedded areas of liner plate (the inaccessible areas) should not be explicitly considered for aging management review based on the evidence cited in Sections 3.3.2.6 and 3.3.3.6.
4.8 Containment Inservice Inspection Plan 4.8-1 In Section 4.8, Duke states three options for implementing the containment inservice inspection plan. Any one of the options is acceptable for inspection during a specific interval. However, for options 2 and 3, prior NRC approval is required. Describe your implementation plans for the following:
- a. The interval starting with the first inspection period as per 10 CFR 50.55a(g)(vi)(B),
- b. the subsequent inspection intervals in the current license, and
- c. the inspection intervals in the renewed license period.
4.8-2 In Section 4.8, Duke Power describes the extent of application of ASME Section XI, Subsection IWE for the examination of containment liner and penetrations.
Subarticle IWE 1240 of the ASME subsection requires augmented examinations of suspect areas. In light of the operating experience of the Oconee containments (e.g., Section 3.3.3.6), describe what areas are identified for augmented examinations for the current and extended license periods.
4.8-3 In Section 4.8.2 (Subsection IWL Examinations), you state that sample size is not applicable for an existing program. For examination Category L-B (unbonded post tensioning tendons), sample size plays a significant role. In 1996, you performed the first inspection of the Oconee containment post-tensioning tendons with randomly selected tendons. Oconee has a limited database for performing a
3 trending analysis. Under the current license, with the normal sample size (i.e., 2% of the population), you would have no more than 6% of the tendon prestress forces available for each containment for trending analysis.
With this limited database, in 1997, you performed trend analyses of the tendon prestressing forces in the three units, and discovered that vertical tendon forces in Unit 1 and Unit 2 would go below the required minimum value at 30 years and 10 years after the Units' structural integrity testing. In light of the operating experience, provide justification why the sample size should not be increased to systematically assess the need for retensioning of tendons prior to the license renewal period so that the tendon prestressing forces could be later managed as a limited aging analysis (TLAA).
4.8-4 In Section 4.8.2 (Subsection IWL examinations), you state that acceptance criteria are specified in IWL-3000. For concrete and areas around tendon anchorages, IWL 3211 leaves the acceptance criteria to the judgement of the Responsible Engineer.
In light of the operating experience about concrete cracking and degradations around tendon anchorages (Refs. 2, 3, 4), describe the acceptance criteria established by the responsible engineer that would be used during the current license period, as well as, the renewed license period.
5.3 Time-Limited Aging Analyses (TLAA) for the Reactor Building (Containment) Structural Components 5.3.1-1 Section 5.3.1 of the license renewal application states that the design of the containment penetrations meets the general requirements of the ASME Boiler and Pressure Vessel Code,Section III, "Nuclear Vessels," 1965, for thermal cycling. The application assumes 500 thermal cycles due to containment heat up and cooling, but is silent regarding specific cumulative usage factors (CUF).
- a. Provide the highest CUF determined for each containment penetration on the basis of 500 thermal cycles combined with relevant mechanical cyclic loading, and loading due to anchor motion.
- b. Provide the basis for the statement "The projected number of cycles for each Oconee unit through 60 years of operation has been determined to be less than the original 360 cycle design limit."
- c. Describe the approach that was used in the fatigue analysis of the containment penetrations to account for environmental effects.
5.3.1-2 State if containment penetrations fall under the Oconee Thermal Fatigue Management Program, described in Section 5.4.1.3 of the license renewal application. If not, provide justification for the exclusion. If yes, provide the following information for each location monitored by the program:
- a. The location monitored;
- b. The parameters monitored;
4
- c. The method used to compare the monitored data to the fatigue analysis of record.
5.3.2-1 The prescribed lower limit lines (PLL) in Figures 1, 2, and 3 of Appendix 16.6-2 indicate that you are using the same values for elastic shortening losses and the time-dependent losses (i.e., creep and shrinkage of concrete, and relaxation of prestressing steel) for calculating the anchorage forces for each group of tendons.
In fact, the elastic shortening loss varies from tendon to tendon, and could vary between 0% to 6% of the anchoring force, and is not time-dependent (see Regulatory Guide 1.35.1, July 1990). Describe, how you account for this discrepancy in comparing the PLL with the measured lift off force of a randomly selected tendon.
5.3.2-2 Figures 1, 2, and 3 (Appendix 16.6-2 of UFSAR supplement) do not include information regarding the operating experience or trend lines for the existing tendon prestressing forces for the three containments at Oconee. The TLAA should be based on the most reliable information available. Provide information (including the actions you plan to take) that would ensure that the actual individual tendon forces will exceed corresponding PLL for the extended lives of the three units.