ML15112A315
| ML15112A315 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/18/1998 |
| From: | Joseph Sebrosky NRC (Affiliation Not Assigned) |
| To: | Mccollum W DUKE POWER CO. |
| References | |
| NUDOCS 9811250042 | |
| Download: ML15112A315 (9) | |
Text
November 18, 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 Civil Engineering and Geosciences Branch (ECGB) regarding the following Sections of OLRP-1001: 2.7, 3.7.1, 3.7.2, 3.7.5, 3.7.7, and 4.28.
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, 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 DOCUMENTNAME:G:\\SEBROSKY\\RAI4.WPD OFFICE LA PDLR PDLR:D NAME Berry
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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. O. 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 Chattooga River Watershed Coalition Washington, DC 20006-3708 P. O. Box 2006 DJW@NEI.ORG Clayton, GA 30525
Distribution: Hard copy PUBLIC Docket File PDLR RF M. EI-Zeftawy, ACRS T2E26 F. Miraglia Hans Ashar J. Roe Tom Cheng 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 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.
2.7 Structures and Structural Components 2.7-1 Section 2.7.2 of OLRP-1001 provides a list of concrete structural and steel components that are within the scope of license renewal and subject to aging management review (AMR). With regard to the scoping of structures and structural components (concrete and steel), address the following questions:
- a. Are there any electrical duct banks and steel structural frames at Oconee? If yes, provide basis for not including these structural components in the scope of AMR.
- b. Provide basis for not including crane columns, trolleys and mechanical cables in the scope of AMR.
- c. Section 2.7.2.2.1 provides a description of various types of pipe supports. Are there any safety-related piping systems supported by structural frames? If yes, provide an explanation how these frames are covered in the AMR.
- d. Provide basis for not considering the steel bracings between steel columns as steel components in an air environment in the AMR.
2.7-2 In Section 2.7.2 of the application, pipe piles (Section 2.7.2.1), piles (Section 2.7.2.2), and masonry block and brick walls (Section 2.7.2.1) are listed as structure components that are within the scope of license renewal and are subject to aging management review. Tables 2.7-1 through 2.7-8 of the application describe the intended functions of these structure components. Also, Section 2.7.9.1 (Switchgear Enclosures as part of the turbine building) of the application includes a statement that the transformer and switchgear enclosure is supported by battered pipe piles.
The pipe piles were filled with concrete during construction of the structure. With regard to the description of structures and structure components, address the following questions:
- a. In the description of structures and structure components (Sections 2.7.3 through 2.7.10), provide discussion of piles and masonry block and brick walls regarding their use (including the design information) and how they are included in the AMR. Have piles (not pipe piles) been used to support equipment or structures? If not, why are they included in the scope of license renewal?
Enclosure
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- b. As stated above, a description is provided in Section 2.7.9.1 (that the transformer and switchgear enclosure is supported by battered pipe piles.
Provide a clarification why the intended functions of pipe piles are listed in Table 2.7-8 (yard structure components), but not in Table 2.7-7 (Switchgear Enclosures as part of the turbine building).
2.7-3 Section 2.7.3 (Auxiliary Buildings) of the application includes a statement that all below grade construction joints in exterior walls are protected by cast in place water stops. It is the staffs interpretation that water stops are included in the scope of license renewal and are subject to the AMR. Regarding the AMR for the water stops, address the following questions:
- a. Why are water stops not listed in Section 2.7.2 (structure components within the license renewal scope) nor described in Table 2.7-1 (intended functions of structure components of the auxiliary buildings)?
- b. Based on the staffs review experience of other operating plants, water stops are commonly used in the embedded exterior walls of reinforced concrete structures.
Explain why water stops were only described in Section 2.7.3 for the auxiliary buildings.
- c. Describe the aging effects on these structure components and provide a discussion of how the aging effects will be addressed for the extended period of operation. (Water stops are not addressed in Section 3.7 of the application).
2.7-4 With regard to Table 2.7-3 of OLRP-1001, address the following:
- a. Trash racks and screens and some component supports are listed as "steel in fluid environment" that require aging management review. Are there any anchorages and embedments (with exposed surfaces and submerged in the water) to which these trash racks and screens and equipment component supports are attached? If yes, provide basis for not considering these anchorages and embedments as "steel in fluid environment" in the aging management review.
- b. Provide an explanation why the intake structure reinforced concrete roof slab is not subject to aging management review.
2.7-5 During the Oconee license renewal scoping and screening process overview meeting held on October 1, 1998, the staff was informed that tanks (including the vertical tanks erected in the field) are considered as mechanical components.
However, the tank foundation and anchorage systems are considered as structural components. With regard to the scoping process for the vertical tanks, address the following concerns:
- a. Provide a basis for not including tank supports in the discussion of OLRP Section 2.7.2, "Structural Components."
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- b. Provide the definition of the boundary (or interface) between tanks (mechanical components) and tank supports (structural components) which are usually welded to the tanks.
2.7-6 As a common industry practice, the portions of piping between the boundary of the safety-related piping and non-safety related piping and the first seismic anchors (or equivalent) beyond the boundary are treated as piping segments that provide structural supports. Provide clarification on how these segments of piping will be included and evaluated in the aging management review.
2.7-7 In Table 2.7-5, the reinforced concrete beams, etc. (under concrete components) includes primary and secondary shield walls. Explain why the intended function 1, (i.e., provides pressure boundary/or fission product barrier) is not applicable to these components.
2.7-8 Section 2.7.7.4 states that the post-tensioning components (i.e., tendon wires and anchorages) of the secondary shield wall (SSW) are subjected to aging management review. In addition, the operating experience database described in Section 3.7.7.4.3 related to the SSW post-tensioning system describes significant degradation of tendon wires. Provide justification why the prestressing forces in the tendons and prestressing losses should not be subjected to time-limited aging analysis (.TLAA).
2.7-9 Provide justification why the settlement (and differential settlement) of in-scope structures and their consequences should not be considered in the aging management review.
2.7-10 Provide sketches of the in-scope structures showing boundaries of the structures included in the AMR (e.g., Intake Structure with other water control structures and buried service water piping).
3.7 Aqing Effects for Structural Components (corresponds to Sections 3.7.1 and 3.7.2 of OLRP-1.001) 3.7.1-1 In view of the fact that expansion joints, caulking, and sealants (other than those for fire barrier) are not 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 Sections 2.7 and 3.7.
3.7.1-2 Section 3.2, referred to in Sections 3.7.1 and 3.7.2, provides tables for structures and components subjected to thermal and radiation environment. The section does not provide a systematic discussion of structures and components subjected to high humidity/moisture/water. High humidity/moisture presents challenging environment for concrete and steel structures and components, as well as for the caulking and sealants. Explain why the process to identify applicable aging effects in Sections 3.2 and 3.7 should not consider high humidity/moisture as one of the dominant challenging environments.
4 3.7.2-1 Section 3.7.2.1.6, "Oconee Operating Experience." cites one example of cracking (in the Auxiliary Building), justifies its existence, and makes a conclusion: "No additional aging effects were identified from this review beyond those identified in this section."
In Table B9 of NUREG-1557, "Summary of Technical Information and Agreements from NUMARC Industry Reports Addressing License Renewal, October 1996, in discussion of NUMARC/NRC agreement, NRC proposes a one time focussed plant specific inspection of (in-scope) structures and components for identification and resolution of potential age related degradation mechanisms (ARDMs). In 1994-1996 time frame, all safety-related structures and components within the scope of the maintenance rule (10 CFR 50.65) would have gone through baseline inspections.
Explain why the results of these inspections including the identification of ARDMs, and their resolution (corrective actions) are not cited as Oconee Operating experience in identifying the aging effects.
3.7.5 Intake Structure 3.7.5-1 The intake structure concrete and components are exposed to water from Lake Keowee, and to backfill and groundwater. Section 3.2.2.3 provides a partial list of the chemicals in these waters. Provide a complete list of chemical properties of these waters, including pH value range, and range of ions such as chlorides, nitrates, and sulfates which could potentially cause corrosion of reinforcing bars in concrete structures and steel structures.
3.7.5-2 Provide a summary of results (observations, identified degradations, corrective actions taken) of the Intake Structure baseline inspection (see also RAI 3.8.3.3) performed in accordance with Section 4.19. Also provide the results of the inspection related to the condition of caulking, seals, and expansion joints in the Intake Structure.
3.7.5-3 Are there any parts of the intake structure that are inaccessible for inspection? If so, describe what aging management program will be relied upon to maintain the integrity of the inaccessible areas. If the aging management program for the inaccessible areas is an evaluation of the acceptability of inaccessible areas based on conditions found in surrounding accessible areas, please provide information to show that conditions would exist in accessible areas that would indicate the presence of or result in degradation to such inaccessible areas. If different aging effects or aging management techniques are needed for the inaccessible areas, please provide a summary to address the following elements for the inaccessible areas: (1) Preventive actions that will mitigate or prevent aging degradation. (2)
Parameters monitored or inspected relative to degradation of specific structure and component intended functions. (3) Detection of aging effects before loss of structure and component intended functions. (4) Monitoring, trending, inspection, testing frequency, and sample size to ensure timely detection of aging effects and corrective actions. (5) Acceptance criteria to ensure structure and component intended functions. (6) Operating experience that provides objective evidence to demonstrate that the effects of aging will be adequately managed.
5 3.7.7 Reactor Building (subsection 3.7.7.4 - Post Tensioning System) 3.7.7-1 Table 3.2-1 (Oconee Thermal Environment) indicates the bounding temperatures in steam generator cavities range between 46.7 0C (116 0F) and 55.6oC (132oF). The industry experience data compiled by Ashar, H., Costello, J., Graves, H.; "Prestress Force Losses in Containments of U.S. Nuclear Power Plants, Proceedings of WANO/NEA Workshop on Loss of Prestress in NPP Concrete Containments, Civaux NPP, Poitier, France, August 25-26, 1997, indicate that at temperatures above 30'C, the loss due to relaxation of prestressing steel starts increasing at an escalated rate resulting in lower prestressing forces in tendons. Discuss how the aging management program for the secondary shield wall vertical and hoop tendons addresses prestressing forces (including the original prestress levels, design assumptions regarding losses in prestressing forces, and subsequent lift-off testing) in elevated temperatures seen in the steam generator cavities.
4.28 Tendon-Secondary Shield Wall - Surveillance Program 4.28-1 The program description of the surveillance program for SSW tendons in Section 4.28.1 indicate that the sample size is not applicable for an existing program. Under the description of "frequency," it is indicated that a random sample of tendons are lift-off tested every other refueling cycle. Please provide a more detailed description of the current SSW tendon surveillance program with regard to the size of the sample of tendons tested and the basis for the associated frequency that is currently used (this information is not described in the UFSAR), and the basis upon which you concluded that this surveillance program is adequate for the period of extended operation.
4.28-2 What population of the tendons in the SSW is inaccessible for visual examination and lift-off testing? How are you assessing the integrity of the inaccessible tendons?
Describe what aging management program that will be relied upon to maintain the integrity of the inaccessible tendons. If the aging management program for the inaccessible tendons is an evaluation of the acceptability of inaccessible tendons based on conditions found in neighboring accessible tendons, please provide information to show that conditions would exist around accessible tendons that would indicate the presence of or result in degradation to such inaccessible tendons.
If different aging effects or aging management techniques are needed for the inaccessible tendons, please provide a summary to address the following elements for the inaccessible tendons: (1) Preventive actions that will mitigate or prevent aging degradation. (2) Parameters monitored or inspected relative to degradation of specific structure and component intended functions. (3) Detection of aging effects before loss of structure and component intended functions. (4) Monitoring, trending, inspection, testing frequency, and sample size to ensure timely detection of aging effects and corrective actions. (5) Acceptance criteria to ensure structure and component intended functions. (6) Operating experience that provides objective evidence to demonstrate that the effects of aging will be adequately managed.
Nobember 18, 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 Civil Engineering and Geosciences Branch regarding the following Sections of OLRP-1001: 2.7, 3.7.1, 3.7.2, 3.7.5, 3.7.7, and 4.28.
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, 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
- See previous concurrence DOCUMENTNAME:G:\\SEBROSKY\\RAl4.WPD OFFICE LA PDLR PDLR:D NAME LBerry JSebrosky CIGrime DATE 11/17/98*
o / ft 19 it/ 1%98 OFFICIAL RECORD COPY