ML15218A127
| ML15218A127 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 01/09/1995 |
| From: | Wiens L Office of Nuclear Reactor Regulation |
| To: | Hampton J DUKE POWER CO. |
| References | |
| TAC-M89366, NUDOCS 9501120106 | |
| Download: ML15218A127 (4) | |
Text
January 9, 1994 Mr. J. W. Hampton Vice President, Oconee Site Duke Power Company P. 0. Box 1439 Seneca, SC 27679
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON ISI REQUEST FOR RELIEF 94-01 OCONEE NUCLEAR STATION, UNIT NO. 3 (TAC NO. M89366)
Dear Mr. Hampton:
By letter dated April 4, 1994, Duke Power Company submitted Request for Relief No. 94-01 from certain ASME Code requirements for Oconee Unit 3. The NRC staff, with assistance from its contractor, Idaho Engineering Laboratory, is reviewing and evaluating the relief request.
Additional information is requested, as identified in the enclosure, to enable the staff to complete its review. We are requesting that a response be provided within 60 days of the date of this letter. In addition, to expedite the review process, we suggest that a copy of your response be transmitted to the NRC staff's contractor at the following address:
Mr. Boyd W. Brown INEL Research Center 2151 North Boulevard P. 0. Box 1625 Idaho Falls, Idaho 83415-2209 This requirement affects fewer than ten respondents, and therefore, it is not subject to the Office of Management and Budget review under P.L.96-511.
Sincerely, original signed Dy Leonard A, Wipns, Senior Project Manager Project Directorate 11-3 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation Docket No. 50-287 DISTRIBUTION Docket Files OGC 0-15 B18
Enclosure:
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 9, 1994 Mr. J. W. Hampton Vice President, Oconee Site Duke Power Company P. 0. Box 1439 Seneca, SC 27679
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON ISI REQUEST FOR RELIEF 94-01 OCONEE NUCLEAR STATION, UNIT NO. 3 (TAC NO. M89366)
Dear Mr. Hampton:
By letter dated April 4, 1994, Duke Power Company submitted Request for Relief No. 94-01 from certain ASME Code requirements for Oconee Unit 3. The NRC staff, with assistance from its contractor, Idaho Engineering Laboratory, is reviewing and evaluating the relief request.
Additional information is requested, as identified in the enclosure, to enable the staff to complete its review. We are requesting that a response be provided within 60 days of the date of this letter. In addition, to expedite the review process, we suggest that a copy of your response be transmitted to the NRC staff's contractor at the following address:
Mr. Boyd W. Brown INEL Research Center 2151 North Boulevard P. 0. Box 1625 Idaho Falls, Idaho 83415-2209 This requirement affects fewer than ten respondents, and therefore, it is not subject to the Office of Management and Budaet review under P.L.96-511.
- Seely, Leonard A. Wiens, Senior Project Manager Project Directorate 11-3 Division of Reactor Projects -
I/II Office of Nuclear Reactor Regulation Docket No. 50-287
Enclosure:
Request for Additional Information cc w/enclosure:
See next page
Mr. J. W. Hampton Duke Power Company Oconee Nuclear Station cc:
A. V. Carr, Esquire Mr. Ed Burchfield Duke Power Company Compliance 422 South Church Street Duke Power Company Charlotte, North Carolina 28242-0001 Oconee Nuclear Site P. 0. Box 1439 J. Michael McGarry, III, Esquire Seneca, South Carolina 29679 Winston and Strawn 1400 L Street, NW.
Ms. Karen E. Long Washington, DC 20005 Assistant Attorney General North Carolina Department of Mr. Robert B. Borsum Justice Babcock & Wilcox P. 0. Box 629 Nuclear Power Division Raleigh, North Carolina 27602 Suite 525 1700 Rockville Pike Mr. G. A. Copp Rockville, Maryland 20852 Licensing - ECO50 Duke Power Company Manager, LIS 526 South Church Street NUS Corporation Charlotte, North Carolina 28242-0001 2650 McCormick Drive, 3rd Floor Clearwater, Florida 34619-1035 Dayne H. Brown, Director Division of Radiation Protection Senior Resident Inspector North Carolina Department of U. S. Nuclear Regulatory Commission Environment, Health and Route 2, Box 610 Natural Resources Seneca, South Carolina 29678 P. 0. Box 27687 Raleigh, North Carolina 27611-7687 Regional Administrator, Region 1I U-S. Nuclear Reaulatory Commission 101 Marietta Street, NW. Suite 2900 Atlanta, Georgia 30323 Max Batavia, Chief Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 County Supervisor of Oconee County Walhalla, South Carolina 29621
REQUEST FOR ADDITIONAL INFORMATION SECOND 10-YEAR INTERVAL INSERVICE INSPECTION REQUEST FOR RELIEF NO. 94-01 In Request for Relief No. 94-01, the licensee requested relief from the Code requirements for a number of welds and areas in the reactor pressure vessel (RPV), pressurizer, steam generators, and core flood tank. Based on the above review, the staff has concluded that the following information and/or clarification is required in order to complete the review of Request for Relief No. 94-01:
A. For RPV circumferential head Weld 3RPV-WR34 (Item B1.21),
examination coverage was limited to 43.5% due to the core catcher lugs. This coverage appears low and the limitation is not adequately described nor justified. Provide a detailed technical description of the limitation and/or a drawing depicting the location of the weld and the dimensions and locations of the obstructions.
B. For RPV shell-to-flange Weld 3RPV-WR19 (Item B1.30), the volumetric
,examination coverage was limited to 67.9% due to clad patches. The extent and exact nature (e.g., surface roughness, attenuation) of this limitation is not clear. Provide a detailed technical description and/or drawings of the examination area and the applicable limitations. In addition, it appears that the weld was examined entirely from the vessel interior surface. Can coverage be increased by examining the weld from the flange face?
Provide a technical discussion regarding this option.
C. For RPV outlet nozzle-to-shell Welds 3RPV-WR13 and 3RPV-WR13A (Item D3.9U),
examinatiun LUcea Ig I IIeLU LU I.%0 due to component geometry. This coverage appears low and is not adequately justified. Are angles other than 0* being used to increase coverage from the nozzle bore? Provide a detailed technical description of the limitations encountered when examining these welds.
D. For steam generator inlet nozzle-to-vessel Welds 3SGA-WG25 and 3SGB WG25 (Item B3.140), component geometry, specifically the support skirt, is listed as the limitation. However, this limitation should only apply to the outlet nozzle at the base of the vessel.
Provide clarification regarding the limitation for the subject steam generator inlet nozzles.
ENCLOSURE