ML20097D165

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Forwards Revised ISI Relief Request NDE-16 Re Volumetric Exam of Pressurizer nozzle-to-vessel Weld & Nozzle Inner Radius Section
ML20097D165
Person / Time
Site: North Anna Dominion icon.png
Issue date: 02/07/1996
From: Ohanlon J
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
96-031, 96-31, NUDOCS 9602120219
Download: ML20097D165 (13)
v • d • e


Text

.

s VIRGINIA ELECTHIC AND POWER COMPANY RicuMown,Vinoin A cac61 February 7,1996 U.S. Nuclear Regulatory Commission Serial No.96-031 Attention: Document Control Desk NL&OS/EJW i

Washington, D.C. 20555 Docket No.

50-339 License No.

NPF-7 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 2 INSERVICE INSPECTION REQUEST FOR RELIEF North Anna Unit 2 is presently in the second period of the second ten-year interval.

Examinations are being conducted in accordance with the requirements of Section XI to the 1986 Edition of the ASME Boiler and Pressure Vessel Code.

By letter Serial No.92-193, dated March 27,1992, we re-submitted the North Anna Unit 2 Second Ten-Year Interval inservice Inspection Program (original was submitted by letter Serial No.90-662, dated November 1,1990). A request for relief (Relief Request NDE-16) pertaining to volumetric examination of the pressurizer nozzle-to-vessel weld and nozzle inner radius section was included in that re-submittal. Relief Request NDE-16 was denied by letter dated November 5,1992, " Safety Evaluation (SE) for North Anna Power Station, Unit No. 2 (NA-2)/Second 10-Year Inservice Inspection (ISI) Program with Associated Relief Request (TAC No. M79147)." The SE Report recommended that a remote visual examination be performed on the internal surface of the pressurizer nozzle. However, a remote examination is impractical based on restricted access to the weld area and the lack of any significant problems reported for similarly configured welds.

In addition, similar relief requests have subsequently been approved for North Anna Unit 1 and Surry Unit 1.

The relief request has been revised to include a more detailed basis for relief and is being re-submitted pursuant to 10 CFR 50.55a(g)(5).

If you have any further questions, please contact us.

Very truly yours, J. P. O'Hanlon Senior Vice President - Nuclear Attachment 960212021? 960207 PDR ADOCK 05000339 s

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U.S. Nuclear Regulatory Commission Region ll 101 Marietta Street, N.W.

Suite 2900 Atlanta, Georgia 30323 R. D. McWhorter NRC Senior Resident inspector North Anna Power Station 1

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RELIEF REQUEST NDE-16 (Revised) l 1.

IDENTIFICAT!GN OF COMPONENTS Nozzle-to-Vessel Welds Nozzle Inner Radius Section Pressurizer: (2-RC-E-2)

Drawina No.

Weld No.

12050-WMKS-RC-E-2 9

9NIR ll.

IMPRACTICABLE CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code,1986 Edition, requires in Category B-D, item No. B3.110 and B3.120, that the Pressurizer Surge Line Nozzle-to-Vessel Weld and Nozzle inside Radius Section to be volumetrically examined. Relief is requested from these requirements.

Ill.

BASIS FOR RELIEF The North Anna Unit 2 pressurizer surge nozzle is surrounded by 78 heater penetrations (see attached drawings). Engineering recommends that the l

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heater cables be disconnected prior to the removal of insulation.

This recommendation is due to the possibility of damage to the heater element connections if the insulation is removed while the cables are connected.

Based upon the most recent survey of the applicable area, the dose rate is 300 MR/ hour in the general area, 500 MR/ hour at one foot, and 1500 to 2000 MR/ hour contact.

Based upon estimates provided by site Electrical Maintenance, Insulation Removal, and ISI/NDE, it would require ten man-hours to disconnect and reconnect the heater cabies, four man-hours to remove and reinstall the reflective insulation and seven man-hours to prepare and examine the nozzle-to vessel weld and nozzle inside radius section. The resulting dose estimate for these examinations is 15.3 man-rem.

Based upon a review of the fabrication drawings, the estimated percentage of the required volume that could be examined on the pressurizer surge line nozzle to-vessel weld (9) is as follows:

EXAMINATION ANGLEi PERCENTAGE EXAMINED 45 Degrees 60%

60 Degrees 40%

0 Degrees 80%

The examination coverage of the nozzle inside radius section (9NIR) would be somewhat larger values, however; we feel that the confined access to the nozzle as a result of the pressurizer skirt, surge line piping and heater penetrations, and area dose rates would result in only a "best effort" examination in either case. Therefore, it is felt that the gain in system integrity is not commensurate with the exposure received from the examinations.

A remote visual examination of the inside surface was considered as an alternative but was eliminated for the following reasons:

1) Access to the pressurizer surge nozzle is restricted by a strainer basket. The basket has ! inch holes on a j inch triangular pattern. The distance from the manway to the nozzle is over 40 feet long. It would be difficult to position a borescope to obtain 100% coverage.
2) This examination was performed at another utility with a similar arrangement. They were not able to pass a borescope through the strainer.

The surface of the area below the strainer was reported to be murky and d'd not give the best picture. The examination did reveal some cracks ir.1.no cladding. These cracks were evaluated by UT and determined to be very shallow. The cladding cracks were determined to be caused by thermal shock of cold water flowing into the pressurizer while it was empty with the heaters on.

3) Aside from the event referenced in item 2, we know of no other problem reported by industry or the NRC concerning cracking in the area of the pressurizer nozzle.

A similar relief request was approved for North Anna Unit 1 and Surry Unit 1.

IV.

ALTERNATIVE EXAMINATION A visual (VT-2) examination of the pressurizer surge line nozzle-to-vessel weld will be performed during the normally scheduled system leakage test each refueling.

In addition, Technical Specifications requires that the Reactor Coolant System Leak Rate be limited to one gallon per minute unidentified leakage. This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Additionally, the containment atmosphere particulate radioactivity is monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

No additional attemative requirements are deemed necessary.

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