ML18153A761
| ML18153A761 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 09/12/1995 |
| From: | Matthews D NRC (Affiliation Not Assigned) |
| To: | Ohanlon J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| TAC-M93565, NUDOCS 9509150047 | |
| Download: ML18153A761 (4) | |
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I UNITED STATES NUCLEAR REGULATORY COMMISSION Mr. J.P. O'Hanlon WASHINGTON, D.C. 20555-0001 September 12, 1995 Senior Vice President - Nuclear Virginia Electric and Power Company 5000 Dominion Blvd.
Glen Allen, Virginia 23060
SUBJECT:
VIRGINIA ELECTRIC AND POWER COMPANY, (VEPCO}, SURRY POWER STATION UNIT 1 ASME SECTION XI RELIEF REQUEST (TAC NO. M93565}
Dear Mr. O'Hanlon:
We have completed our review of the Virginia Electric and Power Company submittal dated Sept~mber 11, 1995, concerning the leakage in the "A" residual heat removal (RHR} pump casing in the Surry Power Station, Unit 1.
Pursuant to 10 CFR 50.55a{g)(5}(iv), you requested relief from the ASME Section.XI (Code) pressure boundary leakage and flaw evaluation requirements to permit a determination of operability of the pump.
The pump casing is cast ASTM A351~
Grade CF-8A stainless steel. The leak rate has been measured at less than two drops per minute from a site that is approximately 0.5 inches in diameter on the suction side of the pump casing.
The leak site appears to have occurred by a slow corrosion mechanism at a casting defect such as shrinkage, porosity, or sand inclusion.
No linear indications are displayed.
Growth of the defect by stress*corrosion cracking is not anticipated due to the low stress in any mode of loading including seismic events, the low operating temperature, and the excellent resistance of this grade of stainless steel to primary water stress corrosion cracking.
In the above cited submittal, you stated that you had considered two repair options.
The first was to maintain the plant in the current condition, isolate the pump and repair the leak. This was considered to be impractical, since isolation to the pump is insufficent to assure an acceptable repair.
The second option involved immediate depressurization of the reactor coolant system but postponement of the repair of the pump for approximately 10 days*
until the fuel has been off-loaded from the reactor vessel. This would allow the 11A 11 train of the RHR system to be available for decay heat removal in the unlikely event that the "B" train of the RHR system degrades before decay heat removal is no longer required.
You also assessed the risk associated with both options and found the second to be preferred from a.risk perspective.
Two structural analyses of a representative cross section of the RHR pump were conducted by VEP.CO to verify the ability of the pump to carry the design basis loading with the identified flaw.
Limit load analysis was performed to determine the largest sized crack the* pump can sustain without ductile rupture when subjected to the design basis loading, even though the defect did not show any evidence of linear flaws.
The limiting flaw size determined by analysis was 12.00 inches compared to the flaw size of 0.50 inches. Although this -is a cast material and, perhaps, has had its toughness reduced by thermal aging, sufficiently large margins to failure are available for the time period l
- in question.
The second analysis performed was to determine the largest un-reinforced opening that the pump body could sustain before ductile tearing could occur.
The analysis indicated that the opening would be 4.0 inches, significantly larger than the 0.50 inch flaw identified.
Initially, you had made a determination that RHR pump "A" was inoperable since it contained a through wall leak and, therefore, exceeded the flaw evaluation requirements of ASME Section XI.
Subsequently, you examined Generic Letter 91-18, "Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions on Operability," for clarification on operability.
Generic Letter 91-18 refers to Generic Letter 90-05, "Guidance for Performing Temporary Non-Code Repair on ASME Class 1, 2, and 3 Piping," for conducting the analysis to verify the ability of a component to carry the design basis loading with an identified flaw.
You completed the structure analyses and verified that the RHR "A" train pump would maintain structural integrity under design basis loadings. Therefore, you concluded and the staff agrees that although the pump is degraded, it is appropiate to consider the pump operable from a functional point of view.
The NRC staff has reviewed the Virginia Electric Power Company submittal requesting relief.from ASME Boiler and. Pressure Vessel Code,Section XI requirements as impractical for the pump where the RHR system may be required to meet the shutdown cooling requirements.
Imposing the requirements on the facility with the existing plant condition could result in an unacceptable repair because of insufficient isolation with the subsequent unavailability of the RHR pump for plant cooldown.
Accordingly, the NRC staff finds that the.
Code requiements are impractical for these plant conditions and that relief is granted pursuant to 10 CFR 50.55a{g){6){i). Further, the NRC staff finds that such relief will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Based on the evaluation, the staff concurs that the RHR pump can be considered degraded but operable since integrity will be maintained and the leakage will not affect the hydraulic performance of the RHR pump until a code repair can be safely performed.
Further, the staff concurs that risk will be minimized by implementation of your second* proposed repair option.
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- This completes our efforts on this issue and we are, therefore, closing out TAC No. M93565.
Contact:
- 8. Buckley, NRR 415-1452 cc:
See next page Distribution Oocket fi 1 e--:;
PUBLIC PDII-1 RF S. Varga OFFICE NAME DATE COPY J. Zwolinski ACRS (4)
M. Verrelli, RII Sincerely,
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I
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\\David 8. Matthews, Director Rroject Directorate II-1 Dtvision of Reactor Projects - I/II Office of Nuclear Reactor Regulation OGC*
EHoller 09/12/95 Yes/No
Mr. J. P. O'Hanlon Virginia Electric and Power Company cc:
Michael W. Maupin, Esq.
Hunton and Williams Riverfront Plaza, East Tower 951 E. Byrd Street Richmond, Virginia 23219 Mr. David Christian, Manager Surry Power Station Post Office Box 315 Surry, Virginia 23883 Senior Resident Inspector Surry Power Station U.S. Nuclear Regulatory Commission 5850 Hog Island Road Surry, Virginia 23883 Chairman Board of Supervisors of Surry County Surry County Courthouse Surry, Virginia 23683 Dr. W. T. Lough Virginia State Corporation Commission Division of Energy Regulation Post Office Box 1197 Richmond, Virginia 23209 Regional Administrator, Region II U.S. Nuclear Regulatory Commission 101 Marietta Street N.W., Suite 2900 Atlanta, Georgia 30323 Robert B. Strobe, M.D., M.P.H.
State Health Commissioner Office of the Commissioner Virginia Department of Health P.O. Box 2448 Richmond, Virginia 23218 Surry Power Station Units 1 and 2 Office of the Attorney General Commonwealth of Virginia 900 East Main Street Richmond, Virginia 23219 Mr. M. L. Bowling, Manager Nuclear Licensing l Programs Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, Virginia 23060