ML20034C515
| ML20034C515 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 04/27/1990 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20034C518 | List: |
| References | |
| 90-183, GL-87-12, GL-88-17, NUDOCS 9005040062 | |
| Download: ML20034C515 (4) | |
Text
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.i VIRGIN!A Ut,1?CTHIC AND POWl:H COMPANY HICIIMOND,VIkotNIA 20061 April 27, 1990 United States Nuclear Regulatery Commission Serial No.
90 183 Attention: Document Control Desk NAPS /PAK/TAH:bgp Washington, D.C. 20555 Docket Nos.
50 338 50 339 License Nos. NPF 4 NPF-7 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGE RESIDUAL HEAT REMOVAL SYSTEM Pursuant to 10 CFR 50.90, Virginia Electric and Power Company requests an l :
amendment in the form of changes to the Technical Specifications (T.S.) of the Operating Licenses Nos. NPF 4 and NPF 7 for North Anna Power Station Units 1 and 2, respectively.
These changes will enhance Residual Heat Removal (RHR) system reliability by removing the automatic isolation requirement for the RHR Suction valves from Technical Specifications 3/4.7.9. A requirement to verify these valves closed and deenergized prior to exceeding 500 psig in the Reactor Coolant System (RCS) will be added.
Currently, Technical Specifications require an automatic closing function for the RHR suction valves whenever the RCS pressure exceeds 660 psig. This ensures double valve isolation between the high pressure RCS and the low pressure RHR system, but industry experience has shown it to be a major cause of inadvertent losses of decay heat removal capability, it has also been addressed specifically as being a major 1
contributor to Loss of Decay Heat Removal events in Generic Letters 8712 and 8817.
Deleting the interlock will increase the reliability of the RHR System during cold shutdown conditions while ensuring the isolation of the RHR System from the RCS will i
still be required by Technical Specifications.
The proposed change will also modify the surveillance testing requirements of the RHR pumps to be in accordance with Technical Specification 4.0.5 which invokesSection XI of the ASME Boller and Pressure Vessel Code for inservice testing..
In addition, there are several administrative changes to specifications 3/4.7.9.1 and 3/4.7.9.2.
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a The proposed changes are provided in Attachments 1 and 2 for the current Technical Specifications and the MERITS Technical Specifications, respecti"ely.
The discussions, safety evaluations and the significant hazards determinations for these i
changes are enclosed in Attachment 3.
This request has been reviewed and approved by the Station Nuclear Safety and Operating Committee. It has been determined that the proposed changes do not involve an unreviewed safety question as defined in 10 CFR 50.59 or a significant hazards consideration as defined in 10 CFR 50.92.
Very truly yours, k
W.. tewart Senior Vice President Nuclear Attachments
- 1) Proposed Technical Specification Changes - STS Format
- 2) Proposed Technical Specification Changes MERITS Format
- 3) Discussion, Safety Evaluation and Significant Hazards Determinations c
cc:
United States Nuclear Regulatory Commission Region 11 101 Marietta Street, N.W.
Suite 2900 Atlanta, GA 30323 Mr. M. S. Lesser NRC Senior Resident inspector Nort~n Anna Power Station Commissioner Department of Health Room 400 109 Governor Street Richmond, Virginia 23219 l
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COMMONWEALTH OF VIRGINIA )
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COUNTYOFHENRICO
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The foregoing document. was acknowledged before me,'in and for the County and Commonwealth aforesaid, today by W. L. Stewart who is' Senior Vice President - Nuclear, of Virginia Electric and Power Company.
He 'is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to.the best of his knowledge and belief.
Acknowledged before me this 27 day of
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