ML18153C630
| ML18153C630 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 05/23/1991 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 91-269, NUDOCS 9106030009 | |
| Download: ML18153C630 (5) | |
Text
L VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 May 23, 1991 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 INTERIM ASME SECTION XI RELIEF REQUEST SYSTEM PRESSURE TEST PROGRAM 91-269 Serial No.
NO/ETSR8 Docket Nos.
50-280 50-281 License Nos. DPR-32 Surry Unit 2 is currently in a refueling outage and is performing Code required examinations. Surry Unit 1 is currently at full power, having completed its latest refueling outage on December 20, 1990.
Previously by letter dated February 15, 1989 (Serial No.89-083), Virginia Electric and Power Company withdrew relief requests RR-22 and RR-21 from the ISi Program.
These relief requests addressed the concerns associated with performing visual examinations (VT-2) of the underside of the reactor vessel and the instrumentation welds in that area as required by ASME Section XI.
The examinations required as a result of the withdrawn relief requests will be conducted during the current refueling outage for Surry Unit 2. However, due to an administrative oversight these same examinations were not conducted during the Surry Unit 1 refueling outage. Therefore, pursuant to 1 O CFR 50.55a paragraph g(5),
interim relief is being requested from the specific ASME Section XI requirements not performed in the previous Surry Unit 1 refueling outage. This relief is requested until the next scheduled refueling outage currently scheduled to commence in April, 1992.
If these examinations prove to be impractical to accomplish due to the constraints of our subatmospheric containments, we may need to reconsider a permanent relief request in the future. The attachment to this letter provides the basis for the interim relief.
This interim relief request has been reviewed and approved by the Station Nuclear Safety and Operating Committee.
If you have any further questions or require additional information, please contact us.
Very truly yours, Jl <;fM=-
- w. L. Stewart Senior Vice President - Nuclear Attachment r-~9:;-:;:lr0~6-;;;:(l~:::::-:;::o-:;;;:o-.;;:-o,:;;;:-;,-,::--;, 1:--(=--::)::"=-=,~=-=.o:::c---: --.
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U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station
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ATTACHMENT INTERIM RELIEF REQUEST ASME SECTION XI RELIEF REQUEST SYSTEM PRESSURE TEST PROGRAM
INTERIM RELIEF REQUEST SYSTEM PRESSURE TEST PROGRAM I. Identification of Component System: Reactor Coolant Components:Partial Penetration Welds Bottom of the Reactor Vessel (1-RC-R-1)
ISi Class: 1 II. Code Reguirement Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition, Winter 1980 Addendum, Category 8-E, item 84.13, requires reactor vessel partial penetration welds to have a visual (VT-2) examination. In addition Category B-P, item 815.1 O requires a visual (VT-2) examination of the bottom of the reactor vessel during the System Leakage Test (IWB-5221 ), which is required following a refueling outage.
111. Basis for Relief Surry Unit 1 is currently operating at 100% rated power. In order to meet the pressure and temperature requirements of the System Leak Test, and reduce the radiation level to allow personnel access, the reactor would have to be brought to the Hot Shutdown condition (reactor subcritical and Tavg greater than or equal to 547°F). Shutting down Surry Unit 1 to perform these examinations is unnecessary to meet the intent of the Code, in that the alternate testing and monitoring as noted below provides the ability to establish and monitor the reactor vessel integrity as intended by the Code examinations. These testing and monitoring activities are performed routinely as part of normal plant operations. In addition, during the last Unit 1 refueling outage when testing was being performed on the loose parts monitoring system under the vessel by engineering personnel, no obvious visual indications of a leak (i.e., build up of boron crystals on the vessel or floor) were observed.
IV. Alternate Testing It is our intention to perform these missed Code requirements at the next outage of sufficient duration, when the Code System Leakage test can be reasonably conducted. If these examinations prove to be impractical to accomplish due to the constraints of our subatmospheric containments, we may need to consider a permanent relief request in the future.
In the interim, other measures will be used during normal operation to provide the Code intended assurance of integrity.
Technical Specifications require that the Reactor Coolant System (RCS) Leak Rate be limited to one (1) gallon per minute unidentified leakage. This value is calculated at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Tech. Spec.
Table 4.1-2A, Item 10). If unidentified leakage increases by greater than or equal to 0.2 gallons per minute, Operations Department personnel will initiate a review of other control room indications and containment air samples and undertake system walkdowns, if appropriate, to identify the source of the increased leakage.
I I
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The containment atmosphere is continually monitored by particulate and gas radiation monitors, which provide both an Alert and High Level Alarm in the Control Room. The containment atmosphere radiation monitors provide an early indication of a small RCS pressure boundary leak. The operability and setpoints for these monitors are verified daily. In addition, containment air samples are taken on a weekly frequency and analyzed for isotopes indicative of RCS leakage. Until the inspections are performed, the frequency of this sampling and analysis will be increased to daily. The sample data will be trended and copies of the sample analysis results and trends will be provided to the Operations Department. Evidence of increasing RCS leakage will be evaluated promptly by the operating shift as part of the leakage monitoring activities discussed above.
These diverse tests, indications and conservative actions will independently identify integrity concerns within the Reactor Coolant System pressure boundary and initiate timely operator action.