ML18153A140
| ML18153A140 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 06/02/1997 |
| From: | Saunders R VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 97-286, TAC-M97686, NUDOCS 9706060318 | |
| Download: ML18153A140 (7) | |
Text
CATEGORY 1 REGULATO.INFORMATION DISTRIBUTION.TEM (RIDS)
ACCESSibN N.BR:*9706060318 DOC.DATE: 97/06/02 NOTARIZED: NO FACJI1:50-281 Surry Power Station, Unit 2, Virginia Electric & Powe A(Jhl. NAME AUTHOR AFFILIATION SAUNDER*s,R.F.
Virginia Power (Virginia Electric & Power Co.)
RECIP.NA.;~E RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)
SUBJECT:
Forwards res~onse to request for addl info ;e ASME Section XI Relief Requests SR-018 through SR-024.
DOCKET t 05000281 DISTRIBUTION CODE: A047D COPIES RECEIVED:LTR l_ ENCL l SIZE:_fu_--=-~-
TITLE: OR Submittal: Inservice/Testing/Relief from ASME* Code -
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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 June 2, 1997 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Gentlemiim:
VIRGINIA ELECTRIC AND POWER COMPANY SURRY !POWER STATION UNIT 2 Serial No.97-286 NL&OS/GDM RO Docket Nos. 50-281 License Nos. DPR-37 ASME SECTION XI RELIEF REQUESTS SR-018 THROUGH SR-024 REQUEST FOR ADDITIONAL INFORMATION (TAC No. M97686)
In our letter dated January 2, 1997, Virginia Electric and Power Company requested relief from certain weld examination coverage requirements of the ASME Section XI Code.
The relief was requested for the third 10-year interval inservice inspection program for Surry Unit 2. The basis for relief was provided in Relief Requests SR-018 through SR-024 in our submittal. In your letter dated April 25, 1997, you requested additional information regarding these relief requests.
The NRG questions and our responses are provided in the attachment. As discussed in the attachment, Relief Request, SR-022 is being withdrawn.
No new commitments are intended as result of this letter. If you have any additional questions, please contact us.
Very truly yours, R. F. Sat:1nders Vice President - Nuclear Engineering and Services Attachment o G o o 15 11\\1111 \\\\Ill \\\\l\\\\)1111 \\\\\\\\\\ )11\\\\l 1111\\11 97060t,0318 970602.
PDR ADOCK 05000281 G
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U. S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. R. A. Musser NRC Senior Resident Inspector Surry Power Station Mr. Michael T. Anderson 11\\lEEL Research Center 2*151 North Boulevard P. 0. Box 1625 Idaho Falls, Idaho 83415-2209
ATTACHMENT REQUEST FOR ADDITIONAL INFORMATION ASME SECTION XI RELIEF REQUEST NOS. SR-018 - SR-024 SURRY POWER STATION UNIT 2
_ _J
REQUEST FOR ADDITIONAL INFORMATION ASME SECTION XI RELIEF REQUEST NOS. SR-018 - SR-024 SURRY POWER STATION UNIT 2 NRC Question A Based on the initial review of VEPCO's submittal, it appears that the appropriate paragraph of the regulations has not been referenced for Request for Relief SR-018.
Relief was requested pursuant to 10 CFR 50.55a(g)(5)(iii), but disassembly of a component constitutes a burden, not an impracticality.
Typically, disassembly of components for the sole purpose of performing examinations is considered a hardship.
Please review this request for relief and provide the appropriate reference of the regulations, along with documentation supporting the regulatory basis.
Response
In the cover letter to our January 2, 1997 submittal, we generically requested relief from Code requirements pursuant to 10 CFR 55.55a (g) (iii). However, this section is not the correct n3ference for Relief Request SR-018 submitted in the letter. The correct section that should have been referenced for this particular relief request is 10 CFR 50.55a(a)(3)(ii). Although Relief Request SR-018 does not specifically request relief pursuant to 10 CFR 50.55a (a) (3)(ii), the request does state that the overall level of plant quality and safety will not be compromised (last paragraph), and that disassembly of the pump would cause a burden without a compensating increase in quality and safety (paragraph IV). Also, Surry Unit 1 has previously received a similar approved relief request for this same component.
During the last refueling outage (March 1997), the "A" reactor coolant pump (RCP) main flange bolting was disassembled for maintenance making the seal injection pipe weld accessible for examination, and the weld was examined. No indications were identified, and the weld was accepted. However, future partial disassembly of reactor coolant pumps for the sole purpose of examining the subject seal injection pipe welds would represent a considerable hardship. Shielding would be required to be installed in the motor room to reduce radiation levels. In addition, it would take two individuals at least twHnty-four hours to remove and reinstall enough RCP bolting to make the pipe weld accessible. Three to four bolts would also have to be removed in the location of the piping weld, as well as another three or four bolts 180° across from the bolts in the weld area.
Furthermore, loosening the bolting could cause gasket problems which would require removal of the remaining bolts, replacement of the gasket, and retensioning of the bolting. Past RCP maintenance activities at Surry indicate that it can take up to six days to properly retension bolting on the RCPs.
Reasonable assurance of system structural integrity is provided by the addition of another weld to the B-J welds examinations to makeup the 25% sample, the system leakage test performE3d every refueling outage, primary system leakage monitoring during routine
operation, and the containment atmospheric radioactivity monitoring system. These assurances are included in Relief Request, SR-018.
Consequently, we request relief from performing the weld examination coverage requirements for the subject welds pursuant to 10 CFR 50.55a(a)(3)(ii) for the reasons cited in our January 2, 1997 response as well as the supplemental basis provided above.
NRC Qu1estion B In Request for Relief, SR-019, VEPCO requested to perform a surface examination of the fillet weld of a reinforcement plate/saddle weld in lieu of the weld obstructed by the saddle plate. An additional part of the proposed alternative is the performance of the system leakage test. Considering that direct access to the Code-required weld is not possible, how will an effective leakage test be performed?
Does the saddle plate include telltale holes which could be used to detect leakage from the underlying weld?
Response
It cannot be determined if there is a telltale hole in the reinforcement plate/saddle weld without removing the insulation from the component and visually inspecting for the telltale hole. This in itself would constitute a hardship without a compensating increase in quality and safety, since a containment entry in a subatmospheric containment would be required. The examination of the reinforcing plate weld is considered adequate to provide assurance for public safety based on the following:
1.
The reinforcing plate reduces the stress on the branch connection weld 4-11 BC by transferring a part of the total stress associated with the joint to the fillet welds that attach the reinforcing plate to the pipe section. The examination of the pipe to reinforcing plate fillets by a.surface technique is subjecting at least part of the total joint stress to examination.
- 2.
Other Class 1 branch connections are examined in the reactor coolant system, which provides assurance that there is no susceptible to common mode failure.
- 3.
The configuration of the piping segment is such that thermal fatigue is not expected to be a concern.
Furthermore, the reinforcing plate fillet weld, i.e., both the plate to main piping and plate to branch section have been examined and found to be acceptable.
While the joint configuration does not allow the system leakage test to monitor weld 4-11 BC directly, it does provide important information regarding joint integrity. The two fillet welds joining the pipe section to the reinforcing plate are bonded metallurgically to
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the RCS pressure boundary. It would be inconsistent with the intent of the ASME Code
-not to monitor these welds as part of the system pressure test program.
NRC Question C In Request for Relief, SR-022, VEPCO states that the Code-coverage requirements cannot be met for the surface examination of safety injection (SI) pump casing Weld 2-06 in Safety Injection Pump 2-SI-P-1 A. In light of the limited examination of Weld 2-06, please describe the examinations performed on the other SI pump casing welds that would support the conclusion that generic degradation is not occurring?
Response
This reliE3f request is being withdrawn at this time. In its present form, Relief Request SR-022 overstates the extent of the weld examined to date. The weld has not yet been examined for its full circumference, i.e., one-third of the weld is scheduled for examination in each inspection period.
Only period 1 examinations have been completed to date.
However, the coverage specified in this relief request is not expected to change since the weld obstruction is well identified. We will resubmit this request for relief when the whole circumference of the weld has been examined.
NRC Question D In Requi3st for Relief, SR-023, VEPCO states that the Code-coverage requirements cannot be met for the volumetric examination of two pressurizer nozzles inside radius sections. The proposed alternative is to perform a best-effort volumetric examination and a visual examination of the IR sections prior to the end of the inspection interval.
However, no indication of the coverage that can be obtained was provided. Please provide 1estimates of the Code-required volume that can be examined.
In addition, provide a discussion regarding the type of visual examination that is proposed.
Response
It is estimated that only five percent of the required volume could be examined for the reason stated in the relief request. The visual examination described in paragraph V of the relief request will be a VT-1 examination. The use of remote visual aids as allowed by IWA-2211 (c) of ASME Section Xl-1989 Edition may be required.
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