ML20101R427
| ML20101R427 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/08/1996 |
| From: | Jain S DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9604160274 | |
| Download: ML20101R427 (11) | |
Text
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er Valley Power Station Shippingport. PA 15077-0004 NNvicNesident raxf4tNsIN P er Divisbn April 8, 1996 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001
Subject:
Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 ISI (Inservice Inspection) Program: Relief Request The purpose of this submittal is to request NRC review and approval of a proposed relief request applicable to the Unit No. 2 First Ten-Year Interval ISI Program. The current 10-year interval ends on November 17, 1997. The ISI Program is based on the 1983 Edition through the Summer 1983 Addenda of ASME Boiler and Pressure Vessel Code,Section XI.
Relief Request BV2-CI.30, Rev. O is attached for review.
The examinations represented by this relief request were originally evaluated by the NRC following completion of preservice inspections. The NRC approved relief request BV2-Cl.10-2 in an August 22, 1991, safety evaluation of the first ten-year interval ISI Program Plan (TAC No. 62873).
Since then the component has become a significant source of radiation exposure during nonnal operation.
It is proposed to substitute visual examinations in accordance with the attached relief request in order to reduce occupational radiation exposure.
A proposed alternative to the requirements, with supporting basis is included in the attached request for relief. 10 CFR 50.55a(a)(3)(ii) provides for NRC approval of proposed alternatives to the ASME Code requirements when it can be demonstrated that compliance would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
It is proposed that the attached relief request satisfies this acceptance criteria.
It is requested that this review be completed before July 8,1996, in order to allow sufficient time to plari and prepare for any follow-up examinations that may be required during the sixth refueling outage tentatively scheduled to begin August 30,1996.
9604160274 960400 PDR ADOCK 05000412 G
pon DELIVERING ouatm
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Beaver Valley Power Station, Unit No. 2 ISI(Inservice Inspection) Program: Relief Request Page 2 If you have any questions regarding this issue, please contact Mr. Roy K. Brosi at r.
(412) 393-5210.
Sincerely, 1
Sushil C. Jain i
P Attachment 4
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Mr. L. W. Rossbach, Sr. Resident Inspector Mr. T. T. Martin, NRC Region I Administrator Mr. D. S. Brinkman, Sr. Project Manager t
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DUQUESNE LIGHT COMPANY Beaver Valley Power Station Unit No. 2 BELIEF REOUEST NO. BV2-CI.30. Rev. O COMPONENT Regenerative Heat Exchanger (2CHS*E23) - Tubesheet to Shell Welds 2, 3, 6, 7,10, and 11 DRAWING NO.
ISI-E-2C SECIIDKXt REOUIREMENT(83S83)
Item No. C1.30 (IWC-2500-1, Category C-A) requires volumetric examination.
BASIS OF RELIEF The general radiation levels in the area of this heat exchanger are between 500 mR/hr and 3000 mR/hr and up to 5000 mR/hr on contact (based on surveys taken in the second and fifth refueling outages). The manrem estimate to prepare, inspect, and reinsulate these six welds is 9800 mrem. This estimate is based on working space dose rates and estimated work duration as noted on the attached manrem estimate. ALARA and leak before break considerations suggest a visual examination for leakage would provide an acceptable inservice inspection for these welds.
The design and function of this component give rise to areas where " hot spots" can readily occur due to corrosion and wear products buildup. Flushing provides little benefit toward reducing radiation fields in this area. The location of the welds, between two branch connections with a welded support between the two welds, make shielding of these areas impractical. Reference attached drawing and preservice photographs.
The proposed visual examination for leakage, in this situation, is a more practical examination for monitoring the integrity of the component than the required examination. The required UT examinations are limited by component geometric restrictions. Approximately 1/3 of the required volume of each weld could not be examined in the preservice inspection due to the adjacent branch connections and the welded pad, plates and lugs of the support located between the two wel6s on each of the ? ee sections of the heat exchanger. (Note: Relief Request BV2-CI.10-2 identified these hmitations and was approved in the SER dated August l
22, 1991.) Preservice ultrasonic examinations performed on welds 2, 3, 6, 7, and 10 had no reportable indications. Examination of weld 11 found a manufacturing defect that was subsequently repaired and reexamined satisfactorily.
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Page 2 of 2 Relief Reguest No. BV2-C1.30. Rev. O Several methods are available to detect leakage from these welds if sufficient weld degradation occurred to cause a through-wallleak. Listed below are some examples:
The control room operators perform Operation Surveillance Test (OST) 2.6.2A a.
" Reactor Coolant System Water Inventory Balance" every three days when the plant is operating at steady conditions. Leakage through the subject welds would be discovered by the conduct of this OST.
b.
Containment airborne radiation monitors continuously sample the containment atmosphere and alarm in the control room. This method is sensitive enough to detect a 1 gpm leak in less than I hour, Two leakage monitoring systems are available in the containment sump. One c.
system uses the flow indication of the containment sump pumps to determine leak rates. A programmable controller monitors the flow and pump operating times so that a 1 gpm leak could be detected by this method in I hour. A second system uses changes in the water level of the sump to determine a leak rate. The containment sump level is continuously monitored by instruments that alarm in the control room. This system is capable of detectin3 a 5 gpm leak in I hour.
The Regenerative Heat Exchanger is readily isolable should a leak occur. Double valve isolation from the reactor coolant system is provided by valves 2CHS*LCV460A and 460B.
ALTERNATIVE EXAMINATION l
A visual examination of these welds for leakage is perfonned in conjunction with the boric acid walkdown, every shutdown while the RCS and associated piping remain at operating pressure and temperature. Also, the regenerative heat exchanger is included in the Mode 3 walkdown e
of the RCS boundary, performed during each startup following refueling outages. Both these activities are performed by qualified VT-2 examiners. These examinations are augmented by l
the leakage detection methods noted above.
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2CHS-E23 EXAMINATIONS WELDS 2, 3,6,7,10 and 11 MANREM ESTIMATE 2CHS-E23 Regenerative Heat Exchanger General area radiation level: 500 mRlhr to 3000 mRIhr Contact radiation level: 1000 to 5000 mR/hr TIME IN EXPOSURE ESTIMATED WORK
- OF WORKERS RAD FIELD RATE EXPOSURE TASK AND JOB CLASS (hrs)
(mR/hr)
(mrem) l CONSTRUCT l
SCAFFOLDING (2)
CARPENTERS X
1 x
600 1200
=
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INSULATION (2)
INSULATORS x
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WELD 1
PREPARATION (2)
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REINSTALL INSTALLATION (2)
INSULATORS x
1 x
600 1200
=
REMOVE SCAFFOLDING (2)
CARPENTERS x
.5 x
600 600
=
TOTAL FOR SIX WELDS:
9800 mrem
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