ML19290B791
| ML19290B791 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 11/30/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 7912140107 | |
| Download: ML19290B791 (2) | |
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Docket No. 50-155 Mr. David Bixel Nuclear Licensing Adminar Consumers Power Company 212 West Michigan Avenu Jackson, Michigan 4920
Dear Mr. Bixel:
By letters dated June 7, August 29, 1979 and October 31, 1979, you submitted reports on thstigation and repair of a reactor vessel control rod drive penetration a31g Rock Point Plant. These reports supplemented your Licensee Event Rep-18 submitted May 2,1979 which reported leakage detected at the F-2 cored drive penetration.
Your August 29, 1979 letated that the repair of the leak was reviewed ur. der the requirements CFR 50.59 and Consumers Power determined that no unreviewed safety qu is involved. Based on that detennination you submitted the report fcnnation only.
We have reviewed your rand conclude that the repair is acceptable and does not involve an unrd safety question or a change to the Technical Specifications. Our asnt of the acceptability of the repair is enclosed.
Sincerely, Dennis L. Ziemann hief Operating Reactors Branch #2 Division of Operatina Reactors
Enclosure:
Assessment cc w/ enclosure:
See next page 1577 261
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Mr. David bixel ?!ovember 30, 1979 cc w/ enclosure:
Mr. Paul A. Perry, Sec retary Consuucrs Power Cut.yony U. S. Environment..1 Protection 212 West H1chison Avenue Agency Federal Activities Branch Jackson, Michigan 49201 Region V Office ATTN:
EIS C00RDIt;ATOR Jucc L. encun, Esquire Consuuers Pcwer Lct.,;any 230 South Dearborn Street Zld.1est dichison Avenue Chicago, Illinois 60604 Jackson, Michisen W201 Herbert Grossuan, Esq., Chairaan nuntun u,,i l l i uas Atonic Safety and Licensing Board Geurge C. Freeman, Jr., Esquire U. S. tiuclear Regulatory Connission
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4 ASSESSMENT BIG ROCK POINT PLANT REPAIR OF REACTOR VESSEL CONTROL ROD DRIVE PENETRATION Introduction On April 20, 1979 during a routine scheduled primary system leakage test at Big Rock Point Plant, visual examination revealed a small amount of leakage at the location where the control rod drive (CRD) housing CRD F-2 penetrates the bottom head of the reactor vessel. The leakage was immediately reported to NRC's Region III Office and the finding was documented in Licensee Event Report 79-18 submitted by Consumers Power Company (CPCo) on May 2, 1979. By letters dated June 7,1979, August 29, 1979 and October 31, 1979, CPCo submitted reports on the investigation and repair of the penetration.
Discussion The April 20, 1979 test was performed prior to startup after refueling outage in accordance with the inservice inspection requirements defined in Section XI of the ASME Boiler and Pressure Vessel Codes and Section 9 of the Technical Specifications of License DPR-6. Subsequent to the detection of the leak several nondestructive examination techniques were employed to determine the cause of the leakage and its path. The examination results showed that the leakage was through a defect in the stub tube-to-inner reactor vessel wall weld and followed the annular path formed by the CRD housing and vessel penetration. Visual examination of the other penetrations showed no leakage or deposits indicative of prior leakage.
Several repair methods were considered after the defect was identified and examined. Sealing the annulus leak by me:hanically rolling the housing into the vessel wall was selected as the method to be used. This method was demonstrated on mockups. Repair procedures, CRD F-2 examinations, mockup test results, and safety considerations are documented in " Report of Investigation and Repair of Reactor Vessel Control Rod Drive Penetration -
Big Rock Point Plant," submitted by letter dated August 29, 1979.
Evaluation A typical CRD housing assembly is shown in the attached Figure 1.
The housing is held in place by the stub tube-to-housing "J-weld" as well as an inter-ference provided by a two mil shrink fit between the housing and stub tube.
The stub tube is welded to the vessel wall which contains an inconel inlay 1577 2o3
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. at the weld location. The stub tube-to-vessel weld has been determined to be the weld which contains the flaw and leak path in the CRD F-2 housing assembly.
The flaw has been determined by the licensee to have propagated between slag inclusions along the fusion line of the stub tube-to-inconel inlay weld.
Ultrasonic and radiographic examination of the reactor vessel base metal inconel inlay, the CFD F-2 housing, stub tube and attachment J-weld revealed no defects. Maximum leakage from the housing was reported to be 18 ml/ minute (approximately 0.005 gal / minute) at 1500 psig and 2000F.
CPCo proposed to seal the leak by rolling the CRD housing into the reactor vessel wall, a method which has been utilized successfully at two other nuclear facilities with similar situations, Oyster Creek in New Jersey and Garigliano in Italy. The procedure and equipment used in the repair was qualified and tested by rolling several simulated CRD F-2 housings. The rolling was performed in the region of the vessel base metal. Thermal cycling to 5700F and hydrostatic tests to 1500 psig were performed on the mock-ups after rolling.
These tests demonstrated that loosening of the rolled joint did not occur and that the leakage from the housing was reduced.
Subsequent to the mock-up testing the F-2 CRD penetration was repaired. A primary system leakage test was then performed and witnessed by an URC inspector. No leakage was observed during the test.
CPCo has stated that inservice inspection of the CRD's will folicw require-ments of Section XI ASME Code. Leak tests will be performed at each refueling cutage prior to startup. Baseline data for the rolled joint will be obtained following the repair and will consist of a leakage test and ultrasonic examination of the housing and reactor vessel wall in the rolled area. These examinations exceed Code requirement and will be performed during each of the next three refueling outages to ensure continued integrity of the "J-weld" and CRD housing and also to ensure that there is no crack propagation into the reactor vessel wall.
We find that the repair method has been adequately tested to demonstrate that the leakage can be decreased or eliminated and has been successfully utilized before.
Failure of the defective weld would not present a significant safety hazard because the stub tube cannot be forced through the reactor vessel penetration and a mechanical restraint is provided beneath all CRD's to ensure that a CRD or housing cannot be ejected. Based on these f cts as well as the leak tests and examinations which will be perf meQ CPCo to ensure the centinued integrity of the CRD F-2 housing ank "ves 1 ha, we conclude that the repair is acceptable.
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