ML14170A336
| ML14170A336 | |
| Person / Time | |
|---|---|
| Site: | Robinson, Brunswick  |
| Issue date: | 05/22/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Jackie Jones CAROLINA POWER & LIGHT CO. |
| References | |
| NUDOCS 7907110084 | |
| Download: ML14170A336 (8) | |
Text
RE9
.UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 MAY 2 2 1979 In Reply Refer To:
RII:JPO 50-325, 50-324 Carolina Power and Light Company ATTN: Mr. J. A. Jones Executive Vice President and Chief Operating Officer 411 Fayetteville Street Raleigh, North Carolina 27602 Gentlemen:
Enclosed is IE Bulletin 79-11 which requires action by you with regard to your power reactor facility(ies) with an operating license or a construction permit.
Should you have questions regarding this Bulletin or the actions required of you, please contact this office.
Sincerely, James P. O'Reilly Director
Enclosures:
- 1.
- 2.
List of IE Bulletins Issued in Last Twelve Months qgg1190' Q
MAY 2 2 1979 Carolina Power and
-2 Light Company cc w/encl:
A. C. Tollison, Jr.
Plant Manager Sox 458 Southport, North Carolina 28461 R. Parsons, Site Manager Post Office Box 101 New Hill, North Carolina 27562 R. B. Starkey, Plant Manager Post Office Box 790 Hartsville, South Carolina 29550
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 May 22, 1979 IE Bulletin No. 79-11 FAULTY OVERCURRENT TRIP DEVICE IN CIRCUIT BREAKERS FOR ENGINEERED SAFETY SYSTEMS Discussion:
We have received information from Westinghouse and an NRC licensee relating to the potential failure of a circuit breaker in an engineered safety system of a nuclear power plant. This circuit breaker had a defect in one of its three time delay dashpots which resulted in a reduced time delay for over current protection.
The defect was a small hairline crack in the end cap of the dashpot. Further investigation by this licensee disclosed that 7 out of 17 spare dashpot end caps and 2 non-engineered safety feature breakers also had similar defects. The circuit breaker is a Westinghouse type DB-75.
Westinghouse type DB-50 breakers also use the same type of dashpot and end cap. DB-50 and -75 breakers are used extensively in PWR's, and some BWR's may also have the same breakers.
Similar make and model circuit breakers, when used for scram purposes, do not require the overcurrent trip feature and thus are not of concern. The end cap crack defect, if severe enough, could result in premature tripping of the circuit breaker because of insufficient time delay in overcurrent protection, i.e., the motor starting (inrush) current could cause the breaker to trip inadvertently and thus prevent the motor start.
The defects reported by the licensee in April 1979, occurred in the replacement end caps which were provided to solve the problem described in IE Bulletin 73-1. The subject of Bulletin 73-1 was end caps made of a black phenolic material. As a result of that Bulletin, the black end caps were replaced with a new type made of fibre-filled polyester material called "navy-gray".
Prior to the April 1979 report, there have been no reports of suspect "navy-gray" end caps either from scheduled testing or unusual behavior in service. The manufacturer of the "navy-gray" and caps believes the crack defects may be linked to a raw material batch problem.
That is, the molding ingredient materials used may have neared the end of their shelf life before use.
It is not believed the end caps, after fabrication, have a significant shelf life limit, due to the low residual stress and low crack propagation probabilities.
7906130037
IE Bulletin No. 79-11 May 18, 1979 Page 2 of 3 Description of Event:
The following infonation was obtained from the Licensee Event Report dated April 12, 1979 and a subsequent meeting with Westinghouse, the NRC staff and the licensee.
During the 1979 surveillance tests, and a review of the previous refueling surveillance test results on a Westinghouse type DB-75 breaker (used as a 480v ESF bus supply breaker) a drift in the overcurrent trip tne from the manu facturer's design minimm value of 6 seconds to 5.50 and 5.12 seconds was observed.
The 1979 test results showed a deviation and inconsistency in the delayed trip timings among three consecutive tests.
The overcurrent devices were remved fra each of the three phases and a visual inspection indicated a hairline crack in the end cap of one of the devices.
That cap was replaced (without checking for a possible crack in the replacement cap), and the breaker was again tested.
This test also showed deviations and inconsistency in the delayed trip timings.
The subsequent inspection of the devices revealed a hairline crack in the end cap which had just been installed.
This prampted an inspection of the in-stock spare caps.
Seven out of 17 caps were found to have similar cracks.
Action Required of all Holders of an Operating License or Construction Permit:
- 1.
Determine whether circuit breakers of the above described manufacturer and type with overcurrent trip devices are in safety related Class IE service or in spares at your facilities.
- 2.
If the subject breakers are in service in safety-related systems:
within 30 days, review the existing test data for all overcurrent trip device calibrations since plant startup or since replacement caps were installed and tested in response to Bulletin 73-1, whichever is mst recent.
Deter mir.e if any delay times are:
(1) outside of the acceptance band; (2) marginally acceptable - on the low side of the acceptance band, or (3) if any significant change in delay time performance has been observed. These breakers should be retested and end caps replaced as necessary to assure no loss of safety function.
- 3. Inspect all end caps in spares for cracks using at least a 3x magnifying glass. Caps having visible flaws should be discarded, or prevented from use in Class IE applications.
- 4. Review test procedures and test schedules for all safety-related circuit breakers to assure that all such breakers are tested at least each refueling outage to onfirm overcurrent time delay protection.
IE Bulletin No. 79-11 May 18, 1979 Page 3 of 3 For facilities with an operating license, a written report of the above actions, including the date(s) when they will be corpleted shall be submitted within 45 days of receipt of this Bulletin.
For facilities with a construction permit, a written report of the above actions, including the date(s) when they will be campleted shall be submitted within 60 days of receipt of this Bulletin.
Approved by GAO, B180225 (R0072), clearance expires 7/31/80.
Approval was given nder a blanket clearance specifically for identified generic problem.
40 IE Bulletin No. 79-11 Enclosure May 18, 1979 Page 1 of 3 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an OL 79-09 Failures of GE Type AK-2 4/17/79 All Power Reactor Circuit Breaker in Safety Facilities with an Related Systems OL or CP 79-08 Events Relevant to BWR 4/14/79 All SWR Power Reactor Reactors Identified During Facilities with an OL Three Mile Island Incident 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 79-06B Review of Operational 4/14/79 All Cambustion Engineer Errors and System Mis-ing Designed Pressurize(
alignments Identified Water Power Reactor During the Three Mile Facilities with an Island Incident Operating Licensee 79-06A Review of Operational 4/14/79 All Pressurized Water Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Incident 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Power Reactors with an aliganents Identified OL except E&W facilitie During the Three Mile Island Incident 79-05A.
Nuclear Incident at 4/5/79 All B&W Power Reactor Three Mile Island Facilities with an OL 79-05 Nuclear Incident at 4/2/79 All Power Reactor Three Mile Island Facilities with an OL and CF
IE Bulletin No. 79-11 Enclosure May 18, 1979 Page 2 of 3 LISTING OF IE BJLLETTNS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
79-04 Incorrect Weights for 3/30/79 All Power Reactor Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 79-03 Longitudinal Welds Defects 3/12/79 All Power Reactor In ASME SA-312 Type 304 Facilities with an Stainless Steel Pipe Spools OL or CP Manufactured by Yourgstown Welding and Engineering Co.
79-02 Pipe Support Base Plate 3/2/79 All Power Reactor Designs Using Concrete Facilities with an Expansion Anchor Bolts OL or CP 79-01 Environmental Qualification 2/8/79 All Power Reactor of Class IE Equipment Facilities with an OL or CP 78-14 Deterioration of Buna-N 12/19/78 All GE BWR facilities Component In ASCO with an OL or CP Solenoids 78-13 Failures in Source Heads 10/27/78 All general and of Kay-Ray, Inc., Gauges specific licensees Models 7050, 7050B, 7051, with the subject 7051B, 7060, 7060B, 7061 Kay-Ray, Inc.
and 7061B gauges78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12A Atypical Weld Material 11/24/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 78-12 Atypical Weld Material 9/29/78 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP
LE Bulletin No. 79-11 Enclosure May 18, 1979 Page 3 of 3 LISTING OF IE BJLLETNS ISSUED IN lAST THELVE MONTHS Bulletin Subject Date Issued Issued Tc No.
78-11 Examination of Mark I 7/21/78 BWR Pmer Reactor Containment Torus Welds Facilities for action:
Peach Bottom 2 and 3, Quad Cities 1 and 2, Hatch 1, Monticello an, Vermont Yankee 78-10 Bergen-Paterson Hydraulic 6/27/78 All W. Power Reactor Shock Suppressor Accumulator Facilities with an Spring Coils OL or CP 78-09 EWR Drywell Leakage Paths 6/14/79 All WR Power Reactor Associated with Inadequate Facilities with an Drywell Closures OL or CP 78-08 Radiation Levels from Fuel 6/12/78 All Power and Research Element Transfer Tubes Reactor Facilities wit a Fuel Elenent transfe tube and an OL 78-07 Protection afforded by 6/12/78 All Power Reactor Air-Line Respirators and Facilities with an DL, Supplied-Air Hoods all class E and F Research Reactors witi an OL, all Fuel Cycle Facilities with an OL, and all Priority 1 Material Licensees 78-06 Defective Cutler-Hamner 5/31/78 All Power Reactor Type M Relays with DC Coils Facilities with an DL or CF