ML20062G366
| ML20062G366 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/19/1978 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Stewart W FLORIDA POWER CORP. |
| References | |
| NUDOCS 7812280025 | |
| Download: ML20062G366 (1) | |
Text
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sb UZlTED STATES o,
NUCLEAR REGULATORY COMMISSION
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r 101 MARIETTA STREET. N.W.
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ATLANTA, GEORGIA 30303 o,
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1978 iQ/
In Reply Refer To:
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RII:JPO 50-302 r
Florida Power Corporation Attn:
Mr. W. P. Stewart, Director Power Production P. O. Box 14042, Mail Stop C-4 f
St. Petersburg, Florida 33733 Gentlemen:
The enclosed Bulletin 78-14 is forwarded to you for information.
No written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely,
,p I
p ames P.
Reilly Directo
Enclosures:
1.
List of IE Bulletins Issued in 1978 cc w/ encl:
G. P. Beatty, Jr.
Nuclear Plant Manager P. O. Box 1228 Crystal River, Florida 32629 781228 cots-(
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UNITED STATES NUCLEAR FIGULATORY C0!NISSION OFFICE OF INSPECTION AND ENFORCD:ENT WASHINGTON, D.C.
20555 Dece=ber 19, 1978 IE Bulletin No. 78-1L DETERIORATION OF BUNA-N COMPONEN'S IN ASCO SOLENOIDS Description of Circumstances:
Following a reactor scra= at the Monticello Nuclear Power Station on July 27, 1978, one control rod was deter =ined to have a slow scra=
insertion ti=e.
While =casuring the average control rod scra insertion ti=es (as required by Technical Specifications), the sa=e control rod failed to scra=.
Subsequent investigation revealed that a Buna-N dise
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in the scra= pilot valve solenoid core asse=bly had fractured. Stall pieces of the Buna-N =aterial vedged between the core asse=bly and the valve pilot body. This prevented proper =ove=ent of the core asse=bly (plune,er), and thus prevented the valve fro = properly venting centrol air iro: the scra= valves. A similar event occurred during scra:
testing on June 5, 1978.
On both occasions the control rod failed te scram during control rod surveillants testing.
In these instances, had a scra= occurred during nor=al operation of the reactor, the affected rods would have inserted by action of the backup scra= solenoid valves.
The reactor was placed in eold shutdown on July 28, 1978.
All of the scra: pilot solenoid valves and scra= backup valves were disasse= bled and inspected.
Six additional fractured Buna-N discs were found, all in scra: pilot solenoid valves.
Northern States Power Co pany replaced the plunger Buna-N discs in all 2k2 scra= pilot valves and the 2 backup scra= valves.
In addition to the Monticello failures, si=ilar events were recently reported fro: the Big Rock Point and Pilgri facilities.
At Big Rock Point, two failures were experienced in October while at Pilgri= one failure was found in July and two in Dece=ber. All of these failures were encountered during surveillance testing.
General Electric had previously issued Service Infor=ation Letters,
SIL 123 (March 31, 1975) and CIL 128 Revision 1 (January 30, 1976),
addressing the deterioration through natural aging of Euna-N co=ponents in CRD seras pilot valves.
These letters reco== ended that a =aintenance program be adopted which replaced certain Buna-N co=ponents in scra:
pilot valves on a regularly scheduled basis.
However, the Buna-N diser 1 of 3 y'sh
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IE Bulletin No. 78-1L Dece=ber 19, 1978 in the sera = pilot valve core asse=blies were not included in these reco==endations. As a result of the Monticello experience and subsequent investigation, General Electric has issued in August 1978, SIL 12S Revision 1, Supplement 1, which reco== ends that all BWR operators take the following actions:
" Establish a preventive maintenance progra= to replace all core asse=blies, diaphrag=s and associated parts in all CRD scra:
pilot valves, backup sera: valves, sera = discharge volu=e vent and drain pilot valves, and scra= discharge volu=e test valves, at periodic intervals. The Buna-N parts used in these valves have at least a seven year shelf and in-service life fro = the packaging date on the rebuild kit.
Thus, the valves should be rebuilt
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periodically to assure that the Buna-N parts are not used in excess of seven years.
It is rece== ended that the work be scheduled in any one outage such that the associated CRDs are selected frc= a distributed checkerboard pattern.
The re=aining valves can be rebuilt during subsequent outages."
For clarity, it is noted that the above seven year li=it reco== ended by GE includes all of the ti=e fro = packaging throu6h installatien to the end of service life.
In addition, it is noted that the March 31, 1975 issue of SIL 128 stated that the design specification for Buna-N specified a =ini=un design life of three years.
Action to be Taken by Licensees:
For all GE BWR power reactor facilities with an operating license:
1.
Review Buna-N =aterial applicatiens in your contrc; rod scran syste=s and deter =ine the ti=e since installatien, and fer installed
=aterial, the time since packaginr.
2.
Report the results of the review set forth in ite I abeve and describe your schedule for replace =ent, both in respense to this Bulletin and for periodie maintenance.
3 Describe the bases for your schedule of replace =ent identified in response to ite: 2 above. Justify any proposed replacement time in excess er three years.
2 of 3 r
IE Bulletin No. 73-lh December 19, 1978 h.
Within h5 calendar days of the date of issue of this Bulletin, report in writing to the Director of the appropriate NRC Regional Office, the results of your review, schedule and bases for replacement with regard to Items 1 through 3.
A copy of your report should be sent to the United States Nuclear Regulatory Commission, Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C.
20555 For all B'4R facilities with a construction permit and all other power reactor facilities with an operating license or construction permit, this Bulletin is for infor:ation only and no written response is required.
Approved by GAO B180225 (R0072); clearance expires 7/31/20. Approval
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vas given under a blanket clearance specifically for identified generie problens.
Enclosure:
List of IE Bulletins Issued in 1978 3 of 3 l
1
O IE Bulletin No. 78-Date:
LISTING OF IE BULLETINS ISSUED IN 1978 Bulletin Subject Date Issued Issued To No.
7 8-01 Fla==able Contact -
1/16/78 All Power Reactor Ar= Retainers in G.E.
Facilities with an CR120A Relays OL er CP 78-02 Ter=inal Block 1/30/78 All Power Reacter Qualification Facilities with an OL or CP r
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78-03 Potential Explosive 2/8/76 All B'G Fever Gas Mixture Accumula-Reactor Facilities tions Associated with with an OL cr CF BWE Offgas Syste:
Operations 78-OL Environ = ental Quali-2/21/78 All Power Reacter fication of Certain Facilities with an Ste: Mounted Limit OL or CP Svitebes Inside Reactor Contaircent 78-05 Malfunctiening of L/1L/78 All Power Reseter Circuit Breaker Facilities with an Auxiliary Contact OL er CP Mechanis:-General s
Model CR105X 78-06 Defective Cutler-5/31/78 All Pever Reacter Ha=ner, Type M Relays Facilities with an With DC Coils OL er CF 78-07 Protection afforded 6/12/78 All Power Reae cr by Air-Line Respirators Facilities with an and Supplied-Air Hood OL, all class E an:1 F Research Reacters with an OL, all Fuel Cycle Facilities with a. OL, and all Prierity '.
Material Licensees Page 1 er 2 1
o IE Bulletin No. 78-Date:
LISTING OF IE BULLETINS ISSUED IN 1978 Bulletin Subject Date Issued Issued To No.
78-08 Radiation Levels from 6/12/78 All Power and Fuel Element Transfer Research Reactor Tubes Facilities with a Fuel Element transfer tube and an OL.
78-09 BWR Drywell Leakage 6/14/78 All BWR Power Paths Associated with Reactor Facilities Inadequate Drywell with an OL or CP Closures 78-10 Bergen-Paterson 6/27/78 All BWR Power Hydraulic Shock Reactor Facilities Suppressor Accumulator with an OL or CP Spring Coils 78-11 Examination of Mark I 7/21/78 BWR Power Reactor Containment Torus Facilities for Welds action:
Peach Bottom 2 and 3, Quad Cities I and 2, Hatch 1, Monti-cello and Vermont Yankee 78-12 Atypical Weld Material 9/29/78 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 78-13 Failures in Source Heads 10/27/78 All General and of Kay-Ray, Inc., Gauges Specific Licensees Models 7050, 7050B, 7051 with Kay-Ray Gauges 7051B, 7060, 7060B, 7061 and 7061B Page 2 of 2