ML20246J641

Supplemental Part 21 & Deficiency Rept L2-89-44 Re Nupro Test Valves,Installed on Instrument Line Racks Furnished by Ge,Not Remaining Leak Tight During DBA Conditions.Initially Reported on 890608.Root Cause Analysis Will Be Performed

ML20246J641
Person / Time
Site:	Limerick
Issue date:	07/07/1989
From:	Kowalski S PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-PT21-89-124-000 CCN-89-11029, L2-89-44, PT21-89-124, PT21-89-124-000, NUDOCS 8907170414
Download: ML20246J641 (4)

Text

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p PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET l.

P.O. BOX SLwu PHILADELPHIA. PA.19101 izisi... sor JUL 71989 S. J. KOWALSKI vic s-en ssmswr CCN 89-11029 NUCLE AM E80 SON S E RING

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S. Nuclear Regulatory Conmission 10CFR50.55Ce)

Attn: Doct. ment Control Desk Washington, DC 20555 Docket No. 50-353

SUBJECT:

Limerick Generating Statlon, Unit 2 Significant Deficiency Report Non-Q NUPRO Valves

REFERENCE:

(1) _Telecon of June 8, 1989, W. Bowers (PECo) to H. Williams (NRC)

FILES:

QUAL 2-10-2 (SDR L2-89-44)

Gentlemen:

By telephone conference call of June 8, 1989, Philadelphia Electric Carpany (PECo) reported a deficiency regarding Limerick Unit 2 nonquali-fled NUPRO valves. This deficiency was identified on Unit 1.

Our assess-ment and corrective action for this deficiency are discussed in the enclosed final significant deficiency report. The corrective actions will be complete prior to initial criticality of Unit 2.

PECo considers this significant de-ficiency resolved.

If you have any further questions at this time, please contact us.

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Enclosure MAM/dk/063089 cc:

W. T. Russell, USNRC, Administrator, Region I T. J. Kenny, USNRC, LGS Senior Resident inspector R. J. Clark, USNRC, LGS Project Manager 8907170414 890707 ADOCK0500gg3

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A. S. MacA!nsh M. S. Iyer (Bechtel)

J. F. O'Rourke H. D. Honan P.'d. Duca M. A.' Miller (TENERA)

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Fnetaeora LIMERICK GENERATING STATION, UNIT 2 SIGNIFICANT DEFICIENCY REPOP,T NON-Q NUPRO VALVES Description of Deficiency Ninety-nine NUPRO brand test valves Installed in Instrument lines on Instru-nent racks furnished by General Electric downstrean of certain excess flow check valves would not remain leak tight during design basis accident condi-tions. These valves rely on a teflon-coated viton seal between the valve plug and body to provide the pressure boundary. A calculation has shown that the radiation dose that would be received by the test valves eight hours after the design basis accident would cause degradation of the valve seals. The test valves are Installed on instruments that monitor pressure, flow, and level following a design basis accident. Of the 99 valves, 15 were identified ar, being required for longer than eight hours.

(The original telephone report inadvertently identified 19 valves.) Because of the assuTed radiation-induced seal degradation, these 15 valves would potentially not be available to perf'onn their post-accident safety functions. The purpose of this instrumentation is to assure acceptable energency response capabilities during and following the course of an accident. These test valves allow simulation of a line break in the instrunent line in order to functionally check the excess flow check valve to ensure it closes when flow is sensed in the line.

Safety implications Two safety implications arise as a result of the failure of the test valves to renaln leak tight. The first is a loss of pressure boundary, and the second Is a partial or total loss of the instrument function.

The loss of pressure boundary could result in a radiological release and consequent contamination of certain areas of the Reactor Enclosure. However, personnel 1

access to the Reactor Enclosure would be restricted after the DBA LOCA due to expected radiation levels resulting from the accident regardless of the condition of the test valve seals. The contaminated fluid that leaked from the valves would be processed by the floor dra!n system and the reactor enclosure recirculation system, and standby gas treatnent systen would l

process the airborne contamination.

These instruments are relied upon during post-design basis accident condi-tions to monitor pressure, flow, and level Indication. This instrumentation requires a leakage free Instrunent ilne to sense its var!able accurately.

The 15 test valves were Installed on instrumentation for pressure vessel level and pressure Indications, neln steam line Isolation valve (MSIV) leakage control system pressure Indications, and reactor recirculation punp flow i

A Indications. These 15 test valves are located on instrunent lines that are connected to Instrumentation regul: 9d by Regulate y Guide 1.97.

With post-accident monitoring capabilities lacking, adequate accident responses nay not be accorp11shed because the operators' ability to verify adequate reactor coolant inventory and pressure, adequate MSIV leakage control system operation, and recirculation pump flow may be affected.

Corrective Action Although only 15 of the 99 valves would be required to perfonn their safety function for a period greater than eight hours when the viton seal is assuned to fall, PECo has elected to replace all 99 valves with valves that do not contain viton seals. This replacement will be accomplished on Unit 2 before Initial criticality. Since the potential failure of these valves occurs only due to radiation, there is no safety impact during the period between fuel load and initial criticality.

Actions Yaken to Prevent Recurrence i

The Installation of these NUPRO valves was originally conpleted as a Unit 1 design change prior to Unit 1 low power IIcensing and subsequently applied to Unit 2 without another detailed review. The cause of this condition was inadequate review of systen design specifications against the reterials in-tended for installation. A detailed root cause analysis will be performed of the design change process by August 31, 1989 to detennine whether further investigation inte a generic concern regarding design changes prior to licensing is necessary.

If this root cause Investigation reveals the need to further Investigate generic concerns, then the results of the complete investigation will be issued in a supplement to Limerick Unit 1 LER 89-034.

The cause of the problem has been corrected for the modification process for operating units by implementation of Administrative Procedure A-14 Just before Unit I licensing. This procedure provides instruction and control throughout the modification process, addret aes the modification review process, and involves the Independent review by several specialized work groups, super-vision and renagement. We have determined that the modification process now in place is adequate and provides the proper instruction to attain the appro-pri ate Independent reviews.

MAM/dk/0630891