ML20235F860

Interim Significant Deficiency Repts SDR-L2-89-03,04 & 05 Re Unavailability of HPCI & RCIC Sys Due to App R Fire. Initially Reported on 890117.Corrective Actions Under Review & Will Be Reported by 890531

ML20235F860
Person / Time
Site:	Limerick
Issue date:	02/17/1989
From:	Pyrih L PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
SDR-L2-89-03, SDR-L2-89-04, SDR-L2-89-05, SDR-L2-89-3, SDR-L2-89-4, SDR-L2-89-5, NUDOCS 8902230032
Download: ML20235F860 (5)

Text

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, d.g PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET l~

P.O. BOX 8699 PHILADELPHIA A. PA.19101 (215) e4140oO L. B. PYRIH NUCLEAR ENG NEER NG DEPARTMENT

-10CFR50.55(e)

February 17, 1989 Mr. W. T. Russell,. Administrator Docket No. 50-353 U.S. Nuclear Regulatory Conmission CPPR-107' Attn: Doctrnent Control Clerk

. Mall Statlon F1-137 Wast.!ngton, DC 20555

Subject:

Limerick Generating Station - Unit 2 Interim Significant Deficiency Report, Unavailability of the HPCI and RCIC Systems Due to an Appendix R Fire.

Reference:

-1) Telecon of PECo to NRC dated January 17, 1989

2) Letter from S. d. Kowalski, PECo, to W. T. Russell, NRC entitled "Significant.

Deficiency Report No. 249-2 Interim Report for the Nonavallability of Safe Shutdown Capabilities from Outside the Control Room in the Event of a Fire" dated November 4, 1988 File:

QUAL 2-10-2 (SDR-L2-89-03,04,05).

Dear.Mr. Russell:

As cmmitted to in reference 1 above, we are submitting the attached interim Significant Deficiency Report which provides a description of the subject concern and a status report of our ccmpleted and ongoing activities.

As part of a self-assessment, PECo is performing a root cause evaluation of all suspected deficiencies with the LGS safe shutdown analysis as cmmitted to in the LGS Fire Protection Evaluation Report (FPER). The results of this assessment should reveal if these nory:on-

' formances are programmatic errors or isolated deficiencies. Based or this assesurent corrective actions to prevent recurrence will be identified. This study should be complete by April 14, 1989, where upon PECo shall provide a response discussing the results of this assessment and the proposed corrective actions in a supplement to SDR

'No. 249-2 (reference 2) by May 31, 1989.

Ybk f00Ck32 890217 Y

05000353 i \\

S PDC I,

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Mr. W. T. Russoll, Administrator February 17, 1989 j

Page Two 3

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

1 Sincerely,

-GBH/sw/02018901 Attachment Copy to:

W. T. Russell, USNRC, Region I Administrator R. A. Granm, USNRC, LGS 2 Senior Resident Inspector R. J. Clark, USNRC, LGS 2 Project Manager l

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w NUCLEAR ENGINEERING l

ENGINEERING DIVISION l

N2-1, 2301 Market Street SIGNIFICANT DEFICIENCY REPORT SDR NO. L2-89-03, 04, 05 UNAVAILABILITY OF THE HPCI AND RCIC SYSTEMS DUE TO AN APPENDIX R FIRE.

NRC CONSTRUCTION PERMIT NO. CPPR-107 DESCRIPTION OF DEFICIENCY As part of an engineering analysis to determine the effects that safe shutdown process parameters and synergistic effects have.upon the availability of safe shutdown systems,'PECo identified three conditions at Limerick Generating Station that could adversely affect the plant's ability to safely shut down in the event of a fire.

i The first condition would cause Reactor Core Isolation Cooling (RCIC) system unavailability in the event of a fire. Control and power cables associated with the RCIC turbine steam supply line inboard containment' Isolation valve could be affected by a postulated fire.

This valve is powered by Division III AC power and will automatically

- close upon receipt of a RCIC Division III isolation signal. However, the Division III control and power cables were not considered necessary to support Shutdown Methods A or R, therefore they were not protected from fire damage in those fire areas for which Methods A or R are relled upon for safe shutdown.

In the event of a fire in these areas, a false ~lsolation signal could be generated, causing the valve to close. The same fire could also damage the Division III power cables such that AC power may not be available to reopen the valve.

If the valve closes and Division III AC power is lost before the valve can be reopened, the RCIC system would be rendered inoperable.

For the second condition, control cables associated with the High Pressure Coolant Injection-(HPCI) system could also be affected by a fire. The HPCI system is reautred to support Shutdown Method B.

The control cables for the HPCI system were not protected from fire damage in those fire areas for which shutdown methods other than Shutdown Method B are relled upon for safe shutdown.

Inadvertent initiation of the HPCI system could occur as a result of fire damage to these cables causing both a false HPCI initiation signal and also the inability to remotely or automatically trip the HPCI system. Addit f or'elly, by design, alternate HPCI system controls are not available at the Remote Shutdown Panel (RSP) to shutdown the system. Because the maximum HPCI

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system flow rate exceeds the ficw rate that is required for reactor l

water makeup during the shutdown transient, the inability to trip the I

J HPCI system would lead (within approximately ten rotnutes) to an exces-sively high water level in the reactor vessel such that water could be carried over into the main steam lines.

Since the steam supply line for the RCIC turbine is connected to the main steam line, water over-flowing into the main steam line would flow into the RCIC turbine steam

Significant Dsficiency Report

,SDR No. L2-89-03, 04, 05.

Page 2 of 3' supply line, rendering the RCIC system inoperable.

This condition could occur during the tine when the RCIC system is being relied upon to accomplish safe shutdown using Shutdown Methods A or R.

Both conditions described above could result in the unavailability of the RCIC system when it is required to support Shutdown Methods A or R, which is contrary to the safe shutdown requirements described in the LGS FPER. These conditions also affect LGS-1 and have been reported to' the NRC per LGS-1, LER 89-002.

I The third condition again involves fire danege to control cables associated with the HPCI systen. The fire could cause the containment Isolation valves to close before manual actuation of the transfer / iso-lation switch. The fire could subsequently cause a loss of_ control and native power to the Division IV valve. This would prevent the capability to renote manually reopen the valve to restore HPCI opera-bility. This condition is not applicable to LGS-1.

SAFETY IMPLICATIONS The postulated fires could affect the operation of several safe shutdown nothods. The first' condition could adversely impact the RCIC system. Fire danage could cause the RCIC turbine steam supply line inboard containment Isolation valve to close. The sane fire could also danage Division III power cables such that AC power ney not be avall-able to reopen the valve.

If the valve closes and Division III AC power is lost before the valve can be reopened, the RCIC systen would be rendered inoperable.

The second coadition could also render the RCIC systen inoperable.

Since the HPCI control cables are not protected in Fire areas where HPCI is not requirei, the HPCI system could be spuriously actuated by fire damage to cablet and the capability to isolate the systen nay not exist. This would lead to an excessively high water level in the reactor vessel such that water could overflow into the main steam lines, damaging the RCIC turbine and rendering the RCIC system inoperable.

The third condition could affect the HPCI system. Fire danage to cables could cause a spurious closure of the inboard containment isolation valve and subsequent dansge could cause the loss of control and motive power to the valve. This would prevent the capability to renote nanually reopen the valve to restore HPCI operability. There-fore, HPCI would be unavailable to support safe shutdown of the plant.

f t Is concluded that these deficiencies, ransining undetected, could have affected the safe operation of the plant.

The proxiante cause of these conditions is personnel er ror cannit-ted during the original SSD Analysis. The condition affecting the RCIC inboard isolation valve was caused by a failure to properly identify and protect essential cables for safe shutdown camponents and systens.

The irboard containment isolation valve in the RCIC turbine steam i

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______________________________________________________._____________________________________.______________J

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,.,p Sl*gnifIcant DefIclency Rtport

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.SDR No. L2-89-03,.04, 05 Page 3 of 3 supply IIne is powered by Division III AC power and will close automat-Ically upon receipt of a Division III RCIC isolation signal. The need to protect Division III control and power cables from damage due to a fire was not considered since both Methods A and R were originally considered to rely on Division'I AC power only. Also, f.or Shutdown l

Method R, credit was improperly taken for manual actuation of the I

isolation / transfer switches at the Remote Shutdown Panel prior to the occurrence of fire-induced circuit faults that could cause both inboard isolation valve closure and loss of AC power to the valve.

The condition involving spurious actuation of the HPCI system was caused by a failure to recognize the Interaction between systems required to support different methods of safe shutdown in the event of a fire. The original safe shutdown analysis did not consider the effects of an inadvertent HPCI initiation on other safe shutdown systems when HPCI is not required, nor did it consider the need for the capability to tenninate spurious operation of the HPCI system at a time when shutdown methods other than Shutdown Method B are required.

The condition affecting the HPCI containment Isolation valves was also caused by a failure to properly Identify and protect essential cables for' safe shutdown components and systems. Credit was improperly taken in the safe shutdown analysis for manual actuation of the Isola-tion / transfer switch prior to the occurrence of fire-induced circuit faults that could cause both inboard Isolation valve closure and loss of AC power to the valve. Therefore, the cables that could be affected were not identified.

CORRECTIVE ACTIONS In accordance with the PECo Deportability Evaluation Process, a deportability evaluation was initiated. On January 17, 1989, the condition was reported to the NRC as a 10CFR50.55Ce) reportable cond!-

tion. The scope of the first two deficiencies also includes Unit 1.

These conditions required pronpt notification on January 4,1989 which was followed by LER 89-002 on February 3, 1989.

The appropriate corrective actions to resolve these conditions are being evaluated. The necessary actions will be identified and a schedule for ccmpletion of these actions will be provided in a supplement to SDR No. 249-2 (reference 2) by May 31, 1989.

ACTIONS TO PREVENT RECURRENCE PECo is performing an assessment of the LGS Safe Shutdewn (SSD) analysis to determine the root cause of recent LERs written against the SSD capabilities as corrmitted to in the LGS FPER. The results of the assessment will Identify if the non-conformances in the LERs are programmatic errors or Isolated deficiencies. The study should be i

completed by April 14, 1989, whereupon PECo shall provide a response discussing the results of this assessment and the proposed corrective actions in a supplement to SDR No. 249-2 (reference 2) by May 31, 1989.

GBH/ss/02158901

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