ML17261B108

LER 90-011-00:on 900619,fire Damper Found Missing During Surveillance Test PT-13.26 Due to Lack of Installation. Caused by Inadequate Design Info.Hourly Fire Watch established.W/900719 Ltr

ML17261B108
Person / Time
Site:	Ginna
Issue date:	07/19/1990
From:	Cavanaugh M, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-011, LER-90-11, NUDOCS 9007250215
Download: ML17261B108 (11)

Text

ACCELERATED D'RIBUTION DEMONSTRATION SYSTEM I

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9007250215 DOC.DATE: 90/07/19 NOTARIZED: NO DOCKET PT FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G'5000244 AUTH. NAME AUTHOR AFFILIATION CAVANAUGH,M.E. Rochester Gas & Electric Corp.

MECREDY,R.C. Rochester Gas & Electric Corp.

RECIP.NAME RECIPIENT,AFFILIATXON R

SUBJECT:

LER 90-011-00:on 900619,fire damper found missing during Surveillance Test PT-13.26 due to lack of installation. D W/9 DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

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NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 A

RECIPIENT COPIES RECIPIENT COPIES D ID CODE/NAME LTTR ENCL XD CODE/NAME LTTR ENCL PD1-3 LA JOHNSON,A 1

1 1

1 PD1-3 PD 1, 1 D INTERNAL: ACNW 2 2 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DS P .2 2 NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB9H3 1 1 NRR/DLPQ/LHFB11 1, 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB11 1 1 NRR/DREP/PRPBll 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR DSTAPLBBD1> 1 1 NRR/DST/SRXB 8E 1 1 E 1 1 RES/DSIR/EIB 1 1

'ILE 01 1 1 EXTERNAL: EG&G BRYCE,J.H 3 3 L ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC MAYS,G 1 1 NSIC MURPHY,G.A 1 1 NUDOCS FULL TXT 1 1 D

D D

NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE iVASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 32 ENCL 32

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'T C i0 C W 0h g AAcA cooE Tlc 546.2700 July 19, 1990 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 90-011, Fire damper found missing during surveillance test PT-13.26, due to not being installed, causes a condition prohibited by Technical Specifications.

R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10CFR50.73, Licensee Event Report System, item (a)(2)(i)(B), which requires reporting of Specifications","Any Operation or Condition Prohibited by the Plants Technical and plant Technical Specifications, section 3.14.6, which requires a 30 day special report, the attached licensee event report LER 90-011 is hereby submitted.

This event has in no way affected the publics health and safety.

Very truly yours Robert C. M cre Division Manager Nuclear Production xc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector 90072.02i5 900719 f3 go gs7zs8 WE RQ If(

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~ LICENSEE EVENT REPORT (lER) TEXT CONTINUATION V.N. HVCLtA1 1COVLATOIIYCOMMINCIOM J

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R.E. Ginna Nuclear Power Plant 0 0 5 0 0 0 1 0 OF VtXr III~MSCS1~. ~ ~SIAC A ~ Il IIn 2 4 4 9 0 1 0 2 NI L PLAN CONDI IO S The plant was at approximately 984 steady state full power with no major activities in progress.

II ESCRIPTION OF EVENT A. Dates and approximate times for major occurrences:

June 19, 1990, 1530 EDST: Event date June 19, 1990, 1530 EDST: Discovery date and time June 19, 1990, 1600.EDST: Fire protection initiated touring the area once per hour in compliance with Technical Specifications B. EVENT:

On June 19, 1990 at approximately 1530 EDST with the reactor at approximately 984 full power, Fire Protection and Safety- along with HVAC personnel were performing PT-13.26 Testing of Fire,Dampers. During the testing of fire damper RR-113-P which is the supply duct for the'ux room HVAC system, no fire damper in the duct.

it was noted that, there was There are fire dampers in the duct between the Mux room and the Relay room. This duct penetrates the Relay room south wall to a stairwell which leads to the Control room. This duct is identified as being a two (2) hour Technical Specification fire wall.

On June 19, 1990 at, approximately 1600 EST, the Fire Protection section assigned a fire watch to perform hourly tours.

C. INOPERABLE STRUCTURES, COMPONENTS OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

Missing fire damper RR-113-P D~ OTHER SYSTEMS OF SECONDARY FUNCTION AFFECTED:

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kAC form %4A IVIII V.I. kVCLIAAAI4ULATOAYCOMOONIIOk LICENSEE EVENT REPORT ILERI TEXT CONTINUATION AffAOVIOOMI kO 31IO&ICO IAflAIIInlrII fACILITYkAMI lll OOCKIT kUIOAKA LTI LIA kIAOIIAI ~ I ~ A4I (II eIOvok<rAL rre v ro lo k hl hr A hh R.E. Ginna Nuclear Power Plant 0 0 0 5 0 0 0 2 4 4 9 0 1 1 0 0 4 QF 0 8 TIxT Irf more krooe e reevveo. voo ooeooohor kAc rrhvrrr ~'llIITI IV LYSI EV N A 1 1/2 hour rated fire damper was not installed in a two hour fire'wall so Technical Specification 3.14.6 could not be met. This event is reportable in accordance with 10CFR50.73, Licensee Event Reporting SYstem, item (a) (2)

(i) (B), which requires reporting of, "Any Operation or Condition Prohibited by the Plant Technical Specification>>,

and (Other) plant Technical Specifications, section 3.14.6, which requires a 30 day special report.

An assessment was performed considering the safety of this event with the following results and conclusions: F There were no operational or safety consequences of implications attributed to the missing fire damper in the

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fire barrier wall because:

The control complex is made up of fire zones AHR, RR and CR. Fire zone AHR was not affected by the missing fire damper. Fire zone RR and CR fire induced effects, safe shutdown methods and proposed compliance methods more similar and are described as follows:

Rector makeup capability - a fire in the area may damage control circuitry to 480V AC buses 14 and 16 providing a loss of charging pumps and safety infection pumps. The alternative shutdown capability is provided by the charging pump 1A control circuitry modification.

Local transfer, circuit isolation and control of the pump at the charging pump room will be accomplished.

Manual, valve V-358 providing RWST suction to the charging, pump must be manually opened in charging pump room.

Reactor reactivity control - will be ensured by maintenance of makeup capability as described above.

Decay heat removal A fire in this area has the potential for damaging power feeds to both motor-driven AFW pumps, both dc trains A and B 480V ac battery charger feeds, and control circuitry for the standby, turbine-'driven and motor-driven AFW pumps.

shutdown capability is provided by operating The'lternative the turbine-driven AFW pump in the Intermediate Building kA4 f Oooo IIIIA II&I

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LICENSEE EVENT REPORT {LER) TEXT CONTINUATION AttlIOVEO OMI <<0, EIEO&ICN E)ctllIES I/314E OOCXET IIUaeEII ITI I,EII IIUIrNEIIIII tAOE IEI NNOU ~ STIAI.

M S R.E. Ginna 'Nuclear Power Plant o 5 o o o 2 4 4 9 0 0 1 1 0 0 0 50F TEXT IN'rrt <<<<er <<<<msst. rNN <<Ssrss<<NNC Arts ~'Il ITl, I

at the pump. The operator will manually isolate dc power to the turbine steam admission valves V-3504A and V-3505A at main dc distribution panels 1A and 1B in fire area BRlA and fire area BR1B. Manual operation start of one of the valves will then be accomplished to the turbine driver. DC power is required for operation of the turbine'-driven AFW lube oil pump. Power to .this pump is supplied from dc train B through the Turbine building dc distribution panel. A transfer switch will is located in the Intermediate building to isolate control circuits to CT and to allow for local oil control'f the dc lube pump.

Manual operation of discharge valves V-3996 and C-4297 (or V-4298) will then be accomplished to permit AFW flow to the, appropriate steam generator. As before, power will be isolated or instrument air isolated and bled off before manual operation is attempted.

'I Long term decay heat removal will be accomplished by use of the RHR system. RHR pump control circuits may be damaged by a fire in the area. Adequate time exists to either repair damaged control circuits or to locally control RHR pump breakers. This analysis assumes that repairs will not be made and that the breakers will be operated locally.

Reactor Pressure Control - pressure control is ensured by automatic operation of pressurizer safety valves and by maintenance of makeup capability as described above.

Support systems - Service water and CCW pump control may be affected by a fire in these areas. Emergency diesel generator control may be affected by a fire .in theses areas. However, alternative shutdown for EDG 1A will allow for complete isolation of all room. necessary .

control circuits from the Control and Relay EDG cooling can be accomplished by manually operating the service water pump breakers in the'creenhouse or via the underground yard fire water system. A local EDG connection exists that allows for a local water supply hook up in case service water is not available. CCW pump control is necessary for cold shutdown.

exists to allow for local control of CCW pump Adequate'ime.

breakers.

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R.E. Ginna Nuclear Power .Plant:

tm aAaeanV SAC Ann ~ Y I I ITI 0 5 0 0 0 2 4 4 9 0 .0 1 1 0 0 0 7 OF 0 diesel generator) and main fuse cabinet 1A (train A dc power). Use of this existing intertie will provide for long-term operation of process monitoring instrumentation located at the ABELIP.

Both redundant trains of source range neutron monitoring may be damaged by a fire in this area. A spare monitor drawer is located on site to provide the capability to.

bypass fire damaged circuits. A procedure is used to connect the 'existing neutron monitor inside the containment to the new drawer. A local power supply

'will be used to power the new drawer.

Based on the above and RG&E Ginna Station Appendix R Alternative Shutdown Report, it can be concluded that the employee's and .the public's health and safety was assured at all times.

V CO ECTIVE TIONS A. Actions taken to return inoperable components to operable status:

An hourly fire watch was established to patrol the area once per hour An evaluation will be conducted to determine a fire damper rating and size A new fire damper will be installed in the duct to bring the fire barrier in compliance with a rated barrier B. ACTIONS TAKEN OR PLANNED TO PREVENT RECUE%ENCE:

Complete a visual inspection and perform a trip test of all Technical Specification and Appendix R fire dampers VI ADDIT 0 A. Failed components I

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114 SvIvI ANNA VA. HVCltAA IIIOVLATOAYCOMMNSOM

%HI L'ICENSEE EVENT REPORT (LER) TEXT CONTINUATION AJP1OVTO OUI N), glgQ&lON TNPIATS 'TlT1IT$

NACIVITY llAMl ill OOCKET HVMISA OV ll1 NVMOI1 I~ l ~ AOl lTI VTAA $ $ 4 V ~ IIT I A g NQVINIOII U U 1 R.E. Ginna Nuclear Power Plant 0 so~0 TECT IN'NvVV Mvvv 1 ~. vM vAKenvI HNC AvtII ~'llllTl 0 5 0 0 0 2 4 4 9 0 0 1 1 0 o s B. Preview LER's on similar events A similar LER event historical search was conducted with the following results: no documentation of similar LER events with the same root cause at Ginna Station could be identified C. Special Comments None

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