ML17261A822

LER 88-010-00:on 881211,two Steam Generator 1B Pressure Channels Began Drifting High.Caused by Freezing of Affected Pressure Transmitter Sensing Lines.Affected Lines Thawed & calibr.W/890110 Ltr

ML17261A822
Person / Time
Site:	Ginna
Issue date:	01/10/1989
From:	Backus W, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
HLWR, LER-88-010, LER-88-10, NUDOCS 8901200173
Download: ML17261A822 (13)

Text

ACCELERATED DISTRIBVIION DEMONSTRATION SYSTEM REGULATORINFORMATION DISTRIBUTION .STEM (RIDE)

ACCESSION NBR:8901200173 DOC.DATE: 89/01/10 NOTARIZED: NO DOCKET N FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION BACKUS,W.H. Rochester Gas & Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 88-010-00:on 881211,simultaneous loss od two "B" SG pressure channels due to sensing line freezing.

W/8 ltr.

DISTRIBUTION CODE: IE22D COPIES RECEIVED:LTR i ENCL TITLE: 50.73 Licensee Event Report (LER), Incident Rpt, etc. j SZEE:

NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 S RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 LA 1 1 PD1-3 PD 1 1 STAHLE,C 1 1 INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 ACRS WYLIE 1 1 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 ARM/DCTS/DAB 1 1 DEDRO 1 1 NRR/DEST/ADE SH 1 1 NRR/DEST/ADS 7E 1 0 NRR/DEST/CEB SH 1 1 NRR/DEST/ESB SD 1 1 NRR/DEST/ICSB 7 1 1 NRR/DEST/MEB 9H 1 1 NRR/DEST/MTB 9H 1 1 NRR/DEST/PSB SD 1 1

,NRR/DEST/RSB SE 1 1 NRR/DEST/SGB SD 1 1

'2 NRR/DLPQ/HFB 10 1 1 NRR/DLPQ/QAB 10 1 1 NRR/DOEA/EAB 11 1 1 NRR/DREP/RAB 10 1 1 NRR/DREP/RPB 10 2 2 N~R-DR- I S B 9A 1 1 NUDOCS-ABSTRACT 1 1 RZ~~ 1 1 R

RES/DSIR/EIB 1 1 DSR/PRAB 1 1 RGN1 FILE 01 1 1 I

EXTERNAL EG&G WILLIAMSI S 4 4 FORD BLDG HOY,A 1 1 H ST LOBBY WARD 1 1 LPDR 1 1 D NRC PDR 1 1 NSIC HARRIS,J 1 1 NSIC MAYS F G 1 1

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NIIC Sorer SOS UA. IIUCLKAR REOULATOIIY COaaMQSION IS42 I ASSROVED OMS NO.SISO DION LICENSEE EVENT REPORT HLER) EXSIRES ~ 0/Sa/() S SACILITY NAME Il) DocxKT IIUMEKR Ql tA R.E. Ginna Nuclear Power Plant 0 5 0 0 0 4 4 1 OF1 0 TITLE (a) Zmu aneOuS SS O earn enera or ressure nne s, e ' ensuing Line Freezin , Causes a Co()mon Mode Failure Condition.

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NAME TELESHONE NVMSER Nesley kI. Backus AREA COOK Technical Assistant to the rations Mana er 3 1 COMtLK'TE ONE LINK SOR EACH CDMSONKNT SAILURK OESCRISED IN THIS REtORT (12)

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D PIT F)l 8 0 Y SVSSLKMKNTAL RESOR'T EXSECTEO (Ia) ~ aONTH DAY YEAR EXtECTEO 1V 2 M1$ 1 ION DATE II E)

YES IIItaa, camtae<<SxtEctf 0 susoNssIOH DA TEI NO ASST R ACT IVmla IO laOO acre<<, I A, chere atara<<at IN<<ee aleteeaaaae Crtrreratree Iwra) II~ I On December 100% full power ll, 1988 at 1124 EST with the reactor at approximately one of the three 1B steam generator pressure channels began drifting high. At approximately 1144 EST a second 1B steam generator pressure channel began drifting high.

The underlying cause of the above events was the freezing of the affected pressure transmitter sensing lines leading to the simultaneous loss of two independent safety related steam generator pressure channels.

Tmmediate operator action was to declare the affected pressure channels inoperable and commence a load reduction to hot shutdown per Technical Specification action statement.

Corrective action taken was to thaw out the affected sensing lines and calibrate and return to service the affected pressure channels.

8'7)01200173 8'7)0110 05000244 PDR ADOCK S

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U.S. NUCLE AR REOULATOIIY COMMISSION NRC /orm 9AAA i9491 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION ARRAOVEO OME NO 9ISOWIOt ER ~ IREE EISI/9$

R AC I LIT Y NAME I I I OOCKET NUMEER 191 LEII NUMSER l ~ I ~ AOE ISI

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~ rUM th VUM to R.E. Ginna Nuclear Power Plant. o s o o o 244 8 0 1 0 000 2 DF 1 0 TExT Illmore NrrNe e eettteer. we ohetteerM HRc lrorhI ~'AI llll PRE-EVENT PLANT CONDITIONS The unit was at approximately 100% steady state full power with no ma)or activities in progress'. A cold weather walkdown per Administrative Procedure A-54.4.1 had been completed at 0842 EST December ll, 1988. r DESCRIPTION OP EVENT A. DATES AND APPROXIMATE TIMES FOR MAJOR OCCURRENCES:

0 December 11, 1988, 1124 EST: ".B" Steam Generator Pressure Channel PT-479 started to was declared inoperable and defeated per Ecpxipment drift high and Restoration Procedure ER-INST.1 (Reactor Protec-tion Bistable Defeat After Instrumentation Loop Failure).

0 December 11, 1988, 1144 EST: "B" Steam Generator Pressure Channel PT-483 started to was declared inoperable.

drift high and 0 December ll, 1988, 1144 EST: Event date and time.

December 11, 1988, 1144 EST: Discovery date and time. I 0 December 11, 1988, 1230 EST: Started unit load reduction.

0 December 11, 1988, 1428 EST: "B" Steam Generator Pressure Channel PT-483 declared operable.

0 December 11, 1988, 1428 EST: Stopped unit load reduction.

0 December Pressure ll, 1988, 1522 Channel PT-479 EST: "B" Steam Generator declared operable.

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U.S. NUCL'EAA AECULATOAY COMMISSION NAC Fons SEEA AFFAOVEO OMS NO, 5IEOWI04 15@El LICENSEE EVENT REPORT ILER) TEXT CONTINUATION EXFIAES SITIIS5 LEA NUMSEA I ~ I ~ AOE ISI FACILITY NAME III SEQVEHTIAL NVM EA R.E. Ginna Nuclear Power Plant o 5 o o o 2 4 4 8 8 0 1 0 0 0 0 3 oF l 0 TEXT IIF mort AMoI A movFMI. vM ~WIC Fame ~'FI llll B. EVENT:

On December ll, 1988 at 1124 EST with the reactor at approximately 1004 full power, steam generator 1B pressure channel PT-479 began drifting high and was declared inoperable and defeated per Equipment Restoration Procedure ER-INST.l, (Reactor Protection Bistable Defeat After Instrumentation Loop Failure).

At, 1144 EST, steam generator 1B pressure channel PT-483 also started drifting high and was declared inoperable.

At this time, the Control Room operators determined that the probable cause of the 1B steam generator pressure channel problems was freezing'of the trans-mitter sensing lines, as all other indication of plant status was normal. Auxiliary operators were dis-patched to the area where the 1B steam generator pressure transmitters were located and reported back that it was extremely cold in the area. This condition was found to be due to outside cold air, (approximately 10 F) being drawn into the building through the inlet air dampers located near the affected pressure transmitters. Actions were immediately taken to reclose the outside air dampers and supply additional heat to the area.

At 1230 EST, a load reduction of 20% per hour was commenced due to Technical Specifications section 3.55 action statement 95 which requires, "at any time the number of operable channels is less than the minimum operable channels required, be at hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at an RCS temperature less than 350 F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, (i.e.

the number of channels at this time was one less than the minimum operable channels required.)

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U.S. NUCLEAR RKOULATORY COMMITSION NRC Forrrr A4A I 045 I LICENSEE EVENT REPORT ILER) TEXT CONTINUATION APPROVED OMR NO. T154&105 TR ~ IRTS d(TI(55 FACILITY NAMK III OOCKCT NUMRKR LTI LCR NUMRIR ldl 1

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At approximately 1300 EST, the 1B steam generator pressure channel PT-479 started decreasing from its highest pressure reached of 1091 psig.

thought at this time that the sensing lines for PT-It was 479 were thawing and that the channel pressure would return to actual steam generator pressure but PT-479 indication continued to decrease below actual steam generator pressure. At approximately 1315 EST, the 1B steam generator pressure channel PT-483 started decreasing from its pegged high position. With PT-479 defeated and PT-483 decreasing, the possibility existed of an inadvertent safety in)ection signal being generated from 2/3 lo lo steam generator pressure of 514 psig if PT-483 continued to decrease in the same manner as PT-479. With PT-479 at approxi-mately 120 psig, .the Operations Shift Supervisor instructed the Instrument and Control (I&C) Technician to insert a simulated signal of approximately 700 psig into the defeated channel PT-479. PT-479 was to remain defeated until it was determined whether PT-483 would decrease past the safety injection initiation setpoint of 514 psig. When PT-483 decreased below 600 psig, PT-479's low and lo lo pressure bistables were reinstated and PT-483 was placed in the defeat mode.

The 1B steam generator pressure channel PT-483 did decrease below 514 psig but safety injection was not actuated due to the simulated signal in PT-479. With PT-483 defeated the safety in)ection coincidence was 1/1 with 1B steam generator pressure channel PT-478 being the only operable channel.

At approximately 1410 EST the I&C department had completed. thawing the sensing lines to PT-479 and .PT-483 and their pressure indication had returned to normal. At approximately 1428 EST the I&C department completed calibration and return to operable status of PT-483. As the Technical Specifications for minimum operable channels was met with the return of PT-483 to operable status the load reduction was terminated. PT-479 was calibrated and returned to operable status at approximately 1522 EST.

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U.S. NUCLfAR RlOULATORY COMMISSION NRC %%dr~ S44A

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION AFFROVEO OMl NO. SI50M<04 E<< ~ <RES 4<SI<SS OOC<<ET NUMSER ISI ~ AOE IS)

FACILITY NAME III I.ER NUMOER I ~ I S~ QUS<<T<AL <ISVIS<0<<

NUM ~R <<UM Sh R.E. Ginna Nuclear Power Plant: o so oo244 8 8 0 0 0 0 5 OF 1 0 TEXT I<F m<<<4 <<M<<4 <<<44<<<44<. <<44 ~ <<<<<h<<F HRC %%dhh JINA'4 I II TI C. INOPERABLE STRUCTURES, COMPONENTS, OR SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None.

E. METHOD OF DISCOVERY:

The event was immediately apparent due to alarms and indication in the Control Room.

F. OPERATOR ACTION:

Immediate operator action was to defeat the inoperable 1B steam generator pressure channel PT-479 and with the failure of the second 1B steam generator pressure channel, to determine the cause of these two pressure channels failing simultaneously.

The Control Room operators, after determining that freezing of the pressure channel sensing lines was the probable cause, took immediate action to reduce the amount of cold air coming into the pressure channel sensing line area and to supply additional heat to this area.

After declaring two of the three 1B steam generator pressure channels inoperable, the Control Room operators started reducing unit power per technical specification action statement.

During this event period, the plant was operating with known system integrity which would not warrant the need for safety injection actuation.

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V.E. NVCLEAN AEOULATOIIYCOMMIEEION NAC rrorrrr 555A IW51 LICENSEE EVENT REPORT ILER) TEXT CONTINUATION AeelIOVEO OME NO, 5150WI05 EAelllE5 EI)1t55 FACILITY NAME III OOCKET NUMEEII IEI LEII NUMEEII I51 ~ AOE 151 55QUErrellL NUM Eo lrUM

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~1 R.E. Ginna Nuclear Power Plant o 5 o o o 244 8 010 0 6 OF 1 0 Tz@T ]It more mooe ~ reooeeeI. rree ~ tooorot HtIC term 5$ $A EI I ITI Because the pressure indication was decreasing toward the safety injection initiation setpoint in the two inoperable 1B steam generator pressure channels, the Operations Shift Supervisor instructed the I&C Technician to insert a simulated signal into one of the inoperable pressure channels to avoid an in-advertent safety injection actuation. The decision that a safety injection actuation was not needed and the simulator should be installed was reviewed and concurred on by the Shift Supervisor (SRO), the Control Room Foreman (SRO), and the Reactor Engin-eer/Technical Manager acting as Shift Technical Advisor. During the time this signal was inserted the Control Room operator's were monitoring the 1B steam generator parameters and reactor coolant system parameters very closely for the steam break accident.

G. SAFETY SYSTEM RESPONSES:

None.

III. CAUSE OF A. IMMEDIATE CAUSE:

The immediate cause of the event was the common mode failure of two of the three 1B steam generator-pressure channels.

B. INTERMEDIATE CAUSE:

The common mode failure of the two 1B steam generator pressure channels was due to the freezing of the pressure channels sensing lines.

C ~ ROOT CAUSE:

The freezing of the two 1B steam generator pressure channels sensing lines was due to inadequate cold weather operations and procedures.

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U,S NUCLEAR RKOULATORY COMMISSION HRC form 054A AttROVSO OM4 NO S150&I04 1045 I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION SXtlh4$ 4ISI/4$

OOCKST NUM44R ~ AOS ISI tACILITYNAMS Ill 141 LKR NUM44R I ~ I U I rr 5 I A 5 II 0HUM AIVISIOH Ih NUM Ih R.E. Ginna Nuclear Power Plant o 5 o o o 24 488 010 00 07 OF 1 0 TTXT (IT mort tooot A neural ow odohtohtt ANC %%de JOSA'tl 1171 IV. ANALYSIS OP EVENT This event is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a) (2)(vii)(D) which requires a report of, "any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a system designed to mitigate the consequences of an accident".

The common mode failure of the two independent 1B steam generator pressure channels was an event where" a single condition caused two independent. channels to become inoperable in a system designed to mitigate the consequences of an accident.

An assessment was performed considering both the safety consequences and implications of this event with the following results and conclusions:

There were no operational or safety consequences attributed to the common mode failure of the two 1B steam generator pressure channels because:

0 The failures and failure mode were determined very quickly.

0 One pressure channel remained fully operable throughout the event.

0 Unit load was reduced in preparation for taking unit off the line in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> per plant Technical Specifica-tion action statement.

0 A simulated signal was inserted into one of the inoperable pressure channels to avoid an inadvertent safety injection actuation and potential for a subsequent detrimental transient on the plant.

The inoperable pressure channels were repaired, calibrated and returned to service quickly.

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U.5. NUCLEAR REOULATORY COMM14410N NAC Sei~ 344A I 043 l LICENSEE EVENT REPORT (LER) TEXT CONTINUATION AeeROVEO OM4 NO, 2ISOW104 ERelRES 4ISIISS OOCRET NUM44R I21 LER NUM4ER ISI ~ AOE ISI SACILITY NAME I)l 55OVINTIAL I A V i@10 H HVM 4R Nvv Ie R.E. Ginna Nuclear Power Plant o 5 o o o 2 4488 0 l 0 0 0 8 OF 1 0 TExT Irt epee Meee N NeveeeL eee edeeeeeve NRC fcnn ~'el I Ill There was the following safety implication attributed to the common mode failure of the two 1B steam generator pressure channels:

o Was the plant still protected against the steam line break accident with the 2 inoperable 1B steam generator pressure channels?

The R.E. Ginna Nuclear Power Plant Updated Final Safety Analysis Report (Ginna/UFSAR) section 15.1 states in'art that the following systems provide the necessary protection against a steam pipe rupture.

1. Safety in)ection system actuation on the following:

a. Two-out-of-three pressurizer low pressure signals.

b. Two-out-of-three low pressure signals in any steam line.

c. Two-out-of-three high containment pressure signals.

Based on this review of the UFSAR, blocking of the steam generator low pressure safety in)ection signal has negli-gible effect on the most limiting transient involved.

Based on the above it can be concluded that safety in)ection actuation for the steam line break would still have taken place from either low pressurizer pressure or high contain-ment pressure thus assuring the public's health and safety at all times.

CORRECTIVE ACTION A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVENT NORMAL STATUS:

o The cold air situation in the area of the affected pressure transmitter sensing lines was immediately rectified.

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o Unit load was returned to full power.

B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

0 Rewrite Administrative Procedure A-54.4.1 (Cold Weather Walkdown Procedure) to provide a better trigger mechanism for initiation and to provide specific quantitative guidance for areas, components, and temperature criteria.

0 Add thermometers in temperature sensitive areas of affected buildings.

0 Add temperature readings of temperature sensitive areas to Aux. Operators Log.

0 Initiate Engineering Work Request to review the adequacy of HVAC systems in the Intermediate Building.

0 Evaluate the integrity of the affected steam generator pressure instrument tubing.

0 Evaluate relocation/replacement of the affected steam generator pressure instrument tubing.

0 Replace existing outside air louver with manual positive sealing louver.

0 Review other potential corrective action to preclude freezing of instrument lines.

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U.S. NUCLKAN XSOULATOAY COMMISSION NIIC Fee~ SttA IOeSI LICENSEE EVENT REPORT (LER) TEXT CONTINUATION AFFIIOVSO OMS NO, SISO&104 S X ~ 11 t$ NISI ISS FACILITY HAMS 111 OOCKST NUMSSII ISI LSII NUMSSII I ~ I ~ AOS ISI YSAn SSQVtNTIAL ntvlSlov vvM n vvM tn R.E. Ginna Nuclear Power Plant 24 488 0 1 0 0 10 DF1 0 TExT llfnxxe NxNe ~ recvee. we eeetNnel HIIc Fenn ~'IlIITI ADDITIONAL INFORMATION:

A. FAILED COMPONENTS:

None B. PREVIOUS LERs 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:

Sensing line for PT-479 also froze on 12-5-80, which did not result in a LER. A main feedwater instrument line also froze on 12-7-84. This entire issue of instrument line operability is discussed in NRC Inspection Report 88-26.

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IL'LCPHO<c inca cooa 7se 546-2700 January 10, 1989 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER-88-010, Simultaneous Loss of Two Safety Related "B" Steam Generator Pressure Channels, Due to Sensing Line Freezing, Causes a Common Mode Failure Condition R.E. Ginna Nuclear Power Plant Docket No. 50-244 (

In accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(vii)(D) which requires a report of, "any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a system designed to mitigate the consequences of an accident",

the attached Licensee Event Report LER 88-010 is hereby submitted.

This event has in no way affected the public's health and safety.

Very truly yours, Robert C. Me redy General Manager Nuclear Production xce U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector