ML17262A680

LER 91-009-00:on 911111,steam Generator Feedwater Isolations Occurred on Both Steam Generators.Caused by Perturbations of Advanced Digital Feedwater Control Sys.Feedwater Regulating Valves Manually controlled.W/911211 Ltr

ML17262A680
Person / Time
Site:	Ginna
Issue date:	12/11/1991
From:	Backus W, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-91-009, LER-91-9, NUDOCS 9112170532
Download: ML17262A680 (22)

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'CCELERATED DEMONRATION SYSTEM DISTRIBUTION INFORMATION DISTRIBUTION SYSTEM (RIDS) 'EGULATORY ACCESSION NBR:9112170532 DOC.DATE: 91/12/11 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME 'AUTHOR AFFILIATION BACKUS,W.H. Rochester Gas & Electric Corp.

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

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 91-009-00:on 911111,steam generator feedwater on both steam generators. Caused by perturbations of isolations'ccurred

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advanced digital feedwater control sys.Feedwater regulating D valves manually controlled.W/911211 ltr.

i(7 (LER), gIncidentg Rpt, etc.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report 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 ID CODE/NAME LTTR ENCL PD1-3 LA 1 1 PD1-3 PD 1 1 D JOHNSON,A 1 1 INTERNAL: ACNW 2 2 AEOD/DOA 1 1 AEOD/DS P/TPAB 1 1 AEOD/ROAB/DS P 2 2

.NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB 7E 1 . 1 NRR/DLPQ/LHFB10 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPB11 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1 NRR~ST/S PLB8 Dl 1 1 NRR/DST/SRXB 8E 1 1 1 1 RES/DSIR/EIB 1 1 N1 LE 01 1 '1 EXTERNAL: EG&G BRYCE,J.H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHY,G.A 1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 D

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NOTE TO ALL "RIDS'ECIPIENTS:

PLEASE HELP US TO REDUCE i'i'ASTE! CONTACT THE DOCUlii!EiiTCONTROL DESK, ROOli I Pl-37 (EXT. 2M79) TQ LILIih!INAl'E YOUR NAiIF. FROii1 DISTRIBUTION LISTS FOR DOCUiiIENTS YOU DON'T NEED!

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, ss@ss ROCHESTER GAS AND ELECTRIC CORPORATION e e9 EAST AVENUE, ROCHESTER N. K 14649-0001 ",

ROBERT C MECREDY TELEPiSONE Vice Psesidens AREA CUE 7 s 6 546 2700 Cfnnn s'ueiess PsodueBun December 11, 1991 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 91-009, Automatic Feedwater Control Perturbations, Due To'lectromagnetic Noise Spikes From Unrelated Relay Actuation, Caused Steam Generator Feedwater Isolation on High Level R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10 CFR 50.73, Licensee Event Report System, item .(a)(2)(iv), which requires a report of, "any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System (RPS)", the attached Event Report LER 91-009 is hereby submitted.

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

Very truly yours, Robert C. Mecredy xco U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector

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MOHTH CAY YIAR IVtfLtMINTALRtt01T IXtlCTIO IIII CXflCTIO Lvll<<I SCION OATI I'III Y Ct lllfer, reer<<rra IX<</CTIO SVS MISSION OATII NO LIITAAOTII<<err a Irof ireea, I ~ .. rterer<<errNf INaee r<<vrre<<rrr trteerarre <<ew ll ~I On November 11, 1991 at approximately 1214 EST, with the reactor at approximately 98% full power, steam generator feedwater isolations occurred on both steam generators. These feedwater isolations were caused by perturbations of the advanced digital feedwater control system which increased feedwater flow to t1 generators. 'team Immediate operator action was to manually control the feedwater regulating valves to reduce steam generator levels and stabilize the plant.

The underlying cause of the event was determined to be electro-magnetic noise spikes affecting the advanced digital feedwater control system.

Corrective action taken was to modify specific relay circuits that were causing these spikes.

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'NOVINT<Al, 'ttV4%8 U A R.E. Ginna Nuclear Power Plant TQTT III~ ~%M A ~. ~ y480OAV NAC Anil ~'il 2 4491 009 00 02or0 I ITI PRE-EVENT PLANT CONDITIONS The plant was at approximately 98% steady state reactor power with no major activities in progress. The Maintenance Department was performing troubleshooting, to determine the source of electromagnetic noise, spikes in the Advanced Digital Feedwater Control System (ADFCS). The troubleshoot-ing was being performed under the guidance of Work Order package f9122181. Unexplained electromagnetic 'noise spike problems were identified previously as coinciding with the start of the diesel fire pump, and which had minor effect on the ADFCS control functions.

The ADFCS was installed during the 1991 "Annual Refueling and Maintenance Outage. These electromagnetic noise spikes were first noticed on June 4, 1991, when a minor feedwater perturbation occurred, following a diesel fire pump start.

Since June 4, spikes have occurred almost every time the diesel 'fire pump has started. The ADFCS .has handled spikes with no noticeable feedwater perturbations, except for two (2) occasions. These occasions, the first on June 4, 1991 and the second on September 13, 1991, were handled by the ADFCS in automatic and no operator action was required.

There has been an ongoing search for the possible source of this electromagnetic noise spike so that corrected. As part of this ongoing search, the Electrical it could be Engineering Department evaluated their cable tray database and identified circuit E174 as a possible source. Circuit E174 is the 125 Volt DC power feed to the fire relay panel and shares some cable trays with ADFCS input cables, most notably, the feedwater header pressure inputs to and P502).

ADFCS'P501

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DESCRIPTION OF EVENT A. DATES AND APPROXIMATE TIMES OF MAZOR OCCURRENCES:

0 November 11, 1991, 1214 EST: Event Date and Approximate Time.

0 November 11, 1991, 1214 EST: Discovery Date and Approximate Time.

o November 15, 1991: Cause of EMP noise spike identified and suppressed to acceptable levels.

WGBFZ:

On November 11, 1991, at approximately 1214 EST, with the reactor at approximately 98% full power, the diesel fire pump was started, as required for ADFCS troubleshooting per Work. Order f9122181.

Approximately thirty (30) seconds after the diesel fire pump was started an "ADFCS System Trouble" alarm (G-22) was received.

The Control Room operator responsible for feedwater control had pre-positioned himself in front of the "A" and "B" S/G Main Feedwater Regulating Valves (FRV) control panel prior to the start of the diesel fire pump.

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AACILITYNASIS I 1 I- OOCKST IIVIAOIAIll L$ 1 NLAN$1 I~ I ~ A4$ ISI SSOUS>S<AL ' ASYCK)U U 1 U A R.E. Ginna Nuclear Power Plant o2 44 0 09 0004 or0 TQ(T /IT mew AIASS 1 ~. ~ e4004HV JYAC AMID ~ Yl I Ill o s o o 9 1 9 At this time, the Control Room operator noticed that both the "A" and "B" Steam Generator (S/G) main feedwater flows were pegged high .with both "A" and "B" S/G Main Feedwater Regulating Valves continuing to open further.

The condensate low pressure heater . bypass valve opened automatically and the standby condensate pump started automa'tically (to increase main feedwater pump suction pressure). Main Feedwater pump suction pressure was decreasing due to the increased feedwater flow to the S/Gs. The "A" and "B" S/G levels continued to increase and before the Control Room operator could shift 'the FRVs to manual, ADFCS automatically shifted the FRVs to manual. While the Control Room operator was manually lowering the setpoints for the FRV controllers, to control S/G level, the following alarms annunciated and feedwater isolation o'ccurred on both S/Gs; G-4 (S/G A HI LEVEL CHANNEL ALERT 67%)

and G-6 (S/G B HI LEVEL CHANNEL,ALERT 67%).

Immediately following the feedwater isolation, the condensate booster pumps tripped on high* pressure. A load de'crease was initiated at 10%/hour to lessen the impact of unstable S/G levels. Main feedwater to the S/Gs was controlled in manual in order to stop secondary system oscillations that were occurring due to the event. During the S/G level stabile.zation, S/G feedwater isolation occurred several times. The S/G levels were subsequently stabilized and main feedwater control was returned to automatic.

After main feedwater control was returned to automatic the load decrease was terminated. Total load decrease was approximately 0.54 full power during the event.

Subsequently, the condensate low pressure heater bypass valve was closed, the condensate booster pumps were restored, and the standby condensate pump was secured and realigned for automatic standby.

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None.'.

OTHER SYSTEMS OR SECONDARY FUNCTIONS AFFECTED:

None.

E. METHOD OF DISCOVERY:

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

OPERATOR. ACTION:

The Control Room operators took immediate manual actions to control S/G levels, reduce power level, and stabilize the plant. Subsequently, the Control Room operators notified higher supervision and the Nuclear Regulatory Commission per 10 CFR 50.72, non-emergency, 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> notification.

G. SAFETY SYSTEM RESPONSES:

The "A" and "B" FRVs closed automatically from the feedwater isolation signal.

III. CAUSE OF >wryly A. IMMEDIATE CAUSE:

The feedwater isolation of the "A" and "B" S/G was due to the "A" and "B" S/G narrow range levels being

)/ = 67%.

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The "A" and "B". S/G narrow range levels were >/= 67%

due to increased feedwater flow to both S/Gs caused by a perturbation of the ADFCS.

The perturbation of th'e ADFCS was apparently due to electromagnetic noise spikes affecting the feedwater header pressure inputs to ADFCS, (i.e. P501 and P502).

C. ROOT CAUSE:

After extensive troubleshooting, it was determined that the spikes that affected the ADFCS feedwater header pressure inputs were caused by the de-energiza-tion of Relay ARSO, located in the fire relay panel.

This relay, which lights the diesel fire pump trouble light, de-energizes approximately 10 to 15 seconds after a diesel fire pump start. During this de-energization, inductive "kickback" causes an electro-magnetic noise spike to be generated and induced into the feedwater header pressure inputs. The signal cables carrying the feedwater header pressure transmitter (PT-.501 and PT-502) inputs share some common cable trays with the 'DC power source for the ARSO relay, and a noise spike was induced from the ARSO relay cable to the feedwater header pressure input cables.

ANALYSIS OF &TENT This event. is reportable in accordance with 10 CFR 50.73, Licensee Event Report System, item (a)(2)(iv), which requires reporting of, "any event or condition that resulted in manual cr automatic actuation of any Engineered Safety Feature (ESF) including the Reactor Protection System (RPS)". The feedwater isolation of the "A" and "B" S/Gs was an automatic actuation of an ESF system.

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R.E. Ginna Nuclear Power Plant 24 49 0- 07 os0 TlxT IA' ~ ~. r ~ ereereMI eIAC Perrrr ~'llI I TI o s o o o 1 9 0 9 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 or implications attributed to the feedwater isolations because:

o The feedwater isolations occurred at the required S/G levels.

o The plant was quickly stabilized and manual control of the FRVs was accomplished to mitigate the transient.

o As the feedwater 'solation occurred as designed, the assumptions of the FSAR for steam line break were met.

Based on the above, -it can be concluded that the public's health and safety were assured at all times.

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

0 The Diesel Fire Pump was temporarily removed from service pending the outcome of root cause troubleshooting and determination. (The pump was returned to service after a noise suppression diode was installed across relay AR80).

0 When S/G levels were stabilized, subsequent to the ADFCS perturbation termination, the FRVs were placed in automatic control.

0 After the plant had been stabilized and the FRVs returned to automatic control, the condensate low pressure heater bypass valve was closed, the condensate booster pumps were restored and the standby condensate pump was secured and realigned for automatic standby.

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• A reverse-biased diode was temporarily installed across the coil of AR80 on November 15, 1991 and subsequent testing determined that the spikes from the AR80 circuit, affecting ADFCS feedwater pressure inputs following diesel fire pump

, starts, were eliminated. This noise suppression diode 'was permanently installed on November 18, 1991.

After reviewing the results of troubleshooting and the discussion with Westinghouse, the following is an outline of the corrective actions being taken or planned in response to the ADFCS noise spiking events:

o Short Term Response a) Operations personnel were made aware that one source of spikes on ADFCS was eliminated, but that spikes from other sources, while reduced in frequency. and magnitude, might occur. Operations will identify any new spikes on the ADFCS by submitting a Work Request/Trouble Report (WR/TR).

b) A WR/TR was submitted for installation of a diode for the fire booster pump relay AR85 (which also produces small spikes on ADFCS). However, these spikes are not of the same magnitude as the noise spikes that were caused by the Diesel Fire Pump starts.

o Intermediate Term Response Electrical Engineering will consult with Westinghouse concerning a database change to increase the ADFCS slew rate filter constant.

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o Long Term Response a) Electrical Engineering will check with Westinghouse for the results of their review of ADFCS arbitration error checking software. This review will determine if the error checking routine for the switching to arbitration values instead of feedwater header pressure field input values is substituting erroneous values for feedwater header pressure used in FRV demand calculations.

b) Electrical Engineering will evaluate the routing of feedwater header pressure input circuits (to the- ADFCS), and will identify any additional modifications that may be required to eliminate the electromagnetic noise spike concern.

VX. ADDITXONAL NFORMATION 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 imilar LER events with the same root cause could be

.identified.

C. SPECIAL CO~BFZS:

None.

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