ML17262A806

LER 92-003-00:on 920229,reactor Trip Occurred Due to Lo Lo Level in Steam Generator B.Caused by Plugged Instrument Tubing for Feedwater Pump a Seal Injection Differential Pressure Switch.Turbine flushed.W/920330 Ltr

ML17262A806
Person / Time
Site:	Ginna
Issue date:	03/30/1992
From:	Backus W, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-003, LER-92-3, NUDOCS 9204080122
Download: ML17262A806 (22)

Text

" 'ACCELERATED DISTRIBUTION DEMONST$&TION SYSTEM REGULA'I . INFORMATION DISTRIBUTIO SYSTEM (RIDS)

ACCESSION NBR:9204080122 DOC.DATE: 92/03/30 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION BACKUSiW.H. Rochester Gas & Electric Corp.

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

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 92-003-00:on 920229,a reactor trip occured due to Lo .Lo level in "B" Steam Generator. Caused by a plugged instrument tubing for "A" Feedwater Pump Seal Injection Differential Pressure switch. Unplugging of instrument.W/910331 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ( ENCL t TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

SIZE: /5 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244

, RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PDl-3 LA 1 1 PD1-3 PD 1 1 JOHNSON;A 1 1 INTERNAL: ACNW 2 2 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 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/DS~T+EBB8D1'EG 1 1 NRR/DST/SRXB 8E 1 1 WLE~=002I 1 1 RES/DSIR/EIB 1 1 RGN FILE 01 1 1 EXTERNAL: EG&G BRYCEEJ.H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHYiG A ~ 1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL "RIDS" RECIPIENTS PLEASE HELP US TO REDUCE WAS'ONTACT 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 30 ENCL 30

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i'!i~Se~,;~/i/lii"l~!it/i,7 ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER N. K 14649.0001 ROBERT C. MECREDY TELEPHONE Vice President AREA COOE 7 t 6 546 2700 Oinna Nuclear Production March 30, 1991 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D'C 20555

Subject:

LER 92-003, Feedwater Transient, Due to Loss of "A" Main Feedwater Pump, Causes Lo Lo Steam Generator Level Reactor Trip 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 92-003 is hereby submitted.

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

Ver truly yours, Robert C. Me edy xco U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector GDO." I:i 9204080122 920330 PDR ADOCK 05000244

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED 0MB NO. 31504)104 EXPIRES: 4(30(92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT {LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH IP630), V.S. NUCLEAR REGULATOAY COMMISSION. WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3(504)(04), OFFICE OF MANAGEMENTANO BUDGET, WASHINGTON, OC 20M3.

FACILITY NAME (I) DOCKET NUMBER l2) PA E R.E. Ginna Nuclear Power Plant 0 5 0 0 0 2 4 4 1 OF 0 9 Feedwater Transient, Due To Loss of "A" Main Feedwater Pump, Causes Lo Lo Steam-EVENT DATE (5)

Generator Level Reactor LER NUMBER (8) REIIORT DATE (7)

Tri OTHER FACILITIES INVOLVED (6)

MONTH DAY YEAR YEAR ops sEQUENTtAL MONTH OAY YEAR FACILITYNAMES DOCKET NUMBER(SI NUMBER 0 5 0 0 0 0 2 2 9'9 2 9 2 0 0 3 0 0 0 3 3 0 9 2 0 5 0,0 0 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE RtQUIREMENTS OF 10 CFR ()s ICrreck onc or more of tne followinp) (11)

MODE (9) 20A02(b) 20.405(cl 50.73(e) (2)(ivl 73.71(5)

POWE R 20.405( ~ )Ill(0 50.36(el Ill 50.73(el(2)(v) 73.71(c)

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LICENSEE CONTACT FOA THIS LER (12)

NAME TELEPHONE NUMBER Wesley'H. Backus AREA CODE Technical Assistant to- the Operations Mana er 3 1 5 5 2 4 - 4 4 4 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBEO IN THIS AEPORT (13)

CAUSE SYSTEM COMPONENT MANUFAC MANUFAC. E'PORTABLE TVREA CAUSE SYSTEM COMPONENT ..w, TVRER TO NPRDS

,. o... QS SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTFD SUBMISSION DATE (15I YES Ilfycs, complere EXPECTED SIJ64IISSIDII DATEI NO ABSTRACT ILlmlt to tc00 sp>>ces, I e., epproxlmetely flfrsen sinpre.specs typswriNsn linNI (18)

On February 29, 1992 at. approximately 1346 EST, with the reactor stable at approximately 974 reactor power, just subsequent to a main feedwater pump trip, a reactor trip occurred due to Lo Lo level (</='74) in the "B" Steam Generator (S/G).

The Control Room operators immediately performed the appropriate actions of Emergency Operating Procedures E-0 (Reactor Trip Or Safety Injection) and ES-0.1 (Reactor Trip Response). Both Main Steam Isolation Valves (MSIVs) were subsequently closed to limit a Reactor Coolant System (RCS) cooldown and the plant was stabilized at hot shutdown.

The underlying cause of the event was plugged instrument tubing for the "A" Feedwater Pump Seal Injection Differential Pressure (D/P) switch which tripped the main feedwater pump. (This event is,NUREG-1022 (X) cause code.)

Corrective action taken was the unplugging of the instrument tubing by flushing and the replacement of one section of tubing and connections. Corrective actions to prevent recurrence are discussed in Section V of the text.

NRC Form 366 (64)9)

c NRC FORM 388A US. NUCLEAR REGULATORY COMMISSION

)689) APPROVED 0MB NO. 3)600104 EXPIRES; 4/30/92

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TIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS

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LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 600 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IP.630), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPFRWORK REDUCTION PROJECT 131600104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY kAME 11) DOCKET NUMBER 121 LER NUMBER LS) PAGE IS)

YEAR

"' SEQUENTIAL .~' REVIS10>

NUM ER 'OO NUM ER R.E. Ginna Nuclear Power Plant 0 6 0 0 0 2 4 4 2 0 0 3 0 0 0 2 oF0 9 TEXT ///mo 4Pooo JP~ ooo d/d/ooo/P/RC ronn 38544/)In

-PRE-EVP2FF PLANT CONDITIONS The plant was at approximately 974 steady state reactor power with no major activities in progress.

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

February 29, 1992, 1346 EST: Event Date and Time 0 February 29, 1992, 1346 EST: Discovery Date and Time 0 February 29, 1992, 1346 EST: Control Room operators verify both reactor trip breakers open, and all control and shutdown, rods inserted.

February 29, 1992, 1350 EST: Control Room operators close both Main Steam Isolation Valves (MSIVs) to limit a Reactor Coolant System (RCS) cooldown.

0 February 29, 1992, 1358 EST: Plant stabilized at hot shutdown'ondition.

.B. EVENT:

On February 29, 1992 at approximately 1346 EST, with the reactor stable at approximately 974 reactor power, the Control Room received Annunciator Alarm H-11 (Feed Pump Seal Water Lo Diff Press 15 Psi) followed in approximately five (5) seconds by a trip of the "A" Main Feedwater Pump. The Control Room operators immediately entered Abnormal Procedure, AP-FW.1 (Partial Or Complete Loss Of Main Feedwater) and performed the immediate actions (i.e. starting all three (3) Auxiliary Feedwater (AFW) pumps, and decreasing turbine power= rapidly to less than 50%.)

NRC Form 388A (84)9)

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NRC FORM 366A (689) t LICENSEE EVENT REPORT ILER)

TEXT CONTINUATION UJL NUCLEAR REGULATORY COMMISSION APPROVED 0MB NO. 31500)05 EXPIRES: E/30/92 TIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 50J) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE IS)

YEAR NUMBER J~

IIa'ECUENTIAL ~Orr REYIBION NUM ER R.E. Ginna Nuclear Power Plant TEXT (Ifrr>>ro BP>>c>> JE ~, lrro JJ(Jor>>J PIC Form 36EA3) (IT) o 5 o o o 24492 0 0 3 0 0 0 3 OF 0 9 During the performance of these immediate actions, a reactor trip occurred due to Lo Lo level (</=174) in the "B" Steam Generator (S/G).

The Control Room operators performed the immediate actions of Emergency Operating Procedure E-0 (Reactor Trip Or Safety Injection) and transitioned to Emergency Operating Procedure ES-0.1 (Reactor Trip Response) when it was verified that both reactor trip breakers were open, all control and shutdown rods were inserted, and safety injection was not actuated or required.

Both MSIVs were subsequently closed at 1350 EST to limit the RCS cooldown. The closing of the MSIVs mitigated the RCS cooldown and the plant was subsequently stabilized i:n hot shutdown at approximately 1358 EST.

The Intermediate Range Nuclear Instrumentation Channel N-35, after tracking consistent with channel N-36 down to approximately 1E-10 amps, had its indication continue to drop below 1E-11 amps. The N-35 channel returned to normal (i.e. 1E-11 amps) approximately ten (10) hours following the trip.

C INOPERABLE STRUCTURES P COMPONENTS P

'R SYSTEMS THAT CONTRIBUTED TO THE EVENT:

None.

D. OTHER SYSTEMS OR SECONDARY FUNCTIONS APPECTED:

None E. METHOD OP DISCOVERY.

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

NRC Form 366A (SJ)9)

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NRC FORM 366A (649)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION US. NUCLEAR REGULATORY COMMISSION t APPROVEO OMB NO. 31504(OS EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 503) HRS. FORWARD COMMENTS REQARDINQ BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F630), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31504104), OFFICE OF MANAGEMENTANO BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER IS) PAGE (3)

YEAR pg SEQUENTIAL NVM Ea ~Q~

REYISION NVMEEa R.E. Ginna Nuclear Power Plant o s o o o 2 4 4 9 2 0 0 3 0 0 0 4 QF 0 TEXT ///IaoIP space /s nqukat, vss ~ /I/oaP/HRC Fcna 356AB/ (12)

F. OPERATOR ACTION:

After the reactor trip, the, Control Room operators performed the actions of Emergency Operating Procedures E-0 (Reactor Trip Or Safety Injection) and ES-0.1 (Reactor Trip Response). The MSIVs were manually actuated closed approximately four (4) minutes after the trip to prevent further plant cooldown. The plant was subsequently stabilized at hot 'hutdown.

Subsequently, the Control Room operators notified higher supervision and the Nuclear Regulatory Commission per 10CFR50.72, Non-Emergency, 4 Hour Notification.

SAFETY-SYSTEM RESPONSES:

None.

III. CAUSE OF EV1PNT A. IMM9)IATE CAUSE:

The reactor trip was due to "B" S/G Lo Lo level

(</=>>~).

INTEFQGPDIATE CAUSE:

The "B" S/G Lo Lo Level (</=174) was due to the imbalance of feedwater flow to steam flow (i.e.

feedwater flow was approximately one half of steam flow) caused by the tripping of the "A" Main Feedwater Pump.

NRC Foaa 366A (6$ 9)

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NRC FORM 366A (SS9)

LICENSEE EVENT REPORT (LER)

(LS. NUCLEAR REGULATORY COMMISSION e APPROVED OMB NO.31600104 6 XP I R ES: O/30/92 ESTIMATED BUADEN PER RESPONSE TO COMPLY WTH 'THIS INFORMATION COLLECTION REQUEST: 608) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS

.TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (315001IM). OFFICE OF MANAGEMENTAND BUDGET. WASHINGTON,DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER IS) PAGE (3)

YEAR SS< BEQUENTIAL

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&4 REYISIoN YA NUM 5 R R.E. Ginna Nuclear Power Plant'EXT o s o o o 2 4 4 9 2 003 0005 oF 0 9

///mon <<w>> /o Ioqo(>>d, ooo ~ d/I/ooo/ NRC hmI 366i3/ (17)

The tripping of the "API Main Feedwater Pump was due to the inadvertent operation of the Feedwater Pump Seal Water Differential Pressure (D/P) switch. This D/P switch senses the feedwater pump suction pressure on the low side and the feedwater pump seal injection pressure on the high side. The requirements of differential pressure to protect the. pump seals is that the high side be >/=15 pounds per square inch pressure greater than the low side.

condition is not met, the main feedwater pump will If the above trip in 5 seconds.

The inadvertent operation of the Seal Water D/P switch was due to the plugging of the high pressure side tubing to the D/P switch, followed by gradual decrease of pressure in the isolated high pressure side tubing to the D/P switch.

C ROOT CAUSE The plugging of the high pressure side tubing to the D/P switch was due to an accumulation of corrosion

-products that built up over several years in an area of no flow through this tubing (i.e. a dead leg).

The pressure decrease in the high. pressure side tubing was most probably due to a combination of plugging of the high side tubing and a slight amount of seepage of fluid from a tubing connection on the high pressure side.

NAC FomI 366A (649)

NRC FORM366A (649)

LICENSEE EVENT REPORT (LER)

UA. NUCLEAR REGULATORY COMMISSION e APPROVED OMB NO. 31500104' XP I 1 ES: O/30 f92 ESTIMATFO BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P430), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (315001(M), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR SEQUENTIAL gg.'EvrsloN

@4 NUMEER :::oS NUMBER R.E. Ginna Nuclear Power Plant o s o o o 2 4 4 9 2 0 0 3 0 0 0 6oF TEXT Il/more eaeoe Je rFFJred, rree aA(rooaor NRC %%drm 35EAB) (IT)

ANALYSXS OP EVENT This event is reportable 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 "B" S/G Lo Lo level reactor trip was an automatic actuation of the RPS.

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

There were no safety consequences or implications attributed to. the reactor trip because:

o The two reactor trip -breakers opened as required.

o All control and shutdown rods inserted as designed.

o The plant was stabilized at hot shutdown.

The Ginna updated Final Safety Analysis Report (UFSAR)

Chapter 15.2.6, "Loss Of Normal Feedwater", was reviewed and compared to the plant response for this event. The UFSAR .transient is a complete loss of Main Feedwater (MFW) at full power, with only one AFW pump available one (1) minute after the loss of MFW, and secondary steam relief (i.e. decay heat removal) through the safety valves only.

The protection against a loss of MFW includes the reactor trip on Lo Lo S/G water .features level and the start of the AFW pumps. These protection operated as designed.

The UFSAR transient resulted in a reactor trip on Lo Lo-S/G water level with S/G levels continuing to decrease and pressurizer (PZR) level and RCS average temperature (TAVG) increasing until the flow from one (1) AFW pump could remove decay heat at approximately 30 minutes into the event. All parameters then trended towards normal.

NRC Form 366A (689)

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NRC FORM 368A US. NUCLEAR REGULATORY COMMISSION (669) APPROVEO OMB NO. 31500'108 EXPIRES: 8/30/92 ESTIMATED BURDEN PEA AESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 600 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND AEPOATS MANAGEMENT BRANCH (P430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, ANO TO 1HE PAPERWORK REDUCTION PROJECT (316001(M), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) I.ER NUMBER (8) PAGE (3) 5(@ 6 8QVCNTIAL REVOION YEAR a.i NVM884 NUM684 RE E. Ginna Nuclear Power Plant o 6 o o o 2 4 4 9 2 0 0 3 0 0 0 7 OF 0 9 TEXT ///more e/>>ce/en//I/rer/ Iree //I/or>>/HRC Form 38843/ (IT)

The plant transient for this event resulted in a PZR level increase to 56.4% and a TAVG increase to 577 F. -

S/G levels remained in the narrow range throughout the transient. This was due to operator's actions to reduce power, steam dump action and the additional AFW flow.

Based on the above evaluation, the plant transient of February 29, 1992 is bounded by the UFSAR Safety Analysis assumptions.

Following the reactor trip, pressurizer level decreased to 0% but began to increase above 0% within approximately five (5) minutes. This is an expected observed transient.

The moderate RCS cooldown did not result in any core voiding. This was confirmed by the Reactor Vessel Level Indicating System (RVLIS), which always indicated a level of 100%.

A slow cooldown occurred during the post trip recovery period. This cooldown was bounded by the plant accident analysis and did not exceed the technical specification limit of 100 F per hour. Additional protection was provided by closure of the MSIVs.

Based on the above and a review of post trip data and past plant transients, it can be concluded that the plant operated as designed and that there was no unreviewed safety questions and that the public's health and safety was assured at all times.

V. CORRECTIVE ACTION

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

o The S/G levels were returned to their normal operating levels by addition of AFW, subsequent to the Reactor Trip.

N/IC Form 366A (669)

NRC FORM 3SSA US. NUCLEAR REGULATORY COMMISSION ISJ)9) APPROVEO OMB NO. 3)504)104 J EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUEST: 50J) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P430), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 20555, AND TO THE PAPERWO)IK REDUCTION PROJECT (3)504)104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME 11) DOCKET NUMBER 12) LER NUMBER 15) PAGE IS)

YEAR SEQUENTiAL PPi('REVSION NUM ER 'W% NUM ER R.K.'inna Nuclear Power Plant 0 5 0 0 0 4 9 2 0 0 3 0 0 0 80FO 9 Jpoco /4 nqaked, u44 ~ //I/ono/HRC %%d SSSA'4/ 117) o As the Intermediate Range NIS Channel N-35 ~

tracked NIS Channel N-36 for its normal operating range and returned to normal approximately ten (10) hours after the trip, no immediate action was deemed necessary. This abnormality has been observed and researched extensively in the past in cooperation with the NSS vendor, Westinghouse.

No technical basis has been identified as to why the 1E-11 idle current does not maintain indication at 1E-11 amps. Rochester Gas and Electric Corporation (RG&E) and Westinghouse

'concurred that the channel was operable and capable of performing all intended functions.

Further evaluations of, the response characteristics of NIS Channel N-35 will be performed during the,1992 Annual Refueling and Maintenance Outage.

0 The "A" Feedwater pump seal water D/P switch tubing was unplugged and flushed completely. This included both the high side and low side tubing.

In addition, similar tubing 'for the "B" Feedwater pump was also flushed completely.

0 The "A" Feedwater, pump seal water D/P switch high side and low side tubing connections were checked for evidence of seepage and serviceability, and one section of tubing and connections was replaced.

0 EWR 4960 was designed and scheduled for installation during the upcoming 1992 outage.

The installation schedule was revised, and this modification was installed as a result of this transient. The feedwater pump seal water D/P switch time delay relays for both pumps were replaced, and the time delay setting was changed from five (5) seconds to a new setting of sixty (60) seconds. This modification provides a better opportunity for the operators to prepare for and mitigate a transient resulting from an impending trip of a MFW pump. In addition, the increased time delay does not compromise reliable operation of the MFW pumps on loss of seal water D/P, and also eliminates inadvertent MFW pump trips due to pressure surges and other short-term transients affecting the seal water D/P switch.

NRC Fono 3SSA )54)9)

NRC FORM388A (669)

US. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION t ~ APPROVED OM 8 NO. 31500104 E XP I R ES: 8/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR

'EGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO SHE PAPERWORK REDUCTION PROJECT (3150d)04l. OFFICE OF MANAGEMFNTAND BUDGET,WASHINGTON, DC 20503.

FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (8) PAGE (3)

YEAR IK SEQVENTIAL NVMSER i~s~R'EYOtON S. / NVMSER R.E. Ginna Nuclear Power Plant TEXT ///Ivory tpooo /o Toqoked. voo ~ NRC hvm 35683/ (17) o 6 o o o2 4 92 003 00 p 9 oF p 9 B. ACTION TAKEN OR PLANNED TO PREVENT RECURRENCE:

As the underlying cause of the event was determined to be the plugging of the instrument tubing to the D/P switch because of a no flow condition, the following actions are being planned:

0 Flushing will be performed on selected secondary system instrument tubing during the 1992 outage.

0 Based on the results of flushing during the 1992 outage, a frequency of periodic flushing of selected instrument tubing will be established.

The areas determined, on a maintenance schedule.

if any, will then be put VI 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 underlying cause at Ginna Station could be identified.

C. SPECIAL COMMENTS~

None.

NRC Form 388A (669)

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