ML17262A902

LER 92-006-00:on 920518,steam Generator High Level Feedwater Isolations Occurred.Caused by Secondary Sys Pressure Transient.After SG Levels Stabilized,Main Feedwater Regulating Valve Returned to Automatic control.W/920617 Ltr

ML17262A902
Person / Time
Site:	Ginna
Issue date:	06/17/1992
From:	Backus W, Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-006, LER-92-6, NUDOCS 9206250008
Download: ML17262A902 (20)

Text

ACCELERATED DISTRJBUTION DEMONSTRATION SYSTEM REGULATORY INFORv&TZON DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9206250008 DOC.DATE: 92/06/17 NOTARIZED: NO DOCKET FACIL:'50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION BACKUSPW.H. Rochester Gas & Electric Corp.

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

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 92-006-00:on 920518,steam generator high level feedwater isolat'ons occurred. Caused by secondary sys pressure transient. After SG levels stabilized, main feedwater ltr.

regulating valve returned to automatic control.W/920617 DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ] ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

J SIZE:

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/PRPBll 2 2. NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1 NRB/ T/SPLB8Dl 1 1 NRR/DST/SRXB 8E 1 1 G FI 02 1 1 RES/DSIR/EIB 1 1 RGN1 FILE 01 1 1 EXTERNAL: EGGG BRYCEEJ.H 3 3 L ST LOBBY NARD 1 1 NRC PDR 1 1 NSIC MURPHYPG.A 1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1

~gg pqyc0') Kpt~

NOTE TO ALL "RIDS" RECIPIENTS:

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

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED:  : TTR 30 ENCL 30

cg ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER N.Y. 14649-0001 ROBERT C MECREDY TELESsSSOt'E Vj<e i'sesident astca COOE7 i 6 546 2700 Cinne Nucleus Production June 17, 1992 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

Subject:

LER 92-006, Feedwater Control Perturbations, Due To A Secondary Side Transient, Causes Steam Generator High Level Feedwater Isolations R.E. Ginna Nuclear Power Plant Docket No. 50-244 In accordance with 10CFR50.73, 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 92-006 is hereby submitted.

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

Very truly yours, Robert C. Mecredy xc'.S. Region I Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Ginna USNRC Senior Resident Inspector cy20617 920b2>000 0500024+

PDR ADQC PD 8

I NRC FORM366 U.S. NUCLEAR REGULATORY COMMISSION (669) APPAOVED OMB NO. 3150d104 E XP IR ES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFOAMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT ILER) COMMENTS REGARDING BURDEN ESTIMATE TO THE AECOADS AND REPORTS MANAGEMENT BRANCH (F630), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITYNAME (Il DOCKET NUMBER (2) PAGE 3 R.E. Ginna Nuclear Power Plant o 5 o o o2 441oF08 Feedwater Control Perturbations, Due To A Secondary Transient, Causes Steam Generator Hi h EVENT DATE(5) LE'R NUMBER (6) REPORT DATE (7I OTHER FACILITIES INVOLVED IS)

MONTH YEAR "q: SEQUENTIAL rrrvrsroN MONTH FACILI1'Y NAMES DOCKET NUMBER(SI OAY YEAR NUMBER &r NVMSER OAY YEAR 0 5 0 0 0 0 518 929 2 006 0 0 0 6 1 7 9 2 0 5 0 0 0 OPERATING THIS REPORT IS SUBMITTED PURSUANT 7 0 THE REQUIREMENTS OF 10 CFR (): IChrcc onr or morr of thr followinp/ (Ill MODE (9) 20A02(itl 20.405(c) 50,73 I~ ) (2) (iv) 73.71(II)

POWER 20A05(el(1)(i) 60.36(cl(1) 50.73( ~ ) (2)(vl 73.71(cl LEVEL 0 20.405(c )II((SI 50.36(c) l2) 50.73 ( ~ ) (2) (vill OTHER /Sorrily in Ahrtrrcr Oriow rnd ln Text, HRC Form 20.405(e) II)(ilil 50.73(e) l2) (Il 60.73( ~ )(2Hviii)(AI 3SSAl k4 20A05(e l(1 l(ivl 50.73(el(2)(ii) 50,73(c) (2((viSI (B)

P44 V4 20.405(e) (1)(v) 50.73(c) (2)(iii) 50.73(el(2) (x)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER Wesley H. Backus AREA CODE Technical Assistant to the Operations Manager 31 552 4-4 446 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBEO IN THIS REPOAT (13)

CAUSE SYSTEM COMPONENT MANVFAC. E PORTA 8 LE MANUFAC.

TVRER TO NPRDS SYSTEM COMPONENT TVRER kKFC~

SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED SUBMISSION DATE (15)

X YES IllyrL complrrr EXPECTED SIJSJIISSIOII DATE/ NO 07 01 ABSTRACT ILimlt to lr00 rprrrr, I r.. rpproxlmrtrly lifrrrn rlnplr.rpecr typrwrlNrn linNJ (16)

On May 18, 1992 at approximately 1339 EDST, with the Reactor at approximately 974 full power, Main Feedwater Isolations occurred on the "A" Steam Generator (S/G). These feedwater isolations were caused by overfeeding the "A" S/G while in manual control, following a secondary system pressure transient.

Immediate operator action was to manually control the Main Feedwater Regulating Valves (FRV) to reduce the steam generator water levels and stabilize the plant.

The immediate cause of the event is due to a secondary system pressure transient. (This event is NUREG-1022 (X) cause code.)

The root cause is undetermined at this time.

Corrective action taken or planned are discussed in Section V of the text.

NRC Form 366 (669)

l NRC FORM 366A (BJ)9)

LICENSEE EVENT REPORT {LER)

UA. NUCLEAR REGULATORY COMMISSION e APPROVED OMB NO. 31600104 EXPIRES: S/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUESTI 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IPJ)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT 13)600)OS), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME Il) DOCKET NUMBER 12) LER NUMBER I6) PAGE 13)

~'jj REVISION YEAR 4 5 6 SSOUSNTIAL NUMSSR N NUMBER R.E. Ginna Nuclear Power Plant o s o o o 2 449 2 006 0 0 2 or- 0 8 TEXT /l/ mom <<Moo /s mqo/t<</ ooo oI/I/ll/ooo/ HRC FomI 36643/ I)T)

PRE-EVENT PLANT CONDITIONS The plant was at approximately 974 steady state power with no major activities in progress. Instrument and Control (I&C) Department personnel were investigating the Advanced Digital Feedwater Control System (ADFCS) due to some feedwater oscillations that had occurred at'pproximately 1110 EDST, May 18, 1992. These feedwater oscillations were being caused by the "A" and "B" Steam Generator (S/G)

Main Feedwater Regulating Valves (FRVs) cycling (between 434 and 48%), at twice their normal frequency and in opposite phase from each other. This feedwater response (instability while in automatic control) was new and unexpected by the operators.

DESCRIPTION OF EVENT A. DATES AND APPROXIMATE TIMES OF MAJOR OCCHBGUBTCES:

0 May- 18, 1992, 1337 EDST: A Secondary Side Transient. occurs.

0 May 18, 1992, 1339 EDST: Event Date and Approximate Time.

o May 18, 1992, 1339 EDST: Discovery Date and Approximate Time.

0 May 18, 1992, 1342 EDST: Control Room operators restore S/G levels to normal and place Main FRVs in automatic.

NRC Form 366A IBB9)

NAC FORM 355A (LS. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150010O (BJ)9)

EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY Wl'H THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 50J) HRS. FORWARD COMMENTS REQARDINQ BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IP430), U.S. NUCLEAR AEGULATOAYCOMMISSION, WASHINGTON, DC 20555, AND TO 1)IE PAPERWOAK REDUCTION PROJECT 131500104), OFFICE OF MANAQFMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LEA NUMBER (5) PAGE (3)

YEAR .WiS SEQUENTIAL Pe NVMOER

/~4'EV SKIN

'.:.: NUMBER R.E. Ginna Nuclear Power Plant o s o o o 24 4 92 0 06 0 0 0 3 OF 0 8 TEXT ///mo>> opo>> b Ioh)II/IPC o>> oCh//Conn/NRC Form 355AB/ (12)

B. EVENT-On May 18, 1992, at approximately 1337 EDST, with the Reactor at approximately 974 full power, a secondary side condensate system decreasing pressure transient occurred. This pressure decrease traveled throughout the entire secondary side and was reflected as low feedwater header pressure and corresponding decrease in differential pressure across both FRVs. The ADFCS responded to the low feedwater header pressure by increasing the open demand signal to the main FRVs in order to maintain the proper feedwater flow. At this same time, the "C" Condensate Pump automatically started on low condensate system header pressure.

Combined with a condensate header pressure low alarm, this alerted the Control Room operators to an abnormal condition. The Control Room operators, after observing that the main FRVs open demand signals were higher than normal and increasing, and that feedwater flow to the "B" S/G was higher than appropriate, took manual control of the Main FRVs to prevent further opening of the valves.

The increased feedwater header pressure was caused by the auto start of the "C" Condensate Pump. Coupled with manual operation of the Main FRVs, the result was momentary overfeeding of the "AL( S/G after the "B" FRV was throttled. During this time period, the operator manually secured the "C" condensate pump to decrease feedwater header pressure. At 1339 EDST, May 18, 1992, Main Feedwater Isolation on high level (i.e. >/= 674 narrow range level) occurred three (3) times to the "A" S/G over a period of ten (10) seconds.

The Control Room operators manually restored S/G levels to normal, and at approximately 1342 EDST restored the "A" and "B" Main FRVs to automatic control.

NAC Form 3$ 5A (559)

NRC FORM SSSA US. NUCLEAR REGULATORY COMMISSION (Beo) APPROVEO OMB NO. 31500(OE E XP I R ES: E/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUESTI 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (PJ)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (3150010E), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITYNAME (1 I DOCKE'T NUMBER (2) LER NUMBER (5) PAGE (3)

YEAR "j SEOUENZIAL,g+b N:: NUM ER REVISION

.PA NUMSER R.E. Ginna Nuclear Power Plant 0 5 0 0 0 2 4 4 2 006 0 0 0 4oF 0 8 TEXT llfmore toete lt ee)rdted, Irte eddldonel iVRC Forrrt 3554't) (17)

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

None.

D. 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 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 hour Notification.

G. SAFETY SYSTEM RESPONSES:

r The "A" S/G Main FRV began to close automatically as a result of the feedwater isolation signal. Due to the short duration that the signal was present, the valve never fully closed.

III. CAUSE OF EVENT A. IMMEDIATE CAUSE:

The feedwater isolation signal to the "A" S/G Main FRV was due to the "A" S/G narrow range level being

>/= 674.

NRC Form 355A (BB9)

NRC FORM 355A UA. NUCLEAR REGULATORY COMMISSION (549) APPROVEO OMB NO. 3'I500I04 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFOAMATION COLLECTION AEGUESTI 500 HAS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F430). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3(504)104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (I l DOCKET NUMBER (2) LER NUMBER (Sl PAGE (3)

YEAR  :'SQ SEOUENTIAL ~~+ REVISION C%> NUMBER '.%'> NUMSER R.E. Ginna Nuclear Power Plant 0 5 0 0 0 2 4 4 9 0 06 00 05 oF 0 8 TEXT ///more SPeoe /I nN)lr/red, oee edd/abns/HRC form 3654'4/ (12)

B. INTERMEDIATE CAUSES:

The "A" S/G narrow range level was >/= 674 due to increased flow to the "A" S/G, caused by the perturbations in main feedwater header pressure during manual operation. This situation resulted in inadvertent overfeeding of the "A" S/G. It should also be noted here that just prior to this event, feedwater automatic control had been experiencing instability with flow oscillations, and the operators were sensitive to abnormal indications with ADFCS in automatic control of the FRVs.

C ROOT CAUSE:

The root cause is believed to be the control instability problems being experienced with the FRVs.

These instabilities caused operators to place ADFCS in manual following observation of FRV response.

After extensive troubleshooting, the" specific root cause has not been determined. Various contributing causes have been identified, and intermediate and long-term corrective actions (see Section V.B) address the need to determine the root cause.

ANALYSIS OP This event is reportable in accordance with 10CFR50.73, Licensee Event Report system, item (a)(2)(iv), which requires reporting 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 feedwater isolation of the "A" S/G was an automatic actuation of an ESF system.

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

NAC Form 35SA (54)9)

NRC FORM 355A US. NUCLEAR REGULATORY COMMISSION (589) APPROVED OMB NO. 31500(04 EXP IR ES: 5/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT {LER) INFORMATION COLLECTION REOUESTI 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION ANO REPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3(SOOIOE), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITYNAME ll) DOCKET NUMBER (2) I.ER NUMBER (5) PAGE (3)

YEAR @3 sEQUENTIAL ':.:5

)g~o REVISION NUMBER NUMBER R.E. Ginna Nuclear Power Plant 0 s 0 0 o 24 492 0 0 600 06 oF0 8 TEXT //IIBo>> EOooo lr>>EUPR/, II>> BI/I/IT/ono/NRC %%drm 35BA3/ (17)

There were no operational or safety consequences or implications attributed to the feedwater isolations because:

o The feedwater isolation signals occurred at the required S/G level.

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

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

Based on the above, health it can be concluded that the and safety was assured at all times.

public's V. CORRECTIVE ACTION A. ACTION TAKEN TO RETURN AFFECTED SYSTEMS TO PRE-EVICT NORMAL STATUS:

o After S/G levels were stabilized, the Main FRVs were returned to automatic control.

ACTION TAKEN OR PLANNED TO PR1W1PNT RECURRENCE:

After reviewing the results of troubleshooting and discussions with Nestinghouse, the following is an outline of the corrective actions being taken or planned in response to the feedwater control instability problems (and ADFCS transient response for secondary system transients):

NRC Fono 355A (589)

NRC FORM3SSA U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 31500104 (SS9)

EXPIRES: S/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUESTI 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPOATS MANAGEMENT BRANCH (F430) U 9 NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1'HE PAPERWORK REDUCTION PROJECT (3(500(04). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME Ii) DOCKET NUMBER (2l LER NUMBER (5) PAGE (3)

YEAR SEOVENTIAL REVISION NVMSSA NVMSSA R.E. Ginna Nuclear Power Plant 0 5 0 0 0 2 4 4 9 0 0 6 00 07 >>0 S TEXT ///moro sPom /s /II/rkrrd, Ir>> odd/done/HRC Form 35649/ (12) 0 Short Term Response:

a) IGC technicians loosened and lubricated the actuator to stem connection block for the "B" FRV.

b) Operations was directed to place the Main FRV bypass valves in full open position and in automatic control, to reduce the feedwater control instability problem.

c) The packing on both FRVs was relaxed to minimum torque, to reduce any potential friction of the valve stem.

0 Intermediate Term Response:

a) Westinghouse has evaluated ADFCS "tuning constants" and has supplied new ADFCS gain values, in an attempt to reduce the ADFCS impact on feedwater control instability problems.

b) New ADFCS gain values have been installed in the ADFCS software. Instability problems have not been eliminated as a result of these new gain values.

c) Diagnostic testing of FRV performance was completed. Testing focused on valve response to changes in current-to-pneumatic (I/P) transducers, valve positioners, and solenoids and associated instrument air tubing.

NRC Form 3OIA (SB9)

NRC FORM 388A UA. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 31500104 (64)9)

E XP I 8 ES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 503) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P4)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, 00 20503.

FACILITY NAME (11 DOCKET NUMBER (21 LER NUMBER (8) PAGE (3)

YEAR w~O SEQUENTIAL Ir~N REVISION N'.6 NUMBER 'AS N U M 8 E R R.E. Ginna Nuclear Power Plant 0 5 0 0 0 2 4 4 9 0 0 6 0 0 08 oFO 8 TEXT ///more epeoe /4 mr/vtred, vee eddldooe/NRC Fomr 3//543/ (17) o Long Term Response:

a) Valve characteristic data (with and without valve shaft packing in place) will be reviewed.

b) The vendors of the Main FRV actuator and positioner will be consulted requesting their recommendations for valve packing and positioner operation, based on a more detailed review of the diagnostic testing.

c) The calibration procedures for the Main FRVs will be revised to include valve response characteristics during scheduled calibrations.

d) EWR 4773 will apply the information obtained from other actions, and is intended to enhance feedwater control and response to plant transients.

improvements will be obtained in both It is envisioned that electronic and pneumatic performance, as a result of these enhancements.

e) After additional actions are ,completed during the 1993 outage, and after a root cause determination is completed, a supplemental report to this LER will be submitted.

ADDITIONAL INFORMATION A. FAILED COMPONENTS:

None.

B. PREVIOUS LERs ON SIMILAR EVENTS:

A similar LER event historical search was conducted with the following results: LER 91-009 was a similar event with a different root cause.

C SPECIAL COMMZXZS:

None.

NRC Form 356A (689)