Text

LER 93-004-00:on 930511,uplanned Actuation of ESF Including RPS Actuation Occurred When High Rv Pressure Signal Initiated Anticipated Trip W/O Scram Signal.Reactor Pressure Stabilized & Personnel Involved counselled.W/930610 Ltr
	ML18036B323
Person / Time
Site:	Browns Ferry
Issue date:	06/10/1993
From:	Austin S, Zeringue O; TENNESSEE VALLEY AUTHORITY
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	LER-93-004-02, LER-93-4-2, NUDOCS 9306170030
	Download: ML18036B323 (20)
	v • d • e

ACCELERATED DOCUMENT DISTIGBUTION SYSTEM REGULATIVE INFORMATION DISTRIBUTIOIISTEM (RIDE)

ACCESSION NBR:9306170030 DOC.DATE: 93/06/10 NOTARIZED: NO DOCKET ¹ FACIL;50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 AUTH. NAME AUTHOR AFFILIATION AUSTIN,S.W. Tennessee Valley Authority ZERINGUE,O.J. Tennessee Valley Authority RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 93-004-00:on 930511,uplanned actuation of ESF including RPS actuation occurred when high RV pressure signal D initiated anticipated trip w/o scram signal. Reactor pressure involved ltr.

stabilized & personnel counselled.W/930610 DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR i ENCL ( SI2E:

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

RECIPIENT COPIES RECIPIENT COPIES D ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-4 1 1 PD2-4-PD 1 1 ROSS,T. 1 1 D INTERNAL: ACNW 2 2 ACRS 2 2 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 NRR/DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRIL/RPEB 1 1 NRR/DRPW/OEAB 1 1 NRR/DRSS/PRPB 2 2 NRR DSS SPLB 1 1 NRR/DSSA/SRXB 1 1 ~R G FIL 02 1 1 RES/DSIR/EIB 1 1 RGN2 3.'ILE 01 1 1 EXTERNAL'G&G BRYCEF J ~ H 2 2 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

NOTE TO ALL"RIDS" RECIPIENTS:

CONTROL DESK, PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT YOUR NAME FROM DISTRIBUTION ROOM Pl-37 (EXT. 504-2065) TO ELIMINATE LISTS FOR DOCUMENTS YOU DON'T NEEDI FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 32 ENCL 32

4l Tennessee Valley Authority. Post Of(ice Box 2000. Decatur. Afabarna 35609.2000 0; J. "Ike" Zerf'ngue Vice President, Browns ferry Nucfear Pfant

'UN $ 0 1993 U.S. Nuclear Regulatory, Commission ATTN: Document Control Desk Washington, D.C. 20555

Dear Sir:

TVA BROWNS FERRY NUCLEAR PL'ANT (BFN) UNITS 1, 2, AND 3 DOCKET NOS. 50-259, 260, AND 296 FACILITY OPERATING LICENSE DPR-33, 52, AND 68 LICENSEE EVENT REPORT 50-260/93004 The enclosed report provides details concerning a high reactor pressure condition that resulted in an Anticipated Trip Without Scram Signal that tripped the Reactor Recirculation Pump and initiated an Alternate Rod Insertion (ARI) signal. The ARI signal resulted in a depressurization of the scram pilot air header and. subsequent scram condition due to low. scram air header pressure.

This report is submitted in accordance with 10 CFR 50.73(a)(2)(iv).

Sincerely,

0. J. Zeringue En'closure cc: See page 2 5ojg.f

'tf30bi70030 'tf30b$ 0 PDR ADQCK 050002b0 8 PDR

f 2

U.S. Nuclear Regulatory Commission JUh i 0 1933 cc (Enclosure):

INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 Paul Krippner American Nuclear Insurers Town Center, Suite 300S 29 South Main Street West Hartford, Connecticut 06107 NRC Resident Inspector Browns Ferry Nuclear Plant Route 12, P.O. Box 637 Athens, Alabama 35609-2000 Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 2900 Atlanta, Georgia 30323'hierry M. Ross U.S. Nuclear Regulatory Commission One White Flint,. North 11555 Rockville Pike Rockville, Maryland 20852

r NRC (6-09)

Form 366 ll LEAR REGULATORY CONIISSION LICENSEE EVENT REPORT (LER) t Approved OMB No. 3'l50-0104 Expires 4/30/92 FACILITY NAME ( 1) (DOCKET NUMBER (2) I P r F FN TITLE (4) High Reactor Pressure Condition Resulted In Anticipated. Trip Without Scram Signal That Tripped The r ' i i '1.

i V V (SEQUENTIAL (REVISION( ( ( ( FACILITY NAMES IDOCKET NUMBER(S)

T Y I I I I I I' ( I I I 4 06 10 93 OPERATING I ITHIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR g:

MODE I I f h 1 w'

N I20.402(b) (20.405(c) (~(50.73(a)(2)(iv) I I73.71(b)

POWER (20.405{a){ l)(l) (50.36(c)(l) I (50.73(a)(2)(v) I (73.>> (c)

LEVEL I (20.405(a)( l)(ii) 150.36(c)(2) I (50 73(a)(2)(vii) (OTHER (Specify in 1 (20.405(a)(l)(iii) (50.73(a)(2)(i)(B)f (50.73(a)(2)('viii)(A) ( Abstract below and in (20.405<a)( 1)<iv) (50.73(a)(2)(ii) ( (50.73(a)(2)(viii)(B) Text, NRC Form 366A)

.4 V N N NAME T N N I AREA CODE I' W. 'n in P N P N I I IREPORTABLEI I I (REPORTABLE(

A Y T MAN 0 I I I I I I I I I I I P P T 4 I EXPECTED I I I SUBMISSION I f m T ABSTRACT (Limit to 1400 spaces, i.e., approximately fifteen single-space typewritten lines) (16)

On May ll, 1993, TVA was performing the Surveillance Instruction (SI) "ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping," and "Functional Test of Instrument Line Flow Check Valve SI." At 2324 hours0.0269 days 0.646 hours 0.00384 weeks 8.84282e-4 months on this date, an unplanned actuation of Engineered Safeguard Features including Reactor Protection System actuation occurred when a high reactor vessel pressure signal initiated an Anticipated Trip Without Scram signal. This event tripped the Reactor Recirculation system pumps and initiated an Alternate Rod Insertion (ARI) signal.

The ARI signal'esulted in a depressurization of the scram pilot air header and subsequent scram condition due to low scram air header pressure.

The root cause of this event is lack of attention to detail in that the Unit Operator did not adequately track the progress of the Instrument Line Flow Check Valve SI.

Corrective actions include counselling personnel involved in the event, and issuance of a briefing on the event. The method for identifying instruments in the Instrument Line Flow Check Valve SI will be evaluated. Additionally, the Functional Test of Instrument Line Flow Check Valves will be designated as a complex infrequently performed test.

NRC Form 366(6-89)

~ i NRC Form 366A (6-89)

U.. UCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER) t Approved OHB No. 3150-0104 Expires 4/30/92 TEXT CONTINUATION FACILITY NAHE (1) iOOCKET NUHBER (2)

I ( I ISE()UENTIAL I )REVISION) )

Browns Ferry Uni t 2 I I I I I TEXT (If more space is required, use additional NRC Form 366A's) (17)

I PLANT CONDITIONS Unit 2 was in the refuel mode, with a moderator temperature of approximately 190 degrees. TVA was performing the American Society of Mechanical Engineering (ASME)Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping Surveillance Instruction (SI). The ASME Section XI test pressure was being maintained using a Control Rod Drive (CRD) [AA]

pump and the Reactor Water Cleanup (RWCU) [CE] reject flow control valve

[FCV]. Also, being performed in parallel with the ASME leak check SI, was the SI for Instrument Line Flow Check Valve (Narotta Excess Flow Check Valve)

Operability. The Reactor Node Switch was in the shutdown position. Units 1 and 3 were defueled.

II. DESCRIPTION OF EVENT

.A. gv~t:

On May 11, 1993, TVA was performing the SI "ASME Section XI System Leakage Test of the Reactor Pressure Vessel and Associated Piping," and "Functional Test of Instrument Line Flow Check Valve" SI. At 2324 hours0.0269 days 0.646 hours 0.00384 weeks 8.84282e-4 months on this date, an unplanned actuation of Engineered Safeguard Features (ESF) [JE] including Reactor Protection System (RPS) [JC] actuation occurred when a high reactor vessel pressure signal initiated an Anticipated Trip Without Scram (ATWS) signal. This event tripped the Reactor Recirculation system [AD] pumps and initiated an Alternate Rod Insertion (ARI) signal. The ARI signal resulted in a depressurization of the scram pilot air header and subsequent scram condition due to low scram air header pressure. Further details of this event are discussed below.

On May 11, 1993, at 1540 hours0.0178 days 0.428 hours 0.00255 weeks 5.8597e-4 months , the Unit 2 Assistant Shift Operations Supervisor,(ASOS) approved performance of the Functional Line Flow Check Valve SI. Following a review of the affected instruments by the licensed unit operator (UO) and Instrument Mechanics (IMs), testing of Group A components was initiated. Control room personnel were provided information concerning instruments being removed from service.

At 1900, the IMs had completed isolation of Group A transmitters, and by 2017, testing of the Group A instruments was completed.

NRC Form 366(6-89)

0 NRC Form 366A U.S. NUCLEAR REGULATORY CONHISSION Approved OHB No. 3)50-0104

{6-B9) Expires 4/30/92 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAHE {1) (DOCKET NUHBER {2)

(SEQUENTIAL ( (REVISION(

Browns Ferry Unit 2 ( ( I I I TEXT {If more space is required, use additional NRC 'Form 366A's) {17)

At 2030, with Group A testing completed, IMs were establishing parameters to test the Group B instrument line check valves. After reviewing the instruments that were to be removed from service, the UO determined that the pressure indicator [PI] utilized to monitor and control reactor pressure during the ASME Section XI test (i.e., 2-PI-3-207) would be out of service and that an alternate method of monitoring pressure would be required. The IMs were not involved in the discussions surrounding this decision; therefore,'hey were not cognizant of the importance of 2-PI-3-207.

The IMs established parameters to test the instrument line flow check valves in the Group B instrument lines between the hours of 2030 and 2150. At approximately 2150, the UO was informed that the reactor low water level transmitter was inoperable requiring a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Limiting Condition of Operation (LCO) action statement to be entered. The UO questioned the IM in the control room whether the entire loop or just the low water level transmitter had been removed from service. The instrument mechanic showed the UO a copy of the step that listed the low watex level transmitter. Based on this information, the UO assumed that he would be informed of each instrument removed from service. The UO did not realize that 2-PI-3-207 had also been isolated.

At 2230 hours0.0258 days 0.619 hours 0.00369 weeks 8.48515e-4 months , the UO gave the IMs permission to commence testing the Group B instrument line check valves.

At 2240 hours0.0259 days 0.622 hours 0.0037 weeks 8.5232e-4 months , the portion of the ASME Section XI test being performed at 1014-1034 psig was completed, and as required by the test, the UO lowered the reactor pressure to 980-1000 psig. By 2306 hours0.0267 days 0.641 hours 0.00381 weeks 8.77433e-4 months , the'O had indication that thi's pressure was established.

At approximately 2310 hours0.0267 days 0.642 hours 0.00382 weeks 8.78955e-4 months , the IMs were verifying the operability of the Group Bl instrument line check valve. Group Bl instrument line also supplied process pressure to 2-PI-3-207. At 2320 hours0.0269 days 0.644 hours 0.00384 weeks 8.8276e-4 months , a test valve [TV]

on the instrument line was opened. Upon doing so, the check valve seated, sealing the instrument line from reactor pressure. This caused the pressure indicator to bleed through its leaking isolation valve, which gave the UO indications that reactor pressure was decreasing. At 2321 hours0.0269 days 0.645 hours 0.00384 weeks 8.831405e-4 months , the UO attempted to maintain pressure in the band prescribed by the ASME Section XI test by lowering RWCU reject flow. The UO and a licensed Shift Operations Supervisor (SOS) realized that the RWCU reject flow control valve movement was excessive so the SOS then directed the UO to evaluate reactor pressure utilizing other instruments available in the control room. By this time, 2-PI-3-207 had attained a low pressure reading of 978 psig and because the instrument mechanics had reisolated their test valve, the pressure had increased to 983 psig.

NRC Form 366{6-89)

4i NRC Form 366A (6-89)

U.. NUCLEAR REGULATORY COHHISSION LICENSEE EVENT REPORT (LER) t Approved OHB No. 3150-0104 Expires 4/30/92 TEXT CONTINUATION FACILITY NAHE (1) [DOCKET NUHBER (2)

I iSEQUENTIAL i iREVISIONJ Browns Ferry Unit 2 I Y I I I I 4 F TEXT (If more space is required, use additional NRC Form.366A's) (17)

At 2324 hours0.0269 days 0.646 hours 0.00384 weeks 8.84282e-4 months , the reactor pressure reached 1118 psig resulting in- an ATWS scram signal that tripped the Reactor Recirculation system pumps and initiated an ARI signal. At 1005 hours0.0116 days 0.279 hours 0.00166 weeks 3.824025e-4 months on May 12, 1993, the ESF actuations were reset and the ASME Section XI test conditions were being reestablished.

This event is reportable pursuant to 10 CFR 50.73(a)(2)(iv), due to an unplanned actuation of an Engineered Safety Feature, including the Reactor Protection System.

B. n t t t t t t th None.

C. t May 11, 1993, at 2324 CDT Reactor pressure of 1118 psig is attained, resulting in an ATWS scram signal.

May 12, 1993, at 0319 CDT 1VA makes a 4-hour nonemergency notification to NRC in accordance with 10 CFR 50.72(b)(2)(ii).

D. th None.

V The ATWS scram and trip of the reactor recirculation pumps was identified by the UO when he received main control room alarms indicating the trip had occurred.

Just prior to the event, the UO'ttempted to maintain the reactor pressure in the band prescribed by the ASME Section XI test by lowering RWCU reject flow. Once the event occurred, actions were taken to stabilize the reactor at 550 psig. These actions included tripping of the CRD pump and securing the RWCU reject flow.

G. t t None.

NRC Form 366(6-89)

Q>

NRC Form 366A (6-89)

U~ . UCLEAR REGULATORY COMMISSION LICENSEE EVENT, REPORT (LER) t Approved OMB No. 3150-0104 Expires 4/30/92 TEXT CONTINUATION FACILITY NAME (1). iDOCKET NUMBER (2)

I [SE()UENTIAL / ]REVISION t Browns Ferry Uni t 2 I I I I I TEXT (If more space is required, use additional NRC Form 366A's) (17)

III. CAUSE OF THE EVENT A.

The immediate cause of the event. was failure to identify specific instrumentation to be removed from service during performance of the Instrument Line Flow Check Valve SI.

S. ~R~tgggg:

The root cause of this event is lack of attention to detail in that the UO did not adequately track the progress of the Instrument Line Flow Check Valve SI. The UO assumed incorrectly that he would be notified prior to each instrument removed'rom service.

The operating crew was aware of the instruments to be taken out of service and had discussed the Instrument Line Flow Check Valve SI among themselves. The ASOSs and UO did not discuss with the IMs the importance of notifying Control Room operators prior to instrument isolation. When it was recognized that 2-PI-3-207 would be affected, operators did not discuss a plan of action with the Senior Reactor Operator or the ASME Section ZI test director.

C. t ta None.

IV. ANALYSIS OF THE EVENT All plant systems and components performed as designed for the actual conditions encountered during the event. The recirculation pumps trip for anticipated trip without scram setpoint is 1118 psig. Post trip data has verified that this trip occurred as designed.

The highest pressure encountered during the event measured at elevation 631 feet,was 1120 psig. Main steam relief valves are centered around elevation 590 feet at which the relief valves would have experienced a pressure of 1138 psig. The lowest pressure relief valve is calibrated to lift at 1105 ~ 1 percent psig and 525 degrees F. During the event, the relief valve was at torus ambient temperature (i.e., approximately 100 degrees F). Cold calibration of 50 psig higher than the operating setpoint is required to achieve the desired setpoint at operating temperature and pressure.

Therefore, during the Section ZI test the lowest set pressure relief valve was expected to lift at 1155 psig instead of 1105 psig and the relief valves performed as expected.

NRC Form 366(6-89)

NRC Form 366A

,(6-89)

U.. UCLEAR REGULATORY'OHHISSION LICENSEE EVENT REPORT (LER) t Approved OHB No. 3150-0104 Expires 4/30/92 TEXT CONTINUATION FACILITY NAHE (1) (DOCKET NUHBER (2)

I ( ( (SEQUENTIAL '( (REVISION(

Browns Ferry Uni t 2 I H I I I I F

TEXT (If more space is required, use additional NRC Form 366A's) (17)

All safety related components operated as expected during the event.

Therefore, the safety of the plant, its personnel, and the public was not compromised.

V. CORRECTIVE ACTIONS tv At The immediate corrective actions included stabilizing, reactor pressure at 550 psig, and restoration of the systems affect'ed.

B. tv t t The personnel involved in this event were counselled on the proper degree of attention to be devoted to monitoring SIs in progress.

A briefing on the lessons learned from the event will be prepared and issued to Operations personnel.

3. The method for identifying instruments in the Instrument Line Flow Check Valve SI will be evaluated.

4. The Functional Test of Instrument Flow Check Valves will be desi'gnated as a Complex Infrequently Performed Test.

VI. ADDITIONAL INFORMATION A.

None.

B. v Ev t None.

NRC Form 366(6-89)

4~

'I

NRC (6-89)

Form 366A U. CLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER) t Approved OMB No. 3150-0104 Expires 4/30/92 TEXT CONTINUATION FACILITY NAME (1) IDOCKET NUMBER (2)

I I SEQUENTIAL f [REVISION(

Browns Ferry Unit 2 I I I I I TEXT (If more space is required, use additional NRC Form 366A's) (17)

VII. CONMIXNENTS

l. A briefing on the lessons learned from the event will be prepared and issued to Operations personnel. This will be accomplished by,July 30, 1993.

2. The method for identifying instruments in the Instrument Line Flow Check Valve SI will be evaluated. This will be accomplished by July 30, 1993.

3. The Functional Test of Instrument Flow Check Valves will be designated as a Complex Infrequently Performed Test. This will be completed by September 1, 1993

'nergy Industry Identification System (EIIS) system and component codes are identified in the text .with brackets (e.g., [ZZ]).

NRC Form 366(6-89)

~I .Qi if I'

ML18036B323

Text

SUBJECT:

Dear Sir:

Navigation menu

Search