ML18153C716

LER 91-007-00:on 910802,ESF Actuated & Safety Injection & Reactor Trip Occurred as Result of High Steam Flow Signal Coincident W/Low Steam Line Pressure Signal.Caused by Loss of Voltage to Protection instrumentation.W/910903 Ltr

ML18153C716
Person / Time
Site:	Surry
Issue date:	09/03/1991
From:	Kansler M VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
91-489, LER-91-007-02, LER-91-7-2, NUDOCS 9109060144
Download: ML18153C716 (8)

Text

. ACCELERATED DlrUTION DEMONS1\A.TION SYSTEM i* REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9109060144 DOC.DATE: 91/09/03 NOTARIZED: NO DOCKET#

FACIL: 50-.281 Surry Power Station, Unit 2; Virginia- Electric & Powe 05000281 AUTH.NAME AUTHOR AFFILIATION KANSLER,M.R. Virginia Power (Virginia Electric & Power Co.)

RECIP.N~ME RECIPIENT AFFILIATION ,

R

SUBJECT:

LER 91-007-00:on 910802,ESFactuated & safety injection & I

.reactor" trip occurred as result of high steam flow signal

,coincident w/low steam line pressure signal.Caused by loss D of voltage to protection instrumentation.W/910903 ltr. ~

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR _l ENCLr_._ SIZE: _ _'_/_ _ s TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

I NOTES:lcy N11SS/IMSB/PM. 05000281 A

RECIPIENT COPIES RECIPIENT COPIES D ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR. ENCL PD2-2 LA 1 1 PD2-2 PD 1 1 D BUCKLEY,B 1 1 INTERNAL: ACNW 2 2 ACRS 2 2 s

AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP *2 2 *NRR/DET/ECMB 9H 1 1 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFBlO 1 1 NRR/DLPQ/LPEBlO 1 1 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPBll 2 2 NRR/DST/SELB SD 1 1 NRR/DST/SICB8H3 1 1

~:~ii~;~~

1 1 NRR/DST/SRXB SE 1

1 1 1 1 R,ES/DSIR/EIB 1 1 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 R NOTES: 1 1 I D

s I

A D

D NOTE TO ALL "RIDS" RECIPIENTS:

s PLEASE HELP US TO REDUCE*WASTE! CONTACT THE DOCUMENT CONTROL DESK_,

ROOM Pl-37 (EXT. 20079) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 34 ENCL 34

Virginia Electric and Power Company Surry Power Station P. 0. Box 315 Surry, Virginia 23883 September_ 3, 1991 U. S. Nuclear Regulatory Commission Serial No~: 91-489 Document Control Desk Docket No.: 50-281 Washington, D. C. 20555 License No.: DPR-37 Gentlemen:

• -Pijrsuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report for Unit 2.

• REPORTNUMBER 91-007-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by the Corporate Management Safety Review Committee.

Very truly yours,

~n--=-rsler~

Station 1vlanager Enclosure cc: Regional Administrator Suite 2900 101 Marietta Street, NW Athmta, Georgia 30323 9109060144 910903 PDR ADOCK 05000281 S PDR

\

NRC FORM 366 (6-891 e U.S. NUCLEAR REGULATORY COMMISSION e APPROVED 0MB NO. 3150-0104 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 (P-5301, U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO l' THE PAPERWORK REDUCTION 'PROJECT (3150-01041, OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME 111 IDOCKET NUMBER 121 .. . I PAGE 13)

Surry Power Station, Unit 2 o 15 Io Io Io 12 18 Id 1 loFb 16'

• 4 TITLE 1 1safety Injection/Reactor Trip From Vital Bus IVA Electrical Fault and Failure of Steam Generator Pressure Transmitter EVENT DATE (51 LER NUMBER (61 REPORT DATE 171

• OTHER FACILITIES INVOLVED (Bl MONTH DAY YEAR YEAR  :\% SE~~t~i~AL t? ~t~~~~ MONTH DAY YEAR FACILITY NAMES DOCKET NUMBERISl OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE R~QUlREMENTS OF 10 CFR §: {CMck one or more of the following/ (111-.

MODE (Bl N 20.402(bl 20.405(cl ,_ X 60,73(all2llivl 73.71(bl 20.405(*111 )Ill 50.311(cll1l 60.731all2lM 73.71(cl

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- 0TH ER {S~cify in Ab1tr*ct b*low *nd In* Toxt, NRC Form 366A/ .

• 20.405(all1 llivl 20.405(al111M

- 60,73(all2llill 50.731all2lliiil ---

LICENSEE CONTACT FOR THIS LER (121 60.731al(21(vilil(Bl 60.73(all2llxl NAME TELEPHONE NUMBER AREA CODE Mr.* M. R. Kansler_, Station Manager 81014 31c; 17 1-11111RIL.

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 MANUFAC . MANUFAC*

CAUSE SYSTEM COMPONENT , TURER TURER X SI B I I P 1T- R13 16 19 y I I I I *I I I I I .I I I I I I I I I *1 I I SUPPLEMENTAL REPORT EXPECTED 1141 MONTH DAY YEAR EXPECTED SUBMISSION ITT YES /If yes. compf*te EXPECTED SUBMISSION DATE/

DATE 1151 111 115 911 ABSTRACT (Limit to 1400 spaces. i.tJ., approximately fiftetm sing/e-spactJ ryptJwritttJn lintJs) (16)

At 1657 hours0.0192 days <br />0.46 hours <br />0.00274 weeks <br />6.304885e-4 months <br /> on August 2, 1991, with Unit 1 at 100% power and Unit 2 at 92% power, an Engineered Safeguards Feature actuated in Unit 2. A safety injection/reactor trip occurred as a result of a high steam flow signal coincident with a low steam line press*ure signal. These conditions did not actually exist, but were generated by a combination of an erratic steam generator pressure channel and an electrical fault in a vital bus distribution panel. Troubleshooting was in progress on both problems, but the actuation occurred before corrective action could be completed. ESF functions performed as expected except fo.r the failure of Emergency Diesel Generator #3 to achieve rated speed and the failure of the Containment Sump Pump Discharge Trip Valve (02-DA-TV-200A) to fully shut. A Notification of Unusual Event (NOUE) was declared at 1710 hours0.0198 days <br />0.475 hours <br />0.00283 weeks <br />6.50655e-4 months <br />, and appropriate reports were made. On-shift operati.ng personnel promptly placed the plant in a stable condition. The NOUE was terminated at 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br />. A Root Cause Evaluation of the Emergency Diesel Generator problem is being conducted. The results of this evaluation will be reported in Licensee Event Report S.1-91-017 .. This event is reportable pursuant to 10CFR50.73(a)(2)(iv).

NRC Form 366 16-891

NRC FORM,366A U.S. NUCLEAR REGULATORY COMMISSION (6-89)

• APPROVED 0MB NO. :ii50-0104 EXPIRES: 4/30/92 .

• ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EV. REPORT (LER). INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO l'HE PAPERWORK REDUCTION PROJECT (3150-0104). OFFICE*

OF MANAGEMENT ANO BUDGET, WASHINGTON,_DC 20503.

FACILITY NAME (11 DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

Surry Power Station, Unit 2 o 1s Io Io Io 12 I s 11 .911 - o Io 11 - o I o o I 2 oF o 16 TEXT (ff mot9 -,,.C. la f'MIU.'rwl, UIIII MlditiofltJ! NRC Fann 3166A'a} (17)

LO DESCRIPTION OF THE EVENT At 1657 hours0.0192 days <br />0.46 hours <br />0.00274 weeks <br />6.304885e-4 months <br /> on August 2, 1991, with Unit 1 at 100% power and Unit 2 at 92% power, an Engineered Safety Feature (ESF) (EIIS-JE) actuated in Unit 2. A safety. injection/reactor trip occurred as a result of a high steam flow signal coincident with a low steam line pressure signal. These conditions did not actually exist, but were caused by an erratic steam generator pressure transmitter (02-MS-PT-2474) (EIIS-SB-PT) and an electrical fault in Vital Bus Distribution Panel IVA (EIIS-UJX-PL). Control and protection instrumentation .for steam flow, feed* flow, first stage turbine impulse pressure, steam generator (line) pressure, T-average control, pressurizer pressure control, steam dump control, and feed regulating valve "C" control are powered from the affected panel.

Problems were first noted at 1441 hours0.0167 days <br />0.4 hours <br />0.00238 weeks <br />5.483005e-4 months <br /> when instrumentation powered by panel IVA began to experience perturbations.

• Concurrent with the problems with the vital bus IVA instrumentation, steam generator pressure channel II failed at 1506 hours0.0174 days <br />0.418 hours <br />0.00249 weeks <br />5.73033e-4 months <br />. Corrective actions were initiated for the failed steam pressure transmitter, and a -

spare transmitter had been located, but work on replacement had not begun because of the continued voltage fluctuations on vital bus IVA.

The vital bus voltage fluctuations were under investigation at the distribution panel when the safety injection/reactor trip occurred. ESF functions performed as expected except for the failure of Emergency Diesel Generator #3 (EOG) (EIIS-EK-DG) to achieve rated speed (835 RPM vice a minimum acceptable speed of 870 RPM) and the failure of the Containment Sump Discharge Trip Valve (02-DA-TV-200A) (EIIS-BD-ISV) to fully shut. A Notification of Unusual Event (NOUE) was -declared at 171 O in accordance with the Emergency Action Level Table, Tab M, "Miscellaneous Abnormal Events", and appropriate reports were made.

Subsequent to the initial ESF actuation, several more actuations _

occurred as a result of continued spiking of the channel IV instrumentation coincident with an actual low value of Reactor Coolant System (RCS) average temperature*(< 543 degrees F). On-shift operating personnel acted promptly to place the plant in a stable condition. The NOUE was terminated at 1840 hours0.0213 days <br />0.511 hours <br />0.00304 weeks <br />7.0012e-4 months <br />.

NRC Form 366A (6-89)

NRC FORM,366A

• U.S. NUCLEAR REGULATORY COMMISSION (6-89)

• APPROVED 0MB NO. 3150-0104 . * .

• EXPIRES: 4/30/92' .

LICENSEE EVE.REPORT (LER) E-TEO BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD

• OOMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION

• AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1'HE PAPERWORK REDUCTION PROJECT (3160-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (21 . LER NUMBER (6) PAGE (31 Surry Power Station, Unit 2.

TEXT (If mor. .,,.co m,-qul.TKI, UStl oddltianal NRC Form .BA :t) (171 2.0 SIG*NIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS The loss of a portion of one of the vital buses is of minor safety concern because of the redundancy of the power supplies for the different channels of control and protection instrumentation. Reactor protective schemes have redundant channels and the power sources are provided from redundant vital bus caqinets. Because of the fail-safe circuitry of the reactor protection instrumentation, a* power source failure to an instrument channel results in a trip signal from the affected channel.

Multiple power supplies are provided to prevent a single power supply failure from initiating a false trip. In this case, a problem occurred on a component powered from one vital bus concurrent with instrument perturbations caused by voltage fluctuations on a second vital bus satisfying the logic for safety injection even though actual plant conditions requiring a safety injection did not exist.* The system is designed so that if such a combination of events occurs, the Reactor

. Protective System (RPS) (EIIS-JC) acts automatically to place the plant in a safe condition.* If, for any reason, the RPS fails to function as designed, NRG-licensed operating personnel are trained to perform the necessary actions in accordance with the station's emergency operating procedures. During this event, ESF functions performed as expected except for the failure of EOG #3 to achieve rated speed and the failure of the Containment Sump Pump Discharge Trip Valve (02-DA-TV-200A) to fully shut.

Surry's emergency electric power system is designed to provide reliable power to engineered safety functions and other essential loads in the*

event of loss of off-site power (LOOP). The *system consists of three 100% capacity diesel generator sets for the two Units. One generator is used exclusively for Unit 1 (EOG #1 ), the second for Unit 2 (EOG #2), and the third (EOG #3) functions as a backup for either Unit. Each Unit has two emergency buses normally fed from independent off-site power sources, with the EDGs functioning as on-site backup power sources.

During this event, the off-site power sources remained available and fed the emergency busses. Although EDG #3 failed to achieve rated speed and would not have automatically loaded onto its bus, EOG #2 functioned as designed and could have carried its emergency bus had

• the need arisen. Also, existing training and procedures guide the operator to take manual control, raise spe.ed as necessary, and place EOG #3 on its bus.

NRC Form 366A (6-89)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION

(~-89) '. APPROVED 0MB NO. 3150-0104

• EXPIRES: 4/30/92 LICENSEE EVE. REPORT (LER). E-ATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING-BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO l'HE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT ANO BUDGET, WASHINGTON, DC 20503.

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

VEAR  :::::::::: SEQUENTIAL *:*>:*:*:* REVISION Surry Power Station, Unit 2  :::::::::: NUMBER  :-:*:*:*:* NUMBER 0 15 IO IO IO I 218 11 911 - 0 10 I 7 - 010 014 OF O I6 TEXT (ff mor. .,,ac. itJ m,ul,._,,, u.. *ddltional NRC Fam, 366A 'a) 1171 Because there was no actual plant condition requiring safety injection and plant conditions remained normal, the failure of the containment

• isolation valve (02-DA-TV-200A) to fully shut had no adverse affect. Also, the redundant isolation valve outside containment (02-DA-TV-200B) functioned properly.

The health and safety of the public were unaffected during the event.

3.0 CAUSE OF THE* EVENT The event was caused by the intermittent loss of voltage to channel IVA protection instrumentation and a _coincident low spiking of channel II -

steam generator pressure protection. The combination of these two conditions momentarily satisfied the high steam flow and low steam line pressure safety injection logic (two out of three high steam line flow signals coincident with two out of three low steam line pressure signals) and produced the ESF actuation .. The cause of the problem in vital bus channel IVA was an inadequate electrical connection at the panel's main breaker with evidence of arcing at the contacts. The suspected cause of the transmitter failure was an internal short causing the instrument to fai.1 high. Subsequently, the shorting condition cleared, producing a negative change in voltage sufficient to trip the low pressure bistable, and make up the safety injection logic matrix.

4.0 IMMEDIATE CORRECTIVE ACTION(S)

Shift operating personnel placed the plant in a stable condition. The electrical connection to the input breaker to Vital* Bus Distribution Panel IVA was found. loose with intermittent contact occurring between the.

breaker and the vital bus panel bus bar. This connection was tightened and channel IVA instrumentation stabilized ..

5.10 ADDITIONAL CORRECTIVE ACTION($)

The Vital Bus Distribution Panel IVA input breaker was bench-tested satisfactorily and reinstalled. This breaker is a two-pole breaker with one pole in use at a time. The reconnection was made at the previously unused pole as a precaution. The three additional Unit 2 vital bus.panels which had been installed during the same plant modification were checked for evidence of a similar condition. They were found to be NRC Form 366A (6-89)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION

(~-89) .... - APPROVED 0MB NO. 3150-0104

.. EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE. REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (61 PAGE (31 Surry Power Station, Unit 2 TEXT'" mom IIPI'"" i& mquimd, U/Jtl llddmona/ NRC Form .1151iA's) 1171 satisfactory. A Root Cause Analysis was initiated to investigate this event.

The failed pressure transmitter was replaced, and the defective instrument is being returned to the vendor for failure analysis.

EOG #3 was declared inoperable, and an. investigation into its performance was initiated. The engine's Woodward UG-8 governor (EIIS-DG-65) was found to be incorrectly set. The governor was returned to the proper setting, and the engine was tested and restored to operable status at 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> on August 3, 1991.

A separate root cause evaluation of the EOG #3 governor problem is being conducted by a team consisting of corporate and key station personnel. Licensee Event Report S1-91-017 will be submitted based on the results of this evaluation.

The Containment Sump Pump Discharge Trip Valve (02-DA-TV-200A) was binding due to excessive spring tension on its valve seat. The valve was repaired, tested, and returned to service at 2132 hours0.0247 days <br />0.592 hours <br />0.00353 weeks <br />8.11226e-4 months <br /> on August 5, 1991. A failure analysis is being conducted on this valve, and a supplement to this report will be submitted when the results of the analysis. have been received.

6.0 ACTIONS TO PREVENT RECURRENCE The vital bus distribution panel main breaker finger connectors and incoming feeds will be inspected and retorqued during the next refueling outage for each unit.

• Breaker maintenance proc~dures will be revised to include the manufacturers torque specifications for finger connectors.

The circuit breaker and associated switchgear preventive maintenance program will be revised to periodically inspect and retorque breaker electrical connections.

Any additional corrective action recommendations resulting from the Root Cause Evaluation will be evaluated and implemented as appropriate.

NRC Form 366A (6-89)

NRC FORM~366A U.S. NUCLEAR REGULATORY COMMISSION

\6.-89) ** ' APPROVEO 0MB NO. 3150-0104 EXPIRES: 4/30/92 .

LICENSEE EVE. REPORT (LER) ES, ..:.IATE.O BUR~EN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARO COMMENTS

- REGAROING BUROEN ESTIMATE TO THE RECOROS TEXT CONTINUATION ANO REPORTS MANAGEMENT BRANCH (P-530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555, ANO TO 1'HE PAPERWORK REOUCTION PROJECT (3150-0104). OFFICE OF MANAGEMENT ANO BUOGET, WASHINGTON, OC 20503. '

FACILITY NAME (1) OOCKET NUMBER (2) LER NUMBER 161 PAGE (31 Surry Power Station, Unit 2 o 1s I o Io I O ,2 1s 11 _9 11 _ o 10 I 7 _ o1o 016 oF o I 6 TEXT /ff mare _ . . la requ/1..J, U$t1 additiofll!! NRC Farm 35liA 's/ (17) 7 ..0 SIMILAR EVENTS Licensee Event Report S1 029-00, "Reactor Trip/Safety Injection Due to Spurious Hi CLS Signal as a Result of a Malfunctioning Relay".

8.0 ADDITIONAL INFORMATION Failed components:

Rosemount Transmitter (02-MS-PT-2474)

Model No. 1153GB9 NRC Form 366A 16-89)