ML18153B305

LER 93-003-00:on 930803,automatic Reactor Trip Occurred Due to Low Steam Generator Water Level Coincident W/Steam/ Feedwater Flow Mismatch Resulting from Spurious Closure of Main Feedwater Regulating Valve A.W/930827 Ltr

ML18153B305
Person / Time
Site:	Surry
Issue date:	08/27/1993
From:	Kansler M VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
93-538, LER-93-003-03, LER-93-3-3, NUDOCS 9309030222
Download: ML18153B305 (8)

Text

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ACCESSION NBR:9309030222 DOC.DATE: 93/08/27 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.NAME RECIPIENT AFFILIATION R

SUBJECT:

LER 93-002-00:on 930803,automatic reactor trip occurred due I to low steam generator water level coincident w/steam/

feedwate~r flow mismatch resulting from spurious closure of D main feedwater regulating valve A.W/930827 ltr~

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I NOTES:lcy NMSS/SCDB/PM. 05000281 A

RECIPIENT COPIES RECIPIENT* COPIES

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ID CODJ~/NAME LTTR ENCL ID CODE/NAME LTTR ENCL*

PD2-2 LA 1 1 PD2-2 PD 1 1

-BUCKLEY,J3 1 1 D INTERNAL: ACNW 2 2 ACRS 2 2

.AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1.

s AEOD/ROA:B/DSP 2 2 NRR/DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/DORS/OEAB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR DRIL/RPEB 1 1 NRR/DRSS/PRPB 2 2 R SPLB 1 1 NRR/DSSA/SRXB 1 1 G FIL

• 02 1* 1 RES/DSIR/EIB 1 1 RGN FILE 01 1 1 EXTERNAL: EG&G BRYCE,J.H 2 2 L ST LOBBY WARD 1 1 R NRC PDR 1 1 NSIC MURPHY,G.A 1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 I NOTES: 1 1 D

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10CFR50.73 Vn-ginia Electric and Power Company SmTy Power Station P.O.Box:315 Surry, Vi_,.;n;a

. - ~ 23883 August 27, 1993 U. S. Nuclear lli~gulatory Commission Serial No.: 93-538 Document Control Desk SPS:RCB Washington, D. C. 20555 Docket No.: 50-281 License No.: DPR-37

Dear Sirs:

Pursuant to Sur:ry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to Surry Power Station Unit 2.

BEPORf NUMBER 50-281/93-003-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be forwarded to the Management Safety Review Committee for its review.

Very truly yours, Enclosure cc: Regional .Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 M. W. Branch NRC Senior Resident Inspector Surry Power Station 0300~u 9309030222 930827 PDR ADOCK 05000281 .*:

i-**

S PDR \ ...

NRC FORM 366 .S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 (S-92) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF (See reverse for:required number of digits/characters for each block) MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

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

Surry Power Station, Unit 2 - . 05000 - 2 81 . 1 OF 5 TITLE (4) Unit 2 Automatic Reactor Trip Due to Low Steam Generator Water Level Coincident With Ste~1.m/Feedwater Flow Mismatch Resulting, *From Snuriou!': r.ln,mrp nf ".A" MFRV EVENT DATE (5) LER NUMBER (6 REPORT NUMBER (7) OTHER FACILITIES INVOLVED (B FACILITY NAME DOCKET NUMBER SEQUENTIAL REVISION MONTH DAY YEAR MONTH DAY YEAR NUMBER NUMBER 05000 08 03 93 ~'I -- 003 -- 00 08 27 93 FACILITY NAME DOCKET NUMBER .

05000 OPERATING THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11 MODE (9)

N 20.402(b) 20.405(c) 73.71(b)

X 50.73(a)(2)(iv)

POWER 20.405(a) (1) (i) 50.36(c)(1) 50.73(a)(2)(v) 73.71 (c)

LEVEL (10) 97% 20.405 (a)(1 )(ii) so:35(c)(2) 50.73(a) (2) (vii) OTHER 50.73(a)(2)(i) 50.73(a) (2) (viii) (A) (Specify in Abstract 20.405(a) (1) (iii)

'.iiiiiI .?:*, ./ 11  !:ii/iii: 20.405(a) (1) (iv) 50.73(a) (2) (ii) 50.73(a)(2) (viii) (8) below and in Text, NRC Form 366A)

......................** ;\!\!\:\ 1\!i!:lii;!!j.:!!*:::t:11:**

• • :;:::*: 20.405(a) (1) (v) 50.73(a)(2)(iii) 50.73(a) (2) (x)

LICENSEE CONTACT FOR THIS LER 12)

NAME TELEPHONE NUMBER (Include Area Code)

M. R. Kansler, Station Manager (804) 357-3184 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER .*,*. CAUSE *sYsTEM COMPONENT MANUFACTURER TO NPRDS ***** TO NPRDS B AA IL M035 N B JB FCV Wl20 y B SB RV Wl20 y SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR I

EXPECTED SUBMISSION YES (If yes, complete EXPECTED SUBMISSION DATE) X NO DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

At 2005 hours0.0232 days <br />0.557 hours <br />0.00332 weeks <br />7.629025e-4 months <br /> on August 3, 1993, with Unit 1 at 100% power and Unit 2 at 97% power, Unit 2 experienced an automatic reactor trip. The trip occ1Jrred when the "A" Main Feedwater Regulating Valve (MFRV) unexpectedly closed, causing a feed flow/steam flow mismatch coincident with a low water level in the "A" steam generator. Reactor Pro'tection System functions actuated as. designed, and post-trip res~>onse was satisfactory. _ The reactor was placed in a safe, hot shutdown condition, and the health and safety of the public were not affected. The cause of the closure of the MFRV was traced to an erratic power supply in the manual/automatic control station. This report is required by 10CFR50. 73(a)(2)(iv).

NRC FORM 366 (5*92)

e e REQUIRED NUMBER OF DIGITS/CHARACTERS FOR EACH BLOCK BLOCK NUMBER OF TITLE NUMBER DIGITS/CHARACTERS 1 UP TO 46 FACILITY NAME 8 TOTAL 2 DOCKET NUMBER 3 IN ADDITION TO 05000 3 VARIES PAGE NUMBER 4 UP TO 76 TITLE 6 TOTAL 5 EVENT DATE 2 PER BLOCK 7 TOTAL 2 FOR YEAR 6 LER NUMBER 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL 7 REPORT DATE 2 PER BLOCK.

UP TO 18 FACILITY NAME 8 OTHER FACILITIES INVOLVED 8 TOTAL - DOCKET NUMBER 3 IN ADDITION TO 05000 9 1 OPERATING MODE 10 3 POWER LEVEL 1

11 REQUIREMENTS OF 10 CFR CHECK BOX THAT APPLIES UP TO 50 FOR NAME 12 LICENSEE CONTACT 14 FOR.TELEPHONE CAUSE VARIES 2 FOR SYSTEM 13 4 FOR COMPONENT . EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES 1

14 SUPPLEMENTAL REPORT EXPECTED CHECK BOX THAT APPLIES 6 TOTAL 15 EXPECTED SUBMISSION DATE 2 PER BLOCK

e

/

e NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 (5-82) EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLV WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSl.::E EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR "TEXT CONTINUATION REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

_FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE*(3)

SEQUENTIAL REVISION YEAR NUMBER NUMBER Surry Power Stc:1.tion, Unit 2 05000 - 281 2 OF 5 93 - 003 00 TEXT (II more space i* required, use addi~'ona/ copies of NRC Form 356A) (17) 1.0 J2gsJ~RIPTION OF THE EVENT On August 3, 1993, at 2005 hours0.0232 days <br />0.557 hours <br />0.00332 weeks <br />7.629025e-4 months <br />, with Unit 2 at 97% power, the Unit

• experienced an automatic reactor trip because of a steam flow/feed flow mi.smatch coincident with a low water level in the "A" steam generator, 2-RC-E-1 A (EIIS-AB, SG). The transient occurred when the "A" Main Feedwater Regulating Valve (MFRV), 2-FW-FCV-2478 (EIIS-JB, FCV) unexpectedly closed. The turbine (EIIS-TA) and main generator (EIIS-TB) t,ripped as designed. The Anticipated Transient Without Scram Mitigation System Actuation Circuitry (AMSAC) also actuated as designed. The Auxiliary Feedwater Pumps, 2-FW-P-2, 2-FW-P-3A, and 2-FW-P-38 (EIIS-BA, P) automatically started on decreasing steam generator level as designed.

Control Room operators responded to the trip in accordance with emenJency and other operating procedures. Plant response was as expec:ted except for the following:

• Individual Rod Position Indicator (IRPI) rod bottom light .(EIIS-AA, IL) forr control rod M-10 was slow in illuminating. (The rod bottom light illuminated at approximately 2055 hours0.0238 days <br />0.571 hours <br />0.0034 weeks <br />7.819275e-4 months <br />.)

• "G" Steam Generator Power Operated Relief Valve (PORV), 2-MS-RV-.

201 C, (EIIS-SB, RV) indicated "intermediate" at 960 psig (design lift setting is 1035 psig). Subsequent investigation showed that the valve did not lift.

The Nuclear Regulatory Commission was notified in accordance with 1OCPR50. 72 at 2221 hours0.0257 days <br />0.617 hours <br />0.00367 weeks <br />8.450905e-4 months <br />. This event is being reported pursuant to 10CFR50. 73(a)(2)(iv) as an automatic actuation of the Reactor Protection System (RPS) (EIIS-JC).

NRC FORM 366A (5-92)

NRC FORM 366A eU.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 (5*92) EXPIRES 5/31/95 ESTIMATED BURDEN PER. RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: . 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB n14), U.S. NUCLEAR

• rEXT CONTINUATION REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

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

SEQUENTIAL REVISION YEAR NUMBER NUMBER Surry Power Station, Unit 2 05000 - 281 3 OF 5 93 - 003 00 TEXT (II more space is required, use additfantJI copies of NRC Form 3561\) (17) 2.0 Sl.GJ\IIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS Upon receipt of the reactor trip, RPS actuations functioned as designed, and all control rods inserted into the core. The electrical buses .

transferred properly and off-site power was maintained throughout the event. The emergency diesel generators remained operable in automatic, but were not required to start. Station operating personnel acted promptly to place the plant in a stable, hot shutdown condition.

The shutdown *margin of reactivity was calculated and found to be satisfactory. The health and safety of the public were hot affected.

3.0 .GA!J SE OF THE EVENT 1

The reactor tripped as designed when the "A" Steam Generator expe!rienced a feed flow/steam flow mismatch coincident with a low steam generator water level. The cause of the decrease in feedwater flow and loss of water level was closure of the "A" MFRV. The valve's positioner received an unexpected demand to shut, and etto*rts on the part of the licensed Control Room Operator to take manual control and open the valve were unsuccessful in reversing the transient. The cause of th1e closure of the MFRV was traced to an erratic power supply in the manual/automatic control station.

4.0 IMM'EDIATE CORRECTIVE ACTION(S)

Operiators acted promptly to place the plant in a safe, shutdown condition in ac:cordance with emergency and other operating procedures. _The licensed Control Room Operator adjusted the "C" Steam Generator PORV set point to clear the intermediate indication.. The Shift Technical _

Advisor monitored the safety function status trees to verify that unit conditions were acceptable. -

• NRC FORM 366A (5*92)

NRC FORM 366A eU.S. NUCLEAR REGULATORY COMMISSION e APPROVED BY 0MB NO. 3150-0104 (5-92) EXPIRES 5/31 /95

• ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 . HRS. FORWARD LICENSIEE EVENT REPORT {LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION

. ANO RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR

'TEXT CONTINUATION REGULATORY COMMISSION, WASHINGTON, DC 2055~0001, ANO TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT ANO BUDGET, WASHINGTON, DC 20503.

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

SEQUENTIAL REVISION YEAR NUMBER *NUMBER Surry Power Statton, Unit 2 05000 - 281 4 OF 5 93 - 003 - 00 TEXT (II more space is raqvired, use addil,onal copies of NRC Form 366A) (1 7) 5.0 Aru>ITIONAL CORRECTIVE ACTION($)

• IRPI rod bottom light for control rod M-10.

- Problems have been experienced with this indication for several years. The Nuclear Steam Supply System vendor has evaluated the condition and concluded that its cause is residual permeability in the control rod drive mechanism housing. This phenomehon is caused by a combination of factors, including material composition and the decrease in reactor coolant system temperature following a trip.

- A hot rod drop test conducted subsequent to the reactor trip showed that control rod M-1 O was fully operable.

• The "C" Steam Generator PORV control loop was investigated and a

• number of discrepancies were detected. The pressure controller and power supply were replaced and the transmitter was recalibrated.

• "A" Main Feedwater Regulating Valve.

.. An extensive troubleshooting effort was conducted in an attempt to identify the source of the signal which caused the "A" main ,

were examined:

-- valve controller inputs

-- manual/automatic control station

-- turbine first stage pressure input summator

-- terminal connections on control system modules

-- valve actuator

-- valve positioner

-- electro/pneumatic converter.

• • A number of components were replaced, including:

-- flow controller

-- manual/auto control station During examination of the MFRV manual/automatic control station which hacl been replaced, it was noted that the - 15 voe power supply was behaving erratically. A failure of this power supply will cause the controller to fail to zero output and inhibit control responses in automatic or manual. Since a lack of controller response in automatic and manual was observed during the event, the controller was determined to be the caU1se of the event.

NRC FORM 366A (5*92)

--1 e e r.:N=R=c=F::::O:::R==M=3=66=A==========:=u;:::;.s;=.c=N;::;:uc="=L"=EA~R'=R'='EG~U~LA====ro~R::':Y';":C;:50:;':M':':M':':l~SS::':IO~N:';'if===:A:"=P~PR;::':O;:':V';;:E;;:'D==e~v==o::=:M:'=B=;'N';';:O:=.'='31='=5'='0-0'='1===0===4===;i

. (5-92)

EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR 1"EXT CONTINUATION REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TD THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

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

SEQUENTIAL REVISION YEAR.

NUMBER NUMBER Surry Power Station, Unit 2 05000 - 281 5 OF 5 93 - 003 - 00 TEXT {II more space is required, use addi~c,na/ copies of NRC Form 366"') (17) 6.0 ACIJ.,ONS TO PREVENT RECURRENCE A Root Cause Evaluation (RCE) was initiated immediately after the reactor trip. Recommendations of the RCE will be evaluated and implemented as appropriate.

7 .0 SIMILAR EVENTS LER S2-86-007 Manual reactor trip due to high steam generator level (metal debris between MFRV plug and valve seat).

LER S2-90-003 Manual reactor trip due to failure of "A" Main Feedwater Regulating Valve (blockage of positioner air supply inlet filter/orifice assembly.)

LER S2-90-004 Manual reactor trip following inadvertent grounding of the "A" Main Feedwater Regulating Valve control signal during testing.

8.0 MANIUFACTURER/MODEL NUMBER IRPI Hod Bottom Light Spang & Co.,

Magnetics Div.

Signal Conditioning Card, EPC-2Nl-13, Part# E 2786 "C" Steam Generator PORV Westinghouse Hagan Controller Model 4111080-001 MFRV Manual/Automatic Control Westinghouse Hagan Statio,n 7100 Series Control System Model 124 Controller NRC FORM 366A (5*92)