ML20044D571

LER 93-002-00:on 930416,automatic Reactor Trip Initiated from Turbine Trip Due to Over Excitation of Main Generator. Emergency Procedure 2-E-0 Entered & Individuals Involved in Event Received Remediated training.W/930514 Ltr

ML20044D571
Person / Time
Site:	North Anna
Issue date:	05/14/1993
From:	Kane G VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-002-02, LER-93-2-2, NUDOCS 9305190337
Download: ML20044D571 (4)

Text

_ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ ._____ _ _-_

J o

VIRGINIA ELECTRIC AND POWE R COMP ANY NO RTH AN N A POW E R S T AT 60 N P. O. BO X 4 02 W IN E R AL, VIRGINI A 23117 l 10 CFR 50.73 May 14,1993 U. S. Nuclear Regulatory Commission NAPS:MPW Attention: Document Contml Desk Docket No. 50-339 Washington, D.C. 20555 License No. NPF-7

DearSirs:

The Virginia Electric and Power Company hereby submits the following Licensee Event Repon applicable to North Anna Unit 2.

Report No. 50-339/93-002-00 This Report has beca reviewed by the Station Nuclear Safety and Operating Committee and will be forwarded to the Corporate Management Safety Review Committee for its :rview.

Very Truly Yours, G. E. ' ne Station Manager

Enclosure:

cc- U.S. Nuclear Regulatory Commission 101 Marietta Street, N.W.

Suite 2900 Atlanta, Georgia 30323 Mr. D. R. Taylor NRC Resident Inspector i North Anna Power Station 180071 9305190337 930514

[M g/

PDR ADOCK 05000339 I S PDR i

NEC FORM 3fE US NJCLEAR RE3JLATORYCnem APPROVED oMB NCL 3150-0104 M EXPIRES: 4T03 ESTIMATED BURDEN PER RESPONSE To COMPLY WITH THIS INFORMATa)N CCALECTON REoUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN UCENSEE EVENT REPORT (LER) ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH (P-533). U.S.

NUCEEAR REGULATORY CoMMISSloN. WASHINGTON. DC 20555. AND To THE PAPERWORK REDUCTION PROJECT (3154010s). oFFCE OF MANAGEMENT AND BUDGET. WASHINGTON.DC 205D3.

F ACauTV NAME (t) DOCAEi NJMillR (2) M1 North Anna Power Station Unit 2 elclSl0l0l31319 1lOFl013 NW AUTOMATIC REACTOR TRIP INITIATED FROM A TURBINE TRIP DUE TO AN OVER EXCITATION OF THE MAIN GENERATOR EVENT DATE (S LER NUMBER (6) REPol.T DATE F) cTHER FACILITES WVoLVED(a)

WONTH DAY YEAR YEAR MONTH DAY YEAR SE,OJE 91slololol i I

~ ~

xCAn uaS) 0 4 1 6 9 3 9 3 0 0 2 0 0 0 5 1 4 9 3 ggglgggjgl l g oPERATNG THIS REPOR1 sS BulwTTED PURSUANT 10 THE RE OJiREWE ATS OF 10 CF R $* fCheca one or mye of me imovnng) 01)

MODE (9) 1 204021b) 2040b(o SoJ1aZ(w) 7331(D)

POWER 20405(a)(1X0 50.3rucin) 503*a)2M 7311(c) l 0l0 -

1. '**""" ''* *#**'**' Ym 20 4054a)JNdi) 507*a)20) 5032a)2rviiixA) 20405%ax1XM SC7Aa)2@ $07&aX2)MA)(B) 20 40Nax1Mv) SCJ*alatily 5071aH2Hz) uCENSEE CONT AC1 FoR THIS LE k 02)

NOME TELEPHONE NUMBER G. E. Kane AREA C00E 7l0l3 8l9l4l-l2l1l0l1 COMPLETE ONE LNE F 04 E ACH COWPONE NT F ALLURE DESCRIBED IN THIS RE PORT 035 CGISE SYSTEv CoWaONENT CAJSE SYSTEM COMPONENT T RE O PROS E T B R G 7" X W ,1 2 0 Y X S J R V C 7 1 0 Y X JlD ClHlAl Wl1l2 0 Y SuPPd WE NI AtkiPokT E pi ciE D pa, X SfJ IlL l Gl0 8 0 N EXPECTED MONTH DAY YEAR SUBWISSloN VEs pre. emap.e tnamto ssaamssom ba'O

%No DATE (15) l l l ABSTRACT nean enaz spenas ie app , mimea enseesse se eams,(16)

On April 16, 1993, at 0717 hours0.0083 days <br />0.199 hours <br />0.00119 weeks <br />2.728185e-4 months <br /> with Unit 2 in Mode 1, 100 percent power, an automatic reactor trip occurred as a result of a turbine trip. Emergency procedures were entered and immediate actions were performed. Subsequently, the operating crew became concerned with the reactor coolant system (RCS) cooldown when temperature decreased to approximately 540 degrees F. To reduce the Steam Generator feedwater addition rate and stabilize the RCS temperature the ATWS Mitigation System Actuation Circuitry was reset and the Auxiliary Feed Water (AFW) pumps were secured before steam generator levels were restored above the automatic start setpoint. Defeating the automatic start capability of the AFW pumps is prohibited by Technical Specifications. A4 hour report was made to the NRC at 1055 hours0.0122 days <br />0.293 hours <br />0.00174 weeks <br />4.014275e-4 months <br /> pursuant to 10CFR50.72 (b) (2) (ii) & (iii) ( A) . The event is reportable as an Engineered Safety Feature System actuation pursuant to 10CFR50.73 (a) (2) (iv) & (v).

The cause of the turbine trip / reactor trip was a malfunction in the main generator voltage regulator circuitry. The cause of defeating the AFW system during the event was a result of personnel error.

No significant safety consequences resulted from the reactor trip because reactor protection safety systems responded as designed. Disabling the AFW pumps did not present a significant safety consequence because the heat sink was maintained throughout the event. Therefore, the health and safety of the public were not affected at any time during this event.

weF um

. . _ _ _ _. - _ _. ~ -.

NRC FoRu am.A es unEAR Rumitafonmmmn appRoeED ous No sisocios l

(66 EKPIRES: ct3n2 '

ESTIWATED BURDEN PER RESPONSE To CoWPLY WITH THIS INFoRMATON l UCENSEE EVENT REPORT (LER) cou. ECTR WEQUEST: 50.0 HRS. FORWARD CoWWENTS REGARDING BURDEN TEXT CONTINUATION ESTsuATE To THE RECORDS AND REPORTS WANAGEMENT BRANCH (P-EXE U.S.

WUCLEAR REGULATORY CowutSSeDN. WASHINGTON. DC 206b6, AND TO THE ,

PAPEkWoRE REDUCTON PROJECT (31540104 oFFcE oF WANAGEWENT AND '

BUDGET. WASHINGTON.DC 23603. ,

F ACILITV NAME (1) 00CEE1 kJup sa LER NJMIER(6) PAGE (3)

SEOJEs(IAL . .m  ;$EvsKm g s. s ,

North Anna Pcwer Station N WWRER M WMER j Unit 2 al t l a l e i n l 31319 913 -lI O l 0 l 2 0 !0 012 OF 013 !

TEKT m mise spes = eeeees ese emas. esc . asen m (17) a 1.0 De s crint i en of the Event ,

1 On April 16, 1993, at 0717 hours0.0083 days <br />0.199 hours <br />0.00119 weeks <br />2.728185e-4 months <br /> with Unit 2 in Mode 1, 100 percent power, an automatic reactor trip occurred from a turbine trip due to a malfunction in .

the main generator voltage regulator circuitry (EIIS System TB, Component TG).

Emergency procedures were entered and immediate actions were perf ormed. The i Auxiliary Feedwater Pumps (APW) (EIIS System BA, Component P) automatically i started on Lo Lo Steam Generator (SG) (EIIS System AB, Component SG) level.

During subsequent recovery actions of the reactor trip response procedure it '

was noted that the reactor coolant system (RCS) was experiencing a cooldown due to feeding the SGs with relatively cold water from the AFW system. The i

operating crew became concerned with the RCS (EIIS System AB) cooldown rate

! when temperature decreased to approximately 540 degrees F. To reduce the SG .

l feedwater (EIIS System SJ) addition rate and stabilize the RCS temperature,

! the ATWS Mitigation System Actuation Circuitry (EIIS System JC) was reset, and >

the AFW pumps were secured in a manner that rendered them inoperable before SG ,

levels were restored above the automatic start setpoint. (

l i

After securing the AFW, Main Feed Water (MFW) was the makeup water source for f l

the SGs. Subsequently, approximately 19 minutes later, the emergency j procedure reader noticed that the AFW pump status did not conform to the l appropriate emergency procedure step and immediately notified the Shift l Supervisor (SS) who directed the pumps to be returned to AUTO. Defeating the '

automatic start capability of the AFW pumps is prohibited by Technical l l Specifications. A 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> report was made to the NRC at 1055 hours0.0122 days <br />0.293 hours <br />0.00174 weeks <br />4.014275e-4 months <br /> pursuant to l 10CFR50.72 (b) (2) (ii) & (iii) ( A) . The event is reportable as an Engineered Safety Feature System actuation pursuant to 10CFR50.73 (a) (2) (iv) & (v). j 2.0 significant safety censequences and Tm lications No significant safety consequences resulted from the reactor trip because reactor protection safety systems responded as designed. No significant safety consequences resulted from disabling the AFW pumps for approximately 19 minutes because the heat sink was maintained throughout the event. The AFW pumps could have been made available immediately by manual operator action.

! The AFW system was always under the cognizance of Licensed Operator. Main feedwater was also available throughout the event and used to provide makeup to the SGs. Therefore, the health and safety of the public were not affected at any time during this event.

3.0 cause of the Event The cause of the turbine trip / reactor trip was the result of a malfunction in the main generator voltage regulator circuitry.

The cause of defeating the AFW system was personnel error. Insufficient command and control of the unit trip response and inadequate communications between the operations crew members resulted in defeating the AFW pump, when a valid start signal was present.

.mc. mea m i

1

3 a NRC FoRW 3fEA US QCLIAR REGJLAT0m Pr"" APPROVED oWB No. 3150a104

$ am EXPIRES: af30W

=

EST!WATED BURDEN PER RESPONSE TO CoWPLY wtTH THIS INFoRWATON UCENSEE EVENT REPORT (LER) cottECTON REouEST: 50 0 HRS. FORWARD CoWWENTS REGARDING BURDEN TEXT CONDNMDON ESiluATE To THE RECORDS AND REPORTS WAAAGEWENT BRANCH (P430L U.S.

NUCLEAR REouLAToRY CoWWISStoN. WASHINGTON. DC 20666. AND To THE  ;

PAPERWORK REDUCTON PROJECT (3160010s). oFFCE oF WANAGEWENT AND acaET.WASHmioN.DC 20m i 5 Act:TV NAuE m D3CK11 maE R m LER NJtEER(6) PAGE m b hEvilKm North Anna Power Station g

E 3 SEQUEWTAL WWKa "Ka L

Unit 2  !

nl c I n ! n l n { 313l 9 913 -

0l0 l2 . O l0 013 0F Ol3 i TEXT p=e ep . wes see emc pen asum 07) l 3_0 Cause of t he Event (centinued)

The policy associated with defeating equipment or system automatic safety functions was misunderstood. In addition, management expectation of communications and problem solving using all crew members was not effectively ,

conveyed.

4.0 T--adiate corrective Actions r Following the reactor trip Emergency Procedure 2-E-0, Reactor Trip or Safety Injection, was entered and the immediate actions performed. The Shift Supervisor immediately directed that the AFW pumps be returned to the  ;

automatic position when the condition was identified.  ;

r 5.0 Additienni corrective Aetiens  !

The individuals involved with the AFW pump condition were coached on the station's policy for defeating equipment automatic functions. These individuals were removed from licensed duties and received remediated training i designed to enhance their control room communication skills and their i understanding of the control room command and control structure during emergency procedure implementation.

i i

f 0 2a+4 m e &6 preype Doe ee m p l

I Requirements are in place to ensure the event is discussed in the Licensed j Operator Requalification Program. A root cause was performed and corrective l actions are being reviewed by management for implementation as appropriate.

( The training reviews and the actions taken regarding the individuals involved are sufficient to preclude recurrence.

7.0 si-41ar Events l

l LER N2-86-008-00 identified a reactor trip from a turbine trip as a result actuation of a main generator differential lockout relay upon loss of an excitation field signal. The signal was caused by failure of the permanent magnet generator in the main generator excitation system.

l B_0 Additsenal Info m tien Component failures resulting from the automatic reactor trip included: Source Range Channel N31 failed low, lA Feedwater Heater relief valve lifted and would not reset until the feedwater heater was isolated and depressurized, and the "B" MFW Pump breaker indicating lights did not work in the Control Room.

Corrective actions included replacement of the Source Range Channel detector, lA Feedwater Heater relief valve, and the "B" MFW Pump breaker lights.

Unit I was in Mode 3, hot standby, returning to power operations following a refueling outage and was not affected by the event.

seeC Fee atu m