ML18153D394

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LER 93-002-00:on 930620,automatic Trip Occurred Due to Low SG Water Level Coincident W/Steam/Feedwater Flow Mismatch Resulting from Main FW Pump Trip.Cr Operators Promptly Initiated Appropriate EOPs.W/930715 Ltr
ML18153D394
Person / Time
Site: Surry Dominion icon.png
Issue date: 07/15/1993
From: Kansler M
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
93-438, LER-93-002-03, LER-93-2-3, NUDOCS 9307230337
Download: ML18153D394 (7)
v • d • e


Text

A1.CCELERATED DOCUMENT DISTRIBUTION SYSTEM

,, - . , REGULA._ INFORMATION DIS_TRIBUTIO-YSTEM (RIDS)

. 1 ACCESSION NBR:9307230337 DOC.DATE: 93/07/15 NOTARIZED: NO DOCKET#

FACIL t 50-*2a1 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 930620~automatic trip occurred due to low I SG water level coincident w/steam/feedwater flow mismatch resulting from main FW pump trip.CR operators promptly D initiated appropriate EOPs.W/930715 ltr.

DISTRIBUTION CODE: IE22T COPIES ~ECEIVED:LTR _J_ ENCL .1._ SIZE:_----'-tJ~~~ s TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

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

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

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D NOTE TO ALL "RIDS" RECIPIENTS:

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

ROOM Pl-37 (EXT. 504-2065) TO ELIMINATE YOUR NAME FROM DISTRIBUTION.

LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL ~rnXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 33 - ENCL 33

e 10CFR50.73 Vn-ginia Electric and Power Company Sony Power Station

. P. 0. Box 815 Surry, Vn-ginia 23883 July 15, 1993 U; S . Nuclear Regulatory Commission Serial No.: 93-438 Document Control Desk SPS:BCB Washington, D. C. 20555 Docket No.: 50-281 License No.: DPR-37

Dear,

Sirs:

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

  • BEPQRI' NUMBER 50-281/93-002-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and 'will be forwarded to the.Management Safety Review Com.nµttee 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

-200iJJ

{~ r,, .. -

9307230337 930715 PDR ADOCK 05000281 S . PDR jV'Y',

N~..;,FORM366 (6-89)

.e U.S. NUCLEAR REGULATORY COMMISSION e APPROVED,OMB N0.3150-0104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT. CLER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS 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) IDOCKET NUMIIER (21 ' PAGE (3)

Surry Power Station, Unit 2 o 1s101010121 811 ,loF 015 TITLE 141 Unit z Automatic Keaci::or 1rip vue to Low :,,:earn uenerai::or wai::er Level Coinci:ient With Steam/Fe!;!dwater Flow Mismatch Resulting From Main Feedwater Pump Trip EVENT CATIE (51 ** LEA NUMBER (61 REPORT DATE (7) OTHER FACILITIES INVOLVED CB:

MONTH DAV YEAR YEAR .}t SE~~~;i~AL { ( ~~i~~ MONTH DAV YEAR FACILITY NAMES DOCKET NUMBERISI OPERATING THIii REPORT IS SUBMITTED PURIIUANT TO THE Rl:.OUIREMENTI OF 10 CFR §: (Chock on* or mor* of th* fol/owin11/ 1111 I

MODE 1111 N 20.402(bl

-- 20.40ll(cl

-- ./\ II0,731ol(2l(M 73.71(bl

  • POWER 20.40lll11111111 II0.311(cll1 I II0.73111C21M 73.711cl L~~~L O 18 I O - 20.40ll(11l11CIII 20.40lll1111 llilll -- II0.311(cl(21 II0.73111121111 II0.731*H21Mil II0.7311H21MIIIIAI OTHER (Spocify in At,itr*ct below *nd in THt, NRC Form 366AJ 20.40lll11C1ICl*I 20.40lll1l11 IM - II0.73(11(211111 II0.73C.IC2111111 LICENSEE'CONTACT FOR THIS LEA 1121 lio.73(11C2llvtllllBI II0.73(11(21hd NAME TELEPHONE NUMBER AREA CODE M. R. Kansler, Station Manager COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 MANUFAC- MANUFAC-CAUSE SYSTEM COMPONENT TURER TURER V

A I . I M I O A I 11 8 I O . y I I I I I I I I I I I I I I I I I I I I I SUPPLEMENTAL REPORT EXPECTED 1141 MONTH DAV YEAR EXPECTED .

n YES (If ya, compl~re EXPECTED SUBMISSION D~TEJ SUBMISSION DATE 1151 I I I Unit 2 was operating at 100% power on June 20, 1993, when Main Feedwater Pump 2-FW-P-1 A tripped at 0415 hours0.0048 days <br />0.115 hours <br />6.861772e-4 weeks <br />1.579075e-4 months <br />. A Unjt 2 automatic reactor trip followed.

at 0416 hours0.00481 days <br />0.116 hours <br />6.878307e-4 weeks <br />1.58288e-4 months <br /> as a result of a low steam generator water _level with steam/feedwater flow mismatch reactor protection signal from Steam Generator 2-RC-E-1A. Appropriate operator actions were taken to ensure the performance of system automatic actions and to respond to abnormal conditions. The Unit was quickly brought to a stable, no-load condition. The event was caused by the occurrence of an instantaneous electrical ground on the 2-FW-P-1 A inboard motor and subsequent pump trip. A Root Cause Evaluation is being performed to determine the cause of the 2-FW-P-1A inboard motor electrical ground and to evaluate the conditions experienced following the reactor trip. This event is being reported pursuant to 10 CFR 50. 73(a)(2)(iv).

NRC Form 3IMI (6419)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-89) a APPROVED 0MB NO, 3150-0104 EXPIRES: 4/30/92 9,,ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVIT 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 1'HE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGE.MENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (11 DOCKET NUMBER (21 tER NUMBER (6) PAGE (31 Surry Power Station, Unit 2 0 1s IO IO IO I 21 8 I 1 913 - 0 I 01 2 - 0 IO 01 2 OF ci I 5 TEXT (ff mom 11/MCtl ~ ,equif9d, uae tlddltional NRC Form 31511A 'a) (17) 1 .o DESCRIPTION OF THE EVENT Unit 2 was operating at 1OOo/o power on June 20; 1993, when Main Feedwater (MFW) Pump 2-FW-P-1A [EIIS-SJ,P] tripped at 0415 hours0.0048 days <br />0.115 hours <br />6.861772e-4 weeks <br />1.579075e-4 months <br />.

MFW Pump 2-FW-P-1 B continued running and a partial loss.of feedwater flow was indicated in the control room. Control room operators responded in accordance with the appropriate abnormal operating procedure by starting a third condensate pump [EIIS-SD,P], bypassing the condensate polishing building [EIIS.,SF], and_ reducing the turbine load (which reduced reactor power). These actions were taken to reduce the feedwater flow requirements to within the capacity of the single operating MFW pump (2-FW-P-1 B).

  • _Control room operators determined at 0416 hours0.00481 days <br />0.116 hours <br />6.878307e-4 weeks <br />1.58288e-4 months <br /> that the steam generator water levels could not be maintained above 20% and decided to manually trip the reactor. The reactor automatically tripped from 80%

power at 0416 hours0.00481 days <br />0.116 hours <br />6.878307e-4 weeks <br />1.58288e-4 months <br /> as the manual reactor trip was being initiated. The automatic reactor trip resulted from a low steam generator water le~el with a steam/feedwater flow mismatch reactor protection* signal from Steam Generator 2-RC-E-1A [EIIS-AB,SG]. This signal generates a reactor trip when the reactor protection system (EIIS-JC] senses a 20%

level in one steam generator coincident with steam flow being 0. 709 E6 pounds mass per hour greater than feedwater flow.

The reactor trip was followed by a turbine trip [EIIS-TA,TRB], generator trip [EIIS-TB,TG], and actuation of the Anticipated Transient Without SCRAM Mitigation System Actuation Circuitry (AMSAC), as designed.

Control room operators promptly initiated the appropriate emergency operating procedures. The reactor trip breakers [EIIS-JC,BKR] were verified to be open and the control rods [EIIS-JD,ROD], except for M-10, were verified to be properly inserted. The Individual . Rod Position Indicator (IRPI) and rod bottom light for control rod M-10 indicated the rod was not fully inserted. Emergency boration was not required by the emergency operating procedures. The M-1 O IRPI. [EIIS-JD,21] condition was later determined to be an indication problem.

NRC Form 366A (6-89)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION 16-89) APPROVED 0MB NO. 3150-0104 EXPIRES: 4/30/92 LICENSEE E V . REPORT (LER)

  • ATED BURDEN PER RESPONSE TO COMPLY WTH THIS RMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TD THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IP-530). U.S. NUCLEAR REGULATORY.COMMISSION, WASHINGTON, DC 20555, AND TO 1"HE PAPERWORK REDUCTION PROJECT 13150-0104). OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (11 DOCKET NUMBER (2)

LER NUMBER (6) PAGE (3)

Surry Power Station, Unit 2 TEXT (ff mon, spaCtJ M n,qui-1, UllO additional NRC Form 3!iSA'&) (17) 1 .O DESCRIPTION OF THE EVENT (Continued)

The motor driven and turbine driven auxiliary feedwater pumps

[EIIS-BA,P] started on low-low steam generator water level and initially provided flow to the steam generators. The main steam dumps

[EIIS-SB,V] automatically opened to admit steam directly to the main condenser [EIIS-SG,COND). Reactor Coolant System (RCS) average temperature (Tave) was reduced to 547 °Fat 0416 hours0.00481 days <br />0.116 hours <br />6.878307e-4 weeks <br />1.58288e-4 months <br /> and the main steam dumps closed, as designed. RCS Tave continued to decrease, reaching a minimum of 532 °F. RCS temperature subsequently stabilized at 547 °F (no load temperature).

The source range nuclear instrumentation detectors [EIIS-IG,DET]

automatically reinstated -as designed at 0427 hours0.00494 days <br />0.119 hours <br />7.060185e-4 weeks <br />1.624735e-4 months <br />. The intermediate range nuclear instrumentation indication was observed to be off-scale low.

  • Other safety systems performed as designed and the Unit was stabilized at hot shutdown. The NRC was notified pursuant to 10 CFR 50.72 at 0708 hours0.00819 days <br />0.197 hours <br />0.00117 weeks <br />2.69394e-4 months <br />. This event is being reported pursuant to 10 CFR 50. 73(a)(2)(iv) as an automatic actuation of the reactor protection system.

2.0 SAFETY CONSEQUENCES AND IMPLICATIONS This event resulted in no safety consequences or implications.

Appropriate operator actions were taken to ensure the performance of system automatic actions and to respond to abnormal conditions. The Unit was quickly brought to a stable, no-load condition. Therefore, the health and safety.of the public were not affected.

NRC Form 366A (6-891

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

LICENSEE EV. . REPORT (LERI *I APPROVED 0MB NO. 3150-0104 EXPIRES: 4/30/92 ATED BURDEN PER RESPONSE TO COMPLY WTH THIS RMATION 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 11) DOCKET NUMBER 12) LEA NUMBER (6) PAGE 13)

Surry Power Station, Unit 2 o 1s Io Io Io 12 1s 11 . 91 3 - 01 o I 2 - o I o o I 4 oF o 1s TEXT /If more spact1 ;,, mquired, use sddition11/ NRC Form 366A's/ 117) 3.0 CAUSE This event was caused by the occurrence of an instantaneous electrical ground on the 2-FW-P-1A inboard motor [EIIS-SJ,MO] and subsequent pump trip. This condition resulted in a main steam flow mismatch with feedwater flow and low Steam Generator 2-RC-E-1A water level. A Root Cause Evaluation is being performed to determine the cause of the 2-FW-P-1A inboard motor electrical ground.

The problems associated with IRPI M-1 O were previously evaluated by the Nuclear Steam Supply System vendor and determined to be the result of residual permeability in the control rod drive* mechanism housing. This phenomenon is caused by a combination of factors, including material composition and the decrease in reactor coolant system temperature following a reactor trip.

4~0 IMMEDIATE CORRECTIVE ACTION{Sl Control room operators promptly initiated the appropriate emergency operating procedures. The reactor trip breakers were verified to be open and the control rods, except for M-10, were verified to be properly inserted.

The Shift Technical Advisor monitored the critical safety function status trees to ensure that plant parameters remained within safe bounds.

5.0 ADDITIONAL CORRECTIVE ACTION{Sl The reactivity shutdown margin was calculated following the RCS cool down to ensure that Technical Specification and administrative shutdown margin limits were satisfied.

The M-1 O IRPI condition was determined to be an indication problem based on the results of a previous vendor evaluation. This condition has.

also been exhibited following previous Unit 2 reactor trips.

The 2-FW-P-1A inboard motor was shipped to a vendor to be repaired.

NRC Form 366A 16-89)

~----

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-89) APPROVED 0MB NO. 3150-0104

,... EXPIRES: 4/30/92 ATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE E V . REPORT (LER)

  • RMATION 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 (6) PAGE (3)

Surry Power Station, Unit 2 0 15 IO IO IO 12 1s 11 9 I 3 - 01 0 I 2 - 0 I O O I 5 OF O 15.

TEXT /ff more 11p11ce i& requimd, use addmono/ NRC Form 36llA '*! (171 s.o ADDITIONAL CORRECTIVE ACTION(S) (Continued)

The Unit was returned to service on June 20, 1993, at 2233 hours0.0258 days <br />0.62 hours <br />0.00369 weeks <br />8.496565e-4 months <br /> and operated at reduced power until July 6, 1993, when 2-FW-P-1A repairs were completed.

A Root Cause Evaluation was initiated on June 21, 1993, to determine the cause of the 2-FW-P-1 A inboard motor electrical ground and to evaluate the conditions experienced following the, reactor trip.

6.0 ACTIONS TO PREVENT RECURRENCE Recommendations resulting from the Root Cause Evaluation will be implemented, as appropriate.

7 .0 SIMILAR EVENTS LER S1-84-015-00 . Unit 1 Automatic Reactor Trip Resulting From Low .

"A" Steam Generator Water Level Following "A" Main Feedwater Pump Trip Due to Loss of Lubricating Oil System Pressure LER S2-84-015-00 Unit 2 Automatic Reactor Trip Resulting From Low "A" Steam Generator Water Level Following "A" Main Feedwater Pump Trip Due to Load Shedding LER S2-84-021-00 Unit 2 Automatic Reactor Trip Resulting From Opening of "B" Main Feedwater Pump Breakers Due to Excessive Drift in Pressure Switches LER S1-85-007-00 Unit 1 Automatic Reactor Trip Resulting From Opening of "A" Main Feedwater Pump Breakers Due to Valve Position Actuator Arm Design

~8.0 MANUFACTURER/MODEL NUMBER Manufacturer: Allis Chalmers Corporation Equipment: 4160 Volt Versa Pak Electric Motor NRC Form 366A (6-89)

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