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LER 97-002-00:on 961213,automatic Reactor Trip Occurred During Planned Shutdown.Caused by Steam Flow/Feedwater Flow Mismatch.Rps Functioned as Designed & Plant Placed in Hot Shutdown
	ML18153A095
Person / Time
Site:	Surry
Issue date:	01/02/1997
From:	Christian D; VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	LER-97-002, LER-97-2, NUDOCS 9701140322
	Download: ML18153A095 (6)
	v • d • e

CAT*EGORY 1 REGULAT. INFORMATION DISTRIBUTIO~fSTEM (RIDS) l ACCESSION NBR:9701140322 DOC.DATE: 97/01/02 NOTARIZED: NO DOCKET#

FACIL:50-281 Surry Power Station, Unit 2, Virginia Electric & Powe - 05000281 AUTH.NAME AUTHOR AFFILIATION CHRISTIAN,D.A. Virginia Power (Virginia Electric & Power Co.)

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 97-002-00:on 961213,automatic reactor trip occurred during planned shutdown.Caused by steam flow/feedwater flow C

mismatch.RPS functioned as designed & plant placed in hot shutdown.

A DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR _L ENCL~' SIZE:__,.!:.)-...;*~~~

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

NOTES: 05000281E G

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL LTTR ENCL 0 ID CODE/NAME PD2-l PD 1 1 EDISON,G. 1 1 R

INTERNAL: ACRS 1 1 2 2 y AEOD/SPD/RRAB 2 2 ~B, 1 1 NRR/DE/ECGB 1 1 * .- EELB 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 1

NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 RES/DET/E_IB 1 1 RGN2 FILE 01 1 1 D

-EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J H 1 1 ---"°_c:Q ___ \_

NOAC MURPHY,G.A 1 1 NOAC POORE,W. 1 1 NRC FDR 1 1 NUDOCS FULL TXT 1 1 C NOTES: 1 1 u M

E N

T NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-S(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 27 ENCL 27

e IOCFRS0.73 Virginia Electric and Power Company Surry Power Station 5570 Hog Island Road Surry, Virginia 23883 January 3, 1997 U. S. Nuclear Regulatory Commission Serial No.: 97-002 Document Control Desk SPS:JDK 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 tiereby submits the following Licensee Event Report applicable to Surry Power Station Unit 2.

REPORT NUMBER 50-281 /96-006-00 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, D. A. Christian Station Manager Enclosure cc: Regional Administrator 101 Marietta Street NW Suite 2900 Atlanta, Georgia 30323 R. A. Musser NRC Senior Resident Inspector Surry Power Station 9701140322 970102 ~

PDR ADOCK 05000281 S PDR 140125

NRC FORM366 (4-95) '

U.S. NUCLEAR REGULATORY COMMISSION e APPROVED BY 0MB NO. 3150-0104 EXPIRES 4/30/98 ESTIMAlcO BIRlEN PER RESPONSE 10 COMPLY WITH THIS MANDATORY INR'.lRMATlON COlLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER) REPORTED LESSONS LEARNED ARE INCORPORAlcO INTO THE LICENSING PROCESS AND FED BACK 10 INDUSTRY. FORWARD COMMENTS REGARDING BURDEN EST1MA1E 10 THE INFORMATION 11<<:J RECORDS MANAGEMENT BRANCH (T-6 F33),

(See reverse for required number of digits/characters for each block) U.S. NUCl.EAR IEGU.ATORY COMMISSION, WASHINGTON, DC 205-1. NlD 10 THE PAPERWORK REDUCTION PROJECT (3150-0104). CFflCE OF MANAGEMENT AND

BUDGET, WASHINGTON. DC 20503.

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

SURRY POWER STATION , Unit 2 05000 - 281 1 OF4 TITLE (4)

Auto Rx Trip due to Stm/Feed Flow Mismatch Coincident with a Low S/G Level EVENT DATE LER NUMBER (6) REPORT DATE (7) *oTHER FACILITIES INVOLVED 8)

SEQUENTIAL REVISION FACILITY NAME DOCUMENT NUMBER MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER 05000-FACILITY NAME DOCUMENT NUMBER 12 13 96 96 -- 006 -- 00 01 02 97 05000-OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check one or more) (11)

MODE (9) N 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii)

- ER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

LEVEL (10) 11 *.* 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 20.2203(a)(2)(ii) 20.2203(a)(4) X 50.73(a)(2)(iv) OTHER 20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v) Specify in Abstract below 20.2203(a)(2)(iv) 50.36(c)(2) 50.73(a)(2)(vii) or in NRC Form 366A LICENSEE CONTACT FOR THIS LER 12)

NAME I TELEPHONE NUMBER (Include Area Code)

D. *A. Christian, Station. Manager (757) 365-2000 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TONPRDS I

SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY YEAR SUBMISSION I YES IX I NO DATE (II yes. complete EXPECTED SUBMISSION DATE).

ABSTRACT (L1mrt to 1400 spaaes. 1.e., approximately 15 single-spaced typewritten lines) (16)

At 0233 hours0.0027 days 0.0647 hours 3.852513e-4 weeks 8.86565e-5 months on December 13, 1996, with Unit 2 at 11 % power, an automatic reactor trip occurred during a planned shutdown. The reactor trip was caused by a steam flow/feedwater flow mismatch in coincidence with low Steam Generator (S/G) level in the 'A' S/G.

Upon receipt of the automatic reactor trip, the Reactor Protection System (RPS) functioned as designed and rod bottom lights were lit for all control rods except 2-RD-RPI-P-6. All electrical buses remained energized by off-site power and all emergency diesel generators were operable. Station operating personnel promptly placed the plant in a stable, hot shutdown condition* in accordance with the. appropriate procedures. The shutdown margin was calculated and found to be satisfactory. The health and safety of th~ public were not affected by this event since all plant parameters remained within the normal range and all required safety equipment operated as designed.

NRC FORM 366 (4*95)

NRC FO~M 366A U.S. NUCLEAR REGULATORY COMMISSION

'(4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6l PAGE (3)

Surry Power Station, Unit 2 05000-281 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 96 -- 006-- 00 20F 4 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

1.0 DESCRIPTION

OF THE EVENT At 0233 hours0.0027 days 0.0647 hours 3.852513e-4 weeks 8.86565e-5 months on December 13, 1996, with Unit 2 at 11 % power, an automatic reactor trip occurred during a planned shutdown. The reactor trip was caused by a steam flowlfeedwater flow mismatch in coincidence with low Steam Generator (SIG) level in the 'A' S/G [EIIS-JE]. Automatic actuation of the Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) occurred as expected, including Turbine Trip by Reactor Trip and Auxiliary Feedwater initiation. Station operating personnel promptly placed the plant in a stable, hot shutdown condition in accordance with the appropriate procedures. The shutdown margin was calculated and found to be satisfactory. The Shift Technical Advisor monitored the critical safety function status trees to verify that satisfactory unit conditions were maintained. -*

In accordance with 10CFR50.72(b)(2)(ii), a 4-hour Non-Emergency Report to the NRG operations center was made at 0425 hours0.00492 days 0.118 hours 7.027116e-4 weeks 1.617125e-4 months due to the Reactor Protection System (RPS) and automatic ESF actuations. This event is being reported pursuant to 10 CFR 50. 73(a)(2)(iv).

Following completion of the repairs during the planned shutdown, Unit 2 was taken critical at 0348 hours0.00403 days 0.0967 hours 5.753968e-4 weeks 1.32414e-4 months and was placed on the line at 1243 hours0.0144 days 0.345 hours 0.00206 weeks 4.729615e-4 months on December 23, 1996. The unit was returned to 100% reactor power, 858 MWe at 1548 hours0.0179 days 0.43 hours 0.00256 weeks 5.89014e-4 months on December. 24, 1996.

2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS Upon receipt of the automatic reactor trip, the RPS functioned as designed and rod bottom lights were lit for all control rods except 2-RD-RPI-P-6. Additionally, four Individual Rod Position Indicators (IRPI) [EIIS-JD-ZI] {2-RD-RPI-D4, F-6, M-4 and P-6) indicated between 10 and 32 steps. Subsequent rod drop tests verified that all Control Rods fully inserted into the core and that the observed irregularities were limited to position indicators. The health and safety of the public were not affected by this event since all plant parameters remained within the normal range and all required safety equipment operated as designed.

3.0 CAUSE OF THE EVENT The preliminary root cause of the reactor trip is determined to be the interface between the operator and the system used to control SIG water level [EIIS-JB] at low power. At low power levels, indication of feedwater flow to the steam generators is not available and feedwater control and steam dumps are in manual in accordance with operating procedures.

While operations personnel are trained for this evolution and normally perform well, the operating team was unsuccessful in mitigating a level decrease prior to a reactor trip on steam flow greater than feedwater flow coincident with low SIG water level. Manual control of SIG level at low power is difficult due to feedwater flow and steam flow instrumentation limitations, and the dynamic steam generator operating characteristics. Although this is a NRr. F()RI.A '\AAA 14-<l'il

NRG FQRM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LE~

TEXT CONTINUATION FACILITY NAME (1) DOCKET LEA NUMBER (6l PAGE (3)

Surry Power Station, Unit 2 05000-281 YEAR l SEQUENTIAL NUMBER l REVISION NUMBER 96 -- 006-- 00 30F 4 TEXT (If more space is required, use additional copies of NRG.Form 366A) (17) difficult evolution, it is routinely performed successfully. The last occurrence of a similar event not caused by equipment problems occurred in 1989.

The operator's response to the dynamics of the SIG prior to the trip are consistent with operating procedures, training and philosophy for this evolution. The preliminary root cause did not identify any inappropriate operator actions and no mechanical equipment failure contributed to the trip.

4.0 IMMEDIATE CORRECTIVE ACTION(S)

Following the automatic reactor trip, the operating team performed the immediate actions of Emergency Operating Procedure 2-E-O, "Reactor Trip or Safety Injection." The reactor was verified to be tripped (rod bottom lights lit except for* P-6, reactor trip and bypass breakers open, neutron flux decreasing). However, IRPls D-4, F-6, M-4 and P-6 did not indicate less than 1O steps initially. The four affected IRPls indicated 32, 20, 12 and 18 steps, respectively. The Source Range Nls automatically re-energized as expected. The turbine was manually tripped at 0233 hours0.0027 days 0.0647 hours 3.852513e-4 weeks 8.86565e-5 months . The minimum RCS temperature of 539 degrees Fahrenheit was reached at 0240 hours0.00278 days 0.0667 hours 3.968254e-4 weeks 9.132e-5 months .

At 02S 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , the team transitioned to 2-ES-0.1, "Reactor Trip Response." The Shift Technical Advisor monitored the Critical Safety Function Status Trees. The operating team paused at Step 4 of 2-ES-0.1 "Verify all control rods - less than or equal to 10 steps" and evaluated IRPI indications. Four IRPls were identified to be between 10 and 32 steps. As directed by procedure, additional boron was injected into the RCS to compensate for the potential of the four control rods not being fully inserted into the core.

5.0 ADDITIONAL CORRECTIVE ACTION(S)

Hot rod drop tests (2-NPT-RX-014) were performed for all control rods. The data collected was satisfactory with no anomalies noted. Specifically, rod drop times were within the Technical Specification limits and were consistent with previous data. These test results confirmed that the control rods were fully inserted into the core and the observed irregularities were limited to IRPI problems. The following actions were taken to correct the IRPI problems:

2-RD-RPI-D4 is located in Control Bank C. Following the trip, this IRPI indicated 32 steps and its rod bottom light was illuminated. The rod bottom light is expected to illuminate at 20 steps. The rod bottom bistable was calibrated and found to be within specifications indicating that the actual rod position was below 20 steps. The IRPI pointer was found rubbing on the faceplate. The IRPI was replaced. Additionally, the signal conditioning module required a zero and span adjustment. The IRPI was calibrated satisfactorily and returned to service.

NRG FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEEEVENTREPORT (LE~

TEXT CONTINUATION FACILITY NAME (1) DOCKET LER NUMBER (6) PAGE (3)

Surry Power Station, Unit 2 05000-281 VEAR I SEQUENTIAL NUMBER I REVISION NUMBER 96 --006-- 00 40F4 TEXT (If more space is required, use additional copies of NRG F.orm 366A) (17) 2-RD-RPI-F6 is located in Control Bank D. Following the trip, this IRPI indicated 20 steps. The signal . conditioning module was replaced. The IRPI was calibrated satisfactorily and returned to service.

2-RD-RPI-M4 is located in Control Bank C. Following the trip, this IRPI indicated 12 steps. The signal conditioning module was calibrated satisfactorily to the baseline calibration voltages. The indication returned to 18 steps when the IRPI was returned to service. This behavior indicates that new calibration voltages were needed. At Hot Shutdown, new bas_eline calibration voltages were developed in accordance with the Analog Rod Position Indication Sys~em Channel Calibration procedure. The IRPI was re-calibrated satisfactorily using the new voltages and returned to service.

2-RD-RPI-P6 is located in Control Bank A. Following the trip, this IRPI indicated 18 steps and the rod bottom light was not illuminated. The signal conditioning module required zero and span adjustmonts. When the mechanical zero cou!d not be adjusted, the IRPI was replaced. The rod bottom bistable was also found to trip out of tolerance, so the bistable was replaced. The IRPI was calibrated satisfactorily and returned to service.

6.0 ACTIONS TO PREVENT RECURRENCE A Category I Root Cause Evaluation is being conducted to identify the direct and contributing causes of this event. Recommendations from the Root Cause Evaluation will be reviewed by management. Approved recommendations will be implemented.

7.0 SIMILAR EVENTS LEA 82-96-004-00, Unit 2 Trip Due to an Automatic High-High S/G Level Trip on the "B" SIG. This event was due to equipment problems with the Feedwater Regulating System.

8.0 ADDITIONAL INFORMATION Unit 1 was operating at 100% and was not affected by this event.

NAr. FOAM 'U\RA /4.Q<;\

ML18153A095

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