ML18153A120

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LER 97-003-00:on 970219,loss of Pressurizer Heaters Resulted in Manual U1 Trip & U2 ESF Actuation.Caused by Loss of Group C Pressurizer Proportional Heaters.Reactor Trip Breakers Were Verified open.W/970319 Ltr
ML18153A120
Person / Time
Site: Surry Dominion icon.png
Issue date: 03/19/1997
From: Christian D
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
97-166, LER-97-003, LER-97-3, NUDOCS 9703260334
Download: ML18153A120 (6)
v • d • e


Text

10CFR50.73 Virginia Electric and Power Company Surry Power Station 5570 Hog Island Road Surry, Virginia 23883 March 19, 1997 U. S. Nuclear Regulatory Commission Serial No.: 97-166

  • Document Control Desk SPS:VLA Washington, D. C. 20555 Docket No.: 50-280 50-281 License No.: DPR-32 DRP-37

Dear Sirs:

Pursuant to Surry Power Station Technical Specifications, Virginia Electric and Power Company hereby submits the following Licensee Event Report (LER)

  • applicable to Surry Power Station Units 1 and 2.

REPORT NUMBER 50-280/50-281 /97-003-00 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, D. A. Christian Station Manager

- --* - - - - -----~--

9703260334 970319 Enclosure PDR ADOCK 05000280 S. P~,,

Commitments contained in this letter: None.

cc: Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 R. A. Musser NRG Senior Resident Inspector Surry Power Station 2G0112 IIIGIIHIDllllllllllll!mlll

  • I 7
  • I I  !!o I 2 C
  • NRC FORM 366 (4-95)
e. NUCLEAR REGULATORY COMMISSION *e APPROVED BY 0MB NO. 3150-0104 EXPIRES 4/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REQUEST: 50.0 HRS.

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSEE EVENT REPORT (LER) LICENSING PROCESS AND FED BACK TO INOUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33),

U.S. MJCLEAR REGULATORY COMMISSION, WASHINGTON, DC.

(See reverse for required number of digits/characters for each block) 20555-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MNlAGEMENT N8J BUDGET, WASHINGTON, DC 20503.

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

SURRY POWER STATION, Unit 1 05000- 280 1 OF5 TITLE(4)

Loss of Pressurizer Heaters Results in Manual U1 Trip and U2 ESF Actuation ENT DATE (5) LER NUMBER (6) RT DATE (7) OTHER FACILITIES INVOLVED 8)

SEQUENTIAL REVISION FACILITY NAME DOCUMENT NUMBER MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NUMBER Surry Unit 2 osooo~2a1 19 . FACILITY NAME DOCUME_NT NUMBER 02 97 97 -- 003 -- 0 03 19 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) X 50.73(a)(2)(1) 50.73(a)(2)(viii)

POWER 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50.73(a)(2)(x)

LEVEL (10) 100  % 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) I 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 TIELEPHONE NUMBER (Include Area Coda)

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 SYSTIEM COMPONENT- MANUF'ACTURER REPORTABLE TONPRDS TONPRDS E AB PMC R305 No B AB ALY B455 No SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAY. YEAR I YES .

IX I NO SUBMISSION

  • DATE (If yes, complete EXPECTED SUBMISSION DATE).

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

At 2040 hours0.0236 days <br />0.567 hours <br />0.00337 weeks <br />7.7622e-4 months <br /> on February 19, 1997, with Unit 1 at 100 percent power and Unit 2 at Hot Shutdown, a Technical Specificcttion 3.12.F.2 two hour Limiting Condition of_ Operation was entered due to Reactor Coolant System (RCS) -pressure being less than 2205 psig. An Abnormal Procedure was entered in response to the decreasing RCS pressure. At 2052 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.80786e-4 months <br />, a Unit 1 shutdown was commenced. At 2259 hours0.0261 days <br />0.628 hours <br />0.00374 weeks <br />8.595495e-4 months <br />, a manual reactor trip was initiated due to a continuing decrease in RCS pressure. Upon receipt of the reactor trip signal, the Reactor Protection System actuated and functioned as designed, and all control rods inserted into the core. Station operating personnel acted promptly to place the plant in a safe, hot shutdown condition in accordance with the proper procedures. The shutdown margin was calculated and the critical safety function status trees were monitored to verify that the unit conditions were acceptable. Plant response was as expected and the unit stabilized at hot shutdown. No conditions adverse to safety resulted from this event and the health and safety of the public were not affected. A detailed Reactor Trip Report and a Root Cause Evaluation is being performed tor this event. The cause of the trip was inability to maintain RCS pressure due to loss of the Group C pressuizer proportional heaters due to failure of the Robicon controller unit. Additional approved recommendatic'"'c: from the Root Cause Evaluation will be implemented. This. event is beinQ reported pursuant t6 10 CFR 50.73(a)(2)(iv) and 10 CFR 50.73(a}(2)(i).

NRC FORM 366 (4-95)

NRG FORM 366A (4-95) e e U.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

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

Surry Power Station, Unit 1 YEAR I SEQUENTIAL NUMBER I ::lEVISION NUMBER 05000- 280 97 -003- 0 20F5 TEXT (If more space is required, use additional copies of NRG Form 366A) (17)

1.0 DESCRIPTION

OF THE EVENT At 2040 hours0.0236 days <br />0.567 hours <br />0.00337 weeks <br />7.7622e-4 months <br /> on February 19, 1997, with Unit 1 at 100 percent power and Unit 2 at Hot Shutdown, a Technical Specification (TS) 3.12.F.2 two hour Limiting Condition of Operation was entered due to Reactor Coolant System (RCS) {EIIS-AB} pressure 'being less than 2205 psig. TSs state that pressurizer pressure must be maintained greater than or equal to 2205 psig. If pressure is less than 2205 psig and not restored within two hours, then thermal power shall be reduced to 5 percent rated power within the next four hours. Abnormal Procedure, 1-

- *AP-31.0o, **increasing or Decreasing RCS Pressure, was entered in response to the a

decreasing RCS pressure. At 2052 hours0.0238 days <br />0.57 hours <br />0.00339 weeks <br />7.80786e-4 months <br />; Unit 1 shutdown was commenced; At 2259 hours0.0261 days <br />0.628 hours <br />0.00374 weeks <br />8.595495e-4 months <br />, with the unit at approximately 53 percent reactor power, a manual reactor trip was initiated due to the continuing decrease in RCS pressure. Upon receipt of the manual reactor trip, the Reactor Protection System (RPS) functioned as designed and all control rods

{EIIS-AA-ROD} inse_rted into the, core as indicated by the rod bottom lighUndication and by all Individual Rod.Position Indicators (IRPls) indicating less than 10 steps. A shutdown margin c~lcul~tion verified adequate shutdown margin.

The RCS pressure and temperature stabilized at no load Tavg following the trip. No primary safety or power operated relief valves were actuated during the event. No secondary safety relief valves or power operated relief valves {EIIS-RV} actuated during the transient. All electrical busses transferred properly following the trip and all emergency diesel generators were operable. However, Reactor Coolant Pump A (RCP) {EIIS-AB-P} tripped during .the Station Service Bus transfer to the Reserve Station Service Transformers due to unexpected actuation of the speed sensing relays. RCP C was secured by the operating* team in accordance with procedures to prevent fu-rther RCS pressure decrease. ,Pressurizier Spray Valve B {EIIS-AB-V} had been previously isolated due to leakby. No safety injection occurred.

Since Unit 2 was in Hot Shutdown at the time of the trip, automatic load shedding was enabled, and the load shed affected station service electrical loads tripped as designed. At the time of the load shed sequence, the Unit 2, Main Feed Pump A (MFP) {EIIS-SJ} was in service, and the Unit 2 MFP B was tagged out for repairs. Unit 2 MFP Atripped on lo~d shed as designed. This resulted in a loss of both MFPs on Unit 2 and resulted in an automatic start of both Unit 2 Motor Driven Auxiliary Feedwater P:.imps.

  • At 2328 hours0.0269 days <br />0.647 hours <br />0.00385 weeks <br />8.85804e-4 months <br /> on 2/19/97, a one hour Non-Emergency report was made to the NRC in accordance with 10 CFR 50.72(b)(1 )(i)(A) for initiation of any plant shutdown required by TS.

This report also included the notification of the Unit 1 RPS actuation following the manual reactor trip. In ac..;ordance with 10 CFR 50.72(br(2)(ii), a separate four hour report was made at 0149 hours0.00172 days <br />0.0414 hours <br />2.463624e-4 weeks <br />5.66945e-5 months <br /> due to the automatic AFW actuation on Unit 2.

  • NRG FOFM 366A (4*95)

NRC FORM 366A (4-95) e eu.s. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

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

Surry Power Station, Unit 1 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 05000- 280 97 -003- 0 30F5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)

There were no radiation releases due to this event, nor were there any personnel injuries or contamination events.

This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv) for a condition that resulted in automatic actuation of the RPS for Unit 1 and automatic actuation of an Engineering Safety Feature (ESF) for Unit 1 and Unit 2. This event is also being reported pursuant to 10 CFR 50.73(a)(2)(i) due to completion of a TS required shutdown for Unit 1.

2.0 SAFETY CONSEQUENCES AND IMPLICATIONS Upor receipt of the _reactor trip signal, the RPS actuated and functioned as designed, and al.I control rods inserted into the core. Station operating personnel acted promptly to place the

.unit in a safe, hot shutdown, condition in accordance with t_he proper procedures. The Shift Technical Advisor calculated .the shutdown margin* and monitored the critical safety function status* trees to verify that the unit conditions were acceptable. Plant response was as expected and the unit stabilized at. hot shutdown. There were no radiation releases due to .

  • this event, nor were there any personnel injuries. or contamination events.: . No cond_itiohs adverse to safety resulted from this event and the health and safety of the public .were not affected. .. . ..

.3.0 CAUSE The cause of the trip was inability to maintain RCS pressure due to loss of the Group C pressurizer proportional heaters. This loss of the proportional pressurizer heaters was due to age related failures of a Robicon controller unit circuit card.

4.0 IMMEDIATE CORRECTIVE ACTION(S)

Following the reactor trip at 2259 hours0.0261 days <br />0.628 hours <br />0.00374 weeks <br />8.595495e-4 months <br /> on February 19, 1997, control room operators initiated the appropriate emergency operating procedures. The reactor trip breakers were verified open and control rods were verified inserted into the core.

The Shift Technical Advisor calculated the shutdown margin and monitored the critical safety function status trees to verify that the unit conditions were acceptable. *

  • Management conducted a post trip review meeting with the operating staff at 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on February 20, 1997.

NRG FORM :;ssA (4*95)

l NRC FORM 366A (4-95) e e.S. NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LEA)

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

Surry Power Station, Unit 1 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 05000- 280 97 -003- 0 40F 5 TEXT (If more space is required, use additional copies of NRC Form 366A) (17) 5.0 ADDITIONAL CORRECTIVE ACTION($)

A Root Cause Evaluation was initiated to investigate the Pressurizer Heater C control circuitry failure. The Robicon controller module was sent to the vendor. The investigation by the vendor revealed age related failures of capacitors, zenor diodes, and resistors in a controller circuit card. The components were replaced, and the controller module was calibrated. . The Robicon controller module was re-installed and satisfactory post maintenance tests returned the proportional heaters to service.

Upon completion of repairs to the pressurizer heaters and the Post Trip review, Unit 1 was taken critical at 0724 hours0.00838 days <br />0.201 hours <br />0.0012 weeks <br />2.75482e-4 months <br /> and on iine at 1342 hours0.0155 days <br />0.373 hours <br />0.00222 weeks <br />5.10631e-4 months <br /> on February 22, 1997. Unit 1 was

. returned to 100% reactor power at 1353 hours0.0157 days <br />0.376 hours <br />0.00224 weeks <br />5.148165e-4 months <br /> on February 23, 1997.

It was determined that the trip of the RCP A was due to the failure of Pro-Star speed sensing relays {EIIS-AB-RLY} due to a manufacturing deficiency. The relay was replaced: Software problems in the computations being performed by the Pro-Star relays were discovered following the Unit 1 trip. The specific causes and corrective actions will be implemented by the corrective action program.

As part of the Root Cause Evaluation, the equipment problems which occurred were evaluated in accordance with the Maintenance Rule and it was determined that two Maintenance Rule Functional Failures occurred. The first was the Robicon heater controller failure and the second was the failure of the speed sensing relays. These are being addressed in ac.cordance with maintenance rule program requirements.

6.0 ACTIONS TO PREVENT RECURRENCE /

A detailed Reactor Trip Report and a Root Cause Evaluation is being performed for this*

event. The Pro-Star relay and the Robicon controller failures are being evaluated. Additional approved recommendations from the Root Cause Evaluation will be implemented in accordance with the corrective action program.

An evaluation of preventive maintenance on the* Robicon controllers is being conducted.

7.0 SIMILAR EVENTS None NRC FORM 366A (4-95)

  • !FNRG FORM 366A (4-95) e eU.S. NUCLEAR REGULATORY COMMISSION

. LICENSEE EVENT REPORT (LER)

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

Surry Power Station, Unit 1 YEAR I SEQUENTIAL NUMBER I REVISION NUMBER 05000- 280 97 -003- 0 50F5 TEXT (If more space is required, use additional copies of NRC Form 366A} (17) 8.0 MANUFACTURER Robicon Controller Unit:

Manufacturer - Robicon Model - Series 413 Pro-Star Relays:

Manufacturer - ABB Model with 1.6 version software NRG FORM 366A (4-95)

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