05000281/LER-1996-004-02, :on 960606,turbine/reactor Trip Occurred.Caused by High Level in Steam Generator B.Placed Plant in Hot Shutdown Condition,Calculated Shutdown Margin & Monitored Critical Safety Function Status Trees

:on 960606,turbine/reactor Trip Occurred.Caused by High Level in Steam Generator B.Placed Plant in Hot Shutdown Condition,Calculated Shutdown Margin & Monitored Critical Safety Function Status Trees

ML18153A014
Person / Time
Site:	Surry
Issue date:	07/02/1996
From:	Christian D VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
96-347, LER-96-004-02, LER-96-4-2, NUDOCS 9607110221
Download: ML18153A014 (7)

LER-1996-004, on 960606,turbine/reactor Trip Occurred.Caused by High Level in Steam Generator B.Placed Plant in Hot Shutdown Condition,Calculated Shutdown Margin & Monitored Critical Safety Function Status Trees

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(iv), System Actuation 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(x)
2811996004R02 - NRC Website

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REGULA~ INFORMATION DISTRIBUTIO.YSTEM (RIDS)

=.,-~: ACCESSION NBR:9607110221 DOC.DATE: 96/07/02 NOTARIZED: NO DOCKET#

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

Virginia Power (Virginia Electric & Power Co.)

RECIP.NAME RECIPIENT AFFILIATION SUBJECT: LER 96-004-00:on 960606,turbine/reactor trip occurred.Caused by high level in Steam Generator B.Placed plant in hot shutdown condition,calculated shutdown margin & monitored critical safety function status trees.W/960702 ltr.

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SIZE: lo TITLE: 50.73/50.9 Licensee Event Report (LER),-Ynciden~Rpt, et-c-.~~~~

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L July 2, 1996 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555

Dear Sirs:

10CFR50.73 Virginia Electric and Power Company Surry Power Station 5570 Hog Island Road Surry, Virginia 23883 Serial No.:

96-347 SPS:JDKNLA Docket No.: 50-281 License No.: DPR-37 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.

REPORT NUMBER 50-281 /96-004-00 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 pc:

Regional Administrator 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 M. W. Branch NRC Senior Resident Inspector Surry Power Station

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9607110221 960702 PDR ADOCK 05000281 s

PDR

NRG FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 (5-92)

EXPIRES 5/31/95 I

ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSING 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 (See reverse for required number of digits/characters for each block)

THE PAPERWORK REDUCTION PROJECT (3150*0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON. DC 20503.

FACILITY NAME (1)

DOCKET NUMBER (2)

II PAGE (3)

SURRY POWER STATION, Unit 2 05000 - 281 1 OF 5 TITLE (4)

Turbine/Reactor Trip due to High Level in the Steam Generator EVENT DATE 5)

LER NUMBER (6 REPORT DATE 7 OTHER FACILITIES INVOLVED 8)

MONTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR FACILITY NAME DOCKET NUMBER NUMBER NUMBER Surry Unit 1 05000- 280 06 06 96 96

-- 004 --

0 07 02 96 FACILITY NAME DOCKET NUMBER 05000 -

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

MODE (9) 20.402(b) 20.405(c)

X 50.73(a)(2)(iv) 73.71(c)

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

LEVEL (10) 16%

20.405(a)(1 )(ii) 50.36(c)(2) 50.73(a)(2)(vii)

OTHER 20.405(a)(1 )(iii) 50.73(a)(2)(i) 50.73(a)(2)(viii)(A)

(Specify in Abstract below and

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20.405(a)(1 )(iv)

50. 73(a)(2)(ii) 50.73(a)(2)(viii)(B) in Text. NRG Form 366A) 20.405(a)(1 )(v) 50.73(a)(2)(iii) 50.73(a)(2)(x)

LICENSEE CONTACT FOR THIS LEA 12)

NAME I (804r357~3n34ing Area Code)

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

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE It>>****

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE TONPRDS TONPRDS h'*

/... '

SUPPLEMENTAL REPORT EXPECTED (14)

I EXPECTED I MONTH I DAY I YEAR I YES (If yes, complete EXPECTED SUBMISSION DATE)

XI NO I

SUBMISSION DATE (15)

I I

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

On June 6, 1996, power escalation was in progress for Unit 2 upon completion of a scheduled refueling outage. Control Room Operators were transferring from manual feedwater flow control with the Main Feedwater Regulation Valve (MFRV) Bypass Valves to automatic feedwater control with the MFRVs. At 2344 hours0.0271 days <br />0.651 hours <br />0.00388 weeks <br />8.91892e-4 months <br />, with Unit 2 at 16% power, a high-high Steam Generator B level signal automatically tripped the running main feedwater pump and the main turbine.

The turbine trip generated a reactor trip signal.

The Reactor Protection System actuated and functioned as designed, and all control rods inserted into the core. Following the trip, Control Room Operators acted promptly to place the plant in a safe, hot shutdown condition in accordance with the 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.

During this event the health and safety of the public were not affected. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv).

NRG FORM 366 (5-92)

.

• rr==:=".::=::='='=::c:=:=':=============9 NRC FORM 366 I *

(5-92)

.S. NUCLEAR REGULATORY COMMISSION LICENSING EVENT REPORT (LER)

(See reverse for required number of digits/characters for each block)

FACILITY NAME C1l DOCKET NUMBER (2)

APPROVED BY 0MB NO. 3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD 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 MANAGEMENT AND BUDGET, WASHINGTON. DC 20503.

LER NUMBER (6)

PAGE C3l YEAR SEQUENTIAL NUMBER REVISION NUMBER SURRY POWER STATION, Unit 2 05000 - 281 96

-- 004 --

0 2 OF 5 TEXT (If more space is required. use additional copies of NRG Form 366A) (17) 1.0 DESCRIPTION OF THE EVENT On June_ 6, 1996, power escalation was in progress for Unit 2 upon completion of a scheduled refueling outage.

Reactor power was 16% based on Reactor Coolant System (RCS) [EIIS-AB] Loop delta T indications, and RCS average temperature was being controlled at 553 degrees-Fahrenheit using control rods in manual and Main Steam Dump Valves [EIIS-SB-TCV] in the steam pressure mode. The main Turbine [EIIS-TA] was rotating at 1800 rpm with turbine control in the operator automatic mode.

Preparations were being made to close the Main Generator [EIIS-TB] output breakers.

Between 2331 and 2333 hours0.027 days <br />0.648 hours <br />0.00386 weeks <br />8.877065e-4 months <br />, Control Room Operators were transferring from manual feedwater flow control with the Main Feedwater Regulation Valve (MFRV) [EIIS-SJ-ISV]

Bypass Valves to automatic feedwater control with the MFRVs. MFRVs A and C were opened first. Steam Generator (SG) [EIIS-AB-SG] water levels were oscillating in all three SGs. The Reactor Operator (RO) assigned to SGs A and B was stabilizing the level on SG A when he noticed SG B level increasing. MFRV B was verified closed, and the demand for the MFRV Bypass Valve B was lowered to reduce feedwater flow.

Level continued to increase, so deniand was lowered to fully close the MFRV B Bypass Valve. Level continued to increase.

At 2344 hours0.0271 days <br />0.651 hours <br />0.00388 weeks <br />8.91892e-4 months <br />, a high-high SG B level signal automatically tripped the running Main Feedwater Pump [EIIS-SJ-P] and the Main Turbine.

The turbine trip generated a reactor trip signal. The Reactor Protection System (RPS) actuated and functioned as designed, and all control rods inserted into the core.

The Motor-Driven Auxiliary Feedwater Pumps [EIIS-BA-P] automatically started as designed. Control Room Operators acted promptly to place the plant in a safe, hot shutdown condition in accordance with the 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. Individual Rod Position Indicator (IRPI) [EIIS-AA-ZI] F-8 initially indicated 13 steps then gradually drifted down to less than 10 steps.

The RCS cooled down below the 547 degree Fahrenheit (no load temperature) and reached a minimum of 539 degrees Fahrenheit.

RCS temperature subsequently stabilized at 547 degrees Fahrenheit after the Main Steam Dump Valves closed and the Auxiliary Feedwater (AFW) Pumps [EIIS-BA-P] were secured.

NRC FORM 366

/5-92)

. NUCLEAR REGULATORY COMMISSION LICENSING EVENT REPORT (LER}

(See reverse for required number of digits/characters for each block)

FACILITY NAME f1\\

DOCKET NUMBER 12\\

SURRY POWER STATION, Unit 2 05000 - 281 TEXT (If more space is required, use additional copies of NRG Form 366A) (17)

APPROVED BY 0MB NO. 3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD 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 MANAGEMENT AND BUDGET. WASHINGTON, DC 20503.

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL NUMBER REVISION NUMBER 96

• • 004 **

0 3 OF 5 In accordance with 10 CFR 50.72(b)(2)(ii), a 4-hour non-emergency report to the Nuclear Regulatory Commission was made at 0325 hours0.00376 days <br />0.0903 hours <br />5.373677e-4 weeks <br />1.236625e-4 months <br /> due to the automatic RPS actuation. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv) due to automatic actuation of the RPS.

2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS

Upon receipt of the reactor trip signal, the RPS actuated and functioned as designed, and all control rods inserted into the core. Control Room Operators acted promptly to place the plant in a safe, hot shutdown condition in accordance with the 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.

No conditions adverse to safety resulted from this event and the health and safety of the public were not affected.

3.0 CAUSE OF THE EVENT

The cause of the reactor trip was a high level in SG 8. This level swell occurred when the SG A stopped steaming and SG 8 picked up steam flow. This level swell combined with the initial level in the SG B caused the level to increase above the 75% high-high SG level trip setpoint. In response to increasing level in SG 8, the RO first lowered and then attempted to secure feedwater flow 30 seconds before the trip.

Investigations following the subsequent startup revealed that erratic Steam Dump Valve operation also contributed to the trip.

Several seconds of oscillating steam dump demand occurred at approximately 9 minutes and again at 3.5 minutes before the trip.

Steam Dump Valves 2-MS-TCV-205A and 8 oscillated in response to the demand signal. These steam dump oscillations increased the magnitude of the SG oscillations.

Also, a zero shift in the controller for 2-'FW-F9R-2558 MFRV Bypass Valve may have contributed to the level increase in SG 8.

The design of the Feedwater Control System requires manual operation of the MFRV Bypass Valve during startup. There are numerous variables involved in controlling SG levels; these variables along with the equipment malfunctions discussed in the previous paragraphs, resulted in the Control Room Operators being challenged to the point where SG level could not be successfully controlled.

u NRC FORM 366

,5-92)

. NUCLEAR REGULATORY COMMISSION LICENSING EVENT REPORT (LER)

(See reverse for required number of digits/characters for each block)

FACILITY NAME 11\\

DOCKET NUMBER 12\\

SURRY POWER STATION, Unit 2 05000 - 281 TEXT (If more space is required, use additional copies of NRG Form 366A) (17) 4.0 IMMEDIATE CORRECTIVE ACTION(S)

APPROVED BY 0MB NO. 3150-0104 EXPIRES 5/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD 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 MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

LEA NUMBER 16\\

PAGE 13\\

YEAR SEQUENTIAL NUMBER REVISION NUMBER 96

• • 004 --

0 4 OF 5 Following the trip, Control Room Operators acted promptly to place the plant in a safe, hot shutdown condition in accordance with the 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.

The SGs were initially fed by the AFW System and were later transferred to the MFRV Bypass Valves.

Decay heat was removed by the Main Steam Dump Valves discharging to the Main Condenser.

5.0 ADDITIONAL CORRECTIVE ACTION(S)

Investigations following the subsequent startup revealed that Steam Dump Valve operation was erratic while in the steam pressure mode.

This erratic operation occurred prior to the trip in the form of spikes in the demand to 2-MS-TCV-205A and B.

During power ascension on June 7, 1996, the steam dump valves were observed to modulate in response to RCS temperature changes. Subsequent investigation by l&C determined that both RCS average temperature and steam pressure voltages were being fed to the valves due to a faulty relay in the steam dump control system. This relay was replaced. The steam dump valves were tested in both the RCS average temperature and steam pressure modes with satisfactory results.

During startup, MFRV Bypass Valve 2-FW-FCV-255B was noted as requiring less demand than the MFRV Bypass Valve A to provide the same flow (e.g., 60% demand on MFRV B provided the same flow as 80% demand on MFRV A). The demand was lowered to zero for MFRV B Bypass Valve in order to isolate feedwater flow to the SG B but level continued to increase. Following the trip, l&C technicians determined that the valve was not going fully closed on demand; however, when instrument air to the positioner was isolated, the valve would move further in the closed direction. Further investigation determined a

slightly elevated pneumatic output from the Electric/Pneumatic (E/P) transducer.

When the cover was removed from the E/P transducer, the input signal wires were found lying against the force beam causing elevated pneumatic output. The wires were relocated away from the force beam and the valve fully closed with E/P input at its minimum.

" rr==N=:=R===c""=F===o===R=M=3=66=========1

.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 EXPIRES 5/31/95 I

f (5-92)

LICENSING EVENT REPORT (LER)

(See reverse for required number of digits/characters for each block)

FACILITY NAME (1)

DOCKET NUMBER (2)

ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD 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 MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.

LER NUMBER !6l YEAR SEQUENTIAL NUMBER REVISION NUMBER PAGE(3l SURRY POWER STATION, Unit 2 05000 - 281 96

-- 004 --

0 5oF5 TEXT (If more space is required, use additional copies of NRG Form 366A) (17)

Further investigation of the MFRV Bypass Valve leakage and the Steam Dump Valve erratic operation is ongoing. Further corrective action related to these items may be recommended as a result of this investigation and will be tracked by the corrective action process.

A calibration check was performed on IRPI F-8 and it was found to be indicating high.

IRPI F-8 was adjusted satisfactorily.

6.0 ACTIONS TO PREVENT RECURRENCE None.

7.0 SIMILAR EVENTS

LER S2-86-003 - Turbine trip/reactor trip from high SG Level due to MFRV bypass valve failing to close on demand

8.0 ADDITIONAL INFORMATION

Unit 1 was operating at 100% during this event.