ML18153A268

LER 98-013-01:on 981122,turbine/reactor Tripped on High Due to Short Circuit in Summator for MSL C Loop Channel III Flow Transmitter.Replaced 1-MS-FT1494 Summator & Module Repair Procedure Revised.With 9903190 Ltr

ML18153A268
Person / Time
Site:	Surry
Issue date:	03/19/1999
From:	Grecheck E VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
98-726A, LER-98-013, LER-98-13, NUDOCS 9903300383
Download: ML18153A268 (6)

Text

10CFR50.73 Virginia Electric And Power Company Surry Power Station 5570 Hog Island Road Surry, Virginia 23883 March 19, 1999 U. S. Nuclear Regulatory Commission Serial No.: 98-726A Attention: Document Control Desk SPS: BCB Washington, D. C. 20555 Docket No.: 50-280 License No.: DPR-32

Dear Sirs:

Pursuant to 10 CFR 50.73, Virginia Electric and Power Company hereby submits the following Licensee Event Report applicable to Surry Power Station Unit 1.

Report No. 50-280/1998-013-01 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, E. S. Grecheck Site Vice President Enclosure Commitments contained in this letter: None 9903300383 990319 PDR ADOCK 05000280 S PDR

e cc: U. S. Nuclear Regulatory Commission Region II Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303 Mr. R. A. Musser NRG Senior Resident Inspector Surry Power Station

e e NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB NO. 3150-0104 EXPIRES 06/30/2001 (6-1998) Estimated burden per response to comply with this mandatory information collection request: 50 hrs. Reported lessons learned are incorporated into LICENSEE EVENT REPORT (LER) the licensing process and fed back to industry. Forward comments regarding burden estimate to the Records Management Branch (T-6 F33),

U.S. Nuclear Regulatory Commission. Washington, DC 20555-0001, and to (See reverse for required number of the Paperwork Reduction Project (3150-0104), Office of Management and digits/characters for each block) Budget, Washington, DC 20503. If an information collection does not display acurrently valid 0MB control number. the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

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

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

Turbine/Reactor Trip on High Steam Generator Level Due to Instrument Failure EVENT DATE (5) LER NUMBER 6) REPORT DATE (7) OTHER FACILITIES INVOLVED 8)

MONTH DAY YEAR YEAR ISEQUENTIAL NUMBER REVISION NUMBER MONTH DAY YEAR FACILITY NAME DOCKET NUMBER 05000 --

FACILITY NAME DOCKET NUMBER 11 22 1998 1998 - 013 -- 01 03 19 1999 n~nnn --

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

MODE {9) N I 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2l(il 50.73(aH2Hviii) 20.2203(a)(1) 20.2203(a)(3)(i) 50.73(a)(2)(ii) 50. 73(a)(2)(x) 0 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)(2Hivl OTHER

-NAME 20.2203(a)(2)(iii) 20.2203(a)(2)(iv) 50.36(c)(1 l 50.36(c)(2)

LICENSEE CONTACT FOR THIS LER (12)

E. S. Grecheck, Site Vice President

50. 73(aH2Hv) 50.73(a)(2)(vii)

TELEPHONE NUMBER (Include Area Code)

(757) 365-2000 Specify in Abstract below or in NRC Form 366A COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

REPORTABLE lie 5 REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX i ~' CAUSE SYSTEM COMPONENT MANUFACTURER TO EPIX Westinghouse ~ 'I B JB IMOD Electric Coro. y SUPPLEMENTAL REPORT EXPECTED (14) 11 MONTH DAY YEAR EXPECTED jYES SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE). Ix NO DATE (15)

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

On November 22, 1998, at 04:30, with Unit 1 at 28% power, control room annunciators alarmed indicating an incorrect level in steam generator (SG) "B" and a feedwater/steam flow mismatch. In response to an apparent increase in steam flow, the main feedwater regulating valve, 1-FW-FCV-1488, began to open further. Although a control room operator began to manually control 1-FW-FCV-1488 to try to prevent a high level condition, the "B" SG reached its high level turbine trip setpoint. The Unit 1 turbine automatically tripped, which was immediately followed by an automatic reactor trip. When the reactor coolant system cooled to the low average temperature (low Tavg) setpoint of 543°F, a safety injection (SI) actuation occurred. The SI initiation resulted from the low T avg condition coincident with an apparent high steam flow condition. The SI actuation was spurious since it resulted from an invalid signal. The event was caused by a short circuit in the summator for the main steam line "C" loop channel 111 flow transmitter. Root Cause Evaluation recommendations, designed to prevent the recurrence of a similar event, were implemented. The NRC was notified pursuant to 10 CFR 50.72 (b)(2)(ii) on November 22, 1998 at 07:20. This report I

is being submitted pursuant to 10 CFR 50.73 (a)(2)(iv).

NRC FORM 366 (6-1998)

e e NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) DOCKET (2) LEA NUMBER 16) PAGE (3)

YEAR I SEQUENTIAL NUMBER REVISION NUMBER SURRY POWER STATION, Unit 1 05000 -- 280 1998 - 013 -- 01 2 OF 4 TEXT (If more space is required, use additional copies of NRG Form 366A) (17)

1.0 DESCRIPTION

OF THE EVENT On November 22, 1998, at 04:30, with Unit 1 at 28% power, control room annunciators

[EIIS-IB] alarmed indicating an incorrect level in steam generator (SG) "B" [EIIS-AB,SG]

and a difference between the feedwater and steam flow parameters. In response to an apparent increase in steam flow, the main feedwater regulating valve [EIIS-SJ,FCV],

• 1-FW-FCV-1488, began to open further. Although a control room operator began to manually control 1-FW-FCV-1488 to fry to prevent a high level condition, the "B" SG reached its high level turbine trip setpoint. As designed, the Unit 1 turbine [EIIS-TA,TRB]

automatically tripped, which was immediately followed by an automatic reactor trip

[EIIS-JC].

The auxiliary feedwater pumps [EIIS-BA-P] started as designed and provided flow to the SGs. When the reactor coolant system (RCS) cooled to the low average temperature I

(low Tavg) setpoint of 543°F, a safety injection (SI) [EIIS-BQ] actuation occurred. The SI initiation resulted from the low Tavg condition coincident with an apparent high steam flow condition (one steam flow instrumentation channel,[EIIS-JB,CHA] for each of SGs "A" and "C" had been placed in the tripped condition, prior to the event, to facilitate system maintenance). Emergency diesel generators (EDG) [EIIS-EK,EDG] Nos. 1 and 3 automatically started upon SI initiation. Following verification that the SI actuation had been spurious, the SI was terminated and the EDGs were shutdown.

The RCS reached a minimum temperature of approximately 535°F and subsequently stabilized at 547°F. The reactivity shutdown margin was calculated following the RCS cooldown to ensure that Technical Specification and administrative shutdown margin limits were satisfied.

The following discrepancies were noted during the post-trip response:

• Intermediate position was indicated in the control room when 1-FW-FCV-1488 was fully closed.

• "C" SG steam line pressure indication was lower than the actual value.

• "B" SG steam line flow was indicated when no flow was present.

The NRC was notified pursuant to 10 CFR 50.72 (b)(2)(ii) on November 22, 1998 at 07:20. This report is being submitted pursuant to 10 CFR 50.73 (a)(2)(iv) as an event that resulted in the automatic actuation of engineered safety features and the reactor protection system.

NRG FORM 366A (6-1998)

e e NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) D0CKET(2) LEA NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER I REVISION NUMBER SURRY POWER STATION, Unit 1 05000 -- 280 1998 - 013 -- 01 3 OF 4 TEXT (If more space is required, use additional copies of NRG Form 366A). (17) 2.0 SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS This event resulted in no safety consequences or implications. The SI actuation was spurious since it resulted from an invalid signal (i.e., an actual high steam flow condition did not exist). Appropriate operator actions were taken in accordance with emergency operating procedures 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 at any time during this event.

3.0 CAUSE

• A Category 1 Root Cause Evaluation (RCE) was initiated on November 22, 1998, to determine the cause of this event and to recommend corrective actions. The. RCE concluded that the event was caused by a short circuit in the summator for the main steam line "C" loop channel Ill flow transmitter [EIIS-JB,FIT], 1-MS-FT-1494. The short circuit resulted in circulating ground currents which caused the "B" loop channel Ill steam flow parameter to be greater than the actual value. The "B" loop channel Ill was affected through its power supply, which is common to the "C" loop channel Ill. As a result of the false steam flow indication, 1-FW-FCV-1488 opened rapidly to increase feedwater flow to the "B" SG. The level in the "B" SG increased to the high level turbine trip setpoint before control room operators could intervene.

The 1-MS-FT-1494 summator had been replaced and was in the process of being returned to service when the event occurred. The RCE investigation revealed that the module repair testing procedure did not include the defective portion of the summator's circuit board. As a result, the fault was not identified before installation.

4.0 IMMEDIATE CORRECTIVE ACTION($)

Following the reactor trip, control room operators acted promptly to place the unit in a safe, shutdown condition in accordance with emergency and other operating procedures.

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

NRC FORM 366A (6-1998)

" e e NRC FORM 366A U-5. NUCLEAR REGULATORY COMMISSION (6-1998)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1) D0CKET(2) LEA NUMBER (6) PAGE (3)

YEAR I SEQUENTIAL NUMBER I REVISION NUMBER SURRY POWER STATION, Unit 1 05000 -- 280 1998 - 013 -- 01 4 OF 4 TEXT (If more space is required, use additional copies of NRG Form 366A) (17) 5.0 ADDITIONAL CORRECTIVE ACTIONS The1-MS-FT-1494 summatorwas replaced. The installation of a new summator corrected the "C" SG steam line pressure and "B" SG steam line flow indication discrepancies.

The limit switches[EIIS-ZIS] for 1-FW-FCV-1488 were adjusted and the valve was tested satisfactorily.

The RCE team evaluated unit conditions and systems response contributing to the RCS cooldown following the reactor trip. The team concluded that the cooldown was normal considering: 1) the decay heat load was approximately 15% of the full-power equilibrium value (the event occurred at the beginning of reactor core life and the unit had been at power, below 30%, for less than one day), 2) SG "B" had been overfed as the unit responded to the apparent increase in steam flow, and 3) the cooling effect of auxiliary feedwater flow.

6.0 ACTIONS TO PREVENT RECURRENCE The module repair testing procedure was revised to include the defective portion of the summator's circuit board. Spare modules were tested in accordance with the revised procedure.

7.0 SIMILAR EVENTS None 8.0 MANUFACTURER/MODEL NUMBER Westinghouse Electric Corporation Signal Summator Assembly No. 4111084-001 9.0 ADDITIONAL INFORMATION Unit 1 was returned to service on November 23, 1998.

Unit 2 was operating at 100% power and was not affected by this event.

NRC FORM 366A (6-1998)