ML18152B206

LER 88-029-00:on 880815,reactor Trip & Safety Injection Occurred.Caused by Spurious Initiation of Train a High Consequence Limiting Safeguard.Relay Replaced & Limit Switch on Steam Dump Valve adjusted.W/880914 Ltr

ML18152B206
Person / Time
Site:	Surry
Issue date:	09/14/1988
From:	Benson D VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
88-045, 88-45, LER-88-029, LER-88-29, NUDOCS 8809200227
Download: ML18152B206 (5)

Text

POW 28-06-01

• • _...RC Form 366 U.S. NUCLEAR REGULATORY COMMISSION (9-831 APPROVED 0MB. NO. 3160-0104 LICENSEE EVENT REPORT (LERI EXPIRES: 8/31/88 FACILITY NAME (11 DOCKET NUMBER (21 I PAGE 131 TITLE (41 Surry Pow.er_ Station, Unit 1 Reactor Trip /Safety Injection Due To Spurious Hi CLS Signal As A Result Of A I0 I 5 I O I O I O I 21 8 K) 1 IOF O I 4 Malfunctioninq Relay EVENT DATE (5) LER NUMB.ER (61 REPORT DATE (71 OTHER FACILITIES INVOLVED (81 MONTH DAY YEAR YEAR }(: SE~~~~~~AL ti ~~1~~~ MONTH DAY YEAR FACILITY NAMES DOCKET NUMBER(SI oI a 1 1s a a aI a - o I 2 I 9 - ol o o I 9 1 14 s I s OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Ch*ck on* or mor* of rh* fol/ovying/ (11)

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ABSTRACT (Limit to 1400 spac*s, i.*., *pproxim1t1/y f/fto*n slnglo-spac* tvfN'writton lin*sl (181 on August 15, 1988 at 0927 hours0.0107 days <br />0.258 hours <br />0.00153 weeks <br />3.527235e-4 months <br />, a Unit 1 reactor trip and safety injection occurred due to the spurious initiation of train "A" High Consequence Limiting Safeguard * (Hi CLS).

The spurious signal was generated during the performance of a periodic test on the train "A" Hi CLS system. A relay malfunction is the suspected cause of the spurious signal.

The relay is actuated by a test terminate button. It is believed that a single relay contact failed to close when the button was depressed, resulting in a Hi CLS signal.

Operators followed appropriate plant procedures and quickly stabilized the plant following the trip. The suspect relay will be analyzed by the system lab and reconunendations will be considered when the report is received.

0009200227 a~iao PDR ADOCK O PNU s

NRC Form 388 (9-83)

POW 28-06-01

,~RC Form 366A U.S. NUCLEAR REGULATORY COMMISSION

,. (9;831 I LICENSEE EVENT REPORT (LERI TEXT CONTINUATION APPROVED 0MB NO, 3150-0104 EXPIRES: 8/31 /88 FACILITY NAME (1) DOCKET NUMBER (21 LER NUMBER (61 PAGE (31 Surry Power Station;--Unit 1 o Is I o I o I o I 21 8 IO 81 8 - o I 219 - oI o o I 2 o F o I 4 TEXT /If more spece is requirrtd, use additional NRC Form 366A's/ (171 1.0 Description of the Event On August 15, 1988 at 0927 hours0.0107 days <br />0.258 hours <br />0.00153 weeks <br />3.527235e-4 months <br />, a Unit 1 reactor trip and safety injection occurred due to the spurious initiation of train "A" High Consequence Limiting Safeguards (Hi CLS). The spurious signal was generated during the performance of a periodic test on the train "A" Hi CLS system.

During and following the trip, plant systems and equipment performed as designed with the following exceptions:

1) One of the eight steam dump valves (turbine bypass to condenser) indicated intermediate position when it was actually fully closed.

2) The number 3 emergency diesel generator (#3EDG) high crankcase pressure alarm was received.

3) Channel 4 trip status light, turbine power greater than 10%, did not clear when turbine power dropped below 10% and the P-7 permissive was not enabled when turbine power and reactor power were less than 10%.

4) The "A" main feedwater bypass valve was unable to be opened when placing main feedwater in service following the reset of the safety injection signal.

In addition, the Refueling Water Storage Tank (RWST) level was noted to have dropped below the Technical Specification minimum as a result of the safety injection in progress.

2.0 Safety Consequences and Implications The Hi CLS engineered safety feature (ESF) is designed to place the plant in a safe condition upon receipt of a containment high pressure signal indicative of a primary or secondary loss of coolant inside the containment vessel. Although the signal in this case was not due to an actual containment high pressure condition, plant systems functioned as designed (with the exception of those items listed above) to leave the plant in a safe and stable condition. Therefore the health and safety of the public were not affected.

NRC FORM 3eaA *u.s. GPO, 1988-520-589100070 (9-831

POW 28-06-01 N RC Form 366A U.S. NUCLEAR REGULATORY COMMISSION 19-831 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVED 0MB NO. 3150-0104 EXPIRES: 8/31/88 FACILITY NAME 11) DOCKET NUMBER (2) LEA NUMBER (6) PAGE 13)

YEAR :;:;:;:;:;: SEQUENTIAL ;:;:;:;:;:; REVISION

NUMBER

:::;::: NUMBER Surry Power Stati~- Unit 1 o I5 1o I o I o I 218 1o 818 - o I2I 9 - o 10 o I 3 oF o 14 TEXT (If mars sp*ce is rsquirsd, use additional NRC Form 366A's/ (17) 3.0 Cause A relay malfunction {EIIS-RLY} is the suspected cause of the spurious train "A" Hi CLS signal. The suspected relay is exercised at the close of the Hi CLS test procedure by depressing a test terminate button.

The test terminate button realigns the circuitry from the test position to the normal position. It is believed that a single relay contact failed to close when the button was depressed resulting in a Hi CLS signal and subsequent ESF actuation.

The failure of the steam dump valve position indication was due to an improperly adjusted limit switch.

The #3 EDG high crankcase pressure alarm is believed to have been caused by a malfunctioning pressure switch.

A failed pressure comparator resulted in the failure of the tri~ status light to clear when turbine power was less than 10% and four failed relays prevented the P-7 permissive from actuating.

4.0 Immediate Corrective Action(s)

-Following the trip, operators utilized the appropriate plant emergency procedures to stabilize the plant. The Shift Technical Advisor (STA) reviewed the critical safety function status trees to ensure plant parameters remained within safe bounds.

5.0 Additional Corrective Action(s)

1) The suspect reiay, in the Hi CLS circuitry, was replaced and will be sent to our systems lab for analysis.

2) The limit switch on the steam dump valve was adjusted.

3) The pressure comparator and the four failed relays in the channel 4, P-7 circuitry, were replaced.

Similar relays in the channel 3, P-7 circuitry, were inspected and found to be correctly positioned.

NRC FORM 3e8A *u.s. GPO, !988-520-589100070 (9-83)

POW 28-06-01 NRC Form 366A U.S. NUCLEAR REGULATORY COMMISSION (9-83)

LICENSEE Ev NT REPORT (LERI TEXT CONTINUAT N APPROVED 0MB NO. 3150-0104 EXPIRES: B/31 /88 FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

Surry Power Station,-- Unit 1 0 15 I o I o I o I 2 la lo a I a - o I 2 I 9 - o I o oI 4 oF oI 4 TEXT (If more spaca is roquif'fJd, usa additional NRC Form 366A's/ (17)

4) The RWST level was returned to the required level.

5) The #3EDG was tested satisfactorily, following the event, and no high crankcase pressure alarm was received. However, the suspect pressure switch will be replaced.

6.0 Action(s) Taken to Prevent Recurrence Recommendations from our systems lab, concerning the suspect Hi CLS relay, will be considered when the report is received. An evaluation of a surveillance program for parallel relays, such as the type used in the P-7 circuitry, will be conducted.

7.0 Similar Events None.

8.0 Manufacturer/Model Number Hi CLS Relay: Westinghouse BFD 44 P-7 Relays: Westinghouse BF 48 Pressure Comparator: Westinghouse NRC FORM 368A *u.s. GPO, 1988-520-589100070 (9-83)

e VIRGINIA ELECTRIC AND POWER COMPANY Surry Power Station P. 0. 8011316 Surry, Virginia 23883 September 14, 1988 U.S. Nuclear Regulatory Commission Serial No. : 88-045 Document Control Desk Docket No. : 50-280 016 Phillips Building Licensee No.: DPR-32 Washington, D.C. 20555 Gentlemen:

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

REPORT NUMBER 88-029-00 This report has been reviewed by the Station Nuclear Safety and Operating Committee and will be reviewed by Safety Evaluation and Control.

Very truly yours,

~:[le 9?lc/t David

• B6'1f6~

Stati n Manager Enclosure cc: Dr. J. Nelson Grace Regional Administrator Suite 2900 101 Marietta Street, NW Atlanta, Georgia 30323

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