ML20198S504

Special Rept 2-SR-86-015:on 860525,reactor Trip Initiated by Reactor Protection Sys as Result of Low Level Condition in Steam Generator 1.Subsequent Depressurization of Steam Generators Resulted in ESF Initiation

ML20198S504
Person / Time
Site:	Palo Verde
Issue date:	05/30/1986
From:	Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
References
2-SR-86-015, 2-SR-86-15, ANPP-36727-EEVB, NUDOCS 8606100448
Download: ML20198S504 (4)

Text

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.P.ECEP/ED "r0 Arizona Nuclear Power Project .i ,:3 ., . , , _ Fy ,,

P.O. BOX 52034 e PHOENIX, ARIZONA 85072-2034 . , , _ .

May 30, 1986 -M""

ANPP-36727-EEVB/PGN/98.05 Mr. John B. Martin, Regional Administrator Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Region V 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596-5368

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Unit 2 Docket No. STN 50-529 (License NPF-51)

Special Report SR-86-015 File: 86-020-404

Dear Mr. Martin:

Attached please find a Special Report (2-SR-86-015) prepared and submitted pursuant to Technical Specifications 6.6.1.a and 6.9.2. This report discussen a Notification of Unusual Event on May 25, 1986.

If you have any questions, please contact me.

Very truly yours, EcNa SnenN j} ,

E. E. Van Brunt, Jr.

Executive Vice President Project Director EEVB/PGN/jle Attachment ec: R. P. Zimmerman (all w/a)

A. L. Hon E. A. Licitra A. C. Gehr INPO Records Center 8606100448 DR 860530 ADOCK 05000529 PDR I

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PALO VERDE NUCLEAR GENERATING STATION UNIT 2 NOTIFICATION OF UNUSUAL EVENT ON MAY 25, 1986  !

Docket No. 50-529 License No. NPF-51 Special Report No. 2-SR-86-015 At 0400 MST on May 25, 1986, the NRC Operations Center was notified via the Emergency Notification System of the declaration of an Emergency Class for Unit 2 of the Palo Verde Nuclear Generating Station. The NOTIFICATION OF UNUSUAL EVENT was declared pursuant to Esergency Plan Implementing Procedure -

02, which requires emergency classification assignment and a written report for reactor trips which are complicated by concurrent or subsequent events or conditions.

Prior to the events which initiated the reactor trip, reactor power was in the process of being reduced from 20% to 10% power by use of the control element assemblies, in order to allow start-up testing of the turbine control and stop valves, and the turbine overspeed protective feature. At the time of the reactor trip, the plant was in Mode 1 (POWER OPERATION) operating at 15%

reactor power, with the Control Element Drive Mechanism Control System in the manual-sequential mode of operation, the feedwater control system in automatic, and the turbine generator aligned to the grid producing approximately 95 MWe.

At 15% power (decreasing), feedwater flow is transferred from the economizer region of the steam generators to the downconer section. The change in feedwater system flowpath, combined with a low bias setting on the feedwater pump speed controller, resulted in a low differential pressure across the feedwater regulating valve, and a temporary and inadvertent termination of feedwater flow to the steam generators.

The termination of feedwater flow to the steam generators resulted in an increase in Reactor Coolant System (RCS) temperature, and an increase in steam generator pressure. As a result of the increased steam generator pressure, the steam bypass control system (SBCS) valves modulated open in order to control steam generator pressure. Actuation of the SBCS resulted in a pressure decrease in the steam generators of approximately 70 psi, which was

sufficient to allow feedwater flow to be re-established, and to cause the steam bypass control valves to close.

1 Following the closure of the steam bypass control valves, RCS temperature and l steam generator pressure increased. As a result of the increasing steaa

generator pressure, feedwater flow was terminated for a second time, again because of the low bias setting on the feedwater pump speed controller, and an j insufficient differential pressure across the feedwater regulating valves.

, The operator on the feedwater controls responded to the termination of i

feedwater flow by increasing the bias on the feedwater pump speed controller.

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NOTIFICATION OF UNUSUAL EVENT ON May 25,1986 q- Page Two 1

Increasing the bias on the feedwater pump speed controller reinitiated

feedwater flow to the steam generators. However, the rapid introduction of a relatively large amount of feedwater to the steam generators caused a j

i depression of the RCS cold leg temperature, and a decrease in steam generator level caused by shrinkage. The operator responded to t.he decreasing temperature and level indications by reducing the bias settings on the 1 feedwater pump speed controllers.

Decreasing the pump speed bias resulted in a third termination of feedwater

{ flow to the steam generators. This termination of feedwater flow, combined with the previous intermittent feedings of the steam generators, and shrinkage caused by the introduction of large amounts of feedwater into the steam

. generators, resulted in a low level condition in the steam generators. At

0313, a reactor trip was initiated by the reactor protection system, as a i result of a low level condition in steam generator number 1.

l The feedwater control system has a refill flow demand feature which, when

i active following a reactor trip, switches feedwater control from a steam generator level and feedwater flow-steamflow mismatch mode to a RCS 3

temperature control mode. Based on RCS temperature, the refill flow demand

! feature sets feedwater flowrate by controlling feedwater pump speed and 4

downcomer valve position. The refill flow demand logic is bypassed when a  :

i reactor trip is not active, and is passive when RCS temperature is below 564 j degrees Fahrenheit. Because of the termination of feedwater flow which

occurred ' immediately prior to the trip, RCS temperature was sufficiently l elevated such that the refill flow demand logic was activated following the

reactor trip.

In addition to operation of the refill flow demand feature, which operated as 1

designed, a continued high steam demand existed after the trip. The high

, steam demand was the result of Unit 1 auxiliary steam loads which were l cross-tied to, . and being supplied by, the Unit 2 steam header, the Unit 2

. auxiliary steam loads, and both main feedwater pumps. The combined effects of the refill flow demand logic, and a continued high steam demand, were i sufficient to cause the overcooling and depressurization of the steam geaerators and the RCS.

1 i The depressurization of the steam generators resulted in the Engineered Safety Features (ESP) initiation of a Main Steam Isolation Signal (MSIS) at 0315, and the RCS depressurization resulted in the ESF initiation of a Safety Injection Actuation Signal (SIAS) and a Containment Isolation Actuation Signal (CIAS) at

! 0316.

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NOTIFICATION OF UNUSUAL EVENT ON May 25, 1986 page Three The MSIS actuation resulted in the isolation of the main steam header, and terminated the steam demand frore the main feedwater pumps and the auxiliary steam loads. Following the ESF actuations, one reactor coolant pump was tripped in each steam generator loop, as required by procedure, and heat removal was accomplished by the use of the steam generator atmospheric dump valves and the auxiliary feedwater system.

Since heat removal was available at all times during this event, and no fission product barriers were challenged or breached, this event did not adversely affect the safe operation of the plant, or the health and safety of the public. During the event all safety systems performed as designed, including the actuation of the reactor protection system, MSIS, SIAS, and CIAS.

As a result of the ESF actuations which complicated this event, a Notification of Unusual Event was declared at 0322. The plant was stabilized, and the MSIS, SIAS, and CIAS signals were reset at 0349, approximately 36 minutes after the reactor trip. The Notification of Unusual Event was terminated at 0410.

This event continues to be evaluated. A Licensee Event Report will be submitted within 30 days to further describe this occurrence.

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