05000312/LER-1979-001, Forwards LER 79-001/01T-0

Forwards LER 79-001/01T-0

ML19263C051
Person / Time
Site:	Rancho Seco
Issue date:	01/26/1979
From:	Mattimoe J SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
Shared Package
ML19263C052	List: 05000312/LER-1979-001, Forwards LER 79-001/01T-0 05000312/LER-1979-001-01, /01T-0:on 790129,RCS Cooldown Occurred at 120 F/H. Caused by Short Circuit Resulting in Reactor Trip.Procedure Mods Developed to Prevent Recurrence
References
FOIA-79-98, TASK-TF, TASK-TMR NUDOCS 7902020207
Download: ML19263C051 (3)

LER-1979-001, Forwards LER 79-001/01T-0

Event date:
Report date:
3121979001R00 - NRC Website

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January 26, 1979 U. S. Nuc] car Regulatory Commission Attn:

Mr. R. H. Engelken, Direct o r Region V, Of fice of Inspection & Enforcenent 1990 North California Boulevard Walnut Creek Plaza, Suite 202 Walnut Creek, CA 94596 RE: Operating License DPR-54 Docket No. 50-312 Reportable Occurrence 79-01

Dear Mr. Engelken:

In accordance with Technical Specifications for Rancho Seco Nuclear Generating Station Section 6.9.4.lb and Regulatory Cuide 1.16, Revision 4, Section L

'(2), the Sacramento Municipal Utility District is hereby sub-nitting a 1 irteen-day f ollowup report to Reportable Occurrence 79-01 which was initially :eported to your office January 5,1979.

This report is bei:.g submitted on a twenty-one day interval as agreed between Mr. R. Rodriguez and Mr. D. Sternberg on January 19, 1979.

On the date of that initial report, a short to ground in the ICS resulted in a reactor trip and subsequent RCS cooldown which exceeded the limits set forth in Technical Specifications Figure 3.1.2-2.

This follow-up report will describe the sequence of~ events surrounding the transient, discuss the reactor vendor review, and set forth the corrective action that will be taken to prevent recurrence.

Prior t o the transient, the plant was operating at 100 per cent steady state power, with all four reactor coolant pumps operating and an average reactor coolant systen temperature of 582* F.

A technician performing a modification in the ICS allowed a wire connected to a terminal in the ICS panel to swing free. The resultant short circuit took out the +25 volt D.C. power supplies. To provide circuitry protection, loss of either posi-tive or negative 25 VDC power will trip both the ICS A.C. supply breakers. ~

This occurred instantaneously, and the loso of logic power to the ICS ran the feedwater valves back to the 50% position.

The dramatically reduced feedwater flow caused the RCS pressure to increase to the RPS liigh Pressure trip limit of 2355 PSIG.

All four RPS channels tripped on high pressure approximately 0.16 seconds into the transient.

Although dramatically reduced (in relation to demand at 100 per cent full power), feedwater flow at this time was estimated as being twice the quantity necessary for removal of the 79020202o7 AN ELICTRIC S Y S 11 M S I RVIN G MORE I ll A N 600.000 IN 11t t tilAR1 Of CAI!!nR'lA

R. II. Engelken Page 2 January 26, 1979 e

decn at being generated.

As a result, approximately 70 seconds into the transient the PCS pressure had dropped to 1900 PSIG the low pressure trip for the P.PS.

At 2 minutes 53 seconds into the t ransient the RCS pressure had dropped to 1600 PSIG and the Safety Features Analog channgels actuated, bringing into service High Pressure Injection, Decay Heat and Reactor Building isolation.

Feedwater flow continued and was augmented by Auxiliary Feedwater brought on by the Safety Features actuation.

Approximately 5 minutes into the transient, and upon return fo ICS power, feedwater flow had increased to nearly 2.2 million lbs./hr.

Operator action terminated the majority of the feedwater flow at this point, and seven minutes into the transient both Main Feedwater Pumps were t. r i p p e d.

Frou that point on feedwater supplied to the OTSG's was limited to that from the Auxiliary Feedwater System.

It should be noted that upon Safety Features Actuation, although both SFAS Channels A and B were actuated, only the H2SFB panel indicated automatic actuation in the control room.

This was investigated and found to have been caused by the Auxiliary transformer 24 volt power supply breaker, located in the Digital Channel A Cabinet, being in the "o f f position. " This 24 volt power, although incapable of preventing an actuation does supply power to the SFAS indicating lights on the H2SFA panel.

Both OTSG water levels increased during the transient.

Subsequent to Safety Features the "01'SG-A" level was brought under control and remained on scale. However, the "0TSG-B", whose Auxiliary Feedwater Sa fety Features valve (SIV-20578) receives its signal f rom the A SFAS channel, was not brought under control as quick]y.

This was mainly due to the lack of indication of a SFAS channel A actuation, and the time involved in identifying and comp (nsating for this.

As a result, "B" OTSG was filled to the top of the operating range and continued at that level. for 10 to 15 minutes.

The filling of the "0TSG-B" was the single most significant cause of the excessive cooldown rate.

Upon recovery of system parameters to nominal hot shutdown values, a review of the data confirned that the pressure / temperature requirements re-garding heatup and cooldown of the RCS had not been violated.

However, the cooldewn rate specification of 100 F/hr. had been.

As a result, detailed data regarding the reactor trip and subsequent cooldown transient was transmitted to the reactor vendor (UE).

The vender revietted and analyzed the data on the adequacy of items relative to returning t.he plant to operation.

The vendor deternined the maximum cooldown rate as being that portion below 550* F.

accorp]ished wit hin any one-hour period.

The lowest RCS ter'perature encountered during the transient was 4 30 F,

therefore, the cooldown rate was 120 F/hr.

Although the Technical Specification cooldown rate was exceeded, the increase in RCS usage factor was judged to be mall when compared to the design cumu-lative usage factor.

Their review confirmed that the previously mentioned pressure / temperature limits for heatup/cooldown had not been violated and that the pressuri:.cr maintained a liquid IcVel t hroughout the transient.

In addition, Reactor Coolant Punp Seal Performance Data was reviewed and no excessive tempera-tures were experienced and all cavity pressures returned to normal values following t he t ransient.

Based on the results of their evaluation, the vendor concurred with the District's intent to return the unit to full power operation.

R.11. Engelken Page 3 January 26, 1979 In retrospect, the reactor trip was caused by the short circuit and resultant Joss of ICS.

The cooldown, in exces, of the Technical Specification limits, was a result of excessive feedwater to the B OTSC from the Auxiliary Feedwater System. The latter being attributable to the lack of SFAS indi-cation on the H2SFA panel and the time element involved in recognizing anI compensating for the condition.

The f remediate corrective action taken was to inform and transmit data to the vendor for engineering analysit. The vendor's concurrence was received prior to returning the unit to operation.

Other areas being pursued by the District include:

1.)

Evaluating whet her the ICS cabinet s should be under " key cont rol."

'This would assure that a)) work on the ICS would be under the cogni:ance of the shift operating personne). This could reduce the tine element involved in recognizing the cause of a trip and/or returning compromised cguipment to service. The evalunt ion as to the feasibility of " key cent.rol" uill he completed prior to January 31, 1979.

If reasonable, impler'enta tion would be c>: pedi t ious.

2.)

!!odification of the operating procedure to include specific instructions to assu re t he PPS/SFAS cabinet s are properly powered up prior to heating up the plant irom a Cold Shutoown condition.

This will be complet ed pi ior t o January 31, 1979.

3.)

Engineering evaluation as to the necessity of Auxiliary reed upon Safety Feat ures Act uation.

This evaluation wonid leave the system as a Safety Features system but would require operator action to admit water to t he OTSG's.

The Engineering evaluation will b2 completed prior to April 30, 1979.

4.)

The vendor is reviewing t he transient and wil] forward the results to the District. The recoreendations will be incorporated into the applicable procedures and correct ive actions taken as appropriate within one r;onth af ter receiving the report.

This transient nec essitated reactor shutdnwn for approximately 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.

Respectfully submitted, 5. ]1LW A

4

/J.

Mattimoe Assistant General Manager and Chie Engineer JJM: RJ R:RWC :lill: s1k Attachment es: Director, Management Information and Program Control (3)

Director, luspection and Enforcement