ML20011F132

LER 90-001-00:on 900124,outboard Steam Supply Line Isolation Valve Unexpectedly Closed During Surveillance Testing,Causing Isolation of HPCI Sys.Caused by Procedural Inadequacy.Test Procedure Will Be upgraded.W/900223 Ltr

ML20011F132
Person / Time
Site:	Cooper
Issue date:	02/23/1990
From:	Horn G, Reeves D NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CNSS903566, LER-90-001, NUDOCS 9003010219
Download: ML20011F132 (5)

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.U. S. Nuclear Regulatory Commission Document Control-Desk- 1

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Dear Sir:

o Cooper Nuclear Station Licensee Event Report 90-001, Revision 0, is being L forwarded as an attachment to this letter. ~

-Sincerely, I R. Horn

Division Manager of Nuclear Operations Cooper Nuclear Station CRH:bj s LAttachment cc:, R.' D.. Martin L.,G. Kunc1 R. E. Wilbur V.-L. Wolstenholm ]!

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... .~. P o *.- ,, ~, um- .. s. a o ei On January 24, 1990 at 8:29 a.m., llPCI MOV M016, the Outboard Steam Supply '

i Line Isolation Valve, unexpectedly closed during surveillance testing. The test in progress was associated with the liigh Pressure Coolant Injection (llPCI) Steam Line Low Pressure switches. Testing had proceeded to the point where switch, llPCI PS 68B, was being checked, At the time of the event, the plant was operating normally at 97 percent power, coasting down in advance of a planned refueling outage in early March.

In setting up for the test of the "B" switch, the Instrument and Control (160)

Technician inadvertently connected a Simpson 260 VOM across terminals for contacts associated with switch IIPCI PS 68D, effectively bypassing its contacts.

As a result of the "*" switch contacts already being closed due to the switch closing thebeing valve.isolated and depressurized, valve closure logic was made up, The cause of the event var due to a lluman Factors deficiency and a procedural inadequacy. While the t.erminal wires for both switches were correctly labeled, they were not readily observable by the Technician due to the height of the terminal box above the floor. Further, specific terminal designators were not clearly specified in the procedure.

Corrective actions taken included restoration of HPCI to service and completion of surveillance testing. The technician involved was counselled regarding connection of test equipment, jumpers, etc. during surveillance testing. Further corrective action includes upgrading the test procedure, reassessing and improving similar test procedures for like concerns and l

reviewing this event with all 160 Technicians during Industry Events Training.

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A, Event Descriotion:

On January 24, 1990 at 8:29 a.m., during performance of a surveillance test of the High Pressure Coolant Injection (HPCI) System Steam Line Low Pressure switches, normally open valve HPCI.MOV M016, the Outboard Steam i Supply Line Isolation Valve, unexpectedly closed. At the time of valve closure, testing of the first pressure switch, HPCI-PS 68A, had been successfully completed at,d testing of the second switch, HPCI PS 68B, had proceeded to the point where the switch was depressurized.

Operations personnel noted the closure of HPCI.MOV.M016 and informed the utility Instrument & Control (16C) Technicians performing the test to stop the test, restore the circuitry to normal and return to the Control Room. An investigation into the unexpected closure was initiated. Upon verifying that no unrelated HPCI problem existed, HPCI MOV M016 was -

reopened, restoring the system to standby status. Subsequently, the i surveillance test was resumed and satisfactorily completed.  !

B. Plant Status In operation at approximately 97 percent power, 780 MWe, coasting down in advance of a planned refueling outage in early March.

C. Basis for Report '

An unplanned challenge (closure) of a component of an Engineered Safety Feature (ESP) System, reportable in accordance with 10CFR50.73(a)(2)(iv).

D. Cause Human Factors deficiency and procedure inadequacy. During performance of the surveillance test of the "B" HPCI Steam Line Low Pressure Switch, HPCI.PS 68B, a Simpson 260 Volt Ohm Meter (VOM) was specified to be i

connected across the normally open contacts of the switch. The Technician, however, mistakenly connected the V0M across the normally open contacts of switch HPCI.PS-68D. In so doing, with the VOM in the Resistance mode of operation, the "D" switch contacts were effectively

-bypassed. With contacts from the "B" switch already closed due to g depressurization of the switch in accordance with a prior procedure

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step, sufficient isolation circuitry was actuated to result in closure of HPCI MOV.M016. Although the switch wires in the terminal box are labeled, the markings were not easily discernable to the Technician due to the height of the box above the floor. Additionally, while procedure guidance was considered adequate, terminal designators to be used to connect the VOM were not clearly specified in the test procedure.

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l0 0l 3 or 0 l4 vaut v ===. . o . ass w mar se, amawim E. Safety Slenificangt None. Within two (2) minutes of valve closure, the llPCI System lineup '

was restored to normal. During the period of time when llPCI was unavailable for automatic actuation, operation of the plant was normal. +

It should be noted that in accordance with Technical Specification Limiting Conditions for Operation, power operation is permitted for up to seven days with llPCI out of service. Therefore, a Technical Specification violation did not occur.

With regard to valve closure, llPCI-MOV M016 functions as a containment isolation valve (Table 3.7.4 of the Technical Specifications) and is one of the valves that automatically closes upon Group 4 actuation. No problems were noted or experienced upon valve closure or upon repositioning the valve to its normally OPEN position.

F. Safety Ienlications

[ As stated in the CNS Technical Specifications (Section 3.5, Basis), the r llPCI System is provided to assure adequate core cooling in the event of i

a small break Loss of Coolant Accident ( W CA) which does not result in  !

rapid depressurization of the reactor vessel. In the analysis of the effects of a small break WCA, loss of IIPCI is the most limiting single failure, llPCI, therefore, is assumed to not be availabic. Reliance, instead is placed on actuation of the Automatic Depressurization System and Low Pressure Emergency Core Cooling Systems by the high drywell '

pressure or low reactor vessel water level initiation signals. This event, without the benefit of IIPCI, results in a much lower Peak Clad Temperature (PCT) when compared to the large break events.

Chapter VI, Core Standby Cooling System and Chapter VII, Control and l Instrumentation, of the CNS USAR, specifies that the llPCI Control system is capable of starting and accelerating the turbine to rated speed within 25 seconds of receipt of an initiation signal,' delivering design llPCI flow of 4250 gpm at reactor pressures between 1120 and 150 psig, For large sized breaks up to and including the design basis LOCA, the llPCI System is assumed to actuate within 30 seconds (5 second sensing time plus 25 second llPCI startup time). The calculated Peak Cladding k Temperatures for large break WCAs, specified in NEDO 24045 and

/ subsequent addenda, are below the 2200'F limit specified in 10CFR50.46.

Injection of a small amount of water from the llPCI System during the design basis (LOCA) is accounted for in the analysis performed by General Electric (CE) for CNS. CE has, however, indicated that the capability of the plant to meet licensing requirements for postulated LOCAs is not very sensitive to the actual llPCI start time or flow rate.

In the case of a large break LOCA, this is due to the rapid depressurization and blowdown of the reactor vessel. The impact, then, of a potentially delayed llPCI startup for such an event is of little Concern.

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With IIPCI MOV M016 closed, automatic actuation of IIPCI would not have occurred until operator actions were taken to reposition the valve to its normally open position. A Control Room annunciator is actuated when the valve is not in its normal standby (OPEN) position to' alert Operations personnel of the off normal condition. Ilad the design basis event occurred at that time, resulting in a llPCI actuation signal, operator action would have immediately been taken to re open the steam supply valve, llPCI flow at the design rate, but possibly not within 30 seconds, would then have been supplied to the reactor vessel. Therefore, while some incremental increase in PCT could be expected, the impact would not be expected to be significant.

C. Corrective Action As noted in Section A, llPCI MOV MOl6 was reopened and the llPCI lineup was restored to its normal standby status. Subsequently, the surveillance test was satisfactorily completed. The technician involved was counselled regarding connection of test equipment, jumpers, etc.

during surveillance testing. o Further corrective action to be taken includes upgrading the surveillance test procedure to precisely identify the terminal strip connections to be made, reassessing and improving other 16C Surveillance

Test procedures in a like manner, and reviewing this event with all 160 l personnel as part of Industry Events Training performed by the Nuclear Training Department.

ll . Past Similar Events Past similar events that involved human factors and/or procedural

( deficiencies associated with incorrect installation of test equipment or jumpers, incorrect lifting of leads or incorrect component selection during surveillance testing include:

LER 89-016 Unplanned Automatic Initiation of an Engineered Safety L Feature While Conducting a Surveillance Test Due to a lluman

Factors Deficiency LER 87-022 Unplanned Closure of a Reactor Water Cleanup System Isolation Valve Due to Personnel Error During Surveillance Testing 1

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