ML17261A138

LER 79-016/01T-0 on 790804:14-h Interval of Reactor Operation Lapsed Prior to Trip During Trip Testing Following Maint Outage & Restoration of Criticality.Caused by Unclear Checkoff procedure.On-shift Training Held

ML17261A138
Person / Time
Site:	Ginna
Issue date:	08/16/1979
From:	Peck C ROCHESTER GAS & ELECTRIC CORP.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML17244A769	List: ML17244A768 ML17261A138
References
LER-79-016-01T, LER-79-16-1T, NUDOCS 7908210524
Download: ML17261A138 (4)

Text

NRC FOQM 366 U. S. NUCLEAR REGULATORY COMMISSION

{7.7I+ 79-016/01T-0 LICENSEE EVENT REPORT (CAR 1231)

CONTROL BLOCK: Q) (PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION)

I 6

~0 7 8 9 N Y R E LICENSEE CODE G 1 14 Q20 15 0 0 0 0 LICENSE NUMBER 0 0 0 0 25 Q3 26 4 1 1 LICENSE TYPE 1 1 30 Q4~57 CAT 58 Qs CON'T

~01 soURcE ~LQB 0 5 0 0 0 2 4 4 7 0 8 0 4 7 9 QB 0 8 1 6 7 9 Qe 7 8 60 61 DOCKET NUMBER EVENT DATE 74 75 REPORT DATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES Q10 Following a maintenance outa e and restoration of criticalit a 14 hr. interval of reac-

~03 tor operation lapsed prior to a trip during trip testing, in which A.C. power was on for accumu injection MOV's and RWST delivery to RHR MOV. (T.S. 3.3.1.1g). This con-dition was noted during precritical checkoff performance after trip testing. Valves

~06 were in proper safeguard position during this interval of operation. Before 14 hr. inter val cktbkr positions had been reported but checkoff did not direct repositioning.

Cktbkrs were rechecked b a 2nd o erator on basis of o erabilit for ali nments and nothing unusual was reported. Control operator completed procedure assuming pre-

~08 critical requirement was fulfilled.

7 8 80 SYSTEM CAUSE CAUSE COMP. VALVE CODE CODF SUBCODE COMPONENT CODE SUBCODE SUBCODE

~oe R Qll ~DQ12 ~ZQ13 C K T B R K QE ~AQls ~Z Qs 7 8 9 10 11 12 13 18 19 20 SEQUENTIAL OCCURRENCE REPORT REVISION Q17 LERiRO REPQRT ACTION FUTURE EVENT YEAR

~79 21 22 EFFECT

~

23 SHUTDOWN

~01 24 REPORT NO.

0 26

~W 27

~01 28 ATTACHMENT CODE 29 NPRDQ TYPE

~T 30 PRIME COMP,

~31 NO.

~0 COMPONENT TAKEN ACTION ON PLANT METHOD HOURS Q22 SUBMITTED FORM SUB. SUPPLIER MANUFACTURER

~GQlg ~HQlg ~ZQEo ~ZQET 0 0 0 0 ~YQ33 ~NQ34 ~NQEs W 1 2 0 QER 33 34 35 36 37 40 41 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS Q27 0 Unclear precritical checkoff procedure. Cktbkrs for these valves are closed for pre-startup lineup and stroking during heatup. Later precritical checkoff specifies valve position with cktbkr "open" while cktbkr panel marked "on" and "off." There is no 3

direction to reposition cktbkr. A. C. power was removed. Checkoff was changed to direct repositioning and to give consistent nomenclature. On-shift training on power removal re uirement was held and fu h tai i b

~Q 7 8 9 80 FACILITY METHOD OF STATUS  % POWER OTHER STATUS DISCOVERY DISCOVERY DESCRIPTION Q32 CCl) 6 MQ C 28 0 0 0 2e NA LJQ3) Performing pre-critical rocedure 45 46 80 ACTIVITY CONTENT RELEASED OF RELEASE Q33 Z AMOUNTOF ACTIVITY NA Q NA LOCATION OF RELEASE 36 Q34 7 8 9 10 11 44 45 80 PERSONNEL EXPOSURES NUMBER TYPE DESCRIPTION Q3e 7 ~00 0 Q37 Z Q33 NA 7 8 9 11 12 13 80 PERSONNEL INJURIES 7 8 9 II 12 80 LOSS OF OR DAMAGE TO FACILITY 43 e

TYPE

~zQ42 DESCRIPTION NA 90821O5 Z 8 9 10 80 PUBLICITY NRC USE ONLY ISSUED DESCRIPTION o ~NQ44 NA 8 9 10 68 69 80 o NAME OF PREPARER Carl H. Peck PHQNE 716/'546-2700, ext. 291-205

0 Attachment to LER 79-016/01T-0 Rochester Gas and Electric Corporation R. E. Ginna Nuclear Power Plant, Unit No. 1 Docket No. 50-244 After a maintenance outage the reactor was taken critical August 4, 1979 at 0425 hours0.00492 days <br />0.118 hours <br />7.027116e-4 weeks <br />1.617125e-4 months <br />. Later that day turbine trip testing was performed in accordance with the manufacturer's recommendations, and during this a reactor trip occurred, at 1824 hours0.0211 days <br />0.507 hours <br />0.00302 weeks <br />6.94032e-4 months <br />. During performance of the restart pre-critical checkoff it had been discovered that the A.C. power was "on" for safety injection accumulator injection valves MOV-841 and MOV-865 and the refueling water storage tank delivery to residual heat removal pump suction valve MOV-856. This indicated that a fourteen hour interval had elapsed in which the reactor had been made critical and was operated with the A.C. power available to these valves contrary to Plant Technical Speci-fications which require A.C. power to be removed. During that interval, however, the valves were in their proper safeguards position and would have performed their intended function.

Investigation revealed that, as a result of the precritical check for the first startup, the circuit breaker positions for the valves had been reported, but since the precritical checkoff did not direct changing the breaker positions, they were not immediately repositioned. These positions were necessary to operate these valves to establish proper alignment for safeguards operation prior to heatup and for valve stroking during the heatup.

Another operator was later ordered to check the breaker status for these valves.

He viewed their condition as normal for operability, reported nothing unusual, and did not reposition them for removal of A. C. power.

Further, the precritical checkoff introduced confusion to operators by specifying that the breakers are to be "open" while the breaker panels are marked "on" and "off."

The precritical checkoff has been corrected to direct repositioning of circuit breakers where removal of A.C. power is required before reactor criticality, and to provide nomenclature consistent to the breaker panels.

In addition, the Operations Engineer instructed each shift to conduct on-

'shift discussions on the requirement for removal of A. C. power to these valves. Further formal classroom training devoted to this requirement has been initiated for the operators in their regularly scheduled training periods.

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