ML19273B411

LER 78-033/01T-2 on 781214:reactor Scram Occurred in Range 1 of Intermediate Range Monitors Due to Failure of Emergency Rod in Control Switch.Caused by Bent Switch Stop.Stop Plate Replaced

ML19273B411
Person / Time
Site:	Oyster Creek
Issue date:	03/28/1979
From:	Ross D JERSEY CENTRAL POWER & LIGHT CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML19273B407	List: ML19273B406 ML19273B411
References
LER-78-033-01T, LER-78-33-1T, NUDOCS 7904060306
Download: ML19273B411 (3)

Text

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8 63 DATE 74 75 DATE 7 60 61 DOCKET NUMBER EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h o 2 l During routine startup, a reactor scram occurred in range one of the IRM l o a system. At the time of the scram, rods were being withdrawn for approach l o 4 l to critical. Because of high xenon concentrations, the operator was o s I making the approach using information from the SRM. Since the SRM count ]

0 6 rate had changed very little, rods were beinO withdrawn in the notch over- _j O 7 l ride mode. An attempt to insert the rod using the emergency rod in switch a s l failed and the reactor scram followed. I 80 7 8 9 SYSTEM CAUSE CAUSE COYP. VALVE CODE CODE SUBCODE COMPONENT CODE SU8 CODE SUECODE 7

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41 lYl@ lNl@ lZl919 9@ 44 47 33 34 35 43 42 43 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS i O l The failure of the emergency rod in switch is attributed to a bent switch l 1 i l stop which would not allow contact to be maintained in the fully stroked 7 position of the switch. The mechanical stop plate on the switch was re-i [

i 3 l placed. The appropriate procedures will be revised to require rods to be i 4 l netc.hed out and a lecture will be given to operations personnel on the incident. ,

7 8 9 80 STA S  % POWE r. OTHER STATUS ISCOV RY DISCOVERY DESCRIPTION NA i s lCl@ l0l0l0l@l NA l l A l@l l A TlVI TY CONTENT RE LE ASED OF RE LE ASE AMOUNT OF ACTIVITY LOC ATION OF R LEASE 1 6 dj @ l' Zlgl NA l l N l PtRSONNEL EXPOS ES NuuBER TYPE oESCR PTic 4 @

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S l 2 l o l dSUE l @ DESCRIPTION Weekly news release - January 3, 1979 I Ii1lll!IIIItl{

7 8 9 10 68 69 80. E Donald A. Ross 201-45E-8784 o NAME OF PREPARER .. PHONE: $

rs Jersey Central Power & Light Company b

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i Madison Avenue at Punch Bowl Road Momstown. New Jersey 07960 (201)455-8200 OYSTER CREEK NUCLEAR GENERATING STATION Forked River, New Jersey 08731 Licensee Event Report Reportable Occurrence No. 50-219/78-33/!T-2 Report Date March 28, 1979 (Previous report dated December 28, 1978, and revised January 10, 1979.)

Occurrence Date December 14, 1978 Identification of Occurrence While performing a routine reactor startup following a scram from full power, a reactor period less than five seconds occurred resulting in a scram in range of one of the IRM's. This event is considered to be a reportable occurrence as defined in the Technical Specifications, paragraph 6.9.2.a.L. -

Conditions Prior to Occurrence The plant was in a routine startup.

Moderator temperature - 380*F.

Reactor pressure - 190 psig 4

Recirculation flow - 5.2 x 10 gpm Source range monitor count rate - 450 cps Reactor at peak xenon Rod worth minimizer in service withdrawal sequence Vill A-1 Do.cription of Occurrence On December 14, 1978, at 0415 nouis, a reactor scram occurred in range one of the IRM's. At the time of the scran, control rods were being witt. drawn for approach to critical as part of recovery operations following a scram from full power at approximately 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br /> on December 13 Because of the high xenon concentrations, an accurate estimated critical position was not possible. The operator at the controls was using SRM count rate information as the guide for approach to critical. Since the SRM count rate had changed only slightly (425 to 450 cps) from the start of the rod withdrawal process, it was thought that the reactor was still strongly subcritical; hence, rods were being with-drawn in the " notch override mode." When control rod 10-43 (first rod in Group 9 was withdrawn to notch position (10), the reactor became critical on Jersey Central Power & Light Company is a Member of the Genera! Puoi:c Utet:es System

Reportable Occurrence No. 50-219/78-33/1T-2 Page 2 March 28, 1979 an estimated 2.8 second period. The operator attempted to insert the rod using the " emergency rod in" control switch to no avail. The neutron flux excursion was terminated by a reactor scram in range one of the IRM's.

Apparent Cause of Occurrence The operator at the controls did not expect criticality to occur at this time considering the low SRM count rate. Furthermore, the approach to critical procedure does not provide specific guidance for startup under hot / peak xenon conditions. The reason that the rod did not respond to the " emergency rod in" switch was a failure of the switch to maintain contact in the fully stroked position due to a bent mechanical switch stop.

Analysis of Occurrence There is no safety significance to the fast positive period since it occurred very low in power and did not cause any observed heating of the moderator or changes in reactor pressure. In addition, because of the time constant between neutron flux and heat flux, the fuel cladding integrity safety limit was not violated. It is highly likely that, had the " emergency rod in" switch functioned properly, the short period would have been terminated by manual control.

Corrective Action The mechanical stop plate on the suiten was replaced and the two dogs on the switch shaft that contact the tab on the stop plate were positionally inter-changed to reduce the tendency to bend the tab when the switch is full stroked to the " emergency in" position.

The appropriate procedure (s) will be revised to require non peripheral control (I rods to be notched out when greater then group six (6) on startups.

Reportable Occurrence Report No. 78-33/lT will be placed on the required (I reading list.

The significance of this reportable occurrence will be incorporated into the (2 plant's training program. A lecture will be designed and given to all oper-ations personnel which specifically covers the events surrounding this occurrence and provides additional guidance for all future startups. This guidance will include (1) source monitor response under various reactor conditions, (2) sub-critical multiplication (1cluding the doubling thumb rule, (3) estimating in-sequence control rod worths during a startup, (4) procedural changes incorporated as a result of this occurrence.

Failure Data Type: SRM Control Switch Part Replaced: Front Plate, Catalog No. 127A6753P1