ML20029A005

Ro:On 901130,max Reactor Heatup Rate Exceeded Following Control Rod Withdrawal.Caused by Personnel Error.Control Rods Reinserted in Reverse Sequence to Mitigate Heatup Rate & Surveillance Test Procedure Revised

ML20029A005
Person / Time
Site:	River Bend
Issue date:	01/25/1991
From:	Odell W GULF STATES UTILITIES CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RBG-34384, NUDOCS 9102010033
Download: ML20029A005 (3)

Text

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January 2 5, 1991 RIG- 3 4 3 8 4 l File Nos. G9.5, G9.25.1.3 l

U. S. Nuclear Regulatory Cannission Document Control Desk Washington, D. C. 20555 Gentlenen:

River Bond Station - Unit 1 Docket No. 50-458 Please find enclosed an Informtional Report concerning the exceeding of the mximum heatup rate specified by Technical Specification 3.4.6.1.A at River Bend Station - Unit 1. This report is subnitted to inform the NRC and docu: rent GSU's investigation and corrective actions.

Sincerely, UB

} l W. I. Odell Manager-Oversight River Bend Nuclear Group 1

l h Y hbfV 1 :/PDG/ CAB /DCil JEV/pg cc: U. S. Nuclear Regulatory Camtission j 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 NRC Resident Inspector Post Office Box 1051 St. Francisville, IA 70775 l

INIO Records Center 1100 Circle 75 Parkway Atlanta, CA 20339-3064 Mr. C. R. OMrg Public Utility Ccnmission of Texas 7900 Shoal Creek Blvd., Suite 400 North in.J. tin, TX 78757 h

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Description of Condition ist 1722 on 11/30/90 with the unit in operational Condition 2 (Startup), the nnximum reactor heatup rate of 100 degrees F/hr, specified in Technical Specification 3.4.6.1.A, was exceeded following contml md withdrawal. The peak heatup rate was 119 degrees F/hr. The operator failed to adequately nonitor the heatup rate.

74 tis condition was restored to catpliance with 'Ibchnical Specification 3.4.6.1.A within 30 minutes in accordance with the action statencnt of the 'Ibchnical Specification.

This Infonmtional Report is suhaitted to infonn tl.e NRC of this condition and pmvide corrective actions taken by SSU.

Investigation Prior to the violation of the Technical Specification heatup rate, the reactor power was at range 9 on the irt.catediate range nonitors (IINs) with the heatup rate at approxinctely 60 degrees F/hr. The at-the-controls (AT) operator was attempting to mitigate fluctuating water level caused by shrinkage due to cold feedwater injections. The root cause of this event was twofold.

First, the AK operator was focused on naintaining watnr level by withirawing contml rods and thus increasing the heatup rate.

Second, he relied on the shif t technical advisor (STA) to nonitor the heatup rate. A contributing factor to this event was that the heatup rate was being recorded on a 30 ntinute frequency.

In accordance with the 'Ibchnical Specification action statement, an evaluation has been perfonted by General Electric Company to detennine the impact on the reactor vessel. This evaluation revealed that the heatup occurred at very low pressures (0 to 88 psig). The contribution of pressum stresses is nontally larger than thental stress when considering the stress and fatigue impact. The pressure during this heatup incident was about 10 percent of nontal operating pressure. The evaluation by GE concluded that bri ttle fracture and the stress and fatique im[ncts were within acceptable lintits.

Corrective Action Inmxllate corrective action was taken by reinserting the control rods in reverse sequence to niitigate the heatup ra te. An analysis was perfonax1 by General Electric to detenttine impact on the vessel per Technical Specification 3.4.6.1 as described above.

Surveillance 'Ibst Pnx:edure 050-0700 was revised to increase the frequency of nonitoring heatup/cooldown rates to every 15 ntinutes with inuediate reviews by the control room forenun. In addition, heatup/cooldown rates will be administra tive1y limited to 80 degrees F/hr.

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. Operating crews were briefed by the Plant Panager on this con 6ition when they assuned their next shift. The licensed operator who was tfE NIC operator during this incident was removed frmi licensed duties pending final disposition.

Operational Impact The action requiremants for Technical Specification 3.4.6.1 were net within the allotted time frzute. The evaluation of vessel integrity confitTod that brittle fracture and the stress and fatigue impacts were within acceptatble limits. Therefore this event did not adversely affect the health and safety of the public.

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