ML20083G408

Provides Info Re Thermal Shield Degradation & Reactor Operation at Nominal Operating Temp But Less than 20% Power, in Response to 831212 Telcon.No Thermal Shield Deterioration Detected

ML20083G408
Person / Time
Site:	Fort Calhoun
Issue date:	12/30/1983
From:	William Jones OMAHA PUBLIC POWER DISTRICT
To:	John Miller Office of Nuclear Reactor Regulation
References
LIC-83-321, NUDOCS 8401040233
Download: ML20083G408 (2)

Text

-

l l

Omaha Public Power District -

1623 Harney Omaha. Nebraska 68102 402/536-4000 December 30, 1983 LIC-83-321 Mr. James R. Miller, Chief U. S. Nuclear Regulatory Commission Of fice of Nuclear Reactor Regulation Division of Licensing Operating Reactors Branch No. 3 Washington, D.C. 20555

Reference:

Docke t No. 50-285

Dear Mr. Miller:

Fort Calhoun Station Th _ mal Shield In a telephone conversation on December 12, 1983 between E. G. Tourigny of your staff and K. J. Morris of Omaha Public Power District, it was requested that the District provide information concerning reactor operation at nominal operating temperature but less than 20% power.

The estimated number of hours which the Port Calhoun reactor has been operated at nominal operating temperature but less than 20% power is 5,856 to date. This estimate was prepared by Combustion Engineering using the same methodology as was applied in preparing a similar history for the St. Lucie Unit No. 1.

At the time of the thermal shield visual inspection con-ducted during the 1983 refueling outage, the Port Calhoun unit had accumulated approximately 5,550 hours0.00637 days <br />0.153 hours <br />9.093915e-4 weeks <br />2.09275e-4 months <br /> of operation at nominal operating temperature but less than 20% power.

During the inspection, there were no signs of deterioration of the thermal shield or its supports. The period of oper-ation at less than 20% power prior to the inspection is ap- '

l proximately 32% greater than that at which it is theorized the St. Lucie unit had experienced prior to suffering a significant loss of preload on the thermal shield position-ing pins.

8401040233 831230 1 PDR ADOCK 05000285 '

P PDR h00l f 45 5124 Employment with Equal Opportunity Male. Female

Mr. James R. Miller LIC-83-321 Page Two Although the basic design of the St. Lucie and the Fort Calhoun reactors is similar, caution must be exercised in comparing their operation. The Fort Calhoun thermal shield i and core support barrel experience lower reactor coolant flow velocities than plants where thermal shield degradation has been observed. Additionally, the radius of the core support barrel and thermal shield are smaller at Port Calhoun and the annulus be tween these components is also smaller. This results in lower anticipated mechanical loads induced by the hydraulic forces in the reactor vessel.

At the present time, the District is pursuing methods to analytically quantify dif ferences between Fort Calhoun and the other plants. This should then yield a measure of the j design margins present at Fort Calhoun.

Sincer,elv, f'y d W . - C . Jones Division Manager l Production Operations i

l WCJ/DJM:jmm cc: Mr. J. E. Gagliardo, Director Division of Resident Reactor Project

& Engineering Programs i

U. S. Nuclear Regulatory Commission l Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 LeBoeuf, Lamb, Leiby & MacRae 1333 New Hampshire Avenue, N.W.

Washington, D.C. 20036 Mr. E. G. Tourigny, Project Manager Mr. L. A. Yandell, Senior Resident Inspector I

l i

i r

I