ML20090D332

Ro:On 730829,four Fuel Bundle Segments Were Operating Slightly in Excess of Allowable Average Planar Linear Heat Generation Rate.Adjustments Made in Power Distribution to Obtain Optimum Power Distribution & Output from Reactor

ML20090D332
Person / Time
Site:	Nine Mile Point
Issue date:	09/07/1973
From:	Bartlett J NIAGARA MOHAWK POWER CORP.
To:	Skovholt D US ATOMIC ENERGY COMMISSION (AEC)
Shared Package
ML20090D333	List: ML20090D332 ML20090D336
References
6844, NUDOCS 8302280015
Download: ML20090D332 (2)

Text

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. ,- . m m NIAGARA MOHAWK POWER CORPORATION NtAGARA MOHAWK **.

300 ERIE SOULEVARD WEST SYRACUSE,N Y 43202 September 7, 1973 x

Mr. Donald J. Skovholt o86p/p5l ear[lbkh .'s s

Assistant Direct'or for Reactor Operations t^t kED/c, Sg' ~i Division of Reactor Licensing 9' United States Atomic Energy Commission Washington, D. C.

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Re: Provisional Operating License: DPR-17 Docket No.: 50-220

Dear Mr. Skovholt:

At approximately 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br /> on August 29, 1973 following a routine calibration of the Local Power Range Monitors (LPRM) and subsequent core power d,istribution calculations, four fuel bundle segments were found to be operating slightly in excess of the allowabic average planar LilGR as shown in Technical Specifications Figure 3.1.7 (change 9). This event was reported to Region I Compliance on August 30, 1973 at 1130 hours0.0131 days <br />0.314 hours <br />0.00187 weeks <br />4.29965e-4 months <br />.

Technical Specification 3.1.7 a states:

a. Average Planar LIIGR ,

During steady state power operation, the average linear heat generation rate (LIIGR) of all the rods in any fuel assembly, as a function of average planar exposure, ati any axial location shall not exceed the maximum average planar LIIGR shown in Figure 3.1.7.

f The LPRM's are routinely calibrated following extensive rod manipulations or upon the completion of each full power month. This consists of physically

adjusting the gain of each LPPJi amplifier by a factor determined thru a i computer calculation. Following the physical adjustment of all LPPJ1 amplifiers the Traversing Incore Probe (TIP) system is used -to traverse each LPRM string and supply the current axial flux distribution to the computer system.

Using this axial flux distribution base and the current LPRM readings along with the other thermal hydraulic parameters, a highly accurate power distribution can then be obtained from the computer.

Prior to the 27th of August, steps were taken to achicire compliance with the fuel densification specification i limiting both local LilGR and average planar LilGR within the reactor core and at the same time minimize possible electrical ourmit 1- #--

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o Mr. Donald J. Skovholt September 7,.1973 U.S. Atomic Energy Commission Adjustments in the power distribution were made to obtain an optimum power distribution and power output from the reactor. In conjunction with this, control rods were shifted to move the. axial peaks higher in the core into more voided regions and thus flatten the axial power distribution. In order to closely monitor the power distribution, calculations were performed following each step and reactor power was reduced to provide an additional degree of conservatism.

Upon completion of preliminary rod movements on August 26-27, 1973 the new axial flux distributions were supplied to the process computer system via the TIP system. -The subsequent calculation of power distribution on August 27, 1973 showed that all bundles were within prescribed limits and the reactor power could be increased. In an attempt to. optimize plant power output more control rods were inserted thus increasing the margin between the existing .

. average KW/FT and the limit. Reactor power was then increased using recir-culation flow. Due to the fact that more control rods had been manipulated following the last axial flux distribution base supplied to the computer system it was necessary to update these distributions using the TIP system. Ilowever due to mechanical problems within the TIP system this was not possible until August 29, 1973. Following the LPR>l calibration and inputing of the new axial flux distributions into the computer system a core power distribution was performed. A slight shift.in calculated power distribution was noted and four fuel bundle segments were found to be slightly higher in maximum average planar Ll!GR than the limit prescribed in Figure 3.1.7 of the Technical Specifications.

The four fuel bundle segments (reload 1) have a nodal exposure of 1930 }!WD/ST and corresponding limit of 10.6 KW/FT (Figure 3.1.7). The four segments were operating .15 KW/FT over this limit. Immediate steps were taken to reduce the maximum average planar LilGR by reducing core thermal power using reactor'recir-culation flow. Subsequent core power distribution calculations showed all fuel types to be operating within limits shown on Figure 3.1.7.

'To prevent reoccurance of this abnormal occurance, the LPRM's will be calibrated once every two full power weeks of operation or twice as often as was previously done. In addition, extension movement of control rods will be done ht a substantially lower power level and once assurance to specification 3.1.7 is obtained, power will be slowly increased so as to assure validity of power distribution calculations. This violation is the first of its kind and evaluation of the safety implications of the incident in light of the cumulative experience obtained previously is not applicable. Ilowever, due to the degree of conservatism built into the calculations it is felt that no undue hazard would have been presented to the general public even had a LOCA occurred during this period.

Very truly yours, y w ...,

t m (oDa.vCCUf James Bartlett . ,

Executive Vice President Operations and Engineering JB:cm -

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