ML20106J912
Text
_.
l To:
James P, O'Rel]Iy DiItctorate of Regulatory Operations Region :
631 Park Avenue King of Prussia, Pennsylvania 19406 4
From:
Jersey Central Power & Light. Company Oyster Creek Nuclear Generating Station Docket #50-219 Forked River, New Jersey 08731' Stbject:
Preliminary Abnormal Occurrence Report No. 73-33 The following is a preliminary report being sthmitted in compliance with the Technical Specifications, l
parsgraph 6.6.2, l
Preliminary Approval:
q h 12/26/73 J. T. Carroll, Jr.
Dato cc: Mr. A. Giaubusso
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/4/2//ga Abn:rmal Occurrzncs
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Report No. 73-33
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SUBJE(T. :
Violation of the Technical Specifications, paragraph 3.1. A.3,
~ he high flow setpoints on the steam and condensate lines of 1
.f the Isolation Condensers wen fond to trip at AP values in excess of those as stated in Technical Specification Table 3.1.1.H Q20 psid - steam, $27 inches AP H2O - Condensate).
This event is considered to be *n abnormal occurance as de-j.
fined in the Technical Specifications, paragraph 1.15A. Nati-ll fication of this event, as required by the Technical Specifi-l cations, paragraph 6.6.2.a, was made to AEC Region I, Directorate of Regulatory Operations, by telephone on Friday, December 21, 1973, at 1635, and by telecopier on Wednesday, December 26, 1973, l
at 0955.
On Friday, Decouber 2',1973, while perfoming surveillance SITuKfl0N:
1 testing on the Isolation Condenser High Flow sensors (two each p,;r condensate and stema line per condenser), four of the sonsors were found to trip.at setpoints in excess of their respective limits. Specifically, the sensors of concem and the.'
corres-i ponding "as found" sctroints are as follows :
Condensat_e "A" COndenaar 1311A1 - 29" of water 1311 A2 31" of water 181182 - 29" of water Steam i
"B" Candenser 2805s1 22 paid 4
4
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Abnormal Occurrence Report No. 73 33 Decembsr 21, 1973 t
CAUSE:
The cause of this event is yet to be dotorwincd.
REMEDI AL ACrlON:
Upon discovery of the condition, the instrument technician per-i forming the surveillance test reset the affected d/p sensors to their required velves Q27 laches H2O on condensate and (20 psid on the steam).
SAFETY SIGNIFX_cANf7.;
The safety iglication of'this event is minimal, since the son-sors were operable and would have perfonned their function.
PAst data has shown that. upcui initiating the condensors, the condensate line senses d/p in exccss of 60 inches of water; thereforc, although the setpoint had drifted, the new value was jf well within the range of the expected signal and the sensor was capable of performing its protective action, i
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.erSey central Power & Lig t Company G.
MADISON AVENUE AT PUNCH BOWL ROAD w MORRISTOWN. N.J.07960.*201-539 6111 asau,evn g,y')),'l, Pubne viinnes corpor iio850 psig.
%e significance of this event is indicated in the bases of th'e Technical Q ecifications.
"W e low pressure isolation of the main steam lines at 850 psig was provided to give protection against fast reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the
. scram feature which occurs when the main steam line isolation valves are closed to provide for reactor shutdown so that high power operation at low reactor pressure does not occur, thus providing protection for the fuel cladding integrity safety limit." % e temperature difference for saturated steam at 850 psig and 800 psig is less than 8'F, thus the resulting vessel cooldown effect is considered
- negligible.
With regards to power operation below 850 psig and the attendant effects on the fuel cladding integrity safety limit, power level must be limited when pressure is less than 600 psig or flow is less than 10% to 354 MWt or approximately 18.3%, of rated. As stated in the Technical Specifi-cations, "The value is applicable to ambient pressure and no flow conditions.
For any greater pressure or flow conditions there is increased margin." We fuel cladding integrity safety limi,t curve has been developed and is applicable for pressure in excess of 600 psig.
% erefore, whether a reactor scram occurs at 850 psig or 800 psig has little safety significance since no severe re-strictions on critical heat flux are imposed unless the reactor pressure is less than 600 psig.
The following actions are planned to avoid recurrence of this event:
1.
Evaluate vendor recommendations as soon as they are available to possibly reduce or eliminate the sensor drift problem.
2.
During various normal plant operating evolutions, measurements of " hydraulic noise" in main steam line sensing lines will be e
made.
Based on the amount of " hydraulic noise" present, an operating set point will be selected above 850 psig which will provide some reasonable operating margin to avoid spurious trips.and still tolerate some downward drift in instrument set point.
Enclosed are forty copies of this report.
' Very truly yours,
&J Donald A. Ross Manager, Nuclear Generating Stations cs J
cc:
Mr. J. P. O'Reilly.. Director
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