ML20041G129

LER 82-005/03L-0:on 820210,at 98% Power,Control Element 24 Inserted Into Core.Emergency Procedure EP-13,CEDM Malfunctions,Implemented & Power Stabilized at 88%.Caused by Erroneous Operating Instruction.Instruction Changed

ML20041G129
Person / Time
Site:	Fort Calhoun
Issue date:	02/22/1982
From:	Gates W OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
Shared Package
ML20041G126	List: ML20041G125 ML20041G129
References
LER-82-005-03L, LER-82-5-3L, NUDOCS 8203190301
Download: ML20041G129 (4)

Text

U. S. NUCLE A3 REIULATORY COMMISSION NRC FORM 46 ,

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LICENSEE EVENT REPORT CONTROL BLOCK: l i

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lh (PLE ASE PRINT OR TYPE ALL REQUIRED INFORMATION)

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67 CAT 68 lg F 8 9 14 15 LICENSE NUMBEH CON'T lolil S$y"E l Igl 0l 5l 0l 0l 0l 2l 8l 5@l 68 69 0l 2lEVENT1lDATE 0l 8l 2l@l0 l2REPORTDATE F4 75 l2 l2 l8 l2 80l@

F S 60 61 DOCK ET NUMBER EVENT DESCRIPTION AND PROB ABLE CONSEQUENCES h O 2 l While operating at 98% power, control element 24 inserted fully into the core. l 0 3 l Emergency Procedure EP-13. "CEDM Malfunctions," was implemented and power stabilized l 10 l l l at 88%. The event occurred at 0938 and the operator started reducing power at 0939.l 0 3 l Technical Specification 2.10.2(4)e requires power to be < 70% within I hour of the l lo Is l l CEA drop. Reactor power was stabilized at < 70% 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 4 minutes following the l 0 7 l CEA drop. l l

10181I 80 7 8 9 SYSTEM CAUSE CAUSE COMP. VALVE CODE CODE SU8 CODE COMPONENT CODE SUSCODE SU8 CODE O 9 8

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9 to 11 12 13 18 19 20 7

SE QUE NTI AL OCCURRENCE REPORT REVISION EVENT YEAR REPORT NO. CODE TYPE NO.

O ,ajg/RO LER 1812l [__j l0l0l5l g l013l 28 23 g

30 g

31 g

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_ 24 22 23 24 2b 27 HOURS 22 SB IT FOR 8. SUP ER MAN FACTURER K N A T ON gON PL NT MET l_C 33,_ }hl J4 A lh 35 h dh 36 l 0l 0l 5l 31 40 l gg 41 gg 42 43 g lZlZlZlZlg 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS h g] The operating instruction OI-RR-2 directs the operator to use a controlled boration l to reduce power. lhis was accomplished by blending boric acid in the CVGb system.

[TTT1 1 Response was not rapid enough and make-up was switched to the safety injection re- l iueling water tank providing > 1700 ppm boric acid solution to the RCS. Power re-j , y l duction rate increased and power stabilized at < 70%. To ensure rapid power re- l

duction 01-KK-Z has been changed to direct the operator to use SIRWI as make-up

, 3 if normal boration is inadequate. l l

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8 9 80 l 7 A S  % POWER oTHER STA TUS ISCO Y DISCOVERY DESCRIPTION

] gl Operator Observation l i 5

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PERSONNEL FxPOSUHES 11 44 45 80 f

NUMBE R TYPE OESCRIPTION i i l0l0l0l@lZl@l NA l 8 8

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C- 'l NuMBE R i a NA l F M 9 l 0 l 0 l 011l@l12 80 LOSS OF OR DAMAGE TO FACILITY l

TYPE OESCRIP floN i o NA l l LZ J@lto 80 i7 8 9 2 O W l Press ISSUED @ DESCRIPTIONrelease on plant outa y. I 88 88 lllllllllllll%2 8 5 l ' 8203190301 820311 PDR ADOCK 05000285 C. Gates PHONE:

402-426-4011 l, l S PDR

LER No.82-005 Omaha Public Power District-Fort Calhoun Station Unit No. 1 Docket No. 05000285 ATTACIDfENT NO. 1 Safety Analysis The reactor power was approximately 71.2% within one hour of drop-ping the CEA. Technical Specification 2.10.2(4)e requires power to be less than 70% within one (1) hour of the CEA insertion. The 70% within one hour limit is imposed to reduce the possible effects of localized power peaking caused by an imbalance in the core flux and attendant

• temporary Xenon redistribution, caused by dropping the most reactive CEA. CEA No. 24 is 60% less reactive than the most reactive dual element CEA; therefore, the four (4) additional minutes required to reduce reactor power to less than 70% did not have a negative effect on fuel performance.

Activity levels in the primary. coolant system were measured and indicated fuel failure did not result from the dropped CEA.

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F LER No.82-005 Omaha public Power District Fort Calhoun Station Unit No. 1 Docket No. 05000285 ATTACllMENT NO. 2 Corrective Action OI-RR-2, " Dropped CEA Recovery," used to respond to the dropped CEA on February 10, 1982 requires reactor power to be reduced by controlled boration. As a result of the February 10 event, OI-RR-2 has been modified to direct the operator to use make-up from the Safety Injection Refueling Water Tank (SIRWT) to accomplish the power reduction if a sufficient rate of power decrease cannot be established by normal boration. During normal boration procedures, there is an approximate 10 to 12 minute delay before the plant starts to react to a boron addition.

There is also a short time period following that before the operator can assess the power reduction rate. Depending upon time in the core life and effects of the rod drop itself, normal boration can reduce power to less than 70% within the one hour time limit. The operator will deter-mine whether or not to use the SIRWT for boration based upon the magnitude of the initial power reduction caused by the dropped CEA and the time in core life. This method of reducing reactor power will ensure power is reduced to < 70% within one hour as required by Technical Specification 2.10.2.

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e LER No.82-005 Omaha Public Power District Fort Calhoun Station Unit No. 1 Docket No. 05000285 ATTAC"".ENT NO. 3 Failure Data This is the first instance at Fort Calhoun Station involving a failure to meet a shutdown or power reduction time limit of a Limiting Condition for Operation. Additionally, this is the first CEA drop event at the Fort Calhoun Station where reactor power was above 70% at the time of drop.

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