ML20045D245

LER 93-003-00:on 930520,plant Experienced Trip from Approx & Continuity Between Fuse & Fuseholder Lost,Resulting in Loss of Selected RCS Average Temp Input to Ics.Faulty Fuse Cap & Fuse replaced.W/930617 Ltr

ML20045D245
Person / Time
Site:	Davis Besse
Issue date:	06/17/1993
From:	Storz L, Turkal M TOLEDO EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
AB-93-0020, AB-93-20, LER-93-003, LER-93-3, NP-33-93-03, NP-33-93-3, NUDOCS 9306280126
Download: ML20045D245 (6)

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• TOLEDO

%ms EDISON ED! SON PLAZA - l 300 MADISON AVENUE J

- TOLEDO, OHIO 43652-o001 1

i AB-93-0020~

NP-33-93-03 l Docket Number 50-346 License Number NPF-3

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June 17, 1993 United States Nuclear Regulatory Commission  !

Document Control Desk j Vashington, D.C. 20555  ;

i Gentlemen: t LER 93-003 Davis-Besse Nuclear Power Station,. Unit Number 1- .

Date of Occurrence --May 20, 1993 Enclosed please find Licensee Event Report.93-003, which is being submitted to' provide 30 days written' notification of'the. subject occurrence. .This LER is being submitted in accordance with 10 CFR 50.73(a)(2)(iv). .

7 Very truly yours, ouis F. Storz Plant Manager Davis-Besse Nuclear Power Station  !

LFS/lkg enclosure cc: Mr. J. B. Martin Regional Administrator USNRC Region III Hr. Stan Stasek DB-1 NRC Senior Resident Inspector 240083

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NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (S ezi .

EXPIRES 5/31/95 E STIMATED BURDEN PE R FISPONSE TO COMPLY WITH THl$ 5

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INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMvENTs ntuRD,Na suRDEN EsmATe TO 1HE iv0nMATcN AND RECORDS MANAGEMENT BAANCH (MNBA 7714), U S. NUCLEAA REGULATORY COMMISSION, W ASHINGTON, DC 20555-0001. AND TO THE PAPERWORK REDUCTION PRCLIECT (31504104), OFFtCE or (See reverse for requwed number of digits / characters for each block) MANAGE MENT AND BUDGET, WASHWGTON, DC POSu FACILITY NAME p) DOCKET NUMDER (2) PAGE (3)

Davis-Besse Unit Number 1 05000-346 1 OF 4 TITLE (4)

Reactor Trip - Loss of ICS Tave Input EVENT DATE (5) LER NUMBER (6i REPORT NUMBER (7) OTHEF< FACILITIES INVOLVED (8)

F ACluTY NAME DOGKET NUMBER SFOUENTAL REVISION MONTH DAY YEAR YEAR NUMBER NUMBER

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05000 FACluTY NAME DOCKET NUMBER

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05 20 93 93 003 00 06 17 93 05000 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 6: (Check one or more) (11)

MODE (9) 1 20.402(b) 20.405(c) X $0.73(a)(2)(iv) 73 71(b)

POWER 20 405(a)(1)(i) 50.30(c)(1) 50.73(a)(2)(v) 73.71(c)

LEVEL (10) 100 20.405(a)(1)(ii) 50.361c)(2) 50.73(a)(2)(vii) OTHER 20.405(a)(1)(ud 50.73(a)(2)(i) 50.73(a)(2)(viii)tA) WP'c@

• Ab**c' 20.405(a)(1)0v) 50,73(a)(2)(n) 50.73(a)(2)(viii)(B) $N6A 20 405(a)(1)(v) [ 50.73(a)(2)(ui) 50.73(a)(2)(x)

LICENSEE CONTACT FOR THIS LER (12)

NAME T& HONE NUMBEH pcmae Area Goce)

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Mark A. Turkal, Engineer - Nuclear Licensing (419) 321-7377 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEM COMPONENT MANUFACTURER . CAUSE SYSTEM COMPONENT MANUFACTUAER ONR l

l SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONT H DAY YLAR YEs SUBMISSION

(" Y+8 ""imee EXPECTED SUGMISS!ON DATE) X DATE (15)

ABSTRACT (Limit to 1400 spaces, i e., approximately 15 single-spaced typewntten Ones) (16)

On May 20, 1993, at 1138 hours0.0132 days <br />0.316 hours <br />0.00188 weeks <br />4.33009e-4 months <br />, the plant experienced a trip from approximately 102% power. Loss of continuity between a fuse and fuseholder resulted in a loss of power to an auxiliary relay in the Non-Nuclear Instrumentation (NNI) System. This caused a loss of the selected RCS average  ;

temperature (Tave) input to the Integrated Control System (ICS). The ICS Tave l input failed to zero VDC indicating that the RCS average temperature was approximately 570 degrees F. Normal RCS average temperature is 582 degrees F. l l

To compensate, the ICS Tave integral increased Reactor Demand, withdrawing control rods and increasing reactor power. In an attempt to reduce reactor l power, operators took manual control of ICS and decreased the Steam Generator / Reactor Demand signal. The primary reactor operator overreacted and lowered demand too far. Primary to secondary heat flow became unbalanced as l feedvater flow followed the lowered demand, but reactor power did not decrease due to the low Tave input. As feedvater flow decreased, RCS pressure increased. The reactor tripped on high RCS pressure.

The faulty fuse cap and fuse vere replaced and other NNI fuse holders and fuses vere inspected. The plant was returned to service on May 21, 1993 at 1308 hours0.0151 days <br />0.363 hours <br />0.00216 weeks <br />4.97694e-4 months <br />.

NGC FORM 366 (5 92)

REQUIRED NUMBER OF DIGITS / CHARACTERS FOR EACH BLOCK ,

BLOCK NUMBER OF TITLE NUMBER DIGITS /CHARACTEDS 1 UP TO 46 FACILITY NAME 8 TOTAL DOCKET NUMBER 3 IN ADDITION TO 05000 3 VARIES PAGE NUMBER 4 UP TO 76 TITLE

^ EVENT DATE 5

2 PER BLOCK 7 TOTAL

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6 LER NUMBER 3 FOR SEQUENTIAL NUMBER 2 FOR REVISION NUMBER 6 TOTAL I REPORT DATE 2 PER BLOCK UP TO 18 - FACILITY NAME OTHER FACILITIES INVOLVED 8 TOTAL - DOCKET NUMBER 3 IN ADDITION TO 05000 9 1 OPERATING MODE 10 3 POWER LEVEL ,

l' REQUIREMENTS OF 10 CFR CHECK BOX THAT APPLIES UP TO 50 FOR NAME 12 LICENSEE CONTACT 14 FOR TELEPHONE CAUSE VARIES 2 FOR SYSTEM 13 4 FOR COMPONENT EACH COMPONENT FAILURE 4 FOR MANUFACTURER NPRDS VARIES SUPPLEMENTAL REPORT EXPECTED CHECK BOX THAT APPLIES 6 TOTAL 15 EXPECTED SUBMISSION DATE 2 PER BLOCK 1

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NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 ts $2;

• EXPlRES 5/31/95 ESTtMATED BURDEN PER RESPONSE TO COMPLY WITH TH!S LICENSEE EVENT REPORT (LER) 1* fj0E#n#2 "RD#fSEETOZ',Nf"NS TEXT CONTINUATION Eu"L[N u E N No"TE*E"'s'N'iiNDI$

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Davis-Besse Unit Number 1 93 003

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Description of Occurrence:  !

On May 20, 1993, the plant was in Mode 1 at approximately 100% reactor power.

The event began at 1135 hours0.0131 days <br />0.315 hours <br />0.00188 weeks <br />4.318675e-4 months <br /> when loss of continuity between a fuse and fuseholder resulted in de-energizing of auxiliary relay 86/RC7A in the Non-Nuclear Instrumentation (NNI) System. Loss of power to the auxiliary '

relay caused the loss of the selected Integrated Control System (ICS-JA)

The ICS Reactor _ Coolant System (RCS-JD) average temperature (Tave) input.

provides automatic coordination of the reactor, steam generator feedvater control, and turbine. The ICS includes four independent subsystems: (1) tht Unit Load Demand Control, (2) the Integrated Haster Control, (3) the Steam Generator Control, and (4) the Reactor Control. The four ICS subsystems were automatically controlling the plant when Tave failed.

The loss of Tave resulted in an erroneous input to the ICS indicating that the RCS average temperature was approximately 570 degrees F. Normal RCS average temperature is 582 degrees F, which.vas the approximate RCS average temperature at this time. As a result, the ICS attempted to recover normal RCS average temperature by withdrawing control rods, increasing reactor power '

to approximately 102%. The ICS includes a Reactor Demand High Power Limiter which prevents the ICS from demanding a power greater than 103% to prevent exceeding an RPS channel high flux trip setpoint.

In a proper attempt to reduce reactor power, operators took manual control of the ICS and decreased the Steam Generator / Reactor Demand signal. The primary reactor operator erred, however, as he overreacted to the conditions and manually lowered demand to approximately 80% of full power without properly informing the balance of the operating crev. Primary to secondary heat flow became unbalanced as feedvater flow decreased without a corresponding decrease in reactor power due to the low Tave input failure. As feedvater flow decreased, RCS pressure increased. The Reactor Protection System (JC) properly tripped the reactor on high RCS pressure (trip setpoint - less than or equal to 2355 psig) at 1138 hours0.0132 days <br />0.316 hours <br />0.00188 weeks <br />4.33009e-4 months <br />. Maximum indicated RCS pressure was 2350 psig based on plant computer data.

Redundant sensors for major system parameters are available to the ICS. The ICS inputs are monitored by 19 Smart Analog. Selector Switch (SASS) channels designed to automatically transfer to a redundant sensor in case of instrumentation problems. The SASS channel monitoring RCS average temptrature did not transfer in this event because the failure was downstream of the SASS.

The plan was returned to service on May 21, 1993 at 1308 hours0.0151 days <br />0.363 hours <br />0.00216 weeks <br />4.97694e-4 months <br />.

NRC FORM 366A (5-92)

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I NRC FORM 3fA6A U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 Hz, '

EXPIRES 5/31/95 l ESTIMATED BURDEN PER RESPONSE TO COMPLY w'TH THIS LICENSEE EVENT REPORT (LER) 'ET@s%&

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SEQUENitA. RE v@ON NUMBER NUMBER 05000-346 3 OF 4

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Davis-Besse Unit Number 1 93 003 00 T ext m .c.c. ww.am. .umi,w .., mC ,,,,n my p r3 l Description of Occurrence: (con't)

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l Initial notification of the reactor trip was made on May 20, 1993, at 1258 hours0.0146 days <br />0.349 hours <br />0.00208 weeks <br />4.78669e-4 months <br />, in accordance with 10 CFR 50.72(b)(2)(ii). This LER is being submitted i in accordance with 10 CFR 50.73(a)(2)(iv).

l l Apparent Cause of Occurrence I

The plant transient was initiated by a loss of continuity between a fuse and fuseholder in the circuit that powers an auxiliary relay for Tave input selection to the ICS. The reactor trip resulted from operator action taken in an attempt to decrease reactor power.

A loss of continuity between a fuse and fuseholder resulted in de-energizing of auxiliary relay 86/RC7A in the NNI system. Loss of power to the auxiliary relay caused the loss of the selected ICS Tave input which provided an l

erroneous input to the ICS indicating that the RCS average temperature was approximately 570 degrees F. Normal RCS average temperature is 582 degrees F.

As a result, the ICS attempted to recover normal RCS average temperature by withdrawing control rods and increasing reactor power to approximately 102%.

In an attempt to reduce reactor power, operators took manual control of ICS and decreased the Steam Generator / Reactor Demand signal. Feedvater flow decreased, however, reactor power did not due to the lov Tave input. As feedvater flow decreased, RCS pressure increased. The RPS functioned properly by tripping the reactor on high RCS pressure.

I Analysis of Occurrence The event reported in this LER has minimal safety significance.

Plant and operating crew response to the trip was satisfactory. The Control Rod Drive trip breakers opened and all control rods inserted on the reactor trip as designed. Steam generator pressure increased due to the closing of the main turbine stop valves. The Turbine Bypass Valves and the Atmospheric Vent Valves (AVVs) properly opened and the Main Steam Safety Valves (MSSVs) lifted in response to the increasing secondary system pressure. The MSSVs and AVVs closed as . steam generator outlet pressure decreased.

Shortly before the reactor trip, there was an undetected lockup of the Safety Parameter Display System (SPDS). As a result, the SPDS was not available to the control room operators post-trip. The failure of the SPDS was not associated with the reactor trip and did not affect the operators' ability to respond to the event. The SPDS was returned to service at approximately 1150 hours0.0133 days <br />0.319 hours <br />0.0019 weeks <br />4.37575e-4 months <br /> and pertinent transient data is available.

NRC FORM 366A @92)

NRC FORM 3fnEA U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 H2> * '

rXPIRES 5/31/95 ESTIMATED BJRDEN PER RESPONSE TO COMPLY w'TH TH:s INFORMATION Cou1C'ON RE QUE ST $3 0 HRS FOAWARD LICENSEE EVENT REPORT (LER) CoMveN s REsiRo,N3 suotN EsT, MITE TO THE ,NFOnMAvoN TEXT CONTINUATION ^'* "'C N' ""^NCH ma nni. u s NuCttAa

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Corrective Action i

The faulty fuse cap and fuse vere replaced and the remaining 37 NNI fuse f holders and fuses associated with SASP instrumentation vere inspected. No additional faulty fuse holders or fuses were discovered, however, seven fuses and one fuse cap were replaced as a precautionary measure.  !

l Operations personrel completed required reading regarding the failure of Tave, l the plant's response, and general guidance for ICS transients.

Subsequent to the event, simulator training on the transient was initiated.

Each operating crev vill complete this training by June 25, 1993. Control room team training, in association with INPO, is scheduled for completion by July 30, 1993.

Failure Data There have been no LERs due to ICS initiated transients in the previous three years. Also, there have been no LERs in the previous three years due to reactor trips resulting from operator action.

NP 33-93-003 PCAO No. 93-0311 1

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NRC FORY 3664 (S 921