ML20155C270

Special Rept on Untimely Submittal of 19 Lers.Training on Safety Responsibilities to Begin Immediately Following Development of Training Plan

ML20155C270
Person / Time
Site:	Fort Calhoun
Issue date:	10/03/1988
From:	Morris K OMAHA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
LIC-88-666, NUDOCS 8810070331
Download: ML20155C270 (27)

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Omaha Public Power District 1623 Ha:ney Omaha. Nebraska 68102-2247 402 536 4000 October 3, 1988 LIC-88-666 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station P1-137 Washington, DC 20555

Reference:

Docket No. 50-285

SUBJECT:

Special Report on the Untimeiy Submittal of Licensee Event Reports Gentlemen.

The Omaha Public Power District (OPPD), holder of Operating Licensa GPR-40, I submits this report to document the untimely submittal of Licensee Event Reports.

l l In November 1987, the process for handling Operations Incident (01) Reports was l revised and computerized to facilitato a more timely review of the new Incident Reports (!R). to assign responsibility for tracking IR's and ensure timely review. In February 1988, OPPD organized a group to close out a backlog of 343 open Ol's which had been generated prior to November 1987. These Ol's had been I reviewed foe reportability at the time of the occurrence, but hau not been i

closed out. The group was tasked to review the circumstances of each incident, I

review again for reportability and perform a safety analysis per 10 CFR 50.59.

l During the review process nineteen incidents were discovered to have been I potentially reportable under 10 CFR 50.72 and/or 10 CFR 50.73 in effect at the l time of occurrence. The LER process is designed for reporting current events l not historical events; therefore, this special report is being submitted to report these events. Further investigation revealed that 15 of these 19 items were reportable under the reporting criteria in effect at the '.ime of the event.

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l V. S. Nuclear Regulatory Coninission LIC-88-666 Page 2 The attachment includes the 19 items provided to the NRC Senior Resident Inspector. Each item is detailed as follows:

a. Event Description - Brief description of the event based on the original O! I and other available references,

b. Immediate Actions Taken - Immediate actions taken as a result of the event based on the original 01 and other references, i.e., Control Room Log.

c. Corrective Actions - Actions proposed or implemented based on information contained in available documents. Not all corrective actions stated could be verified due to the time span since the event occurred. In those instances, corrective actions which have been taken as a result of more recent similar events are included to address the measures taken to prevent future occurrences.

d. Nuclear Safety Significance - An assessment of the safety consequences and implications of the event is given.

e. Basis for Reportability - The applicable section of the Technical Specification, 10 CFR 50.72 and/or 10 CFR 50.73, is stated.

f. Reason for Not Reporting - The reason most of the 15 reportable items were not reported can be attributed to an inadequate O! review and tracking  ;

process. Other reasons are included if they could be determined. j i

Several actions have been taken to improve the review and reporting process of l Incident Reports. These are broken down into two areas. '

1. The IR Tracking Process The IR procest his been computerized since November 1987. The 19 events occurred prior to this tin.e. {

The IR's are initiated on computer data base !

and go through an extensive review process. Initially, the originator and/or the Shift Technical Advisor (STA) write the IR, and the STA consults l i

with the Shift Supervisor on the reportability of the event. The IR is then reviewed by the Incident Evaluation Coordinator for reportability, and t

he assigns an action addressee. The reports are then presented to the Plant Review Committee within the week. Nuclear Licensing & Industry ,

Affairs also reviews IR's periodically and is notified upon the '

determination of a reportable item. Also upon determining an item is reportable, the STA assesses the safety significance of the event if the event places the plan 2 in an abnormal situation or for any plant parameter that affects or reflects an abnormal indication of a safety-related system. This )rocess should preclude an inaccurata determination of reportability )eing made.

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U. S. Nuclear Regulatory Commission LIC-88-666 Page 3 I

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2. The Plant Review Committee (PRC) Training Program

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A PRC training plan will be developed by the end of October 1988. Training on safety responsibilities of PRC mer.bers will begin immediately following I the development of the training plan. PRC member training will begin in January 1989 and will be completed by July 1989. This training will include the determination of reportability and should preclude future errors in this area.

If you have any questions, do not hesitate to contact us.

Sincerely, Wh y Ic.K.J. Morris

/ Division Manager Nuclear Operations KJM/mc Attachment c: LeBoeuf, Lamb, Leiby & MacRae I

1333 New Hampshire Ave., N.W.

Washington, DC 20036 R. D. Martin, NRC Regional Administrator P. D. Milano, NRC Project Manager P. H. Harrell, NRC Senior Resident inspector i

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INADVERTENT ACTUATION OF 480V LOAD SHED I 01 001319

a. Event Description On May 26, 1981, at 0921, at Fort Calhoun Station, channel "A" 480V load l shed was initiated. This occurred during the performance of Surveillance .

Test ST-ESF-2 F.1, "Channel 'A' Safety Injection Actuation Signal Test,"  !

which verifies the proper operation of the initiation circuitry and all l equipment normally operated by channel "A" safety feature actuation  !

signals. This test requires the operator to turn the 480V Load Shed Switch  !

CS-A/LS on Panel Al-30A to 0FF prior to initiating a test signal. However, I the operator did not block the Load Shed prior to initiating the test i i signal and the Load Shed was activated. A)propriate nonessential motors, (

) pumps, etc. on channel "A" were properly sled.

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b. Immediate Actions Taken t

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The loads that were shed during this event had their power sup)1y returned -

i to normal once the test signal was removed and the 480V Load Sled Switch '

4 CS-A/LSwasreturnedtoEMERGENCYSTANDBY(itsnormalposition).

c. Corrective Actions This event occurred because approved procedures were not followed. The i s)ecific corrective action for this event could not be determined because (

tie individual and his immediate supervisor are no longer employed by OPPD.

l, j d. Nuclear Safety Sionificance '

I This event had no effect on nuclear safety. The loads shed were

-!l non-engineered safeguards and are not required for the safe shutdown of the  !

j plant.  ;

i e. Basis for Reportability 10 CFR 50.72(a)(7) (year 1981) required licensees to report, "Any event j resulting in manual or automatic actuation of Engineered Safety Features, i including the Reactor Protection System." Since the 480V Load Shed is an j Engineered Safety Feature and was not actuated as part of a preplanned j sequence, this event was reportable, j f. Reason for Not Reportina j An error was made determining the reportability during the initial review 4

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9 INOPERABLE SNUBBERS ON SHUTDOWN COOLING SYSTEM O! 001384

a. Event Description

! At approximately 1100 on September 29, 1981, during refueling shutdown i

conditions, it was observed that six Technical Specification Snubbers were I

inoperable on the shutdown cooling system. This system is required to be i i operable during Modes 4 and 5. The snubbers had been taken down for redesign and/or replacement.

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b. Immediate Actions Taken GeneratingStationEngineering(GSE)immediatelybeganreturningallofthe inoperable snubbers to service. The last of the six snubbers in question was returned to service at 2000 on September 29, 1981. i

) This was completed within nine hours from the time of discovery.

c. Corrective Actions i

No corrective actions were taken other than the immediate actions taken.

d. Nuclear Safety Sionificance i

1 The shutdown cJoling system is required to be operable during Modes 4 and i 5.

] e. Basis for Reoortability i NOT REPORTABLE l

l There was no Technical Specification requirement for snubber operability 1

while in the refuelin mode of the time of the event. No Technical Specification was vio ated.

Reference:

Fort Calhoun Station Technical I SpecificationAmendment48Section2.18(1).

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f. Reason for Not Reportino l

The incident was not reportable.

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VHRA 000R LEFT OPEN O! 001571

a. Event Description On August 6,1982 at 1415 a door to a very high radiation area was lef'.

open. The door, shield door in RM-11, was left open after the pressure equipment group transferred filters to drums behind the door. The door was logged open at 1415 on August 6, 1982. The same alarm was also verified active on August 7 and August 8 and had not been cleared. The wrong door was visuelly checked shut by both operations and security personnel. Since it appeared that the door was secure and the personnel checking the door were unaware that there were two very high radiation area doors associated with Room 11 on the circuit, a maintenance order was issued to fix the circuit. At 0900 on August 9, 1982 the M0 was reviewed by a Health Physics representative. He had health physics personnel check to see that both doors were shut. At 0930 it was determined the inside door was still o )en. This could not be determined from outside the step off pad, due to t1e distance (about 21 feet) and the lack of illumination,

b. Immediate Actions Taken The door was shut, and the alarm cleared.

c. Corrective Actions Since the occurrence of this incident the following corrective actions have been taken:

1. Red painted padlocks are used to secure doors which allow entrance into a VHRA. The keys are controlled by the on duty shift supervisor, auxiliary building operators, and plant health physicist.

2. A sign has been posted on the VHRA doors stating, "HP Technician Required", "Two Persons Required for Entry", and "FC-647 Form Must Be Completed Upon t.xit".

3. The Radiation Protection Manual has been revised to require the shift health physics technician to perform a documented physical check to ensure the doors are properly locked every four hours.

4. The Radiation Protection Manual now requires that after entry a qualified health physics technician will check the very high radiation area door and ensure that it is closed and locked.

5. The Manual also requires that one of the individuals accompanying the health physics technician into the VHRA will also check that the VHRA door is closed, latched, and locked.

d. Nuclear Safety Sianificance These doors are kept locked and surveillance is maintained to prevent people from accidentally entering the very high radiation area.

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e. Basis for Reportability This was a violation of Technical Specification 5.11.2. This event was report?.ble pursuant to 10 CFR 50.72(a)(4) (year 1982),

f. Reason for Not Reportina An error was made in determining the reportability during the initial review of the 01.

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BOTH DIESEL GENERATORS PLACED IN OFF AUTO CONDITION 0! 001686

a. Event Descriotion, On March 29, 1983 at 0830, at Fort Calhoun Station, Emergency Diesel Generators 0-1 and 0-2 were ) laced in an off auto condition to prevent their starting during trip c1ecks on the 345 kV System. Diesel Generators are designed to provide station )ower in the event of a loss of the 161 kV (offsite)and345kVSystems. T1eReactorCoolantSystem(RCS) temperature was day. atFortCalhounStationTechnicalSpecification2.71)1 180*F at 0830 but was increased to 380'F during(the course of the requires,in part, that the reactor shall not be heated up or maintained at a temperature above'300'F unless both diesel generators are operable. Since the reactor was heated above 300'F with tne Diesel Generators inoperable, TechnicalSpecification2.7(1)1wasviolated,

b. Imediate Actions Taken Both Diesel Generators were switched from off auto to auto standby at 1558, March 29, 1983, and trip check testing of the 345 kV System was suspended uponthediscoveryofTechnicalSpecification2.7(1)1 violation,

c. Corrective Action 1 Since the occurrence of this event, there has been increased emphasis on pencedure compliance. Operator training has been improved with respect to Technical Specification requirements,

d. Nuclear Safety Sionificance Nuclear safety was not impacted due to the fact that the safeguard equipment powered by the diesel generators is not required by the Technical Specifications until the reactor is critical,

e. Basis for Reportability Violation of Technical Specification Limiting Condition for Operations 2.7()1. This event was reportable pursuant to 10 CFR 50.72(a)(5) (year 1983 .

f. 'teason for Not Reportina f An error was made in determining the reportability during the initial review of the 01.

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MISSED SURVEILLANCE TEST 0! 001847 l

a. Event _Descriotion On March 24, 1984 and March 25, 1984, while in the refueling mode, the documentation for ST,SDM-1 was not completed. This surveillance test is used to ensure that the reactor will be maintained subcritical to preclude accidental criticality in the shutdown condition. Upon investigation it was determined that the data for March 24 and March 25, 1984 was taken.

However, the worksheets which are attached to ST-SDM-1 were not completed until March 26, 1S84. This event resulted from a failure to follow approved procedures and resulted in a violation of Technical Specifications 3.10. Technical Specification 3.10 requires that the shutdown margin be determined daily,

b. Immediate Agtion Taken Verified Boron concentration on March 24, 1984 and March 25, 1984 was i adequate from primary chemistry reports,

c. Corrective Actions A new surveillance test tracking procedure using manual scheduling and computer tracktng has been in place on a trial basis since February 1988.

This procedure will be incorporated into Standing Order G-23 following approval of Procedure Change 22548.

d. Nuclear Safety Sianificance l l

The Boron Dilution Event during refueling is analyzed in Section 14.3.2.5 of the USAR. The shutdown margin did not fall below 1900 PPM boron during i this event. This is sufficient baron to keep the reactor suberitical.

Nuclear safety was not affected by this event.

e. Basis for Reportability Technical Specification 3.10 was violated. 10 CFR 50.73(a)(2)(i)(B) (year 1984) stated, "Any operation or condition prohibited by the plant's Technical Specifications."

f. Reasca for Not Reportina It was assumed that the event was not reportable because the tests had actually been performed, the data was satisfactory, and the event concerned only timely documentation.

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MISSED SURVEILLANCE TEST i 01 001869 I i

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a. Event Description The monthly surveillance test that verified that the spare battery charger  ;

, operates correctly when connected to either bus, was not completed within '

l its allowed time. ST-DC-2 was scheduled on April 12, 1984 and was not l completed until April 20, 1984.  ;

l Technical Specification 3.1 states that "A maximum allowable extension not I to exceed 25% of the Surveillance Interval unless otherwise specified".

Technical Specification 3.7.(2).c states that at monthly intervals the third battery charger shall be paralleled in turn to each bus. The surveillance test was not completed within the time allowed. ,

b. Immediate Actions Taken ST-DC-2 was performed on April 20, 1984.

c. Corrective Actions A new surveillance test tracking procedure using manual scheduling and computer tracking has been in place on a trial basis since February of 1988. This procedure will be incorporated into Standing Order G-23 following approval of Procedure Change No. 22548.

d. Nuclear Safety Sionificance The surveillance testing specified verifies that the spare battery charger operates correctly when connected to either bus. The specified testing interval is considered adequate, given the system redundancy to detect and correct any malfunction before it adversely affects the station battery system.

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e. Basis for Reportability This event was This event was a violation of Technical Sp(ecification 3.1.

reportablepursuantto10CFR50.73(a)(2)i)(B)lant'sTechnical(year "Any operation or condition prohibited by the p 1984)whichstated, Specifications."

f. Reason for Not Reportino An error was made in determining the reportability during the initial review of the 01.

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LOSS OF 0FFSITE POWER DUE TO WEATHER CONDITION  :

01 001885

a. f.yint Description ,

On A)ril 29, 1984 Fort Calhoun was in refueling shutdown and station power ,

was )eing supplied by the offsite 161 kV lines. The 345 kV li.1es were tagged out. Off-site power was lost due to lightning. The diesel 3

generators and associated breakers actuated to supply house power.  :

1 b. Immediate Actions Taken Restarted r.ecessary equipment to maintain plant conditions in their nnrmal l refueling stutdown condition. The NRC was notified by phone at 2110, April l 29, 1984.

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c. CorrectiveActjgji, No further corrective action were taken since the system functioned as i designed.

d. Nuclear Safety Significance Theresponseoftheemergenc$'electricalsupplysystemwasasdescribedin l Section 8.1.2 of the USAR.

e. Besis for Reportability 10CFR50.72(b)(2)(ii)and10CFR50.73(a)(2)(iv)(year 1984), require licensees to report, "an event or condition that results in the manual or ,

automatic actuation of a)ny Engineered Safety Feature (ESF), including the l

ReactorProtectiveSystem(RPS)."

f. Reason for Not Reportina .

I There was a failure to follow up and the LER was not sent. }

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SURVEILLANCE TEST PERFORMED LATE 01 001906

a. Event Descriotion Surveillance tests ST-RM-1 F.2 and ST-RM-2 F.2, monthly tests of area and process radiation monitors, scheduled for May 2, 1984, were not completed until May 11, 1984. Technical Specification 3.1 requires these test to be performed monthly with a maximum allowable extension of 25t of the surveillance interval. This allowable extension was exceeded for these tests and Technical Specification 3.1 was violated.

This event occurred because the technicians performed annual surveillance tests before doing the monthly surveillance tests,

b. Immediate Actions Taken The tests were completed on May 11, 1984, and no irregularities were identified,

c. Corrective Actig.01 A new surveillance test t ,iing procedure using manual scheduling and computer tracking has been in place on a trial basis since February 1988.

This procedure will be incorporated into Standing Order G-23 fellowing approval of Procedure Change 22548.

d. Nuclear Safety Sionifican.c_ee Surveillance tests ST-RM-1 F.2 and ST-RM-2 F.2 ensure the operability of the process and area radiation monitors important to nuclear safety.

e. Basis for Reportability This event was a violation of Technical Specification 3.1, Table 3.3. This eventwasreportablepursuantto10CFR50.73(a)(2)(B)(year 1984)which stated, "Any operation or condition prohibited by the plant's Technical Specifications."

f. Reason for Not Reportina An error was made in determining the reportability during the initial review of the 01.

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6 COOLING AIR SUPPLY VALVES FAIL CLOSED O! 001981  ;

a. Event Descriotion l Air Inlet Dampers YCV-871G and YCV-871H for Diesel Generator Room Number 1 are required to open upon the loss of instrument air or loss of power. l However, it was discovered on August 14, 1984 that upon the loss of air, both dampers fail closed. Diesel Generator No. 2 is not invclved.

b. Immediate Actions Taken Both dampers were blocked open.

c. Corrective Action i The field work for design modification FC-84-151 was completed on i Nevember 27, 1984. This modification changed the failure mode of the valves to fail open.  :

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d. Nuclear Safety Sianificance Paragraph 8.4.1.1 of the USAR itates: "The emergency diesel generators are l designed to furnish reliable in-plant a-c power adecuate for safe )lant L shutdown and standby and for operation of engineerec safeguards, w1en no {

energy is available for the 345 and 161 kV systems. For adequate  ;

reliability two units are provided." (

e. Basis for Reoortability

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When YCV-371G and YCV-871H fail close, the design basis of 8.4.1.1 of the  :

USAR is violated. This event was reportable ursuant to I 10CFR50.72(b)(1)(ii)(B)and50.73(a)(2)(ii)p(B)(year 1984).  !

f. Reason for Not Reportina An error was made in determining the reportability during the initial I review of the 01.

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. l MISSED SURVEILLANCE TEST 01 002007 ,

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a. Event Descriotion Surveillance test ST-FP-9, quarterly ins)ection of temporary fire barriers, due April 10, 1984 was not completed. T1e next scheduled test was performed on July 10, 1984, and no irregularities were identified.

Technical S)ecification 3.15(5)a, which requires that this visual inspection )e performed at least once per 18 months, was not violated. '

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b. Immediate Actions T' ken  :

At the time of the missed surveillance test, the plant was in an extended forced outage with the reactor in refueling shutdown condition. The temporary barriers were verified functional before the reactor was made critical.

c. Corrective Actions

, Surveillance Test ST-FP-9 F.1 was revised September 24, 1988 to make it easier to follow,

d. Nuclear Safety Sionificance t

Fire detection and fighting systems are provided to minimize the adverse effects of fires on structures, systems, and components im)ortant to safety. Surveillance testing insures the operability of t1ese systems.

e. Basis for Reportability NOT REPORTABLE.

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f. Reason for Not Reportina '

] TcchnicalSpecification3.15(5)awasnotviolated, j

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HIGH COLD LEG TEMFERATURE-O! 002009

a. Event Descriotion l On intoSeptember service. While 25, 1984, flushingPurificationIonExchankerCH-8Awasbeingplaced the exchanger, t e reactor power and temperature began increasing. The cold leg temperat.ure rose above 545'F for about an hour. Reactor power increased to 1515 MWt.

l l Technic.alSpecification2.10.4(5)(a)(i) requires,inparts that the cold leg temperature be less than or equal to 545'F. If the parameter exceeds the limit, it mst be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reactor ptwer must be reduced to less than 15% of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This event terminated before the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limit was exceeded.

b. Inunediate Actions Taken Group 4 control rods were inserted and Boron was injected into the core to reduce system power,

c. Corrective Actions Operating Instruction 01-CH-4 was revised just prior to thi!. ev6nt. The date of the revision is September 11, 1988. This revision requires flushing of the ion exchanger for five minutes to prevent a dilution event. Since this event, the operator training program has c"atly improved and should reduce the chance of operator error,

d. Nuclear Safety Stanificance The plant was operated within the provisions of Technical Specification 2.10.4(5)(a)(1).

e. gjsis for Reportability NOT REPORTABLE

f. Reason for Not Reoortina ,

This event did not result in a violatfor. of Technical Specifications. The plant was operated within the Limiting Conditions of Operation.

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k SURVEILLANCE TEST ST-RI.T-3 F.1 HISSED 01 002126

a. Event Description

! On July 13, 1984, at Fort Calhoun Station, Surveillance Test ST-RLT-3 F.1, "Reactor Coolant System Leak Rate Test", was not completed. The Reactoe

Conlant S methods.ystem(RCS)leakratecanbecalculatedbyeitheroneoftwo One method uses the Plant Computer, P-250, to calculate the RCS

, leak rate while the other method, used when the computer is unavailable, is a hand eticulation of the RCS Icik rate. ST-RLT-3 F.1 is normally completed curing each night shift and is the responsibility of 0)erations.

Oa this date, the Plant Com> uter was unavailable and the night s1ift operators did not perform t1e hand calculation of the RCS leak rate. This  :

information was not included on the Shif t Turnover Log and the subsequent shifts did not perform ST-RLT-3 F.1. ST-RLT-3 F.1 has a daiiy frequency as rewired by Fort Calhour. Station Technical Specift:ation 3.2, Table 3-5, Item 8.

The reactor was in the start-up mode and the reactor coolant system was l being dileted. The surveillance test gives erroneous results under these cor:ditions.

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b. ! mediate A-tions Taken

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, Sm veillance Test, $T-RLT-3 F.1 was completed on July 12, 1984 using the l j Plane Computer and on July 14, 1984 using the hand calculation method.

c. Carrective Actions  !

l l A new surveillance test tracking procedure using manual scheduling and  ;

c computer tracking has been in place on a trial basis since February of l 1988. This procedure will be incorporated into Standing Order G-23 following approval of Procedure Change No. 22548.

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d. Nuclear Safety Sionificance l Even though this event is a violation of the Fort Calhoun Station Technical Specifications. it does not affect nuclear safety. The plant is equipped with indicators in the control room of reactor coolant leakage. The devices providing this ir.dication are Containment Air Particulate Monitor RM-050 Containinent Gas Monitor RM 051 Dew Point Monitor, and Containtnent '

Sump Pump Operation. Thesedeviceswillprovideadequatewarningtothe  !

operators in the case of excessive reactor coolant leakage,

e. Basis for Reportability

ThiseventwasaviolationofTechnicalSpecification3.2, Item 8. Thiseventwasreportablepursuantto10CFR50.73(a)(2)(1) Table 3-5(B)

(year 1984) which required the licensee to report, "Any operation or condition prohibited by the plant's Technical Specifications."

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f. Feason for Not Reportina An error was made in determining the reportability dJring the initial review of the 01.

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NON-CERTIFIED FILLERS USED IN VA-64 -

O! 002210 I

a. Event Descriotion On September 10, 1985, Surveillance Test ST-IR-1 F.4 for Control Room ,

Filter (VA-64)Replacementwasperformed. Instead of attaching  !

certification of the filters to the Surveillance Test, the QA material i release tags affixed by the storeroom to the filters were attached to the  ;

test. A test engineer tried to find the certifications for the three ,

! filter trays. lwoofthethreefiltertrays(serialnumbers1885and1888)  !

) did not have certifications. The plant continued to operate with these l filters in place until September 29, 1985.

J b. Immediate Actions Taken i

] The non-certified filters were replaced with certified filters.

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c. Corrective Actions

! Personnel awareness of the im)ortance of verbatim compliance with

] procedures is being promoted )y an increased emphasis on training.  :

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1 d. Nuclear Safety Sionificance I  !

1 The Fort Calhoun Station Technical Specifications define "operable," and i

} "operability" as follows:

J "A system, subsystem, train, com)onent or device shall be OPERABLE or have OPERABILITY when it is capa)le of performing its specified I

! function (s)." i Section 7. Paragraph 4.1.3 of the Quality Assurance Plan states, "Until l'

suitable documentary evidence is available to show the equipment or  ;

affected equipment shall be considered to be i

! materialisinconformance}ltyhasbeendeterminedtobeadequate,basedon inoperable, unless o l J

adocumentedreview,gerabi i

] lechnicalSpecification2.12(4)requiresthecontrolroomairtreatment  !

a systemtobeoperable;and,ifitisnot,seven(7)daysareallowedto j

make the inoperable circuit operable. If these conditions are not met, the l reactor shall be placed in cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. l

e. Basis for Reportability Technical Specification Violation. This is a re ursuant to j 10CFR50.72(b)(ii)(B)and10CFR50.73(a)(2)(it$ortableevent(B)(year 1985.

j f. Reason for Not Reportino l An error was made in determining the reprtability during the initial i j review of the 01.  !

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l TWO CHARGING PUMPS INOPERABLE i O! 002455

a. Event Dercription

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I On June 4, 1986, while Fort Calhoun was operating at 100 percent power, ,

charging pump CH-18 was tagged out at 0825 so that M0-862052 could be i done, in the process of tagging out charging pump CH-1B, discharge header

cross connect valve, CH-191, was closed instead of the charging pump CH-1B

i discharge valve, CH-192. This resulted in the loss of the ncrmal flow path l from charging pump CH-10. The Fort Calhoun Station Technical l 1 Specifications define "operable" and "operability" as follows:

"A system, subsystem, train com>onent or device shall be OPERABLE or -

j have OPERABILITY when it is,capa>1e of performing its specified l function (s)." l Thus, there was only one operable charging pump, CH-1A, when Valve CH-191 was closed. f 1

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b. Immediate Actions Taken j

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i This error was discovered upon completion of MO-862052 on Jane 5, 1986 at i

{ 1130, and CH-191 was imediately opened to restore the flow path from '

1 CH-10. Charging pump CH-1B was returned to service at 1235. The other ,

! valve positions associated with the three charging pumps were checked at i j this time and it was determined that CH-191 was the only valve that had -

j been incorrectly set. ,

c. Corrective Actions t

f' The specific corrective action for this event could not be datermined.

However, since this event, there has been increased emphasis on procedure i l

] compliance in the operator training programs. j

d. Nuclear Safety Sionificance l

. TechncalSpecification2.2(2)astates,"Atleasttwochargingpumpsshall l l be operable." TechnicalSpecification2.2(3)a. states,"Oneofthe  !

i operable charging pumps may be removed from service provided two charging l l pumps are operable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." l The basis for the Limiting Conditions for Operation, h chnical l Specification 2.2, states:

I i "The limits on component operability and the time periods for  !

! inoperability were selected on the % sis of the redundancy indicated l above and engineering judgrant."  !

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e. Basis for Reoortability Technical Specification violation. The Danger / Caution Tag Sheet for l June 4,1986showsValveCH-192(CH-191)wastaggedoutat0825on June 4, 1986 and returned to service on 1235 on June 5, 1986. Valve .

CH-191 was closed for 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> and 10 minutes. Technical Specification  !

Limiting Condition for Operation 2.2(3)a was violated. This is a reportableeventpursuantto10CFR50.72(b)(ii)(B)and '

10CFR50.73(2)(ii)(B)(year 1986).

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f. Reason for Not Reportina Poor documentation resulted in an error in determining the reportability !

during the initial review of the OI.

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1 TEMPORARY PENETRATION FIRE BARRIER DISCOVERED MISSING 01 002481 l i

a. Event Descriptioq On July 9, 1986, during a quarterly fire barrier inspection, it was discovered that temporary penetration fire barrier 19-E-56 was completely missing. This barrier seals an opening around a crane rail that penetrates a steel panel door. The door separates the turbine building from an equip7)(ment 2.19 hatch requires thataccess in the auxiliary the nonfunctional building.

penetration beTechnical Specification restored to operable statut within 7 days, or prepare and submit a report to the NRC pursuant withinanaddItional30 to Section days. 5.9.3 ofwas The barrier thenotTechnical replaced Specifications,l and no specia report was submitted,

b. Imediate Actions Taken Upon discovery, an hourly firewatch patrol was initiated,

c. Corrective Actions ,

None were required as discussed in item d.

d. Nuclear Safety Sionificance l

fire detection and protection system of appropriate ca)acity and capability shall be provided and designed to minimize tie adverse effects i of fires on structures, systems, and components important to safety.

4 The current Fort Calhoun Station Fire Hazards Analysis describes this door i and opening as unrated. The door is protected by a water curtain, and is c.onsidered acceptable in its current configuration without a penetration barrier. A subsequent engineering evaluation concluded that a penetration  ;

barrier was not required at this location (reference memo FC-738 88 dated 1 April 21,1988).

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e. Basis for 3.tportability '

NOT REPORTABLE., j

f. Reason for Not Reportino NOT REP 03 TABLE.

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.o WASTE GAS DISPOSAL 01 002485

! a. Event Description On July 4, 1986, the gas in the vent header of the Waste Gas Disposal System was pumped to Waste Gas Decay Tank C. Prior to ) umping, the gas was analyzed for hydrogen and oxygen concentrations. T1e Waste Gas Sampling System, Al-110, was available for use during this operation, but was not put into service. if the Waste Gas Sampling System is not sut into service, Technical Specification 2.9.1(2)d requires that a gra) sample be taken from the waste gas decay tank and analyzed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This analysis was not performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Imediate Actions Taken A grab sample was taken from the waste gas decay tank and analyzed on July 9, 1986. The analysis showed that the flammability limit of hydrogen and oxygen was not exceeded.

c. Corrective Actions Procedure Chan e No. 17891, initiated August 11, 1986, requires the recording of h drogen and oxygen concentrations in the Operators log,

d. Nuclear Safety Sionificance

The basis of Technical Specification 2.9.1(2)d, ' ensures that the concentration of potentially explosive gas mixtures entrained in the gas decay tank (s) will be maintained below the flammability limits of hydrogen and oxygen."

The worst case everi., a gas decay tank rupture, was analyzed in i

Section 14.19 of the USAR which states, in part "that a rupture of a gas decaytankwouldnotinterruptorrestrictpubileuseofareasbeyondthe exclusion aret boundary."

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e. Basis for Reportability This was a violation of Technical Specification 2.9.1(2)d. This event was l reportr.blepursuantto10CFR50.72(b)(ti)(B)and10CFR50.73(a)(2)(ii) .

(B)(year 1986)thatrequirereportingwhenalimitingconditionof I opeiation is not met. {

f. Reason for Not ReDortina An error was made in determining the reportability during the initial review of the 01.

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. o INOPERABLE T0XIC GAS MONITORS 01 002580 l l

a. Event Description On September 16, 1986, the Control Room Ventilation System was placed in the recirculation mode C in accordance with Technical Specification 2.22.

Toxic Gas Monitors 6286 A/B and 6288 A/B were taken out of service so that ST-TCH-1 F.1 could be performed. Upon receipt of VIAS, due to other surveillance tests being performed, the Control Room was switched from recirculation Mode C to filtered makeup Mode B approximately 6 to 12 times from September 16 to October 9, 1986. During this time, the toxic gas ,

monitors were still inoperable. The longest a VIAS signal was ever in was approximately 15 minutes.

b. Immediate Actions Taken By the time the determination was made that the Technical Specification had been violated, the system had been returned to its proper configuration; therefore, no immediate corrective actions were necessary.

c. Corrective Actions Since this event there has been an increased emphasis on operator training which will reduce the chance of this type of event happening again.

d. Nuclear Safety Jianificangg [

The basis of Technical Specification 2.22 states, "If both of the Toxic Gas Monitors are found inoperable, there is no imediate threat to the control room operators and reactor oseration may continue while repairs are being made. During this repair, tie control room ventilation will be switened to internal recirculation mode of operation."

e. Basis _for ReDortability This event constituted a violation of the limiting condition of operation i set forth in Technical Specification 2.22. This event was re ortable '

p(ursuantto10CFR50.72(b)(ii)(B)and10CTR50.73(a)(2)(ii)p(B) year 1986). ,

f. Reason for Not Reportina An error was made in determining the reportability during the initial I review of the O!.

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INADVERTENT ACTUATION OF ENGINEERED SAFETY FEATURES O! 002613

a. Event Descriotion On December 1,1986, at 0835, at Fort Calbrun Station, Raw Water Pum)

AC-100 Low Pressure Safety injection (LPSI) Pump $1-1B, and the Tur)ine Driven Auxiliary feedwater Pump FW-10 were inadvertently started. This occurred during the performance of Surveillance Test ST-ESF-5 F.1, "Automatic Load Sequencer Check," which verifies the proper operation and timer settings of the Automatic Load Sequencers. This test requires the operator to inhibit operation of the equipment connected to the Sequencu by turning all Sequencer Isolation Switches on the particular Sequencer to 0FF. Then the operator must turn the Sequencer Auto Start Test Switch to TEST which trips the Sequencer Lockout Relays and starts the sequencer timers. Next, the operstar should return the Sequencer Auto Test Switch to NORM, reset the Sequuncer Lockout Relays, and return the Sequencer Isolation Switches to ON. While testing Automatic Load Sequencer S2-2, the operator turned the Sequencer Isolation Switches to ON for AC-108, SI-18, and FW 10 prior to resetting Sequencer Lockout Relays 86-1/S2-2 and 86-2/52-2. This resulted in the actuation of the above listed pumps,

b. Immediate Actions Taken A second operator immediately observed control room indication of the pumps' actuation and after confirming that this event was an inadvertent actuation by the first operator, turned pumps AC 10B, S1-18, and FW-10 off at their respective control switches.

c. Correc tive_A.gljaqS, The surveillance test ST-ESF-S r.1 was revised the following day, December 2, 1986, Procedure Cnange No. 18727. ST-ESF-5 f.1 only had one set of procedure steps for both sequencers with a signoff for each sequencer.

This was confusing. The procedure change separated the steps for sequencers 51-2 and 52-2.

d. Nuclear Safetv_Sionificance This event does not affect nuclear safety for the following reasons. The Low Pressure Safety injection pumps and piping are aligned for recircula-tion and will not inject into the Reacter Coolant System until a Safety injetionActuationSignal(SIAS)isreceived. The Auxiliary feedwater p ups and piping are aligned for recirculation and will not provide flow to thesteamSeneratorsuntilanAuxiliaryFeedwaterActuationSignal(AFAS) is receivet. The Raw Water pumps and piping are aligned to permit additional flow from consecutive pumps being started,

e. Ea,sh for Reportability, The actuation was not part of a preplanned action and was reportable pursuant to 10 CFR $0.72(b)(2)(fi) and 10 CFR 50.73(a)(2)(iv) (year 1986).

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f. Reason for Not ReDortina An error was made in determining the reportability during the initial review of the 01.

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I INADVERTENT ACTUATION OF ENGINEERED SAFETY FEATURES 01 002660

a. Event Description On February 2, 1987, at 0915, at Fort Calhoun Station, Raw Water Pumps AC-10AandAC-100,LowPressureSafetyinjection(LPSI)PumpSI-1A,High Pressure Safety injection (HPSI) Pumps SI-2A and S!-2C, and Auxiliary ,

Feedwater Pump FW-6 were inadvertently started. This occurred during the performance of Surveillance Test 51-ESF-5 F.1, "Automatic Load Sequencer Check,' which verifies the proper operation and timer settings of the Automatic Load Sequencers. This test requires the operator to inhibit operation of the equipment connected to the Sequencer by turning all Sequencer Isolation Switches on the particular Sequencer to 0FF. Then the o>erator must turn the Sequencer Auto Start Test Switch to TEST which trips tie Sequencer Lockout Relays and starts the sequencer timers. Next, the operator should return the Sequencer Auto Test Switch to NORM, reset the Sequencer Lockout Relays, and return the Sequencer Isolation Switches to ON. While testing Automatic Load Sequencer $1-2, the operator turned the Secuencer Isolation Switches to ON for AC-10A, AC-100, S!-1A, SI-2A, SI-2C anc FW 6 prior to resetting Sequencer Lockout Relays 86-1/SI-2 and 86-2/51-2. This resulted in the actuation of the abo' i listed pumps,

b. Imediate Actions Taken A second .aperator immediately observed control room indication of the pump's actuation and after confirming that this event was an inadvertent actuation by the first operator, turned pumps AC-10A, AC-10C, S!-1A, SI-2A, SI-2C and FW-6 off at their respective control switches.

c. Corrective Actions t

The surveillance test now requires the operator to verify tnat the OFF AUTO lights are on.

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d. $_yslear Safety Sianificance I

This event does not affect nuclear safety for the following reasons. The i Low Pressure Safety injection and High Pressure Safety Injection pumps and piping are aligned for recirculation and will not inject into the Reactor CoolantSystemuntilaSafetyInjectionActuationSignal(S!AS)is received. The Auxiliary Fcedwater pumps and piping are aligned for recirculation and will not provide flow to the steam generators until an Auxiliary feedwater Actuation Signal (AFAS) is received. The Raw Water pumps and piping are aligned to permit additional flow from consecutive pumps are started,

e. pasis for Reportability l This actuation was not art of a preplanned actuation and was reportable under 10 CFR 50.72(b)(2 (ii) and 10 CFR 50.73(a)(2)(iv) (year 1987).

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f. B.g.aion for Not Rgoortina An error was made in determining the reportability during the initial review of the 01.

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