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, - . _ -.' 4 . -TENNESSEE. VALLEY AUTHORITY

,

6N 38A Lookout' PlaceChattanooga, Tennessee 37402-2801

r

July 9, 1990

.U.S. Nuclear. Regulatory Commission-ATTN Document Control Desk jWashington, D.C. 20555 .|

|'

Gentlemen:

TENNESSEE VALLEY AUTHdRITY - SEQUOYAH NUCLEAR PLANT UNIT 1 - DOCKET NO.50-327 - FACILITY OPERATING LICENSE DPR-77 - LICENSEF EVENT REPORT (LER)-50-327/90011

'

.i

The enclosed LER provides details concerning the nonconservative calibration :of nuclear-instrumentation system intermediate range and power range channels i

as a result of.the misinterpretation of vendor information. 'This event is i

being reported in accordance with 10 CFR 50.73(a)(2)(1) as an operation |prohibited by technical specifications and 10 CFR 50.73(a)(2)(vii) as a single i

cause resulting-in. multiple inoperable.-channels.1

-Very truly yours, ji

TENNESSEE VALLEY AUTHORITY .|

/..-R. Bynum,. ce President 1

Nuclear Power Production

Enclosure "

- cc ' (Enclo'sure ): ;

Mr; J. N.' DonohewU.S. Nuclear Regulatory CommissionOne White Flint, North j11555 Rockville PikeRockville, Maryland 20852 ;

qINPO Records CenterInstitute.of Nuclear Power Operations1100 Circle 75 Parkway, Suite 1500 1Atlanta, Georgia.-30339

j

NRC Resident Inspector'3

Sequoyah Nuclear Plant2600 Igou Ferry RoadSoddy. Daisy,-Tennessee 37379

Mr. B. A. Wilson, Project ChiefU.S. Nuclear Regulatory CommissionRegion II101 Marietta Street, NW, Suite 2900Atlanta, Georgia 30323

900717002'l YVU/UYPDR ADOCK 05000327 / g

S PDC An Equal Opportunity Employer

Li '

*'NRC Form 366 U.S. NUCLEAR REGULATOR 7 COMMIS$10N Approved OMB No.

1150-0104-..

(6-8;)~ - Erpires 4/30/92'

"

LKENSEEEVENTREPORT(LER)

FACILITY NAME (1) |DOCKETNUMBER(2) |PAGE(3)Seouovah Nuclear Plant. Uni t I 10lBl0101013 12 17 1110Fl 01 7

TITLE (4) Nuclear Instrumentation System Intermediate and Power Range Channels Were Nonconservatively CalibratedEgg ule of Misinterpretation of Vendor Info m tion

EVENT DAY (5) I LER NUMBER (6) | REPORT DATE (7) | OTHER FACILITIES INVOLVED (B)| | | | | SEQUENTIAL | | REVISION | | | | FACILITY NAMES |00CKETNUMBER(S)

MONTHl DAY lYEAR lYEAR l | NUMBER | | HUMBER IMONTHI DAY lYEAR I 1015l010101 l lI I I l_| |_I I I I I I ,

Q.,j,_11 01 Bf 91 Ol 91 of I0| 1 | 1 I l 0 1 0 1 01 71 Ol 91 91 of 1015l010101 l l-OPERATING | |THISREPORTISSUBMITTEDPURSUANTTOTHEREQUIREMENTSOf10CFR$:

MODE | |_ICheckoneormoreofthefollowino)(11)(9) I il (20.402(b) |_l20.405(c) |_l50.73(a)(2)(iv) |_l73.71(b)

POWER | |_|20.405(a)(1)(i) |_|50.36(c)(1) |_|50.73(a)(2)(v) |_|73.71(c) !

LEVEL | |_|20.405(a)(1)(ii) |_|50.36(c)(2) |El50.73(a)(2)(vil) |_|0THER(Specifyin_ (10) 1 01 21 41 |20.405(a)(1)(iii)|Ml50.73(a)(2)(i) |_l50.73(a)(2)(viii)(A)| = Abstract below and in

|_|20.405(a)(1)(iv) |_|50.73(a)(2)(ii) |_|50.73(a)(2)(viii)(B)| Text, NRC form 366A) -i

l 120.405fa)(1)(v) l 150.73(a)(2)(iii) I 150.73(a)(2)(x) l |

LICENSEE CONTACT FOR THIS LER (12) |NAME | TELEPHONE NUMBER

'

|AREACODE|'

Russell R. Thomoson. Comollance Licensino Enoineer |6|1 15|Bl4|3|-|7l417 |0 ICOMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

| | | .| REPORTABLE | | | | | | REPORTABLE |CAUSElSYSTEMI COMPONENT |MANUFACTURERl TO NPRDS l |CAUSElSYSTEMI COMPONENT |MANUFACTURERl TO NPRDS l

| I | | | | I I i | |I I l' I I I i i I I I I I I I I I I I I I I I I I

| 1 I I I I I I I I II | | | t i I I I I I I I I I | | t | | | I I I i ,

SUPPLEMENTAL REPORT EXPECTED (141 | EXPECTED |MONTHlDAYlYEAR,_, |_ | SUBMISSION'l | | ,

I YES (If ves. comolete EXPECTED SUBMISSION DATE) | X l NO I DATE (15) l l l l l I |

ABSTRACT (Limit to 1400 spaces, i .e., approximately fif teen single-space typewritten lines) (16) '

On June 8, 1990, at 0100 Eastern daylight time (EDT) with Unit 1 in Mode 1,'it wasdetermined that Unit I had operated in noncompliance with Technical Specification2.2.1, " Limiting Safety System Settings." The actual Nuclear Instrumentation System

-(NIS) power range (PR) detector currents were determined to have been 20 to 31 percent glower than predicted, which would shift the PR trip setpoint actuation 20 to 31 percent '

higher than expected.. This nonconservative calibration of the PR channels was present.from May 31, 1990 (Unit 1 Cycle 5 criticality), to June 6, 1990. The intermediate i

range (IR) channels had been nonconservatively calibrated in the same manner from fMay 31, 1990, to June 1, 1990. The nonconservative calibration resulted from a i

misinterpretation of vendor information when calculating expected NIS IR and PRdetector currents for Cycle 5 operation. Although the nonconservative calibration

Tresulted in IR and PR setpoints being outside of their respective TS allowable values,the plant remained within the. Updated Final Safety Analysis Report (UFSAR) accidentanalyses limits. The prestartup NIS calibration procedures will be revised to require jthe use of the correct prediction methodology. The NIS correction procedures will berevised to provide guidance for performing evaluations of observed NIS deviations inNIS detector indications.

i

NRC Form 366(6-89)

- _ _ _

1

: .; *

NRC F:rm 366A U.S. NUCLEAR REGULATORY COMMISSION Approved OMB No. 3150-0104,

' (6-89) .' Expires 4/30/92

LKENSEEEVENTREPORT(LER) !*

TEXT CONTINUATION

|FACILITY NAME (1) |DOCKETNUMBER(2)l LER NUMBER (6) | | PAGE (3)

| | | |$EQUENTIAL| | REVISION | | | | |Sequoyah Nuclear Plant Unit 1 | |YEARI I NUMBER l | NUMBER l | | | | |

1015l0101013 12 17 19 10 |--I 0 1 1 l1 |-| 0 1 0 1 01 2l0Fl 01 7TEXT-(If more space is required, use additional NRC Form 366A's) (17)

Description of Event '

On June 8, 1990, at 0100 Eastern daylight time (EDT) with Unit 1 in Mode 1 (24 percentreactor power, reactor coolant system (RCS] pressure at 2,235 pounds per square inchgauge (psig}, and RCS temperature at 555 degrees Fahrenheit (F)), it was determinedthat. Unit I had operated in noncompliance with Technical Specification (TS) 2.2.1, j

" Limiting Safety System Settings " because the nuclear instrumentation system (NIS)intermediate range. (IR) and pc er range (PR) channels (EIIS Code IG) had beennonconservatively calibrated. TS 2.2.1 requires the reactor trip system -!instrumentation and interlock setpoints to be set consistent with trip setpoint values j

of TS Table 2.2-1. This requires the NIS PR channels to have high and low setpoints of I|

|. less than or equal to 109 and 25 percent reactor power, respectively; and allowableI values of less than or equal to lil A and 27.4 percent reactor power, respectively.'' -The NIS IR channels are required to have setpoints of less than or equal to 25 percent

reactor power; and allowable values of less than or equal to 30 percent. reactor power. jIf a reactor trip system instrumentation or interlock setpoint is found to be less jconservative than the value shown in the allowable values column of Table 2.2-1, the

;

corresponding channel is to be declared inoperable, and the applicable action i

provisions of Limiting Condition for Operation applied. *

During the timeframe of this event, Unit 1 was restarting- f rom its Cycle 4 refueling !

outage. The unit entered Mode 2 at approximately 0540 EDT on May 31, 1990, achievedcriticality at 1033 EDT on May 31, 1990; and entered Mode 1 at approximately 1728 EDTon June 1, 1990. On June 2,1990, Unit 1 tripped from approximately 11 percent powerat 0802 EDT. The unit was again critical at 2104 EDT on June 2, 1990, and enteredMode 1 at 2233 EDT on June 2,'1990. The reactor was returned to Mode 2 at 1608 EDT on !

!June 3, 1990, for maintenance on a main turbine governor valve. After the completionof the maintenance activities, power escalation was initiated, and the unit entered -|Mode 1 at 1406 EDT on June 5, 1990. Power escalation continued from this point. '

Prior to startup from the Unit 1 Cycle 4 refueling outage IR and PR NIS detectorcurrents were predicted, and adjustments were made based on the past cycle low leakagecore operation and the even lower leakage core loading pattern for this cycle. Due to !

uncertainty introduced by the re. load and installation of Gamma-Metrics for the IR

detectora, the IR high flux trip setpoint was conservatively adjusted to one-half the !

normal trip setpoint, i.e., from 25 percent to 12 percent.

On June 1, 1990 (4 percent actual power), Periodic Instruction (PI) 0-PI-NXX-092-002.0,"Poststartup NIS Calibration Following Core Load," was performed. This PI usescalorimetric data to check the output of the IR channels. The performance was done tofurther check the NIS detectors and to then allow readjustment of the IR trip setpointsto the normal 25 percent nominal value. Deviations of approximately 5 percent voltagespan were observed, i.e., the IR detectors indicated approximately one-half decadebelow actual reactor power as determined by calorimetric data. The accuracy of thecalorimetric data was questioned because of low secondary system flowrates. IIoweve r ,the IR channels were adjusted consistent with the more conservative callometricindicated power. The deviations were not considered to be atypical for an initialstartup from a refueling outage. No anomolies were identified in the PR indication atthis time; however, problems would not likely be apparent at this low of a power level.

NRC Form 366(6-89)_ - . . . _ _ - _ _ _ - _ - _ _ _ - - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - - _ - . _ _ _ _ . - - - _ _ _ _ _

.. i*-

HRC Fora 366A~ U.S. NUCtEAR REGULATORY COMMISSION Appro n d OMB No. 3150-0104,

(6-89) Empires 4/30/92*

LEENSEEEVENTRENRT(LER)'

TEXT CONTINUATION

FACILITY NAME (1) | DOCKET NUMBER (2) | LER NUMBER (6) | | PAGE (3)

| | | | SEQUENTIAL | | REVISION | | | | |Siquoyah Nuclear Plant Unit I- | lYEAR I | HUMBER l l HUMBER l | | | |

101$10101013 12 17 19 10 |--! 0 | 111 |-I01 0 l 01 310FI 01 7TEXT (If more space is required, use additional NRC Form 366A's) (17)

Description of Event (Continued)

On June 6, 1990 (24 percent actual power), Operations performed Surveillance *

Instruction (SI) 78, " Power Range Neutron Flux Channel Calibration by Heat BalanceComparison Daily," to adjust the PR channels to within 2 percent of the secondary sidecalorimetric full power value. Calorimetric data at this power level was consideredreliable, and deviations of approximately 20 to 31 percent between actual and-prestartup estimated detector currents were noted. Again the deviations were notconsidered to be atypical for a startup from a refueling outage.

At 0100 EDT, on June 8, 1990 (24 percent actual power), during the performance of-PI 0-PI-NXX-092-001.0,,"Incore-Excore Detector Calibration," it was determined byReactor Engineering that the projected normalized top and bottom excore PR detectorcurrents for 100 percent power indicated that the PR channels had been initiallycalibrated with nonconservative values prior to startup from the Unit I refuelingoutage due to an incorrect equation used to calculate preliminary detector currentvalues. 'This error caused the PR channels to indicate lower than actual power by 20 to31 percent of the power level, and resulted in the NIS reactor trip setpoints being set20 to 31 percent higher than their TS-required values during startup evolutions(May 30-June 6). The nonconservative calibration values were noted when "new" PR fullpower detector currents calculated in 0-PI-NXX-092-001.0 were compared to theprestartup calibration detector currents calculated in 1-PI-NXX-092-001.0, "PrestartupNIS Calibration Following Core' Load." Subsequent evaluation also determined that thesame error had been made-in predicting the prestartup IR detector currents resulting inthe IR detectors indicating approximately one-half decade below actual reactor power.

Procedure 1-PI-NXX-092-001.0 is used to calculate preliminary NIS IR and PR detectorcurrent values. The-PR and IR channels are then calibrated using the preliminaryvalues prior to startup. The procedure utilized power distributions from beginning oflife (BOL) of the previous cycle (PR old), power distributions from BOL of the newcycle (PR new), and end-of-life (EOL) detector currents from the previous cycle (I old)to calculate the new preliminary current values. (I old x PR new/PR old = I new) Thiscalculation methodology was believed to have been in accordance with WestinghouseElectric Corporation information. However, it was determined that the information hadbeen misinterpreted. The reason for the resultant nonconservative values was the useof BOL power distributions with E0L detector currents. Power moves towards the outercore over core life, which increases the leakage neutron flux. For this reason -EOL-

detector currents tend to be latger than BOL detector currents (20 to 31 percent forthis cycle), and this resulted in larger nonconservative preliminary detector currentvalues being calculated. More accurate preliminary current values would have beencalculated by utilizing detector currents and power distributions from the same point I

in core life from the previous cycle (both BOL or both EOL). Utilization of EOL valueswas recommended by Westinghouse in discussions held subsequent to this event.

i

No deviations outside of TS allowable values were noted during previous startups fromrefueling outages at SQN, although the initial prediction of full power NIS IR and PRdetector currents were made utilizing the equation described above.

NRC Fom 366(6-89)

.~ _

+ . "

NRC F,orm 366A U.S NUCLEAR REGULATORY COMMISSION ApprovGd OMB No. 3150-0104,

(6-89) Expires 4/30/92.

llCENSEEEVENTREPORT(LER)'

TEXT CONTINUATION

FACILITY NAME (1) |DOCKETNUMBER(2) | LER HUMBER f6) I | PAGE f3)| | | |$EQUENTIAL| | REVISION | | | | |

Sequoyah Nuclear Plant Unit 1 | lYEAR | | NUMBER l I NUMBER I | | | |10151010101312171910 l-l 0 | 111 |-|0l0 1 01 410Fl 01 7

TEXT (If more space is required, use additional NRC form 366A's) (17)Description of Event (Continued)

In the past, the IR detectors indicated in amperes and the correlation between IRindication and actual reactor power was difficult to observe directly. The status ofthe IR indications has been specifically monitored since the Unit 2 IR detectormispositioning_ event (LER 50-328/89006). No corrections of the IR indications wouldhave been made routinely until the performance of the full power (90 to 100 percent)calorimetric and incore-excore cross-calibration. Additionally, the IR reactor trip isblocked at 10 percent reactor power.

For the NIS PR channels, daily adjustments of PR indication are made as necessary byOperations' personnel when reactor power is greater than 15 percent. These dailyadjustments match PR indications to plant calorimetric data. Pe.st reviews of theadjustment _ procedures.between 15 and 30 percent focused primarily on "as-left" data.Calibrations of the PR channels would have been performed as-required after performanceaf the 30 percent and full power incore-excore cross-calibrations. For both the IR andPR channels, no accurate calorime:.ric data has been available typically prior to30 percent reactor power for comparison. A possible reason why a deviation was moreevident during this startup may be related to the point in the operating cycle wherethe EOL detector current was obtained. For this specific case, the last cycle detectorcurrent was taken at the extreme E0L condition, thereby maximizing the error.

Cause of Event

The preliminary IR and PR calibration values were incorrectly calculated based onmisinterpreted vendor information. A Westinghouse letter on low leakage loading

3(L P) schemes stated. that "The expected values of these currents can be estimatedfrom the last previous cycle's Incore-Excore calibration and the predicted X-Y power

3distribution for the L P cycle at HFP conditions." The Reactor Engineering group'

interpreted this to mean the NIS currents from the last-(EOL) incore-excore calibration-performed in the previous' cycle, in combination with BOL power distributions from bothcycles.

A contributing cause to the duration of the event may have been the lack of proceduralguidance concerning the evaluation of deviations.- Disagreement between predicteddetector currents and calorimetric indicated power was considered to be solely afunction of the inherent uncertainty involved in the prediction and measurementprocesses,'i.e., the refueling had included a lower leakage loading pattern,gamma-metrics had been installed for the IR detectors, secondary side flow measuringdevices had been cleaned and recalibrated, and accuracy of the secondary side

L calorimetric data of low power levels is considered inaccurate. As a result, potentialfor disagreement was considered likely and to some extent unavoidable; resolution ofdeviations consisted of readjustment of NIS to match best estimate power particularly,during low power escalation. More detailed evaluation of the cause of the observeddeviation in the IR detectors on June 1 may have identified the potential for error inthe PR detectors at an earlier point in time.

NRC form 366(6-89)

).. . . - -

,NRC Fprm 366A . U.S. NUCLEAR REGULATORY COMMIS$10N Approved OMB No. 3150-0104

(6-89) '* Expires 4/30/92

LKENSEEEVENTRENRI(LER)'

TEXT CONTINUATION

FACILITY NAME (1) |DOCKETNUMBER(2)| LER NUMBER (6) l l PAGE (3)

| | | |$EQUENTIAt| | REVISION | | | | |Sequoyah Nuclear Plant Unit 1 | lYEAR I I NUMBER | | NUMBER | | | | |

1015l0101013 12 17 19 10 l--| 0 | 1 1 1 |--| 0 1 0 1 Ol_$10Fl 01 7TEXT (If more space is required, use a.iditional NRC Form 366A's) (17)

Analysis of Event

This event is reportable in accordance with 10 CFR 50.73(a)(2)(i) as a conditionprohibited by TSs and in accordance with 10 CFR 50.73(a)(2)(vii) as a single causeresulting in multiple inoperable channels.

Section 15.2.1.1 of the Updated Final Safety Analysis (UFSAR) states that the reactortrip for a postulated uncontrolled rod cluster control assembly (RCCA) bank withdrawalfrom suberitical conditions is assumed to be initiated by PR high neutron flux (lowsetting). The most adverse combination of instrument and setpoint errors, as well asdelay for trip signal actuation and RCCA release, is taken into account. A 10 percentincrease is assumed for the PR flux trip setpoint raising it from the nominal.value_of25 percent to 35 percent. Previous results, however, show that rise in the neutronflux is so rapid that the effect of errors in the trip setpoint on the actual time atwhich the rods are released is negligible. In addition, the reactor trip insertioncharacteristic is based on the assumption that the highest worth RCCA is stuck in itsfully-withdrawn position.

Prior to 24 percent' power, the PR channels were adjusted such that the low flux tripbistables would have actuated at 33 percent (worst case), which is less than the35 percent setpoint used in the analysis. The plant safety analysis takes no creditfor the IR trips, which would have tripped at 38 percent power. Because the IR

N channels were newly installed equipment (Gamma-Metrics), for conservatism, the reactortrip biotables had-been set to one-half of their normal 25 percent power setpoint(12 percent). Even with this additional conservatism, during a transient,' reactorpower could have reached 38 percent before the first IR channel would have tripped.-

Above 10= percent power, the PR low flux trip and IR trips were blocked as required by. plant procedures. Between 10 and 24 percent power, the PR positive rate trip wouldhave provided reactor protection for a postulated control rod drive rupture accident.-Although the positive rate trip was affected by the misalignment (6.2 percent setpoint-instead of 5 percent), it was still within the TS allowable of.6.3 percent _and wouldhave had little effect on the analysis due to the high rate of power increase predicted

~

for this event.

The uncontrolled rod withdrawal at power event (UFSAR, Section 15.2.2) relies on the PRhigh flux trip at 109 percent power. With the PR channels improperly calibrated, thehigh flux trip setpoint was approximately 135 percent, which was outside of the safetyanalysis-of 118 percent power. Another reactor trip used to mitigate this event is;over temperature delta temperature (T), provided the reactivity addition rate is withincertain limits. Over temperature delta T is used to mitigate slower reactivityinsertion rates while the high flux trip is used in faster transients.

-NRC form 366(6-89)

_ _ ____._ ___

-# ..

NRQ T,orm 366A U.S. NUCLEAR REGULATORY COMMIS$10N Approved OMB No. 3150-0104.

(6-89) Expires 4/30/42

UCENEE WENT RPORT REO' TEXT CONTINUATION

( FACILITY NAME (1) (DOCKETNUMBER(2)l LER NUMBER (6) l I PAGE (3)

| | | |$EQUENTIAL| | REVISION | | | | |Sequoyah NvClear Plant Unit 1 | lYEAR l I NUMBER l I NUMBER | | | | |

1015l0101013 12 17 19 10 l--! O l 1 11 |--I 0 1 0 I of 610Fl 01 7TEXT (If more space is required, use additional NRC Form 366A's) (17)

Analysis of Event (Continued)

During power operation between 10 and 24 percent power, control bank D was at least130 steps out of the core, (range is 0 to 230 steps). For conservatism, if 100 stepsout of the core is used, the intergral rod worth for control bank D (from 100 to230 steps) is 660 percent milli (pcm). Maximum rod speed is 72 steps per minute.Therefore, it will take 130 steps at 72 steps per minute, which equals 1.8 minutes(108.3 seconds) to withdraw bank D. This corresponds to 660 pcm in 108.3 seconds,which equals 6.09 pcm per second (6.09 E-5 delta k per second) for the averagereactivity insertion rate. In accordance with UFSAR Figures 15.2.2.6 and 15.2.2.7 for

,

60 and.10 percent, a 6.09 E-5 delta k per second rate falls well within the protectionenvelope of over temperature delta T, and the high flux trip would not have beenrequired. The peak reactivity insertion rate for this scenario is approximately 7.50E-5 delta k per second. Again, the high flux trip would not have been required.

The RCCA misalignment event (UFSAR, Section 15.2.3) utilizes the PR negative rate tripto mitigate the effect of a dropped group of rods. As with the positive rate trip, thenegative rate trip was within the TS allowable value of 6.3 percent.

For the postulated startup of an inactive RCS loop (UFSAR Section 15.2.6), the P-8(single loop loss of flow) interlock setpoint provides protection because the corepower level increases to a power level above P-8 when the inactive loop is started,before loop flow reaches a value sufficient to clear the low-low trip setpoint.However, SQN procedures restrict the startup of an inactive RCS icop to reach powerlevels less than 10 percent. This restriction provides sufficient margin to theaccident analysis limits that a reactor trip signal is not generated. Therefore, theshift in the P-8 setpoint resulting from the nonconservative NIS PR calibration wouldhave no impact on this event.

Postulated excessive heat removal events from feedwater system malfunctions (UFSAR,Section 15.2.10), are bounded by the uncontrolled RCCA bank withdrawal from suberiticalconditions analysis.

In summary, although the reactor trip setpoints were out of TS limits, the consequenceswere bounded by the UFSAR accident analyses. Additionally, normal administrativecontrols were in place and utilized to adjust the NIS channels during startup such thatthe nonconservatisms were corrected by 24 percent power. Therefore, the health andsafety of plant personnel or the general public was not adversely affected by thisevent.

Corrective Action

The NIS IR channels were corrected at 4 percent reactor power on June 1, 1990, by theperformance of 0-PI-NXX-092-002.0; and the NIS PR channels were corrected at 24 percentreactor power on June 6, 1990, by the performance of SI-78. These corrections are partof the normal NIS corrections performed during power escalation from a refueling

i

NRC Fonn 366(6-89).-

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|. HRf-form 366A= U.S. NUCLEAR REGULATORY COMMIS$10N Approved OMB No. 3150-0104

(6-89)fr Empires 4/30/92

LKENSEEEYENTREPORT(LER)TEXT CONTINUATION

FACILITY NAME (1) |DOCKETNUMBER(2)i LER NUMBER (M l l PAGE (3)>

| | | |$EQUENTIAL| | REVISION | | | | |'Sequoyah Nuclear Plant Unit 1 | lYEAR l I NUMBER l I NUMBER l | | | j

|0l510101013 12 17 19 10 |- I O l 1 I ) 1-!OI0 1 01 710Fl 01 7TEXT-(If more space is required ' use additional NRC Form 366AN (17)

Corrective Action (Continued)

outage. The NIS PR channels were recalibrated to reflect the new 100 percent reactorpower detector currents on June 8, 1990. To prevent recurrence, Pts 1- and2-PI-NXX-92-001.0 will be revised to require the use of EOL NIS detector currents andfuel assembly power fractions from the last cycle and the corresponding BOL fuelassembly power fractions for the upcoming cycle.

.

To enhance capability for early identification and correction of NIS anomolles, SI-78and 0-PI-NXX-092-002.0 will be revised to provide guidance for performing evaluationsof observed deviations in NIS detector indication.i

Additional Information

A similar event was reported by LER 50-328/89006, which reported the mispositioning ofthe NIS IR detectors for Unit 3. The NIS PR detectors were not affected during thissimilar event. Corrective actions for this event were directed at configurationcontrol and the status of Ik indications during power escalation. Although thisprevious event further heightened sensitivity of station personnel to potential for NISanomolies, corrective actions would not have been expected to have prevented thisevent. These events and similar events experienced at other utilities, emphasize theneed for continued diligence in improving NIS prediction, monitoring, and assessmentperformance.

Commitments

1. Periodic Instructions 1- and-2-PI-NXX-92-001.0 will be revised before startup fromUnit 2 Cycle 4 refueling outage to require the use of "end-of-life" NIS detector, currents and fuel assembly power fractions from the last cycle and thecorresponding "beginning-of-life" fuel assembly power fractions for the upcoming-cycle.

2. SI-78 and 0-PI-NXX-092-002.0 wi;1 be revised before startup from the Unit 2 Cycle 4refueling outage to provide guidance for performing evaluations of observeddeviations in NIS detector indications.

0921h-

NRC form 366(6-89).,,,-h . , - . . . . . . . . . . . _ , , , . . . . . , , . - , - - - , - . . -