ML20044A936

LER 90-011-00:on 900608,determined That Actual Nuclear Instrumentation Sys Power Range Detector Currents Were 20% to 31% Lower than Predicted.Caused by Calibr Values Being Incorrectly Calculated.Channels corrected.W/900709 Ltr

ML20044A936
Person / Time
Site:	Sequoyah
Issue date:	07/09/1990
From:	Bynum J, Thompson R TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-011, LER-90-11, NUDOCS 9007170021
Download: ML20044A936 (8)

Text

n a c,c 1 4

4. -TENNESSEE. VALLEY AUTHORITY 6N 38A Lookout' Place Chattanooga, Tennessee 37402-2801 r

July 9, 1990

.U.S. Nuclear. Regulatory Commission-ATTN Document Control Desk j Washington, D.C. 20555 .l l

Gentlemen:

TENNESSEE VALLEY AUTHdRITY - SEQUOYAH NUCLEAR PLANT UNIT 1 - DOCKET NO.

50-327 - FACILITY OPERATING LICENSE DPR 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 l prohibited by technical specifications and 10 CFR 50.73(a)(2)(vii) as a single i cause resulting-in. multiple inoperable.-channels.

j 1

-Very truly yours, i

TENNESSEE VALLEY AUTHORITY

.l

/.

.-R. Bynum,. ce President 1 Nuclear Power Production Enclosure

- cc ' (Enclo'sure ):  ;

Mr; J. N.' Donohew U.S. Nuclear Regulatory Commission One White Flint, North j 11555 Rockville Pike Rockville, Maryland 20852  ;

q INPO Records Center Institute.of Nuclear Power Operations 1100 Circle 75 Parkway, Suite 1500 1 Atlanta, Georgia.-30339 j

NRC Resident Inspector' 3 Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy. Daisy,-Tennessee 37379 Mr. B. A. Wilson, Project Chief U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 900717002'l YVU/UY PDR ADOCK 05000327 PDC

/ g S 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) lDOCKETNUMBER(2) lPAGE(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 Calibrated Egg ule of Misinterpretation of Vendor Info m tion EVENT DAY (5) I LER NUMBER (6) l REPORT DATE (7) l OTHER FACILITIES INVOLVED (B) l l l l l SEQUENTIAL l l REVISION l l l l FACILITY NAMES l00CKETNUMBER(S)

MONTHl DAY lYEAR lYEAR l l NUMBER l l HUMBER IMONTHI DAY lYEAR I 1015l010101 l l I I I l_l l_I I I I I I ,

Q.,j,_11 01 Bf 91 Ol 91 of I0l 1 l 1 I l 0 1 0 1 01 71 Ol 91 91 of 1015l010101 l l

-OPERATING l lTHISREPORTISSUBMITTEDPURSUANTTOTHEREQUIREMENTSOf10CFR$:

MODE l l_ICheckoneormoreofthefollowino)(11)

(9) I il (20.402(b) l_l20.405(c) l_l50.73(a)(2)(iv) l_l73.71(b)

POWER l l_l20.405(a)(1)(i) l_l50.36(c)(1) l_l50.73(a)(2)(v) l_l73.71(c)  !

LEVEL l l_l20.405(a)(1)(ii) l_l50.36(c)(2) lEl50.73(a)(2)(vil) l_l0THER(Specifyin

_ (10) 1 01 21 41 l20.405(a)(1)(iii)lMl50.73(a)(2)(i) l_l50.73(a)(2)(viii)(A)l = Abstract below and in l_l20.405(a)(1)(iv) l_l50.73(a)(2)(ii) l_l50.73(a)(2)(viii)(B)l Text, NRC form 366A) -i l 120.405fa)(1)(v) l 150.73(a)(2)(iii) I 150.73(a)(2)(x) l l LICENSEE CONTACT FOR THIS LER (12) l NAME l TELEPHONE NUMBER lAREACODEl Russell R. Thomoson. Comollance Licensino Enoineer l6l1 15lBl4l3l-l7l417 l0 I COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) l l l .l REPORTABLE l l l l l l REPORTABLE l CAUSElSYSTEMI COMPONENT lMANUFACTURERl TO NPRDS l lCAUSElSYSTEMI COMPONENT lMANUFACTURERl TO NPRDS l l I l l l l I I i l l 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 l 1 I I I I I I I I I I l l l t i I I I I I I I I I l l t l l l I I I i ,

SUPPLEMENTAL REPORT EXPECTED (141 l EXPECTED lMONTHlDAYlYEAR l_ l SUBMISSION'l l l ,

I YES (If ves. comolete EXPECTED SUBMISSION DATE) l X l NO I DATE (15) l l l l l I l 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 was determined that Unit I had operated in noncompliance with Technical Specification 2.2.1, " Limiting Safety System Settings." The actual Nuclear Instrumentation System

-(NIS) power range (PR) detector currents were determined to have been 20 to 31 percent g lower 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 f May 31, 1990, to June 1, 1990. The nonconservative calibration resulted from a i misinterpretation of vendor information when calculating expected NIS IR and PR detector 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) accident analyses limits. The prestartup NIS calibration procedures will be revised to require j the use of the correct prediction methodology. The NIS correction procedures will be revised to provide guidance for performing evaluations of observed NIS deviations in NIS 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) lDOCKETNUMBER(2)l LER NUMBER (6) l l PAGE (3) l l l l l$EQUENTIALl l REVISION l l l l l Sequoyah Nuclear Plant Unit 1 l lYEARI I NUMBER l l NUMBER l l l l l l 1015l0101013 12 17 19 10 l--I 0 1 1 l1 l-l 0 1 0 1 01 2l0Fl 01 7 TEXT-(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 percent reactor power, reactor coolant system (RCS] pressure at 2,235 pounds per square inch gauge (psig}, and RCS temperature at 555 degrees Fahrenheit (F)), it was determined that. 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 been nonconservatively calibrated. TS 2.2.1 requires the reactor trip system -!

instrumentation and interlock setpoints to be set consistent with trip setpoint values j l of TS Table 2.2-1. This requires the NIS PR channels to have high and low setpoints of I

l. less than or equal to 109 and 25 percent reactor power, respectively; and allowable I 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. j If a reactor trip system instrumentation or interlock setpoint is found to be less j conservative 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, achieved criticality at 1033 EDT on May 31, 1990; and entered Mode 1 at approximately 1728 EDT on June 1, 1990. On June 2,1990, Unit 1 tripped from approximately 11 percent power at 0802 EDT. The unit was again critical at 2104 EDT on June 2, 1990, and entered Mode 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 completion  !

of the maintenance activities, power escalation was initiated, and the unit entered

-l 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 detector currents were predicted, and adjustments were made based on the past cycle low leakage core 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 uses calorimetric data to check the output of the IR channels. The performance was done to further check the NIS detectors and to then allow readjustment of the IR trip setpoints to the normal 25 percent nominal value. Deviations of approximately 5 percent voltage span were observed, i.e., the IR detectors indicated approximately one-half decade below actual reactor power as determined by calorimetric data. The accuracy of the calorimetric data was questioned because of low secondary system flowrates. IIoweve r ,

the IR channels were adjusted consistent with the more conservative callometric indicated power. The deviations were not considered to be atypical for an initial startup from a refueling outage. No anomolies were identified in the PR indication at this 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) l DOCKET NUMBER (2) l LER NUMBER (6) l l PAGE (3) l l l l SEQUENTIAL l l REVISION l l l l l Siquoyah Nuclear Plant Unit I- l lYEAR I l HUMBER l l HUMBER l l l l l 101$10101013 12 17 19 10 l--! 0 l 111 l-I01 0 l 01 310FI 01 7 TEXT (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 Balance Comparison Daily," to adjust the PR channels to within 2 percent of the secondary side calorimetric full power value. Calorimetric data at this power level was considered reliable, and deviations of approximately 20 to 31 percent between actual and-prestartup estimated detector currents were noted. Again the deviations were not considered 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 by Reactor Engineering that the projected normalized top and bottom excore PR detector currents for 100 percent power indicated that the PR channels had been initially calibrated with nonconservative values prior to startup from the Unit I refueling outage due to an incorrect equation used to calculate preliminary detector current values. 'This error caused the PR channels to indicate lower than actual power by 20 to 31 percent of the power level, and resulted in the NIS reactor trip setpoints being set 20 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 full power detector currents calculated in 0-PI-NXX-092-001.0 were compared to the prestartup calibration detector currents calculated in 1-PI-NXX-092-001.0, "Prestartup NIS Calibration Following Core' Load." Subsequent evaluation also determined that the same error had been made-in predicting the prestartup IR detector currents resulting in the 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 detector current values. The-PR and IR channels are then calibrated using the preliminary values prior to startup. The procedure utilized power distributions from beginning of life (BOL) of the previous cycle (PR old), power distributions from BOL of the new cycle (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) This calculation methodology was believed to have been in accordance with Westinghouse Electric Corporation information. However, it was determined that the information had been misinterpreted. The reason for the resultant nonconservative values was the use of BOL power distributions with E0L detector currents. Power moves towards the outer core 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 for this cycle), and this resulted in larger nonconservative preliminary detector current values being calculated. More accurate preliminary current values would have been calculated 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 values was recommended by Westinghouse in discussions held subsequent to this event. i No deviations outside of TS allowable values were noted during previous startups from refueling outages at SQN, although the initial prediction of full power NIS IR and PR detector 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) lDOCKETNUMBER(2) l LER HUMBER f6) I l PAGE f3) l l l l$EQUENTIALl l REVISION l l l l l Sequoyah Nuclear Plant Unit 1 l lYEAR l l NUMBER l I NUMBER I l l l l 10151010101312171910 l-l 0 l 111 l-l0l0 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 IR indication and actual reactor power was difficult to observe directly. The status of the IR indications has been specifically monitored since the Unit 2 IR detector mispositioning_ event (LER 50-328/89006). No corrections of the IR indications would have 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 is blocked at 10 percent reactor power.

For the NIS PR channels, daily adjustments of PR indication are made as necessary by Operations' personnel when reactor power is greater than 15 percent. These daily adjustments match PR indications to plant calorimetric data. Pe.st reviews of the adjustment _ procedures.between 15 and 30 percent focused primarily on "as-left" data.

Calibrations of the PR channels would have been performed as-required after performance af the 30 percent and full power incore-excore cross-calibrations. For both the IR and PR channels, no accurate calorime:.ric data has been available typically prior to 30 percent reactor power for comparison. A possible reason why a deviation was more evident during this startup may be related to the point in the operating cycle where the EOL detector current was obtained. For this specific case, the last cycle detector current 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 on misinterpreted vendor information. A Westinghouse letter on low leakage loading (L3 P) schemes stated. that "The expected values of these currents can be estimated from the last previous cycle's Incore-Excore calibration and the predicted X-Y power distribution for the L P3 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 both cycles.

A contributing cause to the duration of the event may have been the lack of procedural guidance concerning the evaluation of deviations.- Disagreement between predicted detector currents and calorimetric indicated power was considered to be solely a function of the inherent uncertainty involved in the prediction and measurement processes,'i.e., the refueling had included a lower leakage loading pattern, gamma-metrics had been installed for the IR detectors, secondary side flow measuring devices had been cleaned and recalibrated, and accuracy of the secondary side L calorimetric data of low power levels is considered inaccurate. As a result, potential for disagreement was considered likely and to some extent unavoidable; resolution of deviations consisted of readjustment of NIS to match best estimate power particularly

,during low power escalation. More detailed evaluation of the cause of the observed deviation in the IR detectors on June 1 may have identified the potential for error in the 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) lDOCKETNUMBER(2)l LER NUMBER (6) l l PAGE (3) l l l l$EQUENTIAtl l REVISION l l l l l Sequoyah Nuclear Plant Unit 1 l lYEAR I I NUMBER l l NUMBER l l l l l 1015l0101013 12 17 19 10 l--l 0 l 1 1 1 l--l 0 1 0 1 Ol_$10Fl 01 7 TEXT (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 condition prohibited by TSs and in accordance with 10 CFR 50.73(a)(2)(vii) as a single cause resulting in multiple inoperable channels.

Section 15.2.1.1 of the Updated Final Safety Analysis (UFSAR) states that the reactor trip for a postulated uncontrolled rod cluster control assembly (RCCA) bank withdrawal from suberitical conditions is assumed to be initiated by PR high neutron flux (low setting). The most adverse combination of instrument and setpoint errors, as well as delay for trip signal actuation and RCCA release, is taken into account. A 10 percent increase is assumed for the PR flux trip setpoint raising it from the nominal.value_of 25 percent to 35 percent. Previous results, however, show that rise in the neutron flux is so rapid that the effect of errors in the trip setpoint on the actual time at which the rods are released is negligible. In addition, the reactor trip insertion characteristic is based on the assumption that the highest worth RCCA is stuck in its fully-withdrawn position.

Prior to 24 percent' power, the PR channels were adjusted such that the low flux trip bistables would have actuated at 33 percent (worst case), which is less than the 35 percent setpoint used in the analysis. The plant safety analysis takes no credit for 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 reactor trip biotables had-been set to one-half of their normal 25 percent power setpoint (12 percent). Even with this additional conservatism, during a transient,' reactor power 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 would have 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 would have 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 PR high flux trip at 109 percent power. With the PR channels improperly calibrated, the high flux trip setpoint was approximately 135 percent, which was outside of the safety analysis-of 118 percent power. Another reactor trip used to mitigate this event is

over temperature delta temperature (T), provided the reactivity addition rate is within certain limits. Over temperature delta T is used to mitigate slower reactivity insertion 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) l l l l$EQUENTIALl l REVISION l l l l l Sequoyah NvClear Plant Unit 1 l lYEAR l I NUMBER l I NUMBER l l l l l 1015l0101013 12 17 19 10 l--! O l 1 11 l--I 0 1 0 I of 610Fl 01 7 TEXT (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 least 130 steps out of the core, (range is 0 to 230 steps). For conservatism, if 100 steps out of the core is used, the intergral rod worth for control bank D (from 100 to 230 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 average reactivity 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 protection envelope of over temperature delta T, and the high flux trip would not have been required. The peak reactivity insertion rate for this scenario is approximately 7.50 E-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 trip to mitigate the effect of a dropped group of rods. As with the positive rate trip, the negative 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 core power 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 power levels less than 10 percent. This restriction provides sufficient margin to the accident analysis limits that a reactor trip signal is not generated. Therefore, the shift in the P-8 setpoint resulting from the nonconservative NIS PR calibration would have 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 suberitical conditions analysis.

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

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

NRC Fonn 366(6-89)

l. 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) lDOCKETNUMBER(2)i LER NUMBER (M l l PAGE (3) l l l l$EQUENTIALl l REVISION l l l l l

'Sequoyah Nuclear Plant Unit 1 l lYEAR l I NUMBER l I NUMBER l l l l j l0l510101013 12 17 19 10 l- I O l 1 I ) 1-!OI0 1 01 710Fl 01 7 TEXT-(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 reactor power detector currents on June 8, 1990. To prevent recurrence, Pts 1- and 2-PI-NXX-92-001.0 will be revised to require the use of EOL NIS detector currents and fuel assembly power fractions from the last cycle and the corresponding BOL fuel assembly power fractions for the upcoming cycle.

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

Additional Information A similar event was reported by LER 50-328/89006, which reported the mispositioning of the NIS IR detectors for Unit 3. The NIS PR detectors were not affected during this similar event. Corrective actions for this event were directed at configuration control and the status of Ik indications during power escalation. Although this previous event further heightened sensitivity of station personnel to potential for NIS anomolies, corrective actions would not have been expected to have prevented this event. These events and similar events experienced at other utilities, emphasize the need for continued diligence in improving NIS prediction, monitoring, and assessment performance.

Commitments

1. Periodic Instructions 1- and-2-PI-NXX-92-001.0 will be revised before startup from Unit 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 the corresponding "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 4 refueling outage to provide guidance for performing evaluations of observed deviations in NIS detector indications.

0921h-NRC form 366(6-89)

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