ML20196E859

TMI-1 Cycle 7 Startup Rept

ML20196E859
Person / Time
Site:	Three Mile Island
Issue date:	12/05/1988
From:	Hukill H GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
C311-88-2163, NUDOCS 8812120107
Download: ML20196E859 (38)

Text

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TMI-1 CYC1E 7 STARIUP REPORT f

331212Q107 OC1'EOb DR ADOCK 0000 ((9 g $ (p

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TABE OF CDtTIT2RS PAGE 1.0 CORE PGEDBMANCE - MEASURDG2RS AT ZIPO IUdIR - SLMiARY . . . . . . 1.

2.O CORE PERFDBMANCE - MEASURDE2fIS AT REER - SUM 1ARY . . . . . . . . . . .4.

3.0 CORE PERFDPMANCE - MEASURDE2ES AT ZEPO POWER . . . . . . . . . . . . . . . . 6.

3.1 Initial Criticality ..................................... 6 3.2 Nuclear Dutrumentation Overlap . . . . . . . . . . . . . . . . . . . . . . . . 11 3.3 Reactimeter Chedout . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

6 12 3.4 ABO Critical Boron Concentration ....................... 13 3.5 Tertperature Coefficient Measurements ................... 14 3.6 Control Rod Group Worth Measuru w.nts ................... 16 3.7 Differential Boron Worth................................ 23 4.O CDRE PERFDFMANCE - MEASURDiENIS AT EDWIF .................. 24 1 4.1 lu:: lear Instrumentation Calibration at Power . . . . . . . . . . . 25 l

4.2 Incore Detector % sting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 4.3 Power Imbalance Detector Correlatico Tbst .............. 28 4.4 Core Power Distribution Verific1 tion ................... 33 4.5 Reactivity Coefficients at Power ....................... 35 t

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1.0 Q2BE PERFDPMANCE - MEASURD4DTIS AT ZERO RVER -

SUMMARY

Core perform noe measurements were conducted & ring the Zexo Pcuer Test Program which began on August 14, 1988 ard eIdsi on Atx3ust 15,1988. This section presents a sumary of the ::ero power measwaments. In all < aw, the appli able test ard Technical Specifications limits weru met. A su:mury of zero power physics test results is included on Table 1-1.

a. Initial Criticality Initialcriticalitywasachievedat1415onAugust 14, 1988.

Reactor conditions were 532 7 and 2155 psig and control rod glaups 1 thrcugh 6 were withdrawn to 100% while group 7 was positioned at 85% withdrawn. Cbntrol Rod group 8 was positioned at 25%

withdrawn. Criticality was achieved by deborating the Reactor Coolant frun 1958 ppn to 1636 ppn. Initial criticality was achieved in an orderly manner and the acceptance criteria of 1577 100 PIN van met.

b. Nucleg_ Instrumentation Overlao At least one rhv% overlap was measured between the source and intermediate ran Specifications. ge neutron detectors as required by Technical

c. Reactimeter Checkout An on-line functional check of the Mod-Ccrp reactimeter (using NI-3) was performed after initial criticality. Reactivity calculated by the rW,timeter was within 5% of the core reactivity determined frus dcublity time measurenents.

d. All Reds out critical Boren Cbncentrat(qn The measured all rods cut critical boron concentration of 1692 pSr3 was within the acx:eptance critaria of 16361 100 ppnB.

e. Te:meraturn Coefficient Maamirenents The measured teperature coefficients of reactivity at 5320F, zero were within the acceptance critaria limits over the range of ren ecocentrations and red positions that the measurements were made.

f. Cbntrol Rod Grrun Worth Maannernents The naasured results for con 1 rod worths of groups 5, 6 and 'i conducted at zero power (532 ) usinJ the bortrVred swap method ure in good agreemer.t w;.th predicted values. The naxirum deviation between measured ard predicted worths was 5.6% which was for (m-7 worth.

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g. Diffe'mitial Boron Worth The predicted differential boron worth at 532 F0 was 5.44% less I than the maamired value. This is within the bounds of the FSAR and B&W supplied limits of i 15%.

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r i TABIE 1-1 Sumary of Zero Power Physics Test Pe_.mdt;g Cycle 7 Acceptance Measured Parameter Critaria value Critical Boron 1577 i 100 ppn 1636 ]:pa NI overlap >l h ada >1.2 h ada Sensible Heat N/A 1.0 x 10~7 m All Reds Out Boron Concentration 1636 100 pgn 1692 ppn Teraperature Coefficient 2.87 (1682 ppn) 2.41 pczV0F i4 rioderator Cbefficient <5.0 pcmVOF 4.09 pcmV0F In Red Worths (532 ) GPS-7 3157 PO4 10% 3070 PCM Group 7 978 PCM i 15% 925 PCM Group 6 971 PCM i 15% 924 PCM Grolp 5 1208 K M 15% 1220 KM Diff Boron Worth 8.74 PC2VPIM ! 15% 9.24 KM/ PIN

i t 2.0 CORE PEFDPMANCE - MEASURDOUS AT PCWER - SUWARY This section su::tnarizes the physics tests conducted with the reactor at power. 1bstir themal power.q was perfomed Operation at power in the power rarge plateaus of 25, began 75, and15, on August 100% core 1988.

Periodic maastuw=nts ard calibrations were performed on the plant nuclear instrumentation during the elation to power. The four power range detector channels were calibrated haamd upon primary and secondary plant heat balance measuresnents. Ibsting of the incore nuclear instrumentatien was perfomed to enetre that all detectors were functioning properly aM that the detector outputs were processed correctly by the plant ccmputer.

Core axial imbalance detemined frun the incore instrumentation s used to calibrate the out of core detector imbalance indication. ystem was The major physics measurernants perfomed during power aamlation consist of detemining the moderator and power doppler coefficients of reactivity arri obta

[NBR and . detailed radial and axial imbalances. Values of minin a linear heat rate were monitored throughout the test program to ensure that corn thermal lianits were not avnaMad.

a. Nuclear Instrumentation Calibration at Power 7he power range channels were calibrated as required dur the startup prcgram based on power as detemined by primary secondary plant heat balance. These calibrations were required due to level, boren and/or control red ccnfiguration charges dur testirg.

b. Incom Detector 7bstirn Tests conducted on the incore detector system denonstrated that all detectors except level 1 of string 49, level 2 of string 31, level 4 of string 4, and level 5 of string 48 were furntionirq as expected.

Symetrical detector raMings agreed within acceptable limits and the plant ctrputer applied the correct backgreurd,1 and depletion correction factnrs. The backup incore reco rs were operational above 80% FP as required by Technical Specifications,

c. Power Irbalance Detector correlation Test The results of the Axial power Shaping Rod (APSR) scans perfomed at 75% FP show that an acceptable incere versus out-of-core offset slcpe of 0.96 is obtained by using a gain factor of 3.684 in the power range scaled difft rence lifiers. The measured values of minimum [NIR and maximum linear t rate for various axial core imbalances indicate that the Reactor Protection Trip Setpoints provide adequate protection to the core. Irbalance calculaticos usirg the backup recorder provide a reliable altemative to ccrputer calculated values.

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d. Cbre Power Distribution Verification (bre power distribution measurements were conducted at 75% ani 100%

full power under steady state equilibrium xenon conditions for specified control rod configurations. 'Ihe maxinna measured and W="a predicted radial and total peaking factors are all in agisiii:nt. 'Ibe maximm differunoe between a measured and p cted value was -2.4% for total peaking at 75% FP. 'Ihis met acceptan critaria of <7.5%.

'Ihe results of the core power distribution measurunents are given in Table 4.4-1. All quadrant tilts and axial core imbalances measured during the power ibution tests were within the Technical Specification and normal operational limitI.

o. Beactivity Coefficients at Power

'Ihe iso terperature coefficient measured at 100% FP was i

-5.25 .

'Ihe measured power doppler coefficient at 100% FP )

was -7.59 pc=V% FP. All 'DK:hnical Specification and Safety Analysis requirements were met.

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3.0 CDRE FERFDPMANCE - MEASUP4fNIS AT ZERO FWER This section presents the detailed results and evaluations of zero power physics testirg. The zero power testirg program included initial criticality, nuclear instrumentation overlap, reactimeter checkout, all rods out critical boren cciceriration, temperature coefficient measurenent, control rod worths, and differential boron worth.

3.1 Initial criticality Initial criticality for Cycle 7 was achieved at 1415 on August 14, 1988. Reactor conditions were 532 F and 2155 psig and control red 1 through 4 waru previously withdrawn during the heatup co 532 .

The initial reactor coolant system (RCS) boron concentration was 1958 [pn.

The approach to criticality began by withdrawiry control rod group 8 to 25% withdrawn, control red groups 5 and 6 to 100% withdrawn, and positi achieved 7 at 85% withdrawn. Criticality was s11h =1uently rating the reactor coolant system to a boron concentrat on of 1636 ppn. The procedure used in the approach to critical is cutlined below in three basic steps:

Step 1 Control Rod Withdrawal Group 8 25% withdrawn Group 5 100% withdrawn Group 6 100% withdrawn Group 7 85% withdrawn Step 2 Deborate using a feed and bleed flow rate of 70 gpn until the inverse ocunt rate is between 0.2 and 0.3. At this point, deborate at a feed and bleed flow rate of 40 gpn.

Step 3 At an inverse ccunt rate of 0.1, stcp deboration and increase letdown flow to maxinum (120 gpn). This enhances mixing between the mkeup tank and the reactor coolant system. Achieve initial criticality and position control rod group 7 to control neutron flux as the reactor coolant systen boron concentration reaches equilibrium.

Thrt.yout the approach to criticality, picts of inverse nultiplicaticn were maintained by two independent persons. Two plots of inverse count rate (IG) versus control red position were maintained during control red withdrawal. S o plots of ICR versus RCS bortx1 cucentration and two plots of ICR versus gallons of demineralized water added were maintained durity the dilution sequence. Dur each reactivity addition (boren dilution and control rod vi wal), count rates were obtained frcra each source rarge neutron detector channel.

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I Durirg control rod witMrawal (step 1) IG plots versus control red group position were maintained fran the outputs of two source rarge detectors. The withdrawal interval for eads control red group was limited to no more than half the reainire predicted distance to criticality as determined from the IG plots.

Deboration of the reactor coolant system was acccmplished in two steps as indicated above. First, deboration fra 1958 ppu was h..iicid using a feed aM bleed flow rate of 70 gpn (step 2) . RCS boron sanples were taken every 30 minutes ard sangles fran the Irakeup tank and the prissurizar were taken hourly. 'De IG plots were maintained vs. lens of danineralized water aMari aM two plots were mainta 30 minutes. Deboration at vs.a RCS letdown letdown w w it. was rate of 70 ration contevery,inued unti one of the IG plots was less than 0.3. At time, demineralized water addition was reduced to 40 gpn. When one of the ICR plots indicated 0.10, the letdown flow rate was increased to 120 gpn in a recirculation mcde to expedite in the RCS (step 3). The reac*or went citical during the pr m a.

The inverse count rate plcts maintained during the aEproach are presented in Figurus 3.1-1 through 3.1-3. As can be seen frtan the plots, the response of the source range channels during reactivity additions was very good. Figure 3.1-1 is the plot of ICR versus control rod group withdrawal. Figure 3.1-2 is the ICR plots versus RCS boron ccrann. ration and Figure 3.1-3 is the ICR picts versus gallons of demineralized water aMui to the RCS.

In sumary, initial criticality was achieved in an orderly manner.

The measured critical boron concentration was within the acceptance criteria of 1577 i 100 FIN.

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0.00 2000.0 1950.0 1900.0 1850.0 1800.0 1750.0 '1700.0 1650.0 RCS BORON CONCENTRATION

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3.2 Nuclear Instrumentation Overlao

a. Purtose Technical Specification 3.5.1.5 states that prior to cperation in theone least intemediate nuclear

• • of overlap instnnentation between the sourc(NI) range, e range NIs at and the intemediate rany nust be r*=tved.

b. Test Method

'Ib satisfy the above overlap requirements, oore power was increased until the interwwiinte rarge channels came on scale.

Detector signal response was then reocrded for both the source rarge and internwiinte range channels. 'Ihis was repeated until the sourt:e range high voltage cutoff value was reached.

c. Test Results h results of the initial NI overlap data at 532 F0 and 2155 psig have shown a >1.2 risen overlap between tho source and intemediate ranges.

d. Cbnclusions h linearity, overlap and absolute output of the intemediate and source detectors are within specifications and performing sa factorily. 'Ibere is at least a one he overlap between the scurre and intemediate ranges, thus satisfying T.S. 3.5.1.5.

3.3 Reactimeter Checkceg

a. Purpose Reactivity calculations durirrJ the Cycle 7 test program were performed using the Mod-Ccmp Reactimeter. After initial criticality and prior to the first Ittysics measurment, an online functioral chock of the reactimeter was perfonned to verify its acuiracy for use in the test program,

b. Test Method After initial criticality and nuclear instrumentation overlap was established, intermediate range channel NI-3 was aivested to the reactimeter and the reactivity calculations were started. After steady state ocniitions with a ccristant neutron flux were establishoo, a small amount of positive reactivity was inserced in the cxas by withdrawing wd.wl red group 7.

Stop watches were used to measure the doubl time of the neutron flux and the reaceivity was de frun the doubling time reactivity cMrves. 'Ihe measurunent was for +22,

+34, 50, and -55 pcm. '1he reactivities determined frca doubling, time measurements sure ccmpared with the reactivity calculated by the reactimete .

c. Test Results

'Ihe measured values were determined to be satisfactory and showed that the reactimeter was ready for startup testing.

d. Conclusions An on-line functional check of the reactimeter was performed after initial criticality. 'Ihe measured data shows that the core reactivity measured by the reactimeter was in good with the values cbtained frun neutzzn flux dcubling

o 3.4 All Rods Out Critical Boron Crsainiuation

a. Purpose The all rods out critical boren u s w/m tion measurement as performed to obtain an accurate value for the excess rinactivity 1* in the TMI Unit 1 core and to provide a basis for the verification of calculated reactivity wor .

This measurement was performed at systen corriitions of 532@"F and 2155 psig.

b. Test Method The Reactor Cbolant Systen was borated to an all rods cut condition and steady stata conditicos worin established.

c. Test Results the measured boren u.a s ra tion with group 7 positicned at 100tND was 1692 ppn.

d. Conclusions The above results show that the measured boren u.astation of 1692 ppn is within the acceptance criteria of 1636 i 100 ppn.

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3.5 Terrerature Coefficlent Meeurener.:s

a. Puruose The moderator tenparature coefficient of reactivity can be positive, dependirg upcn the soluble boron concentration in the reactor coolant. Because of this possibility, the Technical Specifications state that the moderator tmperature coefficient shall not be positive Mille greater than 95% FP. The moderator tecperature coefficient cannot be measured directly, but it can be derived fran the oors taperature coefficient and a known fuel taperature (isothermal Doppler) ocefficient.

b. 14st Methort i Steady state conditions wru established by maintainirs reactor flux, rtactor coolant pressure, turbine header pressure and core average yture constant, with the reactor critical at approximately 10 anpa on the intennediate rarge.

Equilibrium boron c.ac e tration was established in the Reactor Coolant System, make-up tank aM pressurizer to eliminate reactivity effects due to boren changes during the subsequent taperature sw connected to The reactimeter and recordezu were tor selected core parameters with the reactivity value calculated by the reactimeter and the core average ta perature displayed on a two channel strip chart recorder.

Once steady state conditions were established, a heatup rate was started by closing the turbine bypass valves. After the core average tenperature ircreased by abcut 50F core tenperature and flux were stabilized and the pr- was

'ersed by decreasing the core average tenperature by ahmt 10 .

After mre tecperature aM flux were stabilized, tore taperature was returned to its initial value. This pr m dire was empleted with control rod group 7 at 90% wi. Calculation of the taperature coefficient frces the measured data was perforined by dividing the change in core reactivity by the conuspciding change in core taperature over a specific time period.

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c. Test Results The results of the isothermal tenperature coafficient measurements are provided below. De ca.lculated values are included for caparison.

In all maan the measured results conpara feNerably with the calculated values.

RCS MEASURED HEDICIED MEASLU 2D BORN REUJIRED ITC IIc MIC MIC (PRO (PCM/DED F1 (PCM/DED F) .[B2f.IED F) (PCM/DED Fi 1682 2.41 2.87 4.09 <5.0

d. .Qgoclusions The naamwed val - of the temperature coefficient of reactivity at 532 , zero reactor are within the acceptance criteria of i 4.0 of the predicted value.

An extrapolation of the ucderator coefficient to 100%FP indicata that 3.1.7.2.

Specifications it is well within the limits of %:hnical l

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3.6 Control Rod Group Worth hiswas--its

a. Purpose

> i B e total amount of excess reactivity present au beginning-of-cycle (BOC), hot (532"-F), clean conditions is 14.64% olyX. During reactor operations, nearly all of the excess reactivity is controlled by the soluble poison systa m .

Additicrial control is provided by moveable control rrds. mis section provides ecsparism between the calculated and measured results fcr the ocatrol red group worths.

Se location and functicm of each ocotrol red group is shown in Figure 3.6-1. ' Ins grotping of the onntrol rods shown in Fi.gure 3.6-1 will be used throughout cycle ?. Calculated a:xt maamtrud control rod group reactivitv worths for the nomal withdrawal were detertair.ed at reactor conditions ol' zero power, 532 atd 2155 psi. S e measured results w:e obtainod using results of reactM+y and group position frun the strip chart recorders,

b. Test Me*tod Ctntrol red grouo vity worth masurments were performed at zero power, 532 using the borerVred swap method. Both

• the differential and integral reactivity worths of cential rod groups 5, 6, and 7 wars dotarmined.

Se boron swap method consisted of establishing a deboration rute in the reactor coolant systen and ccarpensating for the reactivity changes by inserting the control rod groups in incremented steps. Se reactivity changes that cocurzwl during the measurements were calculated by the reactimeter and differential rod worths were obtained frcra the measured reactivity worth versus the change in red group position. 23 differential red "<orths of each group were then sumed to '

obtain the integral red group worths.

c. Test Results i ,

Control rog group reactivity worths tMire masured at zero power, 532 i' conditicos. Se boron / red swap method was used to detamine differrntial ard integral red worths for control red group 5 - 7 frun 100% to 0% withdrawn.

me. integral reactivity worths for control red groups 5 thrux3h 7 are presented in Figures 3.6-2 through 3.6-4.

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These cines were obtained by integrating the measurrd differential worth alwes. 'Ihe integral worth of 8 was not measured. The calculated worth of group 8 for zero power is 0.19% a IQ/K. '

Table 3.6-1 prwides a ocuparison between the predicted W n==1mi results for the rod worth measurements. The results h4 good agreenant between the measured and predicted red p

worths. 'Ihe mavi== deviation between measured and cted was 5.6%.

d. Conclusions Differential and integral control red group reactivity worths i.ere maamired using the boprod swap method. The maa= M results at zero power, 532"F indicate good hjswiduait With the predicted group worths, i

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Fuel Transfer

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A B' 1 6 1 C 3 5 5 3 D 7 8 7 8 7 E 3 5 4 4 5 3 F 1 8 6 2 6 8 1 G 5 4 2 2 4 5 H W- 6 7 2 4 2 7 -Y 6

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1 2 3 4 5 6 7 8 9 10 11 12 13 14 1S I

t X Group Number Grou0 NO. Of Rods function 1 8 Safety 2 8 Safety 3 8 Safety 4 9 Sa fety 5 12 Control 6 8 Control 7 8 Control 8 8 APSRs

. i FIGURE 3.6-2

, , , INTEGRAL WORTH FOR CRG-5

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Total wuth - 12203 PCM

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FIGURE 3.6-3 1000.00 INTEGRAL WORTH FOR CRG-6

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Total Worth = 924.1 PCM l

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FIGURE 3.6-4

,,, INTEGRAL WORTH FOR CRG-7 l -

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Totd Worth = 925.9 PCM

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s TABIE 3.6-1 CMPARISCN OF PREDIC'IED VS MEASURED 300 WORIHs i

MEMiURED HEDICIED PEPC2Nr W i. WORIH WORIH DIM ERENC2 i LQ.a fM (POO (ti ___

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3.7 Differential Boron Worth

a. Purrese Soluble poison in the fom of dissolved boric acid is added to the moderator to prwide additional reactivi control beyorri that available fran the control rods. The pr , functicn of the soluble poiscn control .W is to control excess reactivity of the fuel thmaghout each core life cycle. The differential reactivity worth of the boric acid was measured during the zero power test.

b. Test Method 0

Vaaen =amits of the differential boren worth at 532 F was perfonned in ecojun::ticn with the control red worth mem;urements. The control rods worths were measured by the boren swap technique in which a deborstion rate was established and the cxx1 trol rods were insertad to c.u@ sate for the charging core reactivity. The reactimeter was used to ptwide a o:ntinuous ruactivity calculation throughout tha measurement. The differential boren worth was then determinod by sunaniny the irsauental reactivity values measured during ,

the red wrth measurements cuer a known boron concentration range. The average differential boren worth is the measured change in reactivity divided by the change in boren concentration.

c. Test Results hnts of the soluble boren differential wrth 'ere ccrpleted at the zero power cxandition of 532T. The measured boren worth was 9.24 penVppnB at en average boren c:noentration 1515 ppnB. The predicted value vas 8.74 pcrVppnB t 15%.

d. Conclusiong

! The meg ered results for the soluble poison differential worth at 532'T was within 15% of the predicted differential worth. l 1

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I 4.0 CORE PGIOPMANCR, - MEASURDE2TPS AT RMER

'Ihis section premnts the results of the physics measurenents that wem cxmducte' sith the reactor at power. Testing was conducted at three rajor power placeaus, 25%, 75), and 100% of 2568 megawatts core themal power, as determined frun primary ard secondary heat balance naasurenents.

Operation in the power rarge began en August 15, 1988. Power escalations occurred as the required testing at each plateau was sumaufully cxxpleted.

Periodic neasurunents and calibrations were performed en the plant nuclear instrunentation during the -laticn to full power. 'Ihe four power rarge detector channels were calibrated haaai upon primary and secondary plant heat balance measurenants. 'I%stirq of the incors nuclear iremtion as p*.rforred to ensure that all detectors were functioning prcperly arri that the detector inputs were proassed correctly by the plant cxztuter.

Core to used axial irbalance calibrate the dttemined out of corefrun the incere detector instrumentation imbalance s inilcation. ystan was

'the major physics measurements perfomed during power elation consisted cf detemining the moderator and power Dcppler coefficients of reactivity and cbtaining detailed radial and axial core pcser distribution measurenents for several core axial imbalarres. Values of minimum EMR and maximum linear heat rata were tenitored throughcut the test program to ensure that core therral limits would not be a_ W _.

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4.1 1Alclear Instrumentation Calibration at Ptmer

a. Purpose The purpose of the Nuclear Instrumentation Calibration at Power '

was to calibrate the power range nuclear instrumentatico inlicaticn to be no less than 2% D of the reactor thermd as detamined by a heat balance and to within 2.5%

systm.

re axial offset as determined by the incore monitoring

b. Test W hod As . tired during power owcalatico, the top and bottm linear anpli..lar gains were adjusted to maintain power range nuclear instrumentation indication to be not less than 2% of tha power calculated by a heat balance.

When directed by the controlling procedure for physias testinJ, the high flux trip birtable setroint was adjusted. The major settings durin; power escalation are given below:

Test Plateau Bistable Setpoint

% FP  % FP 30 40 75 85 100 105.1

c. Tbst Pesults An analysis of test results iniicated that changes in Paactor Coolant Systs boren and :enen buildup or burnout affected the pow r as observed by the nuclear instrumentation. This was as expected since the ,xwer range nuclear instrumentation measures reactor neutron lea: whidi is directly related to the above changes in systen itions. Each time that it was necessary to calibrate the powr range nuclear instrumentaticn, the acceptance criterla of calibration to be nc less than 2.0% FP of the heat balance power was met withcut any difficulty.

Also, each time it was r=,ame n to calibrate the power range '

nuclear instrumentatico, the 2.5% axial offset criteria as determined by the incore monitoring syst m was also met.

l The high flux trip bistable was adjusted to 40, 85 an! 105.1%

FP prior to escalation of power to 30, 75 and 100% FP, respectively.

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d. Corrlugigos h peer range dunnels e calibrated haw on heat bal cal ra N % pwar '

bo control rod configuratim changes during the program. ard/o Acomptance criteria for ruclear instrumentation calibration at power e net in all instimons.

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4.2 Incore Detector Testim

a. Eirroes Self-powered-neutrcm-detectors (incore detector system) monitor the core power densit '

xnitcand and prma==y =1uithin by thethe oors plant and their otmputer tooutputs prwide are accurata rmadings of relative neutron flux.

%sts conducted on the incore detector system were perfomed to:

(1) Verify that the output frun each detector and its response to incruasing reactor power was as expected.

(2) Verify that the background, length and depleticn corrections applied by the plant ocmputer are correct.

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(3) % measure the degree of azimthal symetry of the neutron flux at 25%FP.

b. Test Method ha response of the incore detectors versus power level was detamined and a ocuparison of the symetrical detector outputs made at steady stata reactor power of 25, 75, and 100%FP.

Using the corrected SEND maps, calculations were performed to determine the detect.r <21rrerf. to average detector current values per assenbly for each incore detector versus axial positions. Any detector levels which w.re determined to have failed were deleted fran scan or substituted.

At 75% FP, SP-1301-5.3, Inmre Neutron Detectors-Monthly Check '

was perfomed to calibrate the back-up recorders to its incore, depletion value.

c. Conclusions Incore detector testing during power el.ation demonstrated that all detectors exospt level 1 of string 49, level 2 of str 31, level 4 of strirq 4, and level 5 of string 48 were cnirq as expected. Symetrical detector rewlirgs agreed within acceptable limits ard the ocmputer applia? oornction factors are accurate. We backup incore recorders were calibrated at 75% FP and cperational above 80% FP as Inquired by the %chnical Specifications.

o 4.3 Power Imbalance Detector Cbr23tlation %st

a. PurDose The Power Inbalance Detector Correlation Tbst has four cbjectives:

1. To determine the relationship between the indicated out-of-corn power distribution and the actual incorn power distribution.

2. Ib ^ -i. ruta axial ocotrol using the Axial Power Shapirq Rods (APSR's) .

3. 'Ib verify the WM and accuracy of backup inbalance i calculations as done in AP 1203-7, "hM Calculation for Quadrant Power Tilt and Cors Power Imbalance."

4. Ib determine the core myLnn lhear heat rate and h ENER at various power imbalances,

b. M Method This tzst was ocoducted at 75% FP to determine the relationship between the oors axial inbalance as indicated by the incore detectors and the out-of-core detectors, w upon this correlaticm, it could be verified that the mininum EMR and I myi== linear heat rata limits would not be avnaariai by operatirq within the flux / delta flux / flow envelope set in the Reactor Protection Systen.

As CE-8 was moved to establish the various inbalances, the integrated control systen automatically c.uww.ted for reactivity changes by repositionirg CM-7 to maintain a constant power level.

c. Test Results The relationship between the ICD and OCD offt;et was determined at 75% FP by perfornirq an inbalance scan with the APSR's. The average slope measured on the four out-of-core detectors was 1.050. 7he lowest slope was .995 for NI-7. The scaled i

difference anplifier gain was 3.684.

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l A c:mparison of the incora detector (ICD) offset versus the out-of-core is shown in (OCD) detector offset obtained for each NI channel Table 4.3-1.

Core power distributicn maasnirements were taken in ocnjunction with the met positive and most negative imbalances at 75% FP and the values of minina DE and worst case MI1R and ctrpalwl to the ameptance criteria.

'Ihe worst case values of minian DM and maxina linear heat rata dete: mined at 75% FP are listed in Table 4.3-2.

The worst case D a ratio was greater than the minia n limit and the wayi== value of linear heat rate was less than the fuel melt limit of 20.5 kw/ft after extrapolation to 105.1 FP.

'Ihese results show that 14chnical Specification limits have been met.

Backup offset calculations using AP 1203-7 agree with the ocmputer calculated offset. Table 4.3-3 lists the ocupater calculated offset as well as offsets obtained using the irecre detector backup recorders.

d. Conclusions Backup imbalance calculations p?.rformed in accordance % .th AP 1203-7 prtnide an acceptable altamate method to c@tr calculated values of irbalance. A difference anplifler K factor of 3.684 will prtwide a sicpe greater than or eql. tl to 0.96 when OCD offset is plotted versus ICD offset.

Mininum DR and Maxian Linear Heat Rate prameters were well within 'Ibchnical Specifications limitations.

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o TA31E 4.3-1 INCORE OFESET VS cur-OF-GRE OFFSET INCCRE OFTSEr cur-OF-< IRE OF73Er (%)

(%) NI-5 NI-6 ____ NI-7 NI-8 10.498 11.308 11.318 10.080 10.573 8.247 9.385 9.507 8.545 8.852 5.048 6.332 6.497 5.869 5.999

-0.652 0.670 0.817 0.725 0.684 '

-0.906 0.384 0.481 0.348 0.388

-4.930 -4.798 -4.752 -4.522 -4.497

-10.994 -11.289 -11.165 -10.163 -10.421

-17.824 -18.957 -19.024 -17.418 -17.693

-18.924 -20.395 -20.421 -18.718 -18.980

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TMEZ 4.3-2 WORST CASE [NER AND UR l IMBAIANCE MINIM M EX'IRAICIA2TD WORST CASE IIR f%) IMR M NIEt DCIRAP. MAX. IIR fIGUPT) (16UFT) 14.07 4.612 3.09 10.59 14.42 7.88 4.556 2.91 10.06 13.27 F

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I E E I OFFSET OFF5Er (t) (t) -

-0.652 -1.056 i 10.498 8.044 i

-18.924 -16.647

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4.4 Coze Ibwer Distrib2 tion Verification

a. Purrose 7b maaetre the core power distributicns at 75 and 100 percent full power to verify that the core axial imbalarce, quadrant power tilt maxian linear heat rate arrt minian [NBR do not er ,aad the b specified limits. Also, to capare the measured and predicted power distributicos.

b. Ipst Method Cbre power distribution inaamirements wre performed at 75 and 100% full power, under stead control red configurations. yIbstata conditions, provide for specified the best cxr.parison between raaamitwd and predicted results, three-di2nensional equilibrium xenon conditions were established. Data collected for the measurunents cxmsisted of detailed distribution information at 364 core locations frun the re detector systan and the worst case core thermal coniitions sure calculated using this data. The measured data was ccrpared with calculated results.

c. 14st Pasults A sunnazy of the mama studimi in this report is given in Table 4.4-1 whlch gives the corn power icvel, ocotrol red pattern, cycle himap, boron c w4 ration, axial imbalance, maxirum quadrant tilt, minian [NIR, maximum 1RR and power peakinJ data for each measurunent. The highest Worst Case MIRR was 12.09 at 100% FP which is well below the limit of 20.5 kw/ft. The lowest minian [NIR value was 3.806 at 100% IT whicu is well above the limit.

The quadrant tilt and axial 12nbalance values measured warn all wi the alloweble limits. Table 4.4-1 also gives a ocuparison between the 'nayL=n calculated and prailcted radial and total peaks for an eighth cerc power distribution.

d. Ocnclusions Cbre pcher discribaticn measuriments were conducted at 75% and 100% full power. Ocmparison of measured arti prailcted results show good agreement. The - L= n difference between a measured i and predicted value was -2.4% for total peaking at 7'A FP. '

This met the acceptance critaria of < 7.5%.

The measured values of INER and MIRR were all within the allowable limits. All power tilts and axial core imbalances measured dur the power distribution test were within the Dchnical spec ficatlans and normal cperational limits.

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TABT.E 4.4-1 CORE IOGR DISIRIREICH RIsJLTS POWER PIAEAU 75% 100%

MIT 08-18-88 Actual Power %FP 08-21-88 75.18 99.96 QC 1-6 %WD 100 QG 7 100

%WD 88.6 90.0 '

QC 8 31.0 Cycle &lrrup

(%WD) 32.0 (EFTO) 1.23 3.79 Bortm conc. (PIM) 1320 1211 1

Imbalance (%)

May(m m Tilt

-0.40 -5.74

(%) 0.31 0.37 MNER 5.115 3.806 Worst Chae MIRR (W/FT) 8.93 12.09 Maxima Radial Peak Measured 1.293 Predicted 1.292 1.290 1.280 Difference (%) 0.232 0.93 Acce Mari==ptance Tbtal Peak Critaris (%) 55% 55%

Measured 1.533 Predictal 1.557 1.570 1.$/0 Difference (%) -2.414 -0.33 Acceptance C.titeria (t) 57.5% $7.5%

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4.5 Heactivity Coefficients at Power

a. Purrese The pulpise of this test is to measiire the taperawns and power dogpler coefficients of reactivity at power. Inis informat*.on is then used to amours that 7%ch. Spec. 3.1.7.1, which states that the moderator taperature coefficient shall not be positive at power levels above 95% of rated power, is satis'ied.

b. %st Metbal For measurirg the temperature coefficient of reactivity the average DC taperaturu was decreased and then increaami,by about 3 Degrees F. The reactivity aswiated with each taperature change was cbtained frun the change in controlling red group position, and the values for the coefficient were calculated.

For measur the power doppler coefficient of reactivity, data was extra frun the fast insert / withdrawal sequences.

Differential ocotrolling red worth measurments were also detemined using the fast insert / withdrawal methcd.

c. Test Results At 100% D, taperature and power doppler coefficient measurments were performed. The moderator ture coefficient measured at 100% FP was -3.74 . This verifies that the moderator ta perature coefficient is negative above 95% FP.

The measured power dcypler coefficient at 100% FP was -7.59 pcarVlFP and the measured fuel dc5pler coefficient was -1.10

, pcrV'F. This meets theCacceptance criteria of being nere i

negative than -0.9 pcarV F.

d. Conclusions The naanwed moderator taperature coefficient (MIC) results

! indicate that the MIC will be negative above 95% F.P.

i The naanwed fuel do[pler ocefficient CFVC) results meet the requirement that the FDC be more negattve than -0.9 pcstV0F.

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GPU Nuclear Corporation a Nuci ar  :::: = es<e Middletown. Pennsylvania 17057 717 944 7621 TELEX 84 2386 Writer"s Direct Dial Number:

C311-88-2163 Deccnber 5,1988 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Dear Sir:

Thrco Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 Cycle 7 Startup Report Enclosed is the Startup Report for TMI-1 Cycle 7 operation. The testing addressed by this report was completed and approved as of August 30, 1988. The report satisfies Technical Specification 6.9.1.A. No NRC response to this letter is necessary or requested.

This is being submitted one week late as discussed with R. Conto, TMI-1 Sci..or NRC Resident Inspector.

Sincerely, H. D. H kill Vice President & Director, TMI-l tIDH/MRK Enclosure ec: J. Stolz R. Hernan R. Conte W. Russell

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