ML20043G248

TMI-1 Cycle 8 Startup Rept. W/900612 Ltr

ML20043G248
Person / Time
Site:	Three Mile Island
Issue date:	04/30/1990
From:	Hukill H GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
C311-90-2076, NUDOCS 9006200091
Download: ML20043G248 (38)

Text

. .. _ . . _ _ _ - .._. ._ _ ._

!. p +,

i . .) .

t s

GPU Nuclear Corporation -

Q f Post Office Box 480 Route 441 South l

Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 June.12, 1990 Writer's Direct Dial Number:

.. C311- 90-2 07 6 y ..

U.S.-Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Dear Sir:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)  :

Operating License No. DPR-50 j

.,; Docket No.'50-289 Cycle 8 Startup-Report EnclosedLis GPUN's Startup Report for TMI-1 Cycle 8 operation.

Initial criticality for Cycle 8 was achieved at 0543' hours on i

.-March 3, 1990. Testing addressed by this report was completed '!

and approved as of April 12, 1990. This report _is being submitted in accordance with TMI-1 Technical Specification 6.9;1.A as discussed with NRC Project Manager R. W. Hernan on  ;

'May 14;.1990. No NRC response to this letter is necessary or '

requested. j l

Sincerely, d

. k 11 Vice President & Director,TMI-1 HDH/MRK.

l Enclosure cc: J. Stolz R. Hernan

"; ' F.' Young

-T. Martin

~

L OF h$fDhC P

n

. . GPU Nuclear Corporation is a subsidiary of General Public Utihties Corporation 763d [

i o, GPU Nuclear-l 1

TMI-1 CYCLE 8 STARTUP REPORT TMI-1 Nuclear Engineering I Apri1, 1990 O

a.,I. 0 ,v. ,m,--, ~ . , ,- - - -- , - - - - - - , . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ , _ _ _ _ _ -

'f: ..

.%> -4

-t..

b 4

TABLE OF C0tfl'ENIS PAGE 1.0 CERE PSUUMANCE - ME' A SURDENTS AT ZDO IOGR - SG99JW . . . . . . . 1 2.O CEE PERIMMANCE - MEASURDENIS AT POWER - SG9RRY . . . . . . . . . . . . 4 3.0 CORE PEREGMANCE - MEASURDENIS AT ZDO POWER . . . . . . . . . . . . . . . . . 6 3.1 Initial criticality ..................................... 6 4

3.2 Nuclear Instrumentation Overlap . . . . . . . . . . . . . . . . . . . . . . . . 11  ?

3.3 Reactimeter Checkout .................................... 12 3.4 ARD Critical Boron Crs. maiwation . . . . . . . . . . . . . . . . . . . . . . . 13 3.5 Tenperature coefficient Measurements . . . . . . . . . . . . . . . . . . . 14 3.6 Cbntrol Rod Group Worth Measurements . . . . . . . . . . . . . . . . . . . . 16 3.7 Differential Boron Worth................................ 23 I

4. 0 - CNE PERPWMANCE - MEASURDGNIS AT POWER . . . . . . . . . . . . . . . . . . . . . 24 4.1 Nuclear Instnanentation Calibration at Power . . . . . . . . . . . 25

' 4. 2 '. Incore Detector 'Ibstirq . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 4.3 Power Imbalance Detector Correlation Test . . . . . . . . . . . . . . 28 4.4- Core Power Distribution Verification ................... 33 i 4.5 Reactivity Coefficients at Power ....................... 35 g

I

('

i h

i' ..-

1.0 03E PERPGMANCE - MEASUREMINTS AT ZEED ICWER - SLM4ARY ,

Q:re perfarinance maame were ocmducted during the Zero Power Test Pry. 1d11& began on Mart 212,1990 and ended on Mardt 4,1990. '1his

' 4 sectico 3 resents a samunary of the zero power maamwenants. In all cases, i the applicable test and Te& nical Specifications limits were net. A sunanary of zero power physics test results is included on Table 1-1,

a. ' Initial Criticality Initial criticality was adileved at 0543 on Mardi 3,1990. Reactor l conditions were 532 F and 2155 psig and control rod groups 1 through 6 were withdrawn to 100% while group 7 was positioned at '

73.17% withdrawn. Control Rod group 8 was positicmed at 25%

withdrawn. Criticality was adlieved by dahnrating the Reactor Coolant frun 2164 ppa to 1813 ppm. Initial criticality was adlieved in an orderly manner and the - tance criteria of 1807 100 PPM was met.

b. Nuclear Instrs==rttation Overlan At least one da mla overlap was measured between the source and intermediate range detectors as required by 'nudinical Specifications.

L i, c. ~ 33timeter Checkout 1.

An in-line functicmal &ack of the reactimeter using NI-3 was per.!atined after initial criticality. Reactivity calculated by the reactimeter was within 5% of the care reactivity deterinined frun doubling time naamwoments.

l d. All Rods out critical Baron ocn -rh ation

'1he maamwed all rods out critical baron uac=hation of 1846 ppnB was within the acceptance criteria of 1829 I 100 ppnB.

e. Temperature Coefficient Measurements

, 'Ihe maamwed tamparature coefficients of reactivity at 5320 F, zero

(= power were within the WJmoe criteria limits over the range of baron canoentrations ard rod positicris that the maanwements were l made.

i

f. Control Rod Group Worth Maa=E+ wits

'the measured results for control rod worths of groups 5, 6 and 7  ;

0 conducted at zero power -(532 F) using the baron / rod swap method l were in good agreement with predicted values. 'Ihe mav4== deviation i between maamwed and predicted worths was -7.17% whidi was for CBG-5 '

worth.

l l

l l

, s  ;

i ..

P

g. - DifAvial Bertm Worth 0

The -ired differential boren worth at 532 F was 11.84% nere f than the predicted value. . This is within the bounds of the PSAR and

, B&W supplied limits of 15%.

s

(

/

f s

9 an em

--.- v,--r-,-. , ,

I

• TABLE 1-1 9n=naw of Zero Power Physics ' Inst MHts Cvele 8 i Acceptance Mansan1x1 ,

Parameter Criteria _ Value Critical Baron 18071100 ppa 1813 ppn NI Overlap >l he >1.77 h e Sensible Heat N/A 1,73 x 10'7 g All Rods out Baron i Cbncent. ration 18291100 ppu 1846 ppo Tenparature Coefficient 1.77 pas / F 1.50 pan / 0F -

(1840 ppn) 4 pan /T t.

Moderator Ocefficient <9.0 pcm/0F 3.18 pan /OF 1 Rod Worths (532. ) GPS-7 2975 PO! i 10% 3114.5 PCM

(- Group 7 914 PCM i 15% 974 PCM

1. Group 6 934 PCM i 15% 926.5~PO(

Grayp 5 1127 PCM i 15% 1214 PCM Diff Baron Worth 7.931 PCM/ PPM i 15% 8.996 PO(/ PPM 1

1 x

t.

• 2.0 QCRE PERIUMhNCE - MEMERI!MENTS AT POWER - SLB00JtY

'Ihis section mamarizes the physics tests conducted with the reactor at power. 'hstirg was performed at power plateaus of 10, 30, 75, ard 95%

core thermal power. Operation in the power rarge began an Mar @ 4,1990.

a. Nuclaar L-tir dJitim Ok14hration at P--r l 'Ihe power range channels were calitrated as required during the startxp program based an power as determined by primary and secondary plant heat balanos. 'Ihese calibrations were required due to power level, baron ard/or control rod ocnfiguration charges during tasting.

b. Incom Detectar 'hstirg

'1%sts conducted on the incare detector system damnnstrated the.t all detectors except level 1 of strirq 49, level 2 of string 31, levels 1 ard 4 of string 4, are level 7 of string 32 were functionirg as expected. Synestrical detectar readings agreed within acceptable limits and the plant ocmputer applied the correct background, length and depleticm %.ction factors. . 'Ihe backup inocre ruoarders were operational above 80% FP as required by 'hchnical Specifications.

c. P-r Tw**1anom DeMnr ccrrelation Test

'Ihe results of the Axial Power Shapirg Rod (APSR) scans performed at 75% FP show that an acceptable incare versus out-of-core offset slope of >0.96 is obtained by using a gain factor of 3.684 in the power range scaled difference anplifiers. . 'Ihe naamned values of 4 mininum [NBR and mavimm linear heat rate for various axial core imbalances indicate that the Reactor Protection Trip Setpoints provide adequata protectico to the care. Imbalance calculatlans using the backup recorder provide a reliable alternative to ocaguter calculated values,

d. core Power Distribution Verification '

Core power distribution measurements were ocnducted at 75% and 97%

full power under steady state equilitrium xenon oorviitions for specified c hul rod configurations. 'Ihe maxinum naamtred and mavi== predicted radini and total peaking factors are all in good agreement. 'Ihe largest difference between the mav4== measured and maximum predicted value was -1.53% for total peaking at 97% FP.

'Ihis met acomptance criteria of <7.5%.

'Ihe results of the care power distribution measurements are given in Table 4.4-1. All quadrant power tilts ard axial ocre imhalances maamired durirq the power distribution tests were within the Technical Specificaticas and normal operational limits.

I, *

e. ]hastidd;y_Q3ath

'!he isotheonal tanparature coefficient measured at 95% FP was

-7.04 pczn/CF. 'Iha measured power doppler coefficient at 95% FP was -8.357 pam/% FP. All 'hu::hnical Specification and safety Analysis requirenants were met. .

i t

-I v

l

l r

.-

3.0 CORE PERIUMMME - MEMKEGMENTS AT ZHC PCHIR This section g.e.iits the detailed results ard evaluations of zero power physics testing. The zero power testirq program included initial criticality, ruclear instnmentation overlap, reactimeter checkout, all i

rods out critical baron concentration, temperature coefficient measuresnant, control red worths, ard differential beamn worth.

3.1 Initini critimlity i Initial criticality for Cycle 8 was achieved at 0543 on Mard 3, 1990. Reactor conditions were 532 0F and 2155 psig and m Lul ro:1 gamups 1 through 4 were previously withdrawn during the heatup to 5320F. The initial reactor coolant system (Ras) boaton concentration was 2164 ppn.

The approach to criticality began by withdrawing wibul rod group 8 to 25% withdrawn, wikul rod groups 5 and 6 to 100% withdrawn, and l positioning group 7 at 85% withdrawn. Criticality was W y atly '

adieved by debaratirq the reactor coolant system to a baran concentration of 1813 ppn. The kwidure used in the approad to criticality is outlined below in two basic staps:

Step 1 control Rod Withdrawal Group 8 25% withdrawn ,

Group 5 100% withdrawn l Group 6- 1004 withdrawn l Group 7 85% withdrawn Step 2 Deborate using a feed and bleed flow rate of 70 gpn until l the inverse count rate is at approximately 0.3.- At this point, stop debaration and inannse letdown flow to wwi = = (120 gpn) . This enhances mixing between the makeup tank and the reactor coolant system. Achieve initial criticality and position wikul' rod group 7 to control neutron flux as the reactor coolant system boton concentration readas equilibrium.

Throughout the approach to criticality, plots of inverse I maltiplication were maintained by two independent persons. 'No plots of inverse count rate (ICR) versus -n.cul red positicm were maintained during wikul rod withdrawal. 'No plots of 1CR.versus RCS baron concentration ard two plots of ICR versus gallons of domineralized water added were maintained during the dilution sequence. During each reactivity addition (boron dilution and-wikvl rod withdrawal), count rates were obtained fran each source range neutron detector channel.

r - , t.

• 1 l, e .

l t i

L.

l

'Ihe inverse count rate plots maintained during the WM to criticality are present*1 in Figures 3.1-1 throucA 3.1-3. As can be

' seen frun the plots, the response of the source rarge dannels 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 boren concentraticn and Figure 3.1-3 is the ICR '

plots versus gallons of domineralized water added to the BCS.

In sunnary, initial criticality was adlieved in an orderly manner.

'Ihe maaenwed critical baron u.=hiuation was within the acceptance critaria of 1807 i 100 P1H.

l l

1 l

i 9

. .o .

8 8 N

,r (>

0

/ /

.o

)

O Z o 3

+o ,

@6

-m lt z o

O /

p C4 [ @

NZ O ) o s1 O

ZO ~

o

D m o e>y <

hs~ y

/ ._a q @

8w Z

O O

., o

/

e e

n-l' )

m l" Q k" k k O O O

- 0 O d d n/L 1:3.-

i '

~

FIGURE 3.1-2 1.20 uM VS RCS BORON CONCENTRATION

_ g_3 1.00 M-2 m

A 0.80 T x

\ <m

( 0.s0 ,

x 0.40 NKy 0.20  %

0.00 2200.0 2150.0 2100.0 2050D 2000D 1950.0 1900.0 1850.0 RCS BORON CONCENTRATION d'n- sp p ..e- m. ... --- hm- -

m O

~

FIG.URE 3.1-3 1/M VS GALLONS OF WATER ADDED 1.20 g._ j -

M-2

^

1.00 -

b s 1

0.80 MN

1. N N i ,-

mee s\

A m.0 0.20 0.0 2000.0 4000.0 6000.0 - 8000.0 10000.0 12000.0 14000.0 GALLONS OF. WATER ADDED '

e

.n-. - + < - . . - , ,

-.p,, n -y , - - -

t .

3.2 Nuc1=av hunantation Overian

a. Purpose

' INK:hnical Specification 3.5.1.5 states that prior to operation-in the iMa*4 ate nuclear instrumentation (NI) range, at least one riarwia of overlap between the source range NI's and

/ the intermediate range NI's met be observed.

b. 'hmt Method

'Ib satisfy the above overlap requirements, occe power was increased until the iMa*iate range ctannela came on scale.

Detector signal response was then recorded for both the source range and intermediate range ciannels. 'Ihis was repeated until the source range high voltage cutoff value was reached,

c. Test Results

'Ihe results of the initial NI overlap data at 532 F 0 and 2155 psig have shown a 1.77 riarwia overlap between the source and intermarilate ranges._

d. Conclusions R

.i

'Ibe linearity, overlap and absolute output of the intermartiate and source rarge deta:: tors are within_ specifications ard performing satisfactorily. 'Ihere is at least' a one riarwla overlap between the source and iMWiate ranges, thus satisfying T.S. 3.5.1.5.

i l

.v ,

i.

I x

. _ _~ . _. _ _ _ _ __ . . _ . _ . - . _ . . . ._ _

4

.t. .

. 3.3 Dame +4== tar checkout a.- Purpose Reactivity calculaticms during the Cycle 8 test program were performed using the reactimeter. After initial criticality ard prior to the first physics maannseent, an enline functional check of the reactimeter was perfamed to verify its accuracy for use in the test program,

b. '1hst Method After initial criticality and nuclear instrumentation overlap. I was established, intermaritate range dannel NI-3 was connected  ;

to the reactimeter and the reactivity calculations were '

started. After steady state ocmditions were established, a small' amount of positive reactivity was insertad in the ocre by )

withdrawing ocmtrol rod group 7.~ Stop watches were used to <

maamne the doubling time of the neutron flux and the l reactivity was determined fr a the doubling time reactivity l curves. 'Ihe maamn1onent 'was for +26 and -50.5 pm. 'Ihe reactivities detamined fran doubling time measurements were I canpared with the reactivity calmlated by the reactimeter,

~

c. ' Dest Results l 'Ihe naamwid values were detamined to be satisfactory and showed that the reactimeter was ready for startup testing.

d. Conclusians i

An on-line functional dock of the reimatar was performed ]

after initial criticality. 'Ihe maamwed data shows that-the i l core reactivity maamnsd by the reactimeter was in good  ;

L

' agreement with the values obtained fr a neutron flux.dcubling H mmm. l

)

l l

l

_ , _ , _, 4,m .: s-.4--n 4-- A- - v+M+ * "' 8

• l"""* *O'4d ' * " ^ - - ' - - - ' - " '*"#"'-- -*#--- "

f 3.4 All Rods Out critical Baron CLied ratim 2

a. Purname

'Ihe all rods out critical baron canoentration maamirement was parfanned to obtain an accarate value for the excess reactivity loN in the 'IMI Unit 1 care and to provide a basis for the

'arification of emim1mtad reactivity worths. 'Ihis measurement les performed at system canditions of 532 0F and 2155 psig.

b.* Tag. Method

'Ihn Reacter Coolant System was barated to an all rods out i candition and steady state canditions were established,

c. Test Results l

'Ihe measured baron wehation with group 7 positioned at 1004WD was 1846 ppm.

d. Conclusians

'Ihe above results show that the maantred boten sciuhation of 1846 ppn is within the acceptance criteria of 1829 1 100 ypn.

l j:

L L

i l

i i

l-

=r .. ,

J 3.5 T----- ture Nficient IdamanA - --/J

]

a. Purpose 1

The moderator tarporature coefficient of reactivity'can be J positive, depending upcm the soluble baron ocreantration in the reactor coolant. Because of this possibility, the Technical Specifications state that the wrrierator tanparature coefficient shall not be positive while greater than 95% PP. The moderator.

tanparature coefficient cannot be measured directly, but it can be derived frcan the core tauperature coefficient and a known

f. fuel tanparature (isothermal Doppler) ocefficient. j

b. Test Method .l Steady state conditions were established by maintaining reactor 'l flux, reactor coolant pressure, tuttine bawiar pressure ard care average yture constant, with the reactor critical at-approximately 10 anpa on the irtav'amiiata range.

Muilibrium baron concentratice was established in the Reactor Coolant Systen, make-up tank and pressurizar to eliminate reactivity effects due to baron charges during the subsequent temperature swings. The reactimeter and reoceders were connected to monitor selected core parameters with the~ l roactivity value calculated by tho' r==+4==ter and the care

' average tamperature displayed on a two channel strip chart recorder. "

I once steady state canditions were established, a heatup rate' was started by closing the turbine bypass valves. After the  !

core' average temperature incranaarl by about 50 F care I L teature and flux were stabilized and the crocess was  ;

reversed by decreasing the core average twahlre by about 10 0F, After care tanparature and flux were stabilized, cxre touverauzre was returned to its initial value. Calculation of the tamperature coefficient fran the measured data was perfcomed by dividing the change in care reactivity by the:

cw.w.rv eding change in core tenperature over a specific time

l. period.

I- I L j l

u 1 1:

1. . ._ .. - _ _ . . _ - _ _ _ . - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -

. . . _ _ _ _ . . _ _ - ~ . . . _ . . _ . . _ . - _ _ _ . _ _

o .

. . l

.. j 1

1 c.. ' Inst Results

'Ihe results of the iwU.Es.a1 taperature coefficient . >

measurments are provided below. 'Ihe calculated values are j included for, ocuprison. . i In all cases the measured results ocupare favorably with the calculated values.

RCS MIASURED PREDICI'ED MEMURED REQUIRED BGEN TIC l'IC MIC MIC ,

(PPM) (PCN/DEC F) (PCM/DED F) - (PCN/DED F) - (POUDED F) i 1840 1.50 1.77 3.18 <9.0

d. Conclusicos

'Ihe measured values of the taperature coefficient of reactivity at 532 F,0 zero reactor power are within the acceptance criteria of 14.0 pczn/% of the predicted value.

An extrapolation of the moderator coefficient to 100%FP indicates that it is well within the limits of Tacimical Specifications 3.1.7.2.

1

(

i ,

i j

l i

l l

u 1

1 l

l l

l 1 .# '

1 3

l l

3.6 control Rod Group Worth Man =m=--tm i

a. PurDoes l

'Ihis section provides ocuparison between the rui.mlated-- and -l nanamed results for the control rod group worths. 'Ibe  !

location and function of each wikul rod group is shown in Figure 3.6-1. 'Ihe grouping of the control rods shown in Figure 3.6-1 will be usal throughout Cycle 8. Chimlated and -;

measured control red group reactivity worths for the normal  :

withdrawal sequence were determined at reactor canditions of zero power, 5320 F and 2155 psi. 'Ihe measured results were obtained using results of reactivity and group position from the strip chart recorders. .

b. 'Itst Method control rod group reactivity worth measurements were performed  !

at zero power, 532 F using the baron / rod swap method. Both the differential and integral reactivity worths of m akel rod groups 5, 6, and 7 were determined.

'Iha baron swap method consisted of establishing a debaration  ;

rate in the reactar coolant system and ocupensating for the '

reactivity changes by inserting the control rod groups in - '

incremental steps. 'Ihe reactivity changes that occurred during the maaawaments were calculated by the reactimeter and differential rod worths were obtained from the measured r reactivity worth versus the change in rod group position. 'Iha differential rod worths of each group were then summed to obtain the integral rod group worths.

c. ' Inst Results control red group reactivity worths were maaawed at zero power, 5320F caniitions. The baron / rod swap method was used to determine differential and integral rod worths for O. shul rod group 5 - 7 from 100% to 0% withdrawn.

'Ihe integral reactivity worths for control rod groups 5 through 7 are presented in Figures 3.6-2 through 3.6-4.

y . .- . .

'1hmes curves were obtained bf ntegrating i the amasammd differential worth curves.

Table 3.6-1 ptwides a ocuparisen betws i the predicted and measured results for the rod worth anamurements. 'Ihm results show good agreenent between the measured and predicted rod gg worths. The maxima deviation between measured and predicted uns -7.17%, I

d. Chnclusims Differential and integral control rod grup reactivity worths were measured using the baron / rod swap method. 'Iha measured results at zero power, 532 0F indicate good agreement with the predicted group worths.

1 ,

FIGL5tt 3.6-1 c"tc22 8 comet nod GRxP Ic~xrIONS Fuel Transfer l Canal i A

3 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 6 "Y K 5 4 2 2 4 5 L 1 8 6 2 6 8 1 ti i 3 5 4 4 5 3 N l 7 8 7 8 7 O l 3 5 5 3 P l l 1 6 1 R I i Z

1 2 3 4 S 6 7 8 9 10 11 12 13 14 15

~

X Group Number Group No. of Rods Function 1 8 Safety 2 8 Safety 3 8 Safety 4 9 Sa fe ty .

12 5 Control 6 8 Control 7 8 Control 8 8 APSRs

n. . . . _ _ .

a FIGURE 3.6-2 INTEGRAL WORTH FOR CRG-5 1400.00 1200.00 s' f

1000.00 m

2 /

/

W /

/

v 800.00 p 600.00 /

'Y l W 400.00 /

200.00 /

/

0.00 - -

0.0 MD 40.0 60.0 80.0 100.0 120.0 CRG-5 POSITION (%WD)

Total worth - 1214.0 PCM

, t FIGURE 3.6-3 INTEGRAL WORTH FOR CRG-6 1000.00 j ,

800.00

- /

8 600.00 ,

/

0 ? /

E O

400.00

/

/

u /

200.00 /

/

0.00-0.0 20.0 40.0 60.0 80.0 100.0 120.0 CRG-6 POSITION (%WD)

Total Worth = 926.5 PCM t

. - . - . _ =

. a

, 3 FIGURE 3.6-4 INTEGRAL WORTH FOR CRG-7 1200.00 10 T.00 j

m [

2 800.00 /

O 7

• 6 /

600.00 /

5 /

400.00 l

200.00

/

O.00-D0 40.0 60.0 80.0 'W O 120.0 CRG-7 POSITION (%WD)

Totol Worth = 974 PCM 9

[s s .. .

'.e TABLE 3.6-1 03FARISCW OF PRIDICTED VS MEASURID RCD MRIHS MEASURED FREDICTED PIRCENT CBG. WCRIH WRIH DIFFIRINCE

h. (PCM) (PCM) (4) 5 1214 1127 f 15% -7.17 I 6 926.5 934 i 15% 0.81 7 974 914 i 15% -6.16 5-7 3114.5 2975 I 10% -4.48 I

1 O

s .

3.7 Differer*4=1 % Worth

a. Purname 1

soluble polacr; in the fann of dimanived boric acid is added to )

the moderator to provide a&iitional reactivity wiuul beyond i that available frcan the control rods. The primary function of the soluble poison w,Lvl systen is to ocotrol the excess ,

reactivity of the fuel thruughout each ocre life cycle. The i differential reactivity worth of the boric acid was naasured during the zero power test. .

t

b. Dest Method Measurements of the differential borem worth at 5324 were perfczned in conjuncticm with the cuikel rod worth nonsurements. The wiuvl rods worths were =mamwed by the boat swap technique in which a deboration rate was established '

and the ocntrol rods were inearted to cxmpensate for the changing core reactivity. The reactimeter was used to pnwida a contirmous reactivity calculation throughout the -

measurement. The differential boren worth was then dotarnined  !

by sunsairs;r the incremental reactivity values measured during the rod worth measurshants over a known baron ocncentraticm range. The average differential baron worth is the measured  ;

ctange in reactivity divided by the change in baron conoontration.

c. 'nnst Results Measurements of the soluble baron diffarential worth were cxmpleted at the zero power candition of 5324. The naamwod baron worth was 8.996 pan /ppnB at an aW boren >

canoantration 1652.5 ppmB. The predicted value was 7.931 pan /ppnB i 15%.

d. Cunclusions The measured 0 results for the soluble poison differential worth at 532 F was within 15% of the predicted diffe.Wal worth.

P

.. .- - _- .__ - - - . . . _ . - - _ - - _ .- . - - . ~ _ . _ _ . . - _ . -

s . .

4.0 CIRE i>rxruesNCE - MEASURDB'TS AT IOGR 2

4his section presents the results of the ghysics mamaniments that were conducted with the reactor at power. Testing was ocnducted at power plateaus of 10%, 304, 75%, and 95% of 2568 megawatts core themal power, as detamined frtan primary and seccmdary heat balance measurunents.

Operation in the power range began on Mar d 4, 1990.

Periodic measurenants and calibrations were performed on the plant nuclear instrunentation during the escalation to full power. 'Ihe four power range detector mannels were calibrated based upon primary and secondary plant heat balance measurements. 'msting of the incare nuclear instrumentation was performed to ensure that all detectors were functioning properly and that the detector inputs were prummami correctly by the plant ocmputar.

Core axial imbalance detamined fra the incore instrumentation system was .

used to calibrate the cut of oore detector imbalance indication.

'Ihe major physics measurements performed during power escalation ocrsisted of determining the moderator and power Doppler coefficients of ruactivity and obtaining detailed radial and axial care power distribution measurenants for several core axial imbalances. Values of minisman DNBR and =vi== linear heat rate were monitored throughout the test progran to ensure that oore themal limita would not be me.

J l

l 4

1

. .. . j 4

f, 4.1 Nuct - li V datim Onluratim at Power  ;

a. Purpons The purpose of the Nuclear Instnanentaticri Calibratim at Power was to calibrate the power rarge nuclear instnmentation inilcatim to be no less than 2% N of the reactor thermal power as dotarmined by a heat baianos and to within 2.5% "

incore axial offset as detamined by the incare amitoring '

systan.

b. 'hst Method As required during power escalation, the top and botten linear i anplifier gaism were adjusted to maintain power range nuclear instruonntation indicaticri to be not less than 2% of the power  !

calculated by a heat balance.

i 1 hen directed by the ocritrolling rh for physics testing, the hiW flux trip bistable setpoint was adjusted. Themajor j mettirgs durirq power escalation are given belews Test Plateau Bistable Setpoint t FP %N 30 50 75 85 100 105.1

c. 'mst Results  :

An analysis of test results indicated that m anges in Reacter -

Coolant system battri arx1 xenon build @ ce burnout affected the power as observed by the nuclear instnmentation. 1his was as expected since the powar range nuclear instn mentation measures reactor neutron leakage whim is directly related to the above

& anges in system conditions. Ea& time that it was panammar'y to calibrate the power range nuclear instnanentation, the acceptance criteria of calikration to be no less than 2.0% N ,

of the heat balance power was met without any difficulty.

Also, and time it was nanaamar'f to calibrate the power range nuclear instrumentation, the f 2.5% axial offset critaria as datamined by the incore mmdtoring system was also met.

The high flux trip bistable was adjusted to 50, 85 and 105.1%

@ prior to escalation of power to 30, 75 and 1004 N, 1e-pinM yely.

4

d. Chnclusiens a

'Iha power range dannels were calibrated based cm heat balance power several times during the startup program. 'Ibene calikrations were required due to power level, bortm, and/cr control rod cxmfiguration danges during the program.

W_anca c:ritaria fer ruclear instrumentation onlibration at power were met in all instances.

l l

i

4.2 Incme Detectp:r heting

a. Purpose Self%= -nsui wMe (incare detectar systen) nonitor the care power density within the care and their autguts are monitored and r ---H by the plant ocaguter to pewide accurate readings of relative neutzen flux.

hsts coniacted on the incare detector system were performed

'to:

(1) Verify that the output frm and detectar and its response to increasing reactor power was as expected.

(2) Verify that the background, length and depletion cua.t. ions applied by the plant ocmputer are carrect.

(3) 7b maassww the degree of azinuthal symmetry of the neutron ,

flux.

b. hat Methnd The response of the incare detectors versus power level was determined and a conparison of the synsastrical detector outputs made at steady stata reactor power of 10, 30, 75, and 100%FP.

Using the carrected SPND maps, calculations were performed to determine the detector current to average detmetor current t l

values per assembly for and incare detecter versus avini positicos. Any detmetor levels whi& were dotarmined to have

, failed were deleted fr a scan or substituted.

t l

At 75% FP, SP-1301-5.3, Ircare Neutrun Detectors-Monthly Check,

! was performed to calibrate the back-up reoarders to its incare l depletion value.

c. Conclusians Incore detector testing during power escalation demonstrated

! that all detectors except level 1 of strity 49, level 2 of string 31, levels 1 and 4 of strirq 4, and level 7 of string 32 were functioning as expected. Synnetrical detector readings ,

agreed within acomptable limits and the ocmputar applied carrection factors are accurate. The har%@ incare recorders were calibrated at 75% FP and operational above 80% FP as required by the Te &nical Specifications.

. - . _ . . - . . - - - - - - - - = - . - - - ~ - - - - - - - - - - - - . - -

-s . ..

l i

4.3 Power T=halancz Detectnr Chrrelation %st l

a. Purpose l

2 e Power Imbalance Detector correlation Test has four -

object 8ves:

1.  % determine the relationship between the indiosted out-of-core power distribution and the actual incare power distribution, t

2.  % demonstrata axial = L ul using the Axial Power Shaping -

Rods (APSR's) .

3.  % verify the MM ard aoouracy of being imbalance calculations as dcne in AP 1203-7, " Hand Calculation for Quadrant Power Tilt and Core Power Tmhalance."

i 4.  % determine the oars ==?4 == linear heat rata and mininum .

INBR at various power imbalances.

b. hst Method his test was conducted at 75% FP to determine the relationship between the core axial imbalance as indicated by the incare detectors and the out-of-core detectors. Based upon this correlation, it oculd be verified that the mininum DNBR and '

maximum linear heat rata limits would not be am by operating within the flux / delta flux / flow envelope set in the Reactor Protection System.

t As CsG-8 was moved to establish the various imhalarmes, the integrated wavl system automatically ocupensated for reactivity changes by repositioning CRG-7 to maintain a constant power level.

c. Test Results '

me relationship between the ICD and OCD offset was determined at 75% FP by performing an imbalance scan with the APSR's. Se average slope measured on the four out-of-core detectors was 1.043. S e lowest slope was 0.987 for NI-7. Se scaled difference amplifier gain was 3.684.

s .. ..

A ocuparison of the incare detector (ICD) offset varuus the out-of-ocre (OCD) detector offset obtained for eacts NI channel is shown in Table 4.3-1.

Core power distribution measurements were taken in conjunction with the most positive and most negative imbalances at 75% PP and the values of minimum DGIR ard worst case MIllR and ocupared to the acceptance criteria.

'!he worst once values of minimum DHHR and maxian linear heat

' rata dotaruined at 754 PP are listed in Table 4.3-2.

'Ihe worst case INBR ratio was greater than the mininum limit and the ==v4== value of linear heat rata was less than the fuel melt limit of 20.5 kw/ft after ELplatics) to 105.1 PP.

These results show that 'Iwchnical Specificatics) limits have been met.

My offset calculations using AP 1203-7 agree with the otsputer calculated c2:fset. Table 4.3-3 lists the ocmputer calculated offset es s.41 as offsets obtained using the inacre detecter backtp ros.aers.

d. Canclusians Backup imbalance calculations parfarined in accordance with AP 1203-7 pcovide an acceptable alternata method to ocmputar calculated values of imbalance. A difference anplifier K factor of 3.684 will provide a slope greater than or equal to 0.96 when OCD offset is plotted versus ICE offset.

Mininum INBR and Mav4== Linear Heat Rate parameters were well within Technical Specifications limitations.

• P s

'*- o ,

TABLE 4.3-1 IHOCRE OFFSI!T VS OUMF-CEE Of7SEP INOME OUDCF-00RE Of7EEP (%)

OFFSET

(%) NI-5 NI-6 NI-7 NI-8 0.24 -0.99 0.21 0.51 0.42 6.78 5.01 6.16 5.67 6.05 3.25 2.04 3.26 3.14 3.26

-2.83 -4.18 -3.01 -2.55 -2.54

-6.17 -8.35 -7.23 -6.41 -6.41

-10.57 -13.28 -12.25 -10.93 -11.01

-20.92 -24.64 -23.69 -21.20 -21.50

-20.04 -23.25 -22.22 -19.85 -20.24 i

l l

I l

l l

a .e a P

a l

l 1

TABLE 4.3-2 WORST CASE INBR AND DR '

l INEAIANCE MINDEN Df5AICIATED W m ST CASE D R IXIRAP. MAX. D R (t) InnR 1ENER (m/Pr) (m /PT)

-15.76 3.898 2.71 10.54 14.12 5.04 4.086 2.67 9.60 12.78 t

5 L

i -.

l 1 . . .

o ++ o

'DJILE 4.3-3 FUIL INO3tE OF7SEP VS BIGUP RB: GOER OI?SI!T FULL INO3tE BIGUP RICOBOER OFFSET OF75I!T (4) (t) 0.24 0.711 6.78 -

6.039 .

-20.92 -16.90 i

1 u r o e, p s

.

• o 4.4 Chru Pouar Distrilmticm Verificatiert

a. Purpose To measure the core power distributions at 75 and 97 WAt full power to verify that the core axial imbalanoe, quadrant power tilt, mavina linear heat rate and minimum INBR do not  ;

auceed their specified limits. Also, to ocupare the measured '

and predicted pcmar distributions. ,

b. 'hst Method

• Core power distributicri measurements were performed at 75 and 97% full power, under steady state aanditions, for specified ocritrol rod configurations. 'Ib ptwide the best ocupariscri between measured and predicted results, Ugc dimensional equilikritan mancri conditions were established. Data collected for the naamarenants ccrisisted of detailed pcuer distrikuticsi information at 364 core locations frcan the inocru detmetor system and the worst ones ocre thennal canditions were  !

calculated using this data. 'Ihe measured data was cxmpu%d with calculated results,

c. Test Results A sunmary of the cases stuiled in this 1.ra.i. is given in Table 4.4-1 whi& gives the core powar level, whel rod pattarn, cycle burrup, bertri cxvicentration, axial imbalance, maximum quadrant tilt, minimum DNBR, maximum IHR and power peaking data for and measurement. 'the hic #iest Worst case MLHR was 11.32 at 97% FP whi& is well balcw the limit of 20.5 kw/ft. 'Ihm lowest minimum DER value was 3.347 at 100% N whim is well above l

the limit.

1

'Ihe quadrant power tilt and axial imbalance values nananed were all within the allowable limits. Table 4.4-1 also gives a ocuparison between the maxinum calculated and predicted radial and total peaks for an eighth oars power distribution.

d. Ckmelusicris l

Cbre power distrikuticri maammments were conducted at 75% and 97% full power. Ocuparison of nammngd arti predicted results show good agreement. 'Ihe largest difference between the maximum measured and maximum predicted value was -1.53% for total peaking at 97% R. 'Ihis met the acomptance critaria of

< 7.5%.

'Iha measured values of IEBR arti MLHR were all within the allowable limits. All quadrant power tilts and axial care imbalances manen"ed during the power distribution test were I within the To&nical specifications arri nonnal operational limits.

)

o..p -

i

(- e o e a

Tall!E 4.4-1 )

CXRE RMR DISTRIBUTICH RESULTS  ;

KMER PIATEAU 75%N 97%N_

DATE 03-18-90 04-04-90  ;

Actual Power (4FP) 74.4 97.01 l C3C 1-6 ( 49 0 ) 100 100 cc 7 (460) 88.8 89.9 OG 8 ( 49 0 ) 31.5 29.0 cycle Burrsp '

(EPPD) 2.35 15.95 Baron Ocmc. (PIM) 1471 1301 Imbalance (%) 0.60 -2.58 Maxinum Tilt (%) 0.50 O.38 M201R 4.390 3.347 Worst Omme MIliR (10f/FT) 8.84 11.32 Mmvim m Radial Peek Mansured 1.297 1.300 Predicted 1.300 1.300 l Difference (%) -0.231 0.00 WJ.noe Criteria (%) 55% $5%

Maxilaam 'D:stal Peak Measured 1.542 1.502 +

Predicted 1.550 1.525 Difference (%) -0.519 -1.53 Acosptance criteria (%) 57.5% $7.5%

i s

cJ?! .

4 O

. r' l

4.5 '

1vity wficients at R--

a. Bgmes I i

'Ihm purpose of this test is to naasure the temperature coefficient of reactivity ard power doppler ocefficient of reactivity at power. 'Ihis infomation is then used to assurm '

that 'IWit. Spec. 3.1.7.1, which states that the noderator taaparature coefficient shall not be positive at power levels above 95% of rated power, is satisfied.

b. 'hst Method Fbr measuring the tanparature coefficient of reactivity, the '

average RCS tamparature is decreased and then increased by l about 5 Degrees F. 'Ihm reactivity associated with eacti j tanparature ctage is obtained fran the ctiange in controlling rod grotp position, and the values for the coefficient are  !

omlaulated. l Pbr maantring the power doppler coefficient of reactivity, data I is extracted frun the fast insert / withdrawal esquences.

Differential controlling rod worth measurements are also determined using the fast insert / withdrawal method.

c. 'hst Results At 95% FP, tanparature and power dopplar coefficient {

measurements were perfomed. 'Ihe noderaior temperaturw ocefficient measured at 95% FP was -5.54 pan /0F. 'Ihis verifies that the mrderator tanparature coefficient is negative )

above 95% FP. I

'Ihe naasured power dcypler coefficient at 95% FP was -8.357  ;

pan /%FP and the measured fuel doppler coefficient was -1.229 pom/0F. 'Ihis meets the acomptance criteria of being more r negative than -0.9 pcn/0F. r

d. Ocnclusions

'Ihe measured modarator tanparature ocefficient (MIC) results indicate that the MIC will be negative above 05% F.P. >

'Ihe mamervid fuel doppler coefficient (FDC) results meet the requiranent that the FDC be more negative than -0.9 pom/0F.

-- _ . . . _ _ _ . _ . - _ . . _ _ . . _ _ _ _ _ _ _ _ _ _ . - - - . _ _ _ _ _ . _ _ _ _ _ _ _ __