ML20092A522

TMI-1 Cycle 9 Startup Rept

ML20092A522
Person / Time
Site:	Three Mile Island
Issue date:	02/04/1992
From:	Broughton T GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
C311-92-2004, NUDOCS 9202100156
Download: ML20092A522 (1)

Text

e t .

I GPU Nuclear Corporation f Q gf Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057 0191 717 944 7621 TELEX 84 2386 Writer's Direct Dial Numtier:

(717) 948-8005 Febrtary 4,1992 C311-92-2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Three Mile Island Nuclear Station, Unit 1 (THI-1)

Operating License No. DPR-50 Docket No. 50-289 Cycle 9 Startup Report Enclosed is the GPU Nuclear Startup Report for TMI-l Cycle 9 operation.

Initial criticality for Cycle 9 was achieved at 1222 hours0.0141 days <br />0.339 hours <br />0.00202 weeks <br />4.64971e-4 months <br /> on November 14, 1991. Testing addressed by this report was completed and approved as of November 20, 1991. This report is being submitted in accordance with TMI-1 Technical Specification 6.9.1.A. No NRC response to this letter is necessary or requested.

1 Sincerely, h$

T. G. Br ghton Ifr Vice President and Director, THI-l MRK l Enclosure cc: Region I Administrator TMI-1 Senior Project Manager THI Senior Resident Inspector 7('Ot ? " ,2023o0336 ,20204

/ '{i (i

PDR ADOCK 05000289 P PDR GPU Nuclear Co poration is a subsidwy of General Pubhc Ulmes Corporat.on

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

.-.-s i

I 4

. GPU . N t2 c l e a r +

4 h

TMI-1 e

CYCLE 9 i

-STARTUP IAEPoltT -i l-l' TMI Nt2 clear ' E ngincori ng-i-

I--

l' l- ..

i -December, 1 9 9 1-e- ~ ,- ,w -

r -m+ ,

i. :

TABLE OF CONTENTS PAGE'

- l '. 0 CORE PERFOR!iANCE - MEASURE!!ENTS AT EERO POWER - SUMfiARY . . . . . . . 1 4

2.0 I CORE PERFORMANCE - MEASUREMENTS AT POWER -

SUMMARY

. . . . . . . . . . . . 3 3.0 CORE PERFORMANCE MEASUREMENTS AT ZERO POWER ........ ........ 4 3,1 Initial criticality .................................-.... 4 3.2 Nuclear Instrumentation Overlap . ....................... 8 >

3.3 Reactimeter Checkout .................................... 9

-3.4 ARO Critical Boron Concentration ........................ 10 3.5 -Temperature Coefficient Measurements .................... 11 3.6 Control Rod' Group Worth Measurements . . . . . . . . . . . . . . . . . . . . 12 s

3. 7. Differential Doron Worth................................. 18

- 4.0 CORE PERFORMANCE --MEASUREMENTS AT POWER ...................... 19 4.1 -Nuclear Instrumentation Calibration at Power ............ 20 4,2 -I ncore Det ec t o r Te st i ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 4.3 Power-Imbalance Detector Correlation Test . . . . . . . . . . . . . . . 22

'4.4 Core = Power Distribution Verification .................... 27 4.5 Reactivity Coefficients at Powor......................... 29 l g-e i

Y i

1s---- - M+ q

• W **

. . _ . _ - . . . m.-. .. _ - . . . - . _ - _ _ _ . _ - _ ... _ _ _ _ _ m._-. m . _ _ _ . _ . . m. . .

e: N 1.0 CORE PERFORMANCE - MEASUREMENTS-AT ZERO POWER -

SUMMARY

Core performance measurements were conducted during the Zero power Test Program which bogan on November- 14,-1991 and ended on November 15, 1991.

This_section presents a_oummary of the zero power measurements. In all cases, the' applicable test and Technical Specifications limits were met.

A-summary of.zero power physics test results appear as Table 1-1. ,

. a, Initial criticality

- Initial- criticality was - achieved at - 1222 on November 14, 1991.

Reactor conditions were 532*F and 2155 poig. Control rod groups I through 6 were withdrawn ' to - 100%; group . 7. was positioned at 91%

withdrawn; group a was positioned at 25% withdrawn. Criticality was

. achieved -by~ deborating the Reactor Coolant from 2428 ppm to 2162 ppm. Initial criticality was achieved in an orderly manner and the acceptance criteria of 2217 1 100 PPM was met, b.- Nuclear Instrumentation Overlay At - least - one decade overlap was measured between-the source and, intermediate- _ range detectors as required by Technical Specifications,

c. Reactimeter Checkout

_An on-line - functional check of the - reactimeter using NI-3 was performed after initial criticality. ' Reactivity calculated by the reactimeter was within 5% of the core roactivity determined _ f rom doubling time measurements.-

d,. All Rods Out Critical Boron Concentration The measured all rods out critical boron concentration of 2157 ppma was-within the acceptance criteria of 2221 1 100 ppmB.

l

- e. Temperature Coefficient Measurements.

- The measured temperature-coefficient of reactivity at 532*F, zero- i power was within the acceptance criteria limit.

f. Control Rod Group Worth Measurements The measured results-,for control rod-worths of groups.-5, 6_and ~7 conducted at zero power (532*F) using the boron / rod swap method were in good - agreement with predicted . values. The maximum deviation between measured and predicted worths was -2,4% which was for CRG-7

- worth.

g. Differential Boron Worth The measured: differential boron worth at 532*F was 9.0% more than-the predicted value. This is within the bounds of the FSAR and B&W~

supplied limits of 115%.

I

~

l i

j_

i.

r - -.. e . ,~,--m,- ,

. s TABLE 1-1 Sununary of Zero Power Physics Test Roeulta  ;

Cycle 9 Acceptance hs sured.

Parameter _Criterla value Devintion critical'Doron -2217 i 100 ppm 2162 ppm -55 ppm NI Overlap >l decade >1.S7 decade ---

Sensible Ileat N/A 8.1 x 10 amps --

All Rods Out Boron-Concentration 2221 1 100 ppm 2157 ppm -64 ppm Temperature Coefficient. 2.17 pcm/*F 3.11 pcm/*F +0.94 pcm/*F (2149-ppm) i 4 pcm/*F Moderator Coefficient <9.0 pcm/*F 4.82 pcm/*F ---

Integral Rod Worths-

-(532*F) GPS-7 2791 pcm i 101, 2839.8 pcm -1.72%

i; Group 7' B68 pcm i 15% 889.3 pcm -2.4%

l Group 6 808-pcm i 15% 816.5 pcm -1.04%

Group 5. 1215 pcm i 15% 1134 pcm ' -1,67%

Diff Boron Worth 7.098 pcm/ ppm 2 15% 7.804 pcm/ ppm - 9 . 01, i

(1951 ppu)

(

?

H

.= i e I

'2.0 CODE PERFORNANCE a MEASUREMENTS AT POWER -

SUMMARY

.This section summarizes the phynien toute conducted with the reactor at power. -Teoting was performed at power plateaus of approximately 10, 30, 75, and-100% core thermal power. Operation in the power range began on November 15, 1991.

Four Westinghouse lead' test f uel assembles (LTA) were rionitored during the ,

startup program to ensure that they - were not the limiting (hottest) assemblien in the core with respect to radial power distribution power peaking. Analyses predict that.the LTA will maintain at least a 3% hot pin

-relative power peaking margin.

a. Nuclear Instrumentation Calibration at Power >

The power range channele were calibrated as requirod during the startup ,

program based on power as determined by primary and secondary plant heat balance. Theno calibrations were required due to power level, baron and/or control rod configuration changea during testing.

b. Incore'Dutoctor Testing Tests conducted-on the incore detector system demonstrated that all-detectore were functioning acceptably. Symmetrical detector readings

• agreed within acceptable limits-and the plant computer applied the correct background, length and depletion correction factors. The backup incore recorders were operational above 80% FP as required by Technical Specifications.

c. Power Imbalance Detector Correlation Test The results of the Axial Power Shaping Rod (APSR) movementa performed at 75% FP show that an acceptable incoro versue out-of-core offoot slope of >0.96 -la obtained by using a gain f actor of 3.684; in the power range scaled differenco-amplifiern. The measured values of minimum DNBR and' maximum linear heat rate for various axial core imbalances indicate that tae-Reactor Protection Trip Setpointa provide adequate protection. to the core. Imbalance calculations using the backup recorder provide a reliable alternative to computer calculated values,

d. Core Power Distribution Verification Core power distribution measurements were conducted ~at approximately

75% and 100% full power under steady stato equilibrium xenon conditions for specified control rod configurations. The maximum measured and maximum predicted radial and total peaking factors are all in good agreement. The largest difference between the maximum measured and maximum predicted value was +2.8% for radial. peaking at 99.98% FP.

This met acceptance criteria of <5.0%.

The results of the core power distribution measurementa are given in Table 4. 4-1. All quadrant power tilta and axial core imbalances

• measured during the power diatribution testo were within the Technical

-Specification and normal operational'11mits, e.- Reactivity Coefficients at Power The isothermal temperature coef ficient measured at approximately .99% FP was -5.91 pcm/*F. The measured power doppler coefficient at approximately 99% FP was -8.78 pcm/t FP. All Technical Specification and Safety Analysin requirements were met.

1 3.0 CORE PERFORMANCE - MEASUREMENTS AT ZERO POWER This section presents the detailed result s and evaluations of zero power physics testing. The zero power testing program included initial criticality, nuclear instrumentation overlap, reactineter checkout, all rods out critical boron concentration, temperature coef ficient measurement, control rod worths, and differential boron worth.

3.1 Initial Criticality Initial criticality for Cycle 9 was achieved at 1222 on November 14, 1991. Reactor conditions were 532*F and 2155 poig.

Control rod groups 1 through 4 were withdrawn during the heatup to 532*F. The initial reactor coolant system (RCS) boron concentration was 2428 ppm.

The approach to criticality began by withdrawing control rod group 8 to 25% withdrawn, control rod groups 5 and 6 to 100% withdrawn, and positioning group 7 at 85% withdrawn. Criticality was subsequently achieved by deborating the reactor coolant system to a boron concentration of 2162 ppm. The procedure used in the approach to criticality is outlined below in two basic steps:

Step 1 Control Rod Withdrawal Group 8 251 withdrawn Group 5 100% withdrawn Group 6 100% withdrawn Group 7 85% withdrawn Step 2 Deborate using a feed and bleed flow rate of 50 gpm until the inverse count rate is at approximately 0.3. At this point, stop deboration and increase letdown flow to maximum (120 gpm). This enhances mixing between t he.

makeup tank and the reactor coolant system. Achievo initial criticality and position control rod group 7 to control neutron flux as the reactor coolant system boron concentration reaches equilibrium.

Throughout the approach to criticality, plots of inverse multiplication were maintained by two independent persons. Count rates were obtained from each source range neutron detector channel. One person used NI-1 and 11, the other used NI-2 and 12. Four plots of inverse count rate (ICR) versus control rod position were maintained during control rod withdrawal. Four plots of ICR versus RCS boron concentration and four plots of ICR versus gallons of demineralized water added were maintained during the dilution sequence.

The inverse count rate plots maintained during the approach to criticality are presented in Figures 3.1-1 through 3.1-3. As can be seen from 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 concentration and Figure 3.1-3 is the ICR plots versus gallons of demineralized water added to the RCS. (

In summary, initial criticality was achieved in an orderly manner. The measured critical boron concentration was within the acceptance criteria of 2217 100 PPM.

o m

C O

o. r- N -

m r- e a .x L T z2 y e . .

f C o m

-o O ..

e o.

a

.N- O u

W o 0 O 4 N 8 r,- CL.

ve

/ o g

o

' ~ ~

6 cd C @ .5 -

21 o

] @

8 a.

9d w

~3 r j

m

' y O

e C

T o O

>- i o O o 3 b

T* C O $

o m N

N o l

O r-m -

m y ,.

C l O o o_

l lE m

~ O l T* .

u L 9 r- m. m 9 N o O r- o o o o o Ou i

O N

-s-

.. e m

C O

D. r C m e r,-

CQ v,-

1 O m z z l

• Q Q v - l W

L._

l H

C s q) n O  !

c  ? a O /  : mE mQ N

r-s

/

l. e/

• 8 a_

a co C .

mO

'Z _

%* =5 u e a J>- he o) O c b~ O ,

~

N8c o

/

CO - . - =

I y~ n O

/[. R s a en W

y i

// aO a m o

E s

2

=

h-[

2 s

i-  % r W .

D  ? N o e o o o o

i; l jl1 l 1Il i e

- s _

n o

p 1 2 s 1 1 e

d R i

N i

N e ~

~

I N =  !

d d

A i i r

e \ 4 )

s t N 1 a

o 0 d i k 7 n

a 3'N 4 s 1 u o

0 1 s h

.f N o

f 3 (

e x i' 4

1 0 d r e u

gn s s d

d i

F 4

A o

1 0

r g 4 n l t e i x

i a

a l

4 m .

1 '

1

o. W d n .

a G

3 i D g f i n

4 o t s

i 1

0 a .

2 s w v 4 l n

o t o

e l

u

~

1 a

M

/ M 1.

0 1

G d

d n

e

/ 4 t 1 1 1 0

a 2 1 8 6 4 2 0 o. p o

r 1 0 0 0 0 d p

e t

e S

l .

3.2 fJuclear Inst unient at ion Overlap

a. Put 1 tre Technical Wrecification 3.5.1.5 etates that prior to operation in the int ermediato nuclear instrumentation ( til ) rango, at least one decado of avorlap between the pource rango !J1's and the int er medi at e rango 141 ' o muut bo observed,

b. '! o ti t tiet hod

't " catisfy the abovo overlap requitements, coro power was irereased until the intermediate tange channele came on ocalo.

DetWt or oignal erponso wau then recorded ior bot h the ocurco range and i rit o rined i a t o rango channoln. This wan opoat ed until the maximum t;ourco rango value wap reachod.

c. Tent Itocults The reaults of the initial fil overlap dat.a at L32*F and 2155 pelg have shown a >1.57 decade overlap bet ween the nource and int armediate ranges.

d. Conclunionn Tho linearity, overlap and abnolut e out put of the intermediato and sourco rango detectors ar e within specificat.iono and perf orming catiofactorily. Thes e in at least a one ducado overlap between t ho t.ot > 'o and intermediato ranges, thus natintying T.S. 3.5.1.5.

i 3.3 _ Hoact imoter Checkout

a. y P pose Reactivity calculations during the cycle 9 Lost - program were performed using the reactimotor. After initial criticality and prior to the first physics measurement, an online functional check of the reactimeter was performed to verif y its accuracy for use in the test program. l

b. Test Hothod Af ter initial criticality and nuclear instrumentation overlap was established, intermodlate rango channul N1-3 was connected to the - ,

reactimeter and the reactivity calculations were started. After i steady state conditions were established, a small amount of

- positivo . reactivity was inserted in the core by withdrawing control rod group 7. Stop watches woro used to measure the ,

doubling time of the neut ron flux and the reactivity was  !

determined from the doubling time- reactivity curves. The l measurements were taken at 4 64.0 and -30.2 pcm. The reactivities determined f rom doubling time measurements were compared with the reactivity calculated by the reactimeter.  ;

i

c. Test Henults ,

The measured values were determined to be satisf actory and showed 3 that the reactimeter was ready for startup testing. .

t

d. Conclusions

- An on-line f unctional check of the reactimeter was perf ormed af t.or i initial criticality. The measured data shows that the core reactivity measured by the reactimeter wac in good agreement with _ e the values obtained from noutron flux doubling times.

T

+

f t

1 i

_ . . _ . . . _ . - - _ _ _ _ - - _ . - - .-______._____.._---....._m_

1 l

l 3.4 All Itodo out critical lloron conennt rat ion a, ,Puriono .

I The all rode out critical boron concentration measurement was porf ormed to obt.aln an accurate value f or t he excoon reactivity i loadod in the TH1 Unit I coro and to provido a bania for the t verification of calculated reactivity wort.he. This rneasurement was perf ormed at system condit ions of $32'F and 2155 poig,

b. Test Hot had i t

The }(eactor Coolant System wau borrted to an all rods out l condit-lon and steady stato conditions were established. l t

'c. Tont Itenu l t s The meauured boron e cens wC h group 7 positicried at 100%WD was 2157 ple.

d.- Conclusionn ,

The above results show that i ho anoanured baron conenntration of 21b7 ppm is within the accopt.6.nce etitoria of 2221 1 100 ppm.

l t

3 D

F J

l I

l l

s

)

1

<E-- -e , + , ~ ,, .-.,,,-.. - + . - . . . . -

..~,e......,,. . -._._.,,,,_.m. , __ . , , , . , . . . , , .~,m_,..__.,..,,_, , , , . , , . _ , . . , , , . , _ _ _ , . , , , , . , . , . ,

4 4

i 3.5 Temlwera*.ure Coe f f leiont Meanusomonto

a. Purlone The niodora tor temperature coefficient of reactivity can be positive, depending upon the soluble boron concentration in the ,

reactor coolant, becaupe of thin possibility, the Technical  !

Specifications etatu that tho niodorator temporat.ure coof ficient shall not bo positive while greater than 95% FP. The moderator ,'

temperature coef ficient cannot be moacured directly, but it can be derived from the inothermal temperaturo cootficient and a known fuel temperaturo (Dopplor) coefficient.

b. Test Hothod Steady utate conditions were establiehod by maintaining roactor flux, reactor coolant precaure, turbine header pressure and coro averago temperature conotant, wit h .the reactor critical at approximately 104' ampo on the intermodlato range. _ Equilibrium boron concentration was cotablichod in the Reactor Coolant Syst.om, mako-up tank and pressuritor to eliminato reactivity ef f ecto due to boron changoo during t he - pubcequent temput ature swings. The reactimotor and recorderu woro connected with the reactivit.y value and the Res average temperature displayed on a two channel atrip chart recorder.

5 Once stoady state condit. ions woro octablished, a heatup rate was start.ed by closing the turbino bypaan valvos. After the core ,

average temperaturu increaned by about 5'F core temperature and i flux were stabilized and the procoon was reverned by decreautng the core average temperature by about IO*F. Atter core  ;

temperaturo and flux were stabilized, core temperature was returned to its initial value. Calculation of the temperature coefficient from the measured data was performed by dividing the chango in core reactivity by the corresponding chango in RCS temperature.

c. Test Fosulto The resulta of the loothermal temperaturo co ef ficient measuremente are provided below. Tho predicted values are included for comparison.

In all canos the measured results compare f avorably with the predicted valuon.

RCS 11EASURED PREDICTED HEASURED REQUIRED DORON ITC ITC HTC HTC (PPM) JPcM/DEG F) (PCM/DEO F) (PCM/DEG F) (PCH/DEG F) 2149 3.11 2.17 4.82 <9.0

d. Conclusionn The moaoured valuon of the temperaturo coef ficient of reactivity  !

at 532*F, zero reactor power are within the acceptarse critoria of i 4.0 pcm/'F of the predicted value. An extrapolation of the Inodorator coef ficj ont to lOOi,FP indicated that it was well within the limits of Technical Specifications 3.1.7.2.

I 3.6 Cont rol Itod Group Wort h lieasus ementa

a. [yttysto This euction providou comparloon bet ween the calculated and ineanut ed result u ior the cunt rol a od group worLhe. The location and f unction of each control rod group in shown in Figure 3.6-J.

The grouping of the control rode uhown in Figure 3.6-1 will be u t:ed throughout cyclo 9. Calculated and meanured control rod group reactivit y wort ha f or the nor mal withdrawal nequence wore determined at react or conditioin of zoao powor, 532*F and 2155 pai. The sneasured resulto were oldained uping results of f' react ivity and group position f rom t ho st rip chart recordern.

b. Tout Hot had i

control rod group reactivity worth measuromonts woro perf ormed at. ,

zero power, $32*F using the boron / rod swap method. Doth tho ,

dilf erential and integral reactivity wortha of control rod groups ,

5, 6, and 7 were determined.  !

The baron / rod owap inothod connioto of optablinhing a deboration  ;

rate in the t oactor cool ant- s yn t em, then compenuating for the  !'

reactivity changen by innerting the control rod groups in incremontal steps.

I The roactivity changos that occurred during thu meanut ements were calculated by the renetimeter. Difforential rod wortha were obtained f rom the measured toactivity worth versus the change in rod group ponition. The dif f erontial rod worths of each group ,

were then summed to obtain the integral rod group wortha.  ;

c. Tent 14enul t n Control rod group reactivity wortha were measured at zero power,

$32*F conditions. The boron / rod awap method was used to determine dif f erential and integral rod wortha for control rod group 5 - 7 from 100s to On withdrawn.

The integral reactivity worths for control rod.groupo 5 through 7 are presented in Figuros 3.0-2 through 3.6-4.

Those curves woro. obtained by integrating t.ho nina pured ditforential worth curven.

Table 3.6-1 provideo a compartoon betwoon the predicted and moauured.results for the rod worth moanuremont a. The results show good agreement between the measured and predicted rod group i worths. Tho tuaximum deviation between measured and predicted wortha t'or a group was -2.4%.

d. Conclunions Differential and integral control rod group reactivity wortha were - measured using the boron / rod owap method. The measured -

reoults at zero power, 532*F indicato good agrooment with the predicted group wortha.

i r!

l 1

I

Figuru 3.6-1 Control Rod Locations and Cf60s Designations for THI l cycle 9 ,

[

• I Fuel Transfer Canal  ;

A -

l 8 1 6 1 C 3 5 5 3 D 1 8 1 8 1 E 3 5 4 4 5 3 F 1 8 6 2 6 8 1 0 5 4 2 2 4 5 H W- 6 1 2 4 2 1 6 ,

-Y

$ 4 2 2 4 5 K

L 1 8 6 2 6 4 1  ;

M i 3 5 4 4 5 3 N l 1 8 1 8 1 0 l 3 5 5l 3 ,

P l l 1 6 1 R i i I ,

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 l X Group Numter

l. Group No. of Rods Function 1 8 Safety  ;

2 8 54faty i 3 8 Safety

4 9 Safety control 5 12 6 8 Control 1 9 Control 3 8 APSRs  ;

f l

Figure 3.6-2

~

Integral Worth for CRG-5

~

i Reactivity (pcm) 1,200  ;

. 1,000 l

/ ~

800 ,

l 1 600

/-

/

/

400 / ,

200 -

[-  !

i

,/ i

/ i 0

O 10 20 30 40 50 60 70 80 90 100 4

CRG-5 Position (% WD) l i

. Total Worth = 1134.0 pcm

Figure 3.6-3 ~

Integra Wort, for CRG-6 Reactivity (pcm) 1,000 d

800

, 600 i

400

/

/ '

lY 200

/

0 ,/ 70 80 90 100 0 10 20 30 40 50 60 <

CRG-6 Position (% WD)

Total Worth = 816.5 pcm

, e

. l f i 4  ;

~

0 0

1

/ I 0

9 _

7 - / 0 8

G R

0 )

7 C -,

D W

r 0 o 6 (

4 6.

3 h f

f,' 0 5

i t

i n

o s

o etr P r

go u / 0 4 7 -

i

/, G FW 0 3

R C

a l

)

m r c p 0 m g (

y 2 c p

e t i

v 0 3

/

t i n t c

a 1 9 8

8 I

e R O =

0 0 0 0 0 O h 0 0 0 0 0 t r

0, 8 6 4 2 o 1

W l

t a

o

? T

l i

,I a 4

TABLE 3.6-1 COMPARISON OF PRF.DICTED VS HEASURED HOD WoltTits

!!EASURED PREDICTED l'ERCEliT CRG. WOltTH Wol4TH DIFFERENCE No.- .(PCH) ,. __ (PCM)  %

5 1134 1115

• 151 -1,67 6- 816.5 000 1 15% -1,04 7 809.3 800 1 15% -2.4 l i

5-7 2839.8 2791 i lot -1.72-l t

i b

.t r

F p

t i

. t.

3.7 Dif f eront t al 110: on won t h n' -EU(2121 soluble poloon in t he f orm of dissolved bot ic acid la added t o the moderator to provido additional s oactivity control buyond that avallable irom tho centrol todo. The primary iunction of the soluble poinori conttol syutom in to control tho excoup s eact.lv i t y of the fuol throughout each core life cycle. The diffotential reactivit.y worth of the boric acid van tuoanured during the r.oro lower teot ,

b. Tout th.t hod lica pu t oment a of tbe difforontial baron worth a t. b32'F won o perf orned in conjunction with the cor. trol rod worth measuromont o.

'Iho control rode wortha weto neauured by t ho boron awap technique in which a deboration rate was octablished and the control rode woro inven ted to cornpensato f or the changing coro reactivity. The reactimotor was used to provido a continuous roac t. l vi t y calculation throughout the meanuronent. The difforential boron worth was then dolormined by summing the incremontal reactivity valuon measured during the rod worth sucanus tnuents over a known borou concentration range. The averago differential baron _ worth in the moanured chango in reactivity divided by the chango in boron concent r ation.

c. Tpat populta 11oanuromont a of the polublo boron ditforential worth were completed at the ' rero power condillon of $32'Y. The sneasured boron worth was 7.804 pcm/ ppm!$ at an averago boron concontration of 1951 ppmil. Tho predicted valuo wan 7.090 pcm/ppmil i 15%.

d. Conclunions Thu moanus ed reaulto f or the soluble poleon dif f orential worth at

$32'P was within 15% of the predicted difforential worth.

i-

~ I Et -

-,..~- - -... - - .-..- - - .---.. - - - - - , . . - - . . - . - - . . . . - . - . . .

l 4.0 Colm Pl<RFORMANcE - MEAsttHEttFNTS AT POWrk This section present o the results of the phyelco moaouremento that were conducted with the reactor at power. Teoting was conducted at power j platonuo of approximately lot, 30%, 75%, and 100% of 2560 megawatts coro )

thermal power, an determined from primary and secondary heat balance tucapuremonto. Operation in the power rango began on November IL, 1991. ,

l Porlodic enoasuremento and calibrations woro performed on the plant nuclear i inotrumentation during the cocalation to f ull powor. The f our power rango l detector channoin wete calibrated based upon primary and secondary plant i heat. balanco measuremento. Testing of the incoro nuclear instrumentation was performed to onoure that all detectors were functioning proporly and '

that - the detect.or inpute were proconsed correctly by the plant computer. i coro axial imbalanco determined f rom the incore instrumentation syntom was usod to calibrate the out of coro detector imbalance indication.  !

The major physico moanuremento performed during power escalation consisted l '

of determining the moderator and power Doppler coefficients of reactivity and obtaining detailed radial and axial coro power diotribution measuremonto f or reveral coro axial imbalances. Values of ininimum DNDR and 4 maximum linear heat rate were monitored throughout the t o o t. program to ensure that. coro thermal limits would not be exceeded.

I

' 4 6

I

+

____..m __ _ -.. _ __ . _ . _ _ - . .

l ie 4.1 Nuclear innt rumentation Calibration at Power

a. Purpoon The purpose of the Nuclear lustrumentation Calibration at Power  ;

was to calibrato the power rango nuclear instrumentation '

indication to be no loan than 24 Fp of the reactor thermal power a' determined by a heat balance and to within i 2.5% incore axial ,

of f oot so determined by the incoro monitoring eyotom.

l

b. Tont flot hod  !

t Ao required during power cocalation, the top and bottom linear -i amplifior gains were adjusted to maintain power range nuclear i instrumentation indication to be not leon than 2% of the power calculated by a heat balance.

When directed by the controlling proceduto for physica testing, the high flux trip blatable set poi nt was adjusted. The major settingo during power occalation aro given below:  ;

Test Plateau liletablo Sotpoint t Fp t FP 30 50 >

75 85 100 105.1  ;

c. Teut Henuit a An analysia of tout results indicated t h a. t changos in Roactor Coolant System boron and xenon buildup or burnout affected the power an observed by the nuclear instrumentation. This was expected since the power rango nuclear instrumentation t.,oasures roactor neutron leakage which in directly related to the above changen in system conditions. Each time that it was nocessary to calibrate the power range nuclear inst rumentation, the acceptanco criteria of calibration to be no leon than 2.0% PP of the huat

. balance power was mot without any dif ficulty. Also, each time it was nocessary to calibrate the power. rango nuclear inntrumentation, tho i 2.5% axial offnot critoria na datormined by the incoro monitoring nyr> tom was also met.

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

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

d. Conclunions

. The power rango channels wuro calibrated baoud on heat balance power several times during the startup program. Those -

calibrations were required due to power lovel, boron, and/or control rod configuration changen during the program. Acceptance ,

criteria for nuclear instrumentation calibration at power were met in all inotances.

1.

I r

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

_.__.____m..-___ .m....-.._..__-._ _ _ _ _ . _ - . - _ _ _ _ _ _ . _ _ _ _ . . _ . . .

4.2 Incore Detector Testing

a. Pu y ue Self-powered-neutron-detectoro (incoro detector system) monitor the core power density within the core and their outputs are monitored and procensed by the plant computer to provido accurate readings of relative neutron ilux.

Tonto conducted on the incoro detector system were performed tot (1) Verify that the output from each detector and its responce ,

to increasing reactor power was an expected.

(2) Verif y that tho background, length and depletion corrections j applied by the plant computer are correct.  ;

(3) To measure the degree of azimuthal synunetry of the neutron ,

flux. l

b. Test Method The response of the incore detectore vercuo power lovel was determined and a comparison of the symmetrical detector outputs mado at steady at.ato reactor power of approximately 10, 30, 75, and 100%FP.

Using the corrected SPND maps, calculations were performed to determine the detector current to averago detector current values por assembly for each incore detector versue axial positions. >

At 76% FP, SP-1301-5.3, Incore fleutron Detectors-Monthly Check, was performed to calibrate the backup reorder detectors to their incore depiction value. ,

t

c. Conclunions Incoro detector testing during power occalation demonstrated that all detectors were functioning as expected. Symmetrical detector l readings agreed within acceptablo limito and the computer applied '

correct. ion factors are accurato. The backup incore recorders

• wero - calibrated at 75% FP and operational above 80%'FP as required by the Technical SpecifAcations. ,

i l

i C

l

_. . ~ - . . - __.

4.3 Power 1mbat ince Det ector Correlat ion Tost

a. Purimpo The Power Imbalance Detector Correlation Test has four objectives:

1. To determine the relationship between the core power distrubtion as measured by the out-of-core detectors and the incore instruments.

2. To demonstrate axial power shaping control using the Axial Power Shaping Rods (APSR's).

3. To verify the adequacy and accuracy of backup imbalance calculations as done in AP 1203-7, " Hand Calculation for Quadrant Power Tilt and Core Power Imbalance."

4. To determine the core maximum linear heat rate and minimum DN!!R at various power imbalances,

b. Test liethod This test was conducted at about 75% FP to determine the relationship between the core axial imbalance as indicated by the incore detectors and the out-of-core detectors. Based upon this correlation, it could be verified that the minimum DHDR and maximum linear heat. rate limitr would not be exceeded by operating within the flux / delta flux / flow envelope set in the Reactor Protection System.

CRG-8 was moved to establish . the various imbalances. The integrated control system automatically compensated for reactivity changes by repositioning CRO-7 to maintain a constant power level.

c. Test Results The relationship between the ICD and OCD of f set was determined at.

about 75% FP by changing axial imbalance with the APSR's. The average slope measured on the four out-of-core detectors was 1.066. The lowest elope was 1.008 for NI-7. The scaled difference amplifier gain was left at 3.604.

A comparison of the incore detector (ICD) offset versus the out-of-core (OCD) detector of f set obtained for each NI channel is shown in Table 4.3-1.

Core power distribution measurements were taken at the most positive and negative imbalances at 75% FP. The values of minimum DNBR and worst caso MLHR were compared to the acceptance

. criteria. .

I' The worst case values of minimum DNBR and maximum linear heat '

rate determined at 75% FP are' listed in Table 4.3-2.

The-worst case DNBR ratio was greater than the minimum limit and the maximum value of linear heat rate was less than the fuel melt limit of 20.5 kw/ft after extrapolation to 105.1 FP. These results show that Technical Specification limits have been met.

Backup offset calculationo u n i rva AP 1203-7 agree with the computer calculated offuet. Table 4.3-3 110tu the computer calculated c't i net as wel1 an of f vet a obtairied uping ihe i ncut u detector backup recordera.

d. Conctoniono Backup i st.ba l a nce calculations performed in accordance wit h AP 1203-7 provido an accolitable alternato method to comput er calculated valuen of imbalance. A dii f orenco amplif ler K iactor of 3.604 will ps ovido a ulope greater than or equal to 0.96 when OCD offoct in plotted veruuo ICD offpet.

111nimum Dimit and 11aximum Linear lleat Ita t o parametern woro well within Technical Specificationn limitations.

I

a l

TAul.E 4.3-1 INCORF. OPJSET VS OUT OF-CORE OFFS!!T INCORE OUT- OF-CORE OFFSET ( % )

OFFSET (Q_ ,.

N!-5 Ni-6 HI-7 HI-0 3.03 3.79 3.98 3.51 3.74 9.70 10.11 10.32 9.00. 9.64 9.62 10.16 10.1 9.19 9.59 6.19 6.90 7 9. 6.40 6.74

-2.03 ~2.11 -1.93 1,96 -a.70

-0.06 -9.21 -9.16 -0.4b -8.41 7

-11.07 -

13.40 -13.48 -12.32 -

12.41 I

-13.46 -

15.10 -15.22 - 13.91 -

14.04 .

I I

i r

h

-l t

. 1 t

i i

a

-y g - - + . . ,, y w .m-..,he . m--we,.'wp,. w%,4, g.,yy , ,,%, ,4 .w.%.,, ew.ym..,, , , _ , , ,, , , w , w .w e ww w. . %--mer-,w,,..e..va+.w_-wgw,,_m..e.--s+w-*ciewa-

1.\lli.E 4. 3 - 2 WORT '?ASE Dl1Drt AllD LilR IMDhlANCE 11111111011 EXTRAPOLATP WoHST CASI: 1.llH E XT!!AP. IthX . LilH 4 - DlillH _ MDtJitR ___(KW/FTl ([W/_Ill 7.33 3.961 2.bb 9.07 12.71

-10.05 4.102 2.76 9.60 13.23 l

l i

. i l

l TAllLE 4.3-3 l FULL INCORE OFFSET VS IIACKUP RECORDER OFFSET.

FULL INColtE IIACKUl* RECOkbER 3 OFFSET OFFSET

( % L_ it) l 3.037 -0.06b i 9.786 6.206

-13.46 -14.06  ;

i a

l 1

r L

F i

t t

y r

Y

-i t

l 4.1 core Power Distribution Vertitcation I

a. Purione  ;

To measure the coro power dist ribut ions at approximately 75 and 100 percent f ull power to verify that the core axial imbalance, quadrant. power tilt, maximum linear heat ruto and minimum DN!!R do not exceed their specified limits. Also, t o compare the measured and predicted power distributions. j

b. Tent liet. hod Coro power distribution measurements woro per f u ed at  ;

approximately 75 and 100% full power, under steady s t.at e

• conditiono, f or specifled control rod configurat. ions. To provido ,

-the best comparloon between moacured and prodicted results, three-dimensional equilibrium xenon conditions weto established.  ;

Data collected ior the moacurementa conoisted of power distribution information--at 364 core locations from the incore detector eyetom. The worst case core thermal conditions waro calculated using this data. The moneured data was compared with calculated results,

c. Test fle nu l t s -

A summary of the caceo studied in thin report is given in Table 4.4-1 which . given the core power level, control rod pattorn, cycle burnup, boron concentration, axial Ambalance, maximum quadrant tilt, minimum DNBR, maximum LitR and power peaking data ~

for each measuromont. The highout Worst Casa ML11R was 12.01 kw/ft at 1001 FP which is well below the limit of 20.5 kw/ft. The lowest minimum DNDR value was 3.001 at 2005 FP which la well above the limit.

The quadrant power tilt and axial imbalance va. lues measured were all within tho allowable limita. Table 4.4-1 also gives a comparison bet. ween the maLimum calculated and predicted radial and total peaks for an eighth core power distribution.

d, conclualonn- ,

-Coro power distribution moacuromonto were conducted at I approximately 75%'and=.100% full power. Comparison of measured and-- prodloted roculta show good agrooment. The largent dif f erence between the maximum measured and maximum predicted peak value was 2.8% for' radial peaking at 1001. PP. This met the acceptance critoria of <5.04.

The nieasured values of DNilR - and MLi!R woro all within the allowable limita. All quadrant power - t ilt s and exjal core imbalancos moauured during the . power distribution- test were within tho . Technical specifications and normal operational limits.

di TADLE 4.4-1 CORE POWER DISTitIBUTION RESULTS POWER-PLATEAU 7 54 Q_ 100%FP DATE 11-16-91 11-18-91 Actual-Power (%FP) 74.71 99.96 CRG_1-6 (%WD) 100 100 CR0 7 (%WD) 92.0 87.7 CR0 8 . (%WD) 31.4 29.7 Cycle Burnup (EFPD) 0.83 2.25 Doron Conc. (PPH) 1835 1663 ~

Imbalance (%)- 2.88 1.94 - .

.itaximum Tilt (%). 0.74 0.65 HDNHH 4.196 3.081 i Worst Caso tittlR -(F.W/rT) 9.00 12,01  ?

Haximum Hadial Peak Honoured 1.327 1.337  ;

-Predicted 1.31 1.30 ,

Difference- (%) 1.28 2.0 l Acceptance criteria (%) (St

~

<5%

~

11aximum Total Peak ,

iteasured 1.550 1.559 +

. Predicted .

1.53- 1.52.

l' Difference (%) 1.29 2.5 Acceptance critoria (%) $7.5%. $7.5%. >

[

N 6

i

> P t

I 4

.; y. .. u ... . . . . . - - . . . - . , , . . . . - . . - . . . _ .._a:......-.__--=_. . . . . . . , . . . . - ~ - . .

I 4.5 P e a c t i v i t v ct + f f i c i e n t rut _Powe t;

a. Di s lu g; The putporo of thio teat in to meanute tho t empe r at u r e coef f 2cient of react iv t t y atni pNe t Derplot confficient of teactivity at power. T h i tt i n f ot n4st i en in then used to ausut e that Tech. Spec. 3.1.7.1, which otatets that the moder at or temperatute c oe f. ! l e i e nt Phall not t>e mit ive at power lovels abovo 95% of rated pwer, aa natintieu.

b. Tout Mothud for toui . ring the temleraturo coefficient of react ivit y, the avet age itc5 t emperature is decr eased and t hen iticreased by about.

5 degreen F. The t eactivity accociated with each temperature change in obtained from the change in controlling rod group poultion, and the values for the evolficient are calculated.

For moanuring the power Dopples coefficlent of react ivity, dat a is extracted from the fast ineert/ withdrawal seguoncos.

Ditforential controlling rod worth meanutemente are also detotmined using the fast incert/ withdrawal method.

c. Teut Heaul,t s Temperaturo and power Doppler coefficient measurements wuro pe r f ornied. At about 9Pt FP t he moaeured moderat or tempet at ure coefficient was ~4.27 pcm/*F. This verifies that the moderator temperat ure cooliicient is negative abovo 95% FP.

The measured power Doppler coefficient at 90% FP was -U.78 pcm/*FP and tho muaoured fuel Doppler coefficient wan -1.34 pcm/*F. This meets tho acceptanco criteria of being more negative than -0.9 pcm/'F.

d. Conclunions The meaoured moderator t emper at u r e coefficient (HTC) results indicate that the MTC is negative above 95% F,P.

The measured fuel Doppler coefficient (FDC) renulta meet the requirement that the FDC be more negat ive t han ~0.9 pcm/

• F.

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