ML20076N024

Cycle 7 Startup Testing Rept

ML20076N024
Person / Time
Site:	McGuire
Issue date:	02/28/1991
From:	Tuckman M DUKE POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9103270065
Download: ML20076N024 (35)

Text

- - - - - -_

__.._________.7 ,

e .

ll l  !

.s <

IMe nnon Ctanpary SIS ItnMo h leur l'wduttenn ik sci \ nr hosknt PO ika lio0i %u ltur Operations Charktte N C lh !Il Itu]l @lHD hil DUKE POWER March 19, 1991 U. S. Nuclear Regulatory Commission ATTN Document Control Desk

-Washington,_D. C. .20555

Subject:

-McGuire. Nuclear Station, Unit 2 -

Docket Number 50-369 Cycle 7 Startup Report Attached is the startup report for McGuire Nuclear Station Unit 2 Cycle 7. -

This report is submitted pursuant to Technical ,

Specification 6.9.1.1, which requires that a report be submitted if

.a modification.is made which may have significantly altered the

. nuclear, thermal,nor' hydraulic performance of the plant. During '.

'the recently_ _ completed . outage, a modification was performed to

. permanently resolve the baffle-jetting problem that has been experienced at McGuire.

If there are any questions, please call Scott Gewehr at (704) 373-17581.-

Very.truly yours, Ih,1 _ _ __

l. M.-S. Tuckman:

M2C7SU/ sag.

g .

l 9103270065 910228

-_POR P- ADOCK 05000369 ' []f l i O u J!. o , POR- x I

I -- .

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

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

' I

• e l

i Nuclear Regulatory commission March 19, 1991 1 Page 2 l

cc Mr. T. A. Reed, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 9H3, OWFN Washington, D. c. 20555  :

Mr. S. D. Ebneter, Regional Administrator l U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, NW - Suite 2900 Atlanta, Georgia 30323 Mr. P. K. Van Doorn i Senior Resident Inspector McGuire Nuclear Station  !

l 1

l l

l t

1 i

1

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

4" i

l DUKE POWER COMPANY l l

McGUIRE NUCIJJR STATION l i

UNIT 2 CYCLE 7 i STARTUP REPORT February 28, 1991 f

i I

l+.

lc l

i

+ . ~ 4. . .:.,.. . _ . . , _ _ _ . . . . - . . . . . _ . . , _ . _ , . . . ,

TABLE OF CONTENT 3 a

pagn i

List of Tables 11 List of Figures 111 1.0 Introduction 1 1.1 Prestartup NIS Realignment Following Refueling - 3 PT/0/A/4600/78

~

2.0 Criticality 5 3.0 Zero Power Physics Testing 6 3.1 Boron Endpoint Measurement - PT/0/A/4150/10 10

.3.2 Isothermal Temperature Coefficient Measurement - 11  ;

PT/0/A/4150/12 1 3.3 -Control Rod Worth Measurement - PT/0/A/4150/11 14

?3.4 : Cont rol Rod Worth Heasurement: Rod Swap - 16 PT/0/A/4150/11A 4.0 -Power Escalation Testing 18 1 4.1 Incore and NIS Recalibration: Post Outage - 21 PT/0/A/4600/02E 4.2 Thermal Power Output Neasurement - PT/0/A/4150/03 - 24 ,

4.3 Reactivity Anomalies Calculation - PT/0/A/4150/04 26 i 4.4' Incore and Nuclear Instrumentation System 27 .i' Correlation Check.- PT/0/A/4600/02A 4.5 Core Power Distribution - PT/0/A/4150/02A 28 i

~

i 1

-i i

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

.- . i l

LIST OF TAlllIS i

- Page

- 1. Overlap Data 7

2. Nuclear lleat 8 )

r

3. Reactivity Computer Checkout 9

4. Control Rod Worth Measurement: Rod Swap 17 5.- Core Power Distribution Results - 30% Full Power 19 r

6. Core Power Distribution Results - 76% Full Power 20

- 7. Quartec Core Flux Map Data for PT/0/A/4600/02E, incore 22 i and NIS Recalibration: Post. Outage 8; Thermal Power Output Measurement Results 25

9. Core Power Distribution Results --100% Full Power 29 1

i~

i s

.I.,..J.-.-,,--..m.,-,,,,b,< -E..-,J.,.,,-,.r , -.wy...- -v~.. -

. _ _ _ _ _ _ _ . _ . _ _ . - _ _ . = . . - .._... .___..__ _ _...__.

-._____.__.m__.._.___._..._.;

LIST OF FIGURES  !

i Page

1. Unit 2 Cycle 7 Core Loading Pattern 2

2. Assemblies to Use for Calculating IR and PR 4 Calibration

3. ITC Heatup and Cooldown Data: First Run 12

4. ITC Heatup and Cooldown Data Second Run 13

• 4

5. Shutdown Bank B' Rod Worth 15

6. Incore and NIS Recalibration Results 23 f

r 3

i 5

a iii

. - _ . , . _11 _ , . . . _ , _ _ . _ - . . . . - . . . . , - . . _ . , _ _ _ _ _ _ . _ _ _ _ _ . _ _ . . _ _ _ __._..~..;...._.._..,

1.0 Introduction Core loading for McGuire Unit 2 Cycle 7 was started on October 13, 1990 and was completed October 16. The core for McGuire 2 Cycle 7 consists of 193 Westinghouse optimized fuel assemblies. To control power peaking and maximize cycle length, 64 Westinghouse Wet Annular Burnable Absorber (WABA) inserts are utilized. Figure 1 gives the Unit 2 Cycle 7 core loading pattern.

Criticality, Zero Power Physics Testing (ZPPT) and Power Escalation Testing (PET) began December 25, 1990. During performance of the Zero Power Physics Test, PT/0/A/4150/11A, Control Rod Worth ?!casurement: Rod Swap, a manual trip occurred on December 27, 1990 at 0142 hours0.00164 days <br />0.0394 hours <br />2.347884e-4 weeks <br />5.4031e-5 months <br /> due to Shutdown Bank E falling into the core. After more than 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of troubleshooting on the Rod Control System to determine the roct cause of Shutdown Bank E falling to its fully inserted position, criticality was achieved again on December 28, 1990. ZPPT and PET were then successfully completed. The unit reached 100% power on January 4, 1991.

1

Figure 1 1 Core Loading Pattern Vc3uire \uclear S ation

.n i : 2 Jycle 7 QUADRANT IAsM #

I 2 l1NS&

4 3 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 T10 U59 T4a U30 T24 U19 T14 316KT $9 58 227kT 89 207KT 88 g T57 T25 U47 T51 UO7 S34 US7 T52 U73 T13 T2B 22SKT R138 4P189K R143 8P22iK R135 SP228K Ri4S #sS7N R108 87 g

TSO TiB U20 T29 U32 S05 T07 sis UOG TS9 U45 T55 T36 Sl2KT 124 #155K R120 12Pil4LR108 216KT R110 12P124 R131 #160K 120 50 C g TBS U41 952 USS S24 U75 TOD U44 SO3 U27 SSO USi Tel4 Rii8 4P181K Rite 12Fitk 08 12Pt 2>.R137 12Pt3D 80 12 Pith,M129 #183K RtBS g

T27 U71 T1B U54 S30 Via S53 U23 SS4 Usc, c21 U49 TS2 U42 TO E 128 #iBOK R102 12P136 228KT 12P139h E it$ 12P113t20lKT 12P118t 20lk'T 12Pi44LR151 #1$$K 70 g USS T34 U50 Sie U70 T71 T35 S13 Tai T54 UO2 901 UO5 T5d U25 91 RtR7 12P132i,204KT 12Pt218;R105 44 R132 titKT R17 12P142a , 43 12Pil26: Ridt 130 F g T09 UO3 S33 U0i S50 T12 SO2 Uli 535 T19 S57 U24 S37 U21 T63 141NT $P227K R81 12P109h 133 71 270KT 12P133s; 55 53 110 12PtP8hRIB4 BP224K S7 g

u31 S2 T33 T15 ue2 SiG uS5 S40 usB SO4 UO4 T20 T30 S44 une H 81 R60 0898 R113 12P140> R119 12P119GR198 12P141 .R140 (2P118i,Rl42 0999 Rig? 94 H i TB4 U57 941 U33 SSB *i90 036 U22 Sit T42 S49 U2B S42 U34 T72 205KT DP22SK R128 12P195t 73 220KT 122 12P10P 249KT 00 22iKT 12P1246 Ril5 8P22kK 45 UOS T53 U48 S47 US2 T31 TOS S25 T41 Til U20 S39 UiB TS7 U70 103 Ril8 12P13De 228KT 12P12P'R103 212KT R123 114 Ridi 12 Pit ta. 83 12P135t R148 51 TBi USS T23 U30 S45 USS S54 UOS S58 U29 SOS V72 T58 U43' T04

( 84 #154K RlB2 12 Pith 215KT 12P129* B0 12Pis16 52 12Pf3h 121 12PlN> R134 #182K 20BKT t

1 TSB U13 SSi UB4 012 UB3 T70 UiB S17 UBn SSS U35 Toi RiO4 #149K Allt 12P108h 127 12 Pith R159 12P10h 11BKT 12Platt:R130 #16tk R163 q

g T3B T05 U40 T32 U14 S46 T22 S27 U17 TSS US9 TO3 T37

, 33tKT 222KT #163K Alti 12P141W A83 219KT R107 12P1226: R125 pl64K 100 334KT N

t g T45 T43 U15 T02 U39 S2B J37 T59 U53 T40 T17 48 Ritt 4PiB2K R139 $P221K R109 kP228K R144 #tB4K RtS0 85 g

q T39 USO T4B U74 T47 U40 T49 i 72 88 100 105 208KT 95 313KT q

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 2

1.1 Prest artup_ NIS Realignment following Refueling - PT/0/A/4600/78 This procedure was perf ortned on Oct ober 10, 1990.

This test was used to calcul te reeliminary calibration data for the intermediate range (IR) and power range (PR) detectors following refueling.

The set of Cycle 7 preliminary oration data was determined by taking the End of Cycle 6 (EOCt ,.ilibration data and adjusting it by a weighted average of the ratio of the sum of the predicted assembly powers for the Cycle 7 loading to the sum of the measured assembly powers from the last Cycle 6 Incore/Excore calibration. The core locatf.ons used to calculate the ratio of the predicted Beginning of Cycle 7 (BOC7) assembly powers to the measured EOC6 values are shown in Figure 2.

The predicted BOC7-to-E006 IR ratio was NO.87; the predicted BOC7-to-EOC6 PH ratio was N0.92. Based on these results, the IR and PR currents were adjusted prior to Cycle 7 Initial Criticality.

3

s t I

Figure 2 Assemblies to Ute for Calculating IR and PR Calibration Setpoints n

e O O o 1

@ X' O X@ a XX XX 3 4

~

5 6

7 8

9 10 11 12 XX XX

~

is

@X O X@

O O O - _

15 P N M L K J H G F E D C B A R

A D

Core locations used for PR calibration l

Q Core locations used for IR calibration l

. 4

2.0 Crit __icalit-y following a Change in Core Nuclear Characteristics -

pT/0/A/4150/28 On December 25, 1990, boron saiaples were taken in preparation f or the approach to criticality. These samples indicated reactor coolant boron to be 1925 ppm. Since it was desired to achieve criticality with N500 pcm of Control Bank D inserted, a target value of 1533 ppm was chosen for reactor coolant boron concentration. This represented the predic' ed 110C, ARD, il7.p, No Xenon, equilibrium Samarium critical boron concentration of 1583 ppm less 50 ppm. Calculations using the unit Data Book (Op/2/A/6100/22) indicated a volume of 13,899 gallons of demineralized water should be added to the system to dilute from 1925 ppm to 1533 ppm.

On December 25, 1990, this dilution of the Reactor Coolant System was started. The dilution was secured after 13,899 gallons of demineralized water had been added to the system. After adequate system mixing, Chemistry samples indicated Reactor Coolant System boron was 1485 ppm.

Since this boron concentration w!,s not within 30 ppm of the desired target reactor coolant baron concentration, four hundred twenty-five (425) gallons of boric acid was added. After system mixing, Chemistry samples then indicated Reactor Coolant System boron was 1521 ppm.

On December 25, 1990, rod withdrawal commenced starting with Shutdown Bank A. As rods were withdrawn, both source range detectors were observed and rod motion was stopped each time either flux level doubled or any control rod bank was fully withdrawn. At these points a set of counts were taken on each source range detector and Inverse Count Rate Ratio (ICHR) was plotted to monitor the approach to criticality. The unit achieved criticality at 1555 hours0.018 days <br />0.432 hours <br />0.00257 weeks <br />5.916775e-4 months <br /> on December 25, 1990, with Control Bank D at 49 steps withdrawn. The predicted critical position per OP/0/A/6100/06, Reactivity Balance Calculation was 76 steps withdrawn on Control Bank D. This represented a reactivity difference of 184 pcm.

Due to the manual reactor trip on December 27, 1990, PT/0/A/4150/28 was performed again to achieve criticality on December 28, 1990. Chemistry samples indicated Reactor Coolant System boron was 1520 ppm. Rod withdrawal commenced on December 28, 1990, and the source range counts were monitored as previously described. The unit achieved criticality at 1234 hours0.0143 days <br />0.343 hours <br />0.00204 weeks <br />4.69537e-4 months <br /> on December 28, 1990, with Control Bank D at 42 steps withdrawn. The predicted critical position per Op/0/A/6100/06, Reactivity Balance Cel ulation was 74 steps withdrawn on Control Bank D. This represented a reactivity difference of 219 pcm.

5 t _ _ _ _ _ _ _ _ _

3.0 Zero Power Physics Testing- (ZPPT)

Zero Power Physics Testing for McGuire 2 Cycle 7 started December 25, 1990, and was completed December 29, 1990. The output of

- Power Range Detector N42 was used as input t.o the reactivity computer for Zero Power Physics Testing. All acceptance criteria for ZPPT were met.

A mininnun of one decade of overlap between the source range and the intermediate range detectore was verified on both December 25 and 28, 1990 i via the Control Board indication, the NIS panel, and the Operator Aid Computer (OAC). The results shown on Table 1 reflect the data from the OAC.

The point of adding nuclear heat was determined December 26, 1990. This j was done by establishing a slow positive startup rate and observing a  !

change in plant parameters such as an increase in Reactor Coolant System average temperatures (Tave) with a change in the reactivity trace and an increase in pressurizer level. The test was performed three times to establish repeatability of the data. Table 2 gives the reruits of the  ;

, second two trials which were used to determine an average nuclear heat  ;

reading.

Nuclear hegt was determined to be at an average flux level of ,

2.73 x 10 -6 amps on the reactivity computer picoammeter (N42) and '

l.84 x 10 amps on Intermediate Range Detector H35 and 2.29 x 10 -6 amps onIntermediateRangeDetectogN36.,Jromtheseresultsthetestbandfor ZPPT was deteemined to be 20 to 10 amps on the reactivity computer.

On both December 26 and 28, 1990, an on line checkout of the reactivity computer was performed. This was done by withdrawing Control Bank D until a positive reactivity insertion of ++25 pcm was indicated on the reactivity computer. The time for the flux level to double was measured l and from this doubling time (DT), the reactor period was calculated (period = DT// -*3). Using the reactor period, the amount of reactivity was determine :ing the predicted data. This reactivity was compared to the reactivity computer indication. The test was repeated for a

- reactivity insertion of s+40 pcm. An on-line negative reactivity checkout L on the reactivity computer'was also performed. This was done.by inserting I Control bank D until a negative reactivity-change of N-40 pcm was

- indicated on the' reactivity computer. The' time for the flux level to halve was; measured and from this halving time (HT), the reactor period was calculated (period = lit /0.693). Using the reactor per.tod,- the amount of l _ reactivity was determined using predicted data. This reactivity was l compared to the reactivity computer indication.- The test was repeated for

-a reactivity change of N-25 pem._ The final results met all acceptance criteria.and are given in Table 3.

An electronics only negative reactivity insertion test was also completed satisfactorily as part of PT/0/B/4600/55, Reactivity Computer Periodic.

-Test.

l ,

6 i

. - - . . , , ~_mu ,,, . mi . . . -_ --_, _ ., _.._._ a m . .~.,,:m. . . . . - , - , - . , , _ . . - , , - ,-,,_-m.

" * " ~"' ~ "

f,ffhf,hk')hb D NIM M W N i

l . .

TABLE 1 Overlap Data on December 25, 1990 Source Range Intermediate Range cps ampa N31 N32 N35 N36 1.4 x 10 1.4 x 10 3 1.5 x 10'II 1.8 x 10'II 4 4 -0 -0 1.8 x 10 1.9 x 10 1.6 x 10 1.9 x 10 Mien SR blocked 2.0 x 10 0

2.1 x 10 1.8 x 10 -10 2.1 x 10 -10 Overlap Data on December 28, 1990 Source Range Intermediate Range eps amps N31 N32 N35 N36 3 3 0.6 x 10 0.7 x 10 1.1 x 10'II 1.2 x 10'II 1.2 x 10 1.4 x 10 1.1 x 10

-0 1.2 x 10 -10 4 -10 -10 When SR blocked 1.9 x 10 2.0 x 10 1.7 x 10 2.0 x 10 7

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

4 4 1 e' .t .

TABLE 2 1

.I Nuclear Heat l

.I Reactivity Computer Intermediate Range  ;

l N42 N3$ N36

-6 1.64 x 10 -6 2.10 x 10 -6 2.38 x 10 '

3.08 x 10 -6 2.04 x 10 ~0 2.48 x 10 -6

-- i

-6

. AVERAGE. 2.73 x 10 amps _ 1.84 x 10 -6 ampo 2.29 x 10 -6 amps I

~0 Test-Band: 10 to 10'I amps on N42.

9

+

4

-t Y

.6 l

l' 8' <

u I

l-. . .

.m TABLE'3 Reactivity Computer Checkout  ; .

Results on December 26,.1990 Doubling or Reactivity Reactivity -+'

Initial Flux Level (Amps) Period . Halving, Time Computer (Ap ) DT(Ap rH DT)(pcm)T(ApHT Reactivity Computer (Seconds) (Seconds). (pcm)  % Error 1.44 x 10~ 199 138 32.5 32.6 .0.3 4.40 x 10_ 97 67 59.3 58.7- 1.0

-8 8.89 x 10 -372 25' -23.0 -22.7 1.3.

~8 8.73 x 10 -248 172 -37.3 -36.2 2.9 e

Results on December 28, 1990 2.91 x 10~ 231 160 28 28.8 2."

~8 3.36 x 10 137 95 44 44.9 2.0

~

6 'O x 10 " -244 169 -37.5 -36.9 1.6 4

7.3' x 10~ -364 25/ -23 -23.2 0.9

+ ap APIrr - ^P c or - @c c 100 OR -

x 100 e c

'4 4 3.1 Boron Endpoint Heasurement - PT/0/A/4150/10 This test wee performed December 26, 1900. Two sets of data were i obtained. In the first set, Control Bank D was initially at 215 steps withdrawn, the Reactor Coolant System bcron concentration was 16?1 ppm and the Pressurizer boron concectration was 1586 ppm.

Control Bank D was pulled to the All Rods Out (AR0) Configuration and the reaulting reactivity change was converted to equivalent boron using the prediated Differential Boron Vorth. Control Bank D was then reir.serted to the just critical condition and the test was repeated three times. After these four test trials, an additional two trials were performed. The initial conditions for these additional two trials were: Control Bank D at 201 steps withdrawn, the Reactor Coolant System boron tvacentration was 1594 ppm and the Pressuri er boron concentration was 1586 ppm.

The results of these reactivity changes were each added to the initial Reactor Coolant System boron concentration to give two values for the ARO Boron Endpoint. All of the values were averaged to give the final result of 1600 ppm. This value met the acceptance cri c-ion for the Hot Zero Power (HZP) ARO Critical Boron concentration : 1583 150 ppm.

10

4 3.2 I,sothermal Temperature' Coefficient-Heasurement -PT/0/A/4150/12

-This test was performed on December 26, 1990. The test measures Isothermal Temperature. Coef ficient (1TC) by plotting Reactivity.

-versus Average Reactor Coolant System Temperature. The Moderator Temperature Coefficient (NTC) is found using the relationship as follows:

MTC (pcm/*F) = ITC - Doppler Temperature Coefficient The acceptance criterion on the ARO ITC was 1.90 12.0 pcm/*F. The predicted Doppler Temperature Coefficient was -1.41 pcm/*F.

The Reactor Coolant System boron concent. ration was 1598 ppm at the start of the test. A heatup/cooldown was-performed while keeping rod position and boron constant to determine reactivity change versus temperature. The heatup/cooldown was performed a seco9d time to est-ablish repeatability of the data. The results are shown in .

Figures 3 and 4. The average ARO ITC was found to be +1.70 pcm/ F.-

This fell lwithin the acceptance criterion band. This gave an ARO MTC of +3.11- pcm/ F which was within acceptable Technical Specification limits.

Following the completion of this test, PT/0/A/4150/31, Determination of Rod Withdrawal Limits to Ensure Moderator Temperature Coefficient .

Within Limits of Technical Specifications was performed. The results of this test-indicated there-were no red withdrawal limits needed for Unit 2 Cycle 7, 1

4 11

Pigure 3 ITC Heggu and Cooldown Da "' First kun avg (*F)

I 554op 555cp S!7*r 1111  ;; !I CGTt

.p.

COOLDOyy Finish- ll llll ll llI

{l Q (Pcm) / l_ llllll h

llllll llll 'S Pcm

" n jl.ff

~S' ' ~3,5 op_

y x.

h

) ,

' HEA7yp v.

Starg k (Pcm) l +5,5 p l ' '

~'

'..g-17..

, it e

~+3,0 0 p f  ;

12

Figure 4 eatuP and congdown Dat"' Second Run T,yg ( F) 3F gg4,F 5;7op

,{56't lll 1 Sta,t j .

i ll f

0--

h'"hi?MPon Q(Pcm)

~5 -

I"I*y y

ll ,

l,J h

f

,Q ~

f'A Pcm ll [j }b

,, '~3.co p 1

/

N ~

c cP..,

.t.,

ay u ..

~""~

!!!!!! l x

///

~

Mll ....., y 4 P

13

l 3.3 Control Rod Worth Measurement - PT/0/A/4150/11 On December 26, 1990, Shutdown Bank B rod worth was measured using the established boration/ dilution method. There were no other rods in the core at the time. Shutdown Bank B was predicted to be the highest worth bank and was measured using this method so as to serve as the reference bank for Control Rod Worth Measurements by Rod Swap.

The measured worth of Shutdown Bank B was 826 pcm. The predicted worth was 860 pcm + 29 pcm. This represented an error of 4.0% and was within the acceptance criterion af 115%. Figure 5 shows the measured integral and differential rod worths for Shutdown Bank B.

Figure 5 McGuire Unit 2 Cwle 7 l Shutdoun Bank 8 Rod (Jorth l Integral and Dif ferential Rod (Jorths Integral Differential ,

.lJorth Cpcm)- (Jorth Cpcm/ step) i 900 - -

9 i 1

~

800 -

8 N -

700

\

7 N _

600

\

6 500

~ '\~ I -

~5

,e

\ ~~_,,  :

g

, N s.

4

\ .'s

\

~ ~

300 - x ,

~3

/ \ \

200

~

' \ '

~

2 N \

--- Di f f ' t ial 00 =/

\

\

\= 1 (J rth Integral

\:

O ' ' ' ' ' ' ' ' O lJorth O 25 50 75 100 125 150- 175 200 225 Bank Position Csteps uithdraun)

I 15

3.4 Control Rod Worth Measurement: Rod Swap - PT/0/A/4150/11A On December 27, 1990, the rod swap method of control rod worth measurement was begun. Shutdown Bank B was used as the reference bank and its worth-was measured by the boration/ dilution method (see Section 3.3).

With the reference bank essentially all the way in and the reactor just critical, each control and shutdown bank was exchanged with the reference bank. The integral worth of the bank being measured (i.e.,

the test bank) was determined from the difference in the critical rod position of the reference bank with and without the test bank in the core.- The following banks were first measured: Control Bank A and Shutdown Banks A, C, D and E. While attempting to measure the rod worth of Control Bank D, Shutdown Bank E was withdrawn to the demand Insition of 225 steps. The Reactor @ erator (RO) noticed that the Digital Rod Position Indication (DRPI) System indicated 228 steps withdrawn. While the RO began to manually insert Shutdown Bank E to 225 steps withdravu, Shutdown Bank E fell into the core to its fully inserted position. The reactor was then manually tripped. Extensive investigation and troubleshooting was_ then performed by station personnel to determine the root cause of Shutdown Bank E falling into the core. No abnormalities were discovered. Criticality was again achieved on December 28,_1990. At this time, the remaining control rod banks were measured and PT/0/A/4150/11A was completed.

The measured bank worths were compared with predicted worths and all -

banks were within the acceptance criteria of i30% or 4200 pcm whichever was greater. The measured total rod worth was >90% of the i

predicted worth which met the acceptance criteria. In addition, all review criteria were met.

The results of the rod exchange test are given on Table 4.

i 16

TABLE 4 Control Rod Worth Measurement: Rod Swap _

Bank Predicted Worth Measured Worth Percent

• Identification pcm pcm++ Difference Shutdown Bank B 860 826* 4.1 (reference)

Control Bank A 295 289 2.1 Contro? Bank B 773 739 4.6 Control ,ank C 830 774 7.2 Control Bank D 494 468 5.6 Shutdown Bank A 309 313 1.3 Shutdown Bank C 455 419 8.6 Shutdown Bank D 455 430 5.8 Shutdown Bank E 482 462 4.3 TOTAL R0D WORTH 4953 4720 4.9

• Measured by boration/ dilution method

, Predicted - Measured x 100 Measured

++ Rounded to nearest pcm i

l 17

4.0 Power Escalation Testing McGuire Unit 2 Cycle 7 Power Escalation testing startel December 29, 1990, at the conclusion of ZPPT and was completed January 7,1991.

The unit went on line December 29 at 1230 hours0.0142 days <br />0.342 hours <br />0.00203 weeks <br />4.68015e-4 months <br />. The unit experienced some holds during power escalation which were scheduled to allow testing '

per PT/0/A/4150/21, Post Refueling Controlling Procedure for Criticality, Zero Power Physics, and Power Escalation Testing, and to allow a secondary

.ide boron soak per Primary Chemistry.

During a hold for turbine testing at N30% power on December 31, i PT/0/A/4150/02A, Core _ Power Distribution, was performed. The results from the test indicated that Core Power Distribution Technical Specification Limits for operation to 94.8% power would not be violated, and all test acceptance criteria were met. Table 5 shows the test results.

Following completion of the_ secondary side boron soak at 30%, power was increased to N50% at N2.5%/hr. From 50% to 80%, PT/0/A/4600/02E, Incore and NIS Recalibration: Post Outage, was performed (see Section 4.1).

At N76% power on January 3,1991, PT/0/A/4150/02A, Core Power

-Distribution, was performed. The test results are given in Tabie 6. All-test acceptance criteria were met. The results from the full core flux map taken were used toLproject a " limiting" power at which F or F Tech Spec peaking factor margin would be maintained. Thisprojeckionikhicated that.F Tech Spec peaking factor margin would be maintained for all power levels AH up to 100% power and that the FnTech Spec peaking factor margin would be maintained for all power levels up to 99.14% power. At S90%

power, Tech Spec RAOC Axial-Flux Difference (AFD) wings were reduced by 1%

based on Design Engineering recommendations te increase power beyond 99.14% and up to 100% power.

The excore detectors were calibrated at +76% power on January 2-3, 1991, and power escalation tnen resumed at a rate of s2.5%. Upon achieving N90%, PT/0/A/4150/03, Thermal Power Output Heasurement, was performed (see Section 4.2).

The remaining tests designated for Hot Full Power Equilibrium Conditions -

were performed on January 7,1991. The tests and their results are described in Sections 4.3 - 4.5.

18

..x TABLE 5 Core Power Distribution Results 10% Full Power. .

NOTE: Axial location 1 is tht ittom of the core.

Axial' location 61 is the top of the core.

Unit 2 Cycle 7 Map FCM/2/07/001 Date/ Time Map Taken 12/30/90 2130 hours0.0247 days <br />0.592 hours <br />0.00352 weeks <br />8.10465e-4 months <br /> Power Level- 30.46%

Cycle Burnup. 0.32 EFPD 13.3 MWD /HTU.

Boron Concentration 1488 ppro

- Control Rod Position Control Bank D at 205/206 steps withdrawn Maximum F : 2.0420'at Axial Loc. 42, Horiz.

9 Loc D-13 Maximum F : 1.2981 at Axial Loc. 42 7

Maximum pin'F 1.4637 at I!nriz. Loc. D-13 Hj Maximum error I gg (from predicted) .

6.447, at Horiz Loc. J - Maximum F /K(Z) 2.2126 at Ax-41 Loc. 42 Maximum % Reduction in Axial Flux None Difference (AFD) Wings Minimum %-Margin to AFD Wings -44.8004% at Axial Loc. 40 Rmax (Tech Spec 3/4.2.3) 0.8128

. Total Reactor Coolant Flowrate 391,328' gallons / minute (Process Computer)

Total Incore Axial Vffset 12.881%

Incore Tilts %:

Upper Core Lower Core Quadrant 1: -1.317% Quadrant 1: -0.779%

_-Quadrant 2: 0.974% Quadrant 2: 1.262%

. Quadrant-3: -0.01 % Quadrant 3: -0.332%

Quadrant 4: 0.352% Qur.drant 4: -0.151%

19

-ve v- - - - - ,- , -- . , - , . . - - . _ - , _ _ _ . _ _ _ _ _ _ _ _ - - _ _ _ - _ _ . . - _ _ . _ . . _ _ . _ _ _ _ _ . _ . _ . _ _ - .

- . ~ - . . .

s

'i i

TABLE 6 Core Power Distribution Results 76% Full Power NOTE: Axial location 1 is the bottom of the core.

AxialLlocation 61.is the top of the core.

Unit 2 Cycle 7 Map FCM/2/07/016 Date/ Time Map Taken 1/3/91 0817 hours0.00946 days <br />0.227 hours <br />0.00135 weeks <br />3.108685e-4 months <br /> Power Level 76.25%

Cycle Burnup 2.25 Ed 0 94 MWD /MTU Boron Concentration 1237 ppm Centrol Red Position Control Bank D at 196 steps withdrawn T

Maximum Fq i 1.8776 at Avial Loc. 35, Horiz.

Loc e Maximum Fg : . 1.2".z3 at Axial Loc. 34 Maxir'um; pin F3g 1.3939 at Horiz. Loc. D-13

-6.20% at'Horiz. Loc.'J-01

- Maximum error F g (from predicted)

Maximum F /K(Z) 2.0102 at Axial Loc. 40-Maximum =% Reduct!.0:. in Axial Flux None Dif ference (AFD) Wings Minimum %-Margin to AFD' Wings -22.8938% at Axial Loc. 36 R,, .(Tech Spec-3/4.2.3) 0.8734 Total Reactor Coolant Flowrate -389,924 gallons / minute (Process' Computer)

Total'Incore Axia. Offset _ 3.521%

- Incore-Tilts'%:

Upper Core Lower Core

! Quadrant 1: -1.361% Qu,drant 1: - -0.132%

.uadrant-2i Q 1.628% Quadrant :2: 0.328%

u Quadrant 3: -0.227% Quadrant 3- -

0.149%

! Quadrant 4: -0.040% Quadrant 4: -

0.047%

E l

p L 20 4 ._ ,-. . ~, n ,, a s e ,,a- w - 4y

,n.

~

4.1 Incore and NIS Recalibration: Post-Outage - PT/0/A/4600/02E This' test was startedion January- 1,1991, and was- run during the  ;

1 power escalation-from:50% to 80% Full Power.: The data obtained from this test were_used to' set the-nuclear-instrumentation system amplifier gains and the axial-flux difference function of the -l

. overpower AT setpoints and to _ determine the correlation between i incore-and.excore axial offsets.

Data collection.was accomplished by taking' quarter core flux maps and associated excore detector currents at eleven different axial offsets as indicated in Table 7. (The quarter core flux map pattern had

previously been verified as: an_ accurate representation of- axial offset through PT/0/A/4150/23,-Quarter Core Flux Map Qualification

-Test); These data were input into a benchmarked off-line-computer program which generated the output-shown in Figure _6. . _The

-appropriate -factors were then put into- the: plant instrumentation '

systems and.al1~ acceptance criteria'were met.

.4 J

4

.i I

~

k l -;

l -

21 E

t-r-~ ~ . - , , . -- , ,y - . -. ,, , y,- ,,r - - --

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

L ,-

TABLE 7 Quarter Core Flux Map Data for PT/0/A/4600/02E, Incore and NIS Recalibration: Post Outage Map Average Thermal Power (%) Incore Axial Of fset (%) s

1) QCM/2/07/003 52.6 13. ;8

2) QCM/2/07/004 55.0 11.198

3) QCM/2/07/005 57.4 8.815

4) QCM/2/07/006_ 59.8 5.754 15 ) . QCM/2/07/007 62.5 3.681

6) QCt1/2/07/008 64.9 1.5t5

7) . .QCM/2/07/009 67.6 0.60J

8) QCM/2/07/010 69.9 -

1.297

9) QCM/2/07/011 72.7 -2.700

10) QCM/2/07/012 75.6 -4.911

-11) QCM/2/07/013 76.5 -5.792

}

l' l:

l l

, i 22 l

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

7 7 , . :r- ; p Figure '6 Incore and NIS Recalibration Results:

ExcoreCurrentsaridVoltagesi fortelated'to 1001 Full Power-at Various Axial.0ffsets 9

-M Unit:2 Cycle 7 FULLP0iERDETECTORCURRENTS(MICR0 AMPS)CORRESPONDIN8TOVARIOUSINCOREAIIALOFFSETS

.IPCORED DETECTORN DETECTOR N-42 DETECTOR N-43 DETECIDR N-44 Atl AL -

i

-0FFSET i. B: 1 B T B B 30.0 260.3 190.0 337.9 237.2 291.6' 209.0 283.9 201.3 20.0 248.8 213.6 -323.9: 266.1 200.0 - 235.0 272.8 225.3 10.'0 237.2- 237.2 309.8- 295.0 268.3 261.0 261.7- 8'9.4 0.0 ^ '225.7- 260.7 295.8 ' 324.0 256.7 287.0 250.( 273.4-

• 297.4

-10.0 214.2 -284.3 281-. B - 352.9 245.1 313.0 239.5

-20.0- -202.6- 321.5-307.8 267.8 ' 381.0 233.4 ~339.0 E28.4 1-30.0L_ -191.l 331.4 253.8. 410.8 . 221.8 ~ _365.I' 217.3- -345.5

?

Lr*- -0.9799 -0'964 'O.9773 -0.9960 0.9741 -0.9967- -0.9735 -0.9961

NORM'. 7ED DETECTOR V011 AGES (VOLTS AT VARIOUS AllAL OFFSETS

. LlhCORE -DETECIDRN-41 DETECTOR N 42- . DETECTOR N 43 ' DETECTORN-44 gyjat .m >

0FFSEI; iTJ 'B' T-B- 1 B= 1-B - - T -- B T-B T- B- ' T B:

~

30.0 9.600: .6.071 s3.537i 9.515- 6.098 : 3.416 e 9.463- 6.066 n 13.397- 9.4?8 ra;133 ,3.305s s20.0- -9.182 6.824--' ' 2.358 - 9.120 -6.842 2;277 9.085 - 6.820s 2.265- 9.069 6.865 2.203-

, C - 10.0 ; L8.1561 J7.577- 1.179 - : 8.725 - - 7.586 1.139 - 8.708 - 7.575: 1;132 : 8.699 - 7.598x :1.102 :

0.0 ; 8.330 L8.330 0.000L 8.330 8.!M : =0.000 8.330 8.330 0.000 8.330 ~ -8.330-.-0.000. , ,

cl0.0 ; 7.904 . 9.083 -l.179; 7.935 9.074 -l.13? 7.952. 9.C85 4 .1321 7.961 9.062 :-l.102~

- L-20.0 :' t7.478 :9.636- -2.358 7.540L 9.818: d 2.277 7.575:- 9.840 1-2,265

'7.591i 9.795 -2.203-J 30.0' '7.052 -:10.589 :-3.537 7.145 -10.562:*-3.416' = 7.197 : 10.594':=-3.397i 7.22E 10.577

-3.305 c

- I y m

AFD:lNCORE/EXCORE hil0S FOR QUADRANTS ! - 4

, r 1

,QUA0 2 ' ' 0UAD 3 .

3 QUAD 4L QUAD t .

N 41 i 'N-42 .N 43' N-44 M = 1.413  : M = 1.463; M = 1.471 M = 1.512.:'

23 g

s.~

1

.f r /'

+ ,,--[. -

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

~ ./

l

=. .

I i

4.2 Thermal-PowetJ_u_tput Measurement - PT/0/A/4150/03 This' test"was used to verify that the primary and accondary heat balances on -the plant computer were consistent with primary and secondary heat balances-on a benchmarked offline computer. The test was run on' January 3, 1991, at s90*; F.P. The results are shown in

-Table 8.

The acceptance criterion of 1% difference betwecn the offline computer and the plant computer was met.

f 3-I 1.

i w -

E n 24 4 . - . . . , .. . .c_.. _ _ .x __ . _ _ _ . . . . . _ _ - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

-!.. + .s; y- .,

i

- TABLE 3

. Thermal Power Output Measurement Results 3 Plant Computer -Off-Line. Computer

% We  % WL Primary Heat Balance 89.43 3050.58 89.64 3057.62 Secondary Heat. Balance 89.92 3067.09 89.99 3069.56 t

e a

f 1

s u

k 1.

s 4

4'

>kI v

s 25

-+ + . _ - --..v - ie., , , , ,

= .

4.3 Reactivity Anomalies Calculation - PT/0/A/4150/04 This test compared the actual core reactivity to the predicted core teactivity by taking into account the actual Reactor Coolant System boron concentration, Xenon and Samarium worths, rod positions and power level and adjusting these to the ARO, Hot Full Power (RFP),

equilibrium Xenon and Samarium condition. Theoretical and actual Reactor Coolant System boron concentration for this conditions were then compared.

The test, performed at s100% on January 7, 1991, indicated that the actual ARO, HFP, equilibrium Xenon and Samarium condition boron concentration was 1Kf2.5 ppm. This ca,mpares to a predicted value of 1138 ppm. The 6.5 ppm difference translated into a 56.6 pcm error an e ' '. and predicted reactivity sorths. This was within the acceptance _riterion for the test of i1000 pcm.

l 26

c a.o 4.4 Incore and Nuclear Instrumentation System Correlation Check -

PT/0/A/4600/02A This test was used to compare the incore axial flux difference as l . indicated by a full core flux map to the axial flux dif ference indicated on the plant computer by the excore detectors. This test also verifies the incore/excore calibration data that had been implemented during PT/0/A/4:'00/02E, Incore and NIS Recalibation:

Post Outage.

The test was performed at $100% on January 7,1991. The indicated incore axial flux dif ference f rom flux map FCM/2/07/017 was 2.436%.

The core average axial flux difference from the excore detectors was 3.950%. These results gave an absolute difference of 1.514% and was within the acceptance criterion of 13% difference i

l l-l l

l l 27 l

l I

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

. ,. - - - . . _ . . _ _ = . . . . . - _ - . _ _. __. _ . . . . _ . . _ . - _ _ . _ . . - . . _ __.

,e' s 4.5 Core Power Distribution - PT/0/A/4150/02A On January 7,1991, PT/0/A/4150/02A, Core Power Dist-ribution, was

- performed to verify the Core Power Distribution Technical Specification Limits for operation would not be violated. The reactor was at N100% Full Power and equilibrium conditions. One result of this' test indicated a 0.8918% margin to the F nTech Spec peaking factor limit. The Tect Spec RAOC AFD wings werd then restored to their 100% power .opete Lion value.

All acceptance criteria for this test were met. Table 9 gives the test results.

l --

L l

l.

l' l.

28 i

. s o TABLE 9 Core Power Distribution Results N100% Full Power NOTE: Axial location 1 is the bottom of the core.

Axial location 61 is the top of the core.

Unit 2 Cycle 7 Map FCM/2/07/017 Date/ Time Map Taken 1/7/91 0935 hours0.0108 days <br />0.26 hours <br />0.00155 weeks <br />3.557675e-4 months <br /> Power Level N100%

Cycle Burnup 16.1 EFPD 253 NWD/MTU Boron Concentration 1139 ppm Control Rod Position Control Bank D at 209 steps withdrawn Maximum Pq : 1.8495 at Axial Loc. 34, lloriz.

Loc. C-12

. a Axial Loc. 34 Maximum /Z:

Maximum pin F 1.3862 at lloriz. Loc. C-12 AH Maximum error FAH ( r m pre ic e . a H r z. . c. E-09 Maximum Fq /K(Z) 1.9634 at Axial Loc. 39 Harimum % Reduction in Axial Flux None Difference (AFD) Wings Minimum % Margin to AFD Wings -0.8918% at Axial Loc. 39 R (Tech Spec 3/4.2.3) 0.9296 Total Reactot Coolant Flowrate 391,117 gallons / minute (Process Comp' iter)

Tov 41 Incore f.atil Offset +2.436%

T.acore Tilts %:

Upper Core Lower Core Quadrant 1: -0.624% Quadrant 1: -0.516%

Quadrant 2: 0.322% Quadrant 2: 0.786%

Quadrant 3: 0.319% Quadrant-3: 0.082%

Quadrant 4: -0.017% Quadrant 4: -0.351%

29

____ ________ _____________ - _ _