ML18150A135

Cycle 7 Core Performance Rept

ML18150A135
Person / Time
Site:	Surry
Issue date:	10/31/1984
From:	Brookmire T, Snow C, Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Varga S Office of Nuclear Reactor Regulation
References
VEP-NOS-12, NUDOCS 8411010242
Download: ML18150A135 (53)

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NOTICE -

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DEADLINE RETURN DATE RECORDS FACILITY BRANCH co SURRY UN-IT 1, CYCLE 7 CORE* PERFORMANCE

REPORT 94*11010242 841031 -

PDR ADOCK 05000280 P

P.DR.

VEP-NOS-12 NUCLEAR OPERATIONS DEPARTMENT Virginia Electric and Power Company

VEP-NOS-12 SURRY UN IT 1, CYCLE 7 CORE PERFORMANCE REPORT BY T. A. BROOKMIRE APPROVED BY:

C. :J.~

C. T. Snow Supervisor, Nuclear Fuel Operation Operations and Maintenance Support Subsection, Nuclear Operations Department Virginia Electric & Power Company Richmond, Virginia October, 1984

CLASSIFICATION/DISCLAIMER The data, techniques, information, and conclusions in this report have been prepared solely for use by the Virginia Electric and Power Company (the Company), and they may not be appropriate for use in situations other than those for which they were specifically prepared.

The Company therefore makes no claim or warranty whatsoever, express or implied,as to their accuracy, usefulness, or applicability. In particular, THE COMPANY MAKES NO WARRANTY OF MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE, NOR SHALL ANY WARRANTY BE DEEMED TO ARISE FROM COURSE OF DEALING OR USAGE OF TRADE, with respect to this report.. or any of the data, techniques,.

information, or conclusions in it. By making this report available, the Company does not authorize its use by others, and any such use is expressly forbidden except with the prior written ~pproval of the Company.

Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein.

In no event shall the Company be liable, under any legal theory whatsoever

.(whether contract, tort, warranty, or strict or absolute liability), for any property damage, mental or physical injury or death, loss of use of property, or other damage resulting from or arising out of the use, authorized or unauthorized, of this report or the data, techniques, information, or conclusions in it.

i

ACKNOWLEDGEMENTS The author would like to acknowledge the cooperation of the Surry Power Station personnel in supplying the basic data for this report.

Special thanks are due Mr. L. J. Curfman.

ii

TABLE OF CONTENTS SECTION TITLE PAGE NO.

Classification/Disclaimer i

Acknowledgements ii List of Tables iv List of Figures V

1 Introduction and Summary.

1 2

Burnup Follow 7

3 Reactivity Depletion Follow 16 4

Power Distribution Follow 18 5

Primary Coolant Activity Follow 40 6

Conclusions 44 7

References.

45 iii

LIST OF TABLES TABLE TITLE PAGE NO.

4.1 Summary of Incore Flux Maps for Routine Operation..... 23 iv

LIST OF FIGURES FIGURE TITLE 1.1 Core Loading Map 1.2 Movable Detector and Thermocouple Locations.

1.3 Control Rod Locations.

2.1 Core Burnup History 2.2 Monthly Average Load Factor 2.3 Assemblywise Accumulated Burnup: Measured and Predicted 2.4 Assemblywise Accumulated Burnup Comparison of Measured and Predicted 2.5 Sub-batch Burnup Sharing 2.6 Sub-batch Burnup Sharing 2.7 Sub-batch Burnup Sharing PAGE NO.

4 5

6 9

.:10

. 11 13 14.

15 3.1 Critical Boron Concentration versus Burnup - HFP-ARO 17 4.1 Assemblywise Power Distribution - Sl-7-11 25 4.2 Assemblywise Power Distribution - Sl-7-19 26 4.3 Assemblywise Power Distribution - Sl-7-62 27 4.4 Hot Channel Factor Normalized Operating Envelope 28 4.5 Heat Flux Hot Channel Factor, F6(Z) - Sl-7-11 4.6 Heat Flux Hot Channel Factor, F6(Z) - Sl-7-19 4.7 Heat Flux Hot Channel Factor, F6(Z) - Sl-7-62 4.8 Maximum Heat Flux Hot Channel Factor, F *P Q,

versus Axial Position......

4.9 Maximum Heat Flux Hot Channel Factor, F-Q(z),

versus Burnup.

4.10 Maximum Enthalpy Rise Hot Channel Factor, F-DH(N),

versus Burnup.

V 29 30 31 32 33 34

LIST OF FIGURES CONT'D FIGURE TITLE PAGE NO.

4.11 Target Delta Flux versus Burnup 35 4.12 Core Average Axial Power Distribution - Sl-7-11 36 4.13 Core Average Axial Power Distribution - Sl-7-19 37 4.14 Core Average Axial Power Distribution - Sl-7-62 38 4.15 Core Average Axial Peaking Factor, F-Z, vers11s Burnup 39 5.1 Dose Equivalent I-131 versus Time 42 5.2 I-131/I-133 Activity Ratio versus Time 43 vi

SECTION 1 INTRODUCTION AND

SUMMARY

On September 26, 1983, Surry l!nit 1 completed Cycle 7. Since the initial criticality of Cycle 7 on May 30, 1983, the reactor core produced approximately 70 x 10 6 MBTU, (11;984 Megawatt days per metric ton of contained uranium) which has resulted in the generation of approximately 6.6 x 10 9 KWHr gross (6.2 x 10 9 KWHr net) of electrical energy. Surry 1, Cycle 7 reached the end of full power reactivity at a core burnup of approximately 11,681 MWD/MTU.

End,of cycle operation was limited to a power level of 88% due to stuck control rod B-6.

The additional reactivity resulting from the power reduction allowed the core to operate for 303 MWD/MTU beyond the end of full power reactivity burnup value.

The purpose of this report is to present an analysis of the core performance for routine operation during Cycle 7. The physics tests that were performed during the startup of this cycle were covered in the Surry 1, Cycle 7 Startup Physics Test Report 1 and, therefore, will not be included here.

The seventh cycle core consisted of ten sub-batches of fuel:

a thrice-burned sub-batch from cycles 2, 3, and 4 (4C3), a twice-burned sub-batch from cycles 4 and 5 (6C3), a twice-burned sub-batch from Surry 2, Cycle 4 and Surry 1, Cycle 6 (S2/6B4), two twice-burned sub-batches from cycles 5 and 6 (7A2 and 7B2), two once-burned sub-batches from cycle 6 (8A and 8B), and three fresh sub-batches (9A, 9B, and S2/9B). The Surry 1, Cycle 7 core loading map specifying the fuel sub-batch identifications, fuel assembly locations, burnable poison locations and source assembly locations is shown in Figure 1.1. Movable detector locations *and 1

thermocouple locations are identified in Figure 1.2.

locations are shown in Figure 1.3.

Control rod Routine core follow involves performance indicators.

These are the analysis of four principal burnup distribution, reactivity depletion, power distribution, and primary coolant activity. The core burnup distribution is followed to verify both burnup symmetry and proper sub-batch burnup sharing, thereby, ensuring that the fuel held over for the next cycle will be compatible with the new fuel that is inserted.

Reactivity depletion is monitored to detect the existence of any abnormal reactivity behavior, to determine if the core is depleting as designed, and to indicate at what burnup level refueling will be required. Core power distribution follow includes the monitoring of nuclear hot channel factors to verify that they are within the Technical Specifications 2 limits, thereby ensuring that adequate margins to linear power density and critical heat flux thermal limits are maintained. Lastly, as part of normal core follow, the primary coolant activity is monitored to verify that the dose equivalent Iodine-131 concentration is within the limits specified by the Surry Technical Specifications 2,

and to assess the integrity of the fuel.

Each of the four performance indicators is discussed in detail for the Surry 1, Cycle 7 core in the body of this report. The results are summarized below:

1.

Burnup Follow - The burnup tilt (deviation from quadrant symmetry) on the core was no greater than +/-0.52% with the burnup accumulation in each sub-batch deviating from design prediction by no more than +/-2.3%.

2.

Reactivity Depletion Follow - The critical boron concentration, used 2

to monitor reactivity depletion, was consistently within +/-0.2% AK/K of the design prediction which is within the +/-1% AK/K margin allowed by Section 4.10 of the Technical Specifications.

3.

. Power Distribution Follow Incore flux maps taken each-month indicated that the assemblywise radial power distributions deviated

• -from the design predictions by an average percent difference of less than 2%.

All hot channel factors met their respective Technical Specifications limits.

4.

Primary Coolant Activity Follow The average dose equivalent

• iodine-131 activity level in the primary coolant during Cycle 7 was approximately O. 0788 µCi/gm.

This is approximately 8% of the Technical Specifications 2 operating limit for the concentration of radioiodine in the primary coolant.

In addition, the effects of fuel densification were monitored throughout the cycle. No densification effects were observed.-

3

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• SECONDARY SOURCE XXP

• BURNABLE POISON ASSEMBLY (XX-NUMBER OF RODS)

C. XDP -

DEPLETED BURNABLE POISON ROD CLUSTER (X*NUMBER OF RODS)

FUEL ASSEMBLY DESIGN PARAMETERS SUB*BATCH 4C3 6C3 7A2 782 IIA 88 9A 98 S2/684 INITIAL ENRICHMENT (W/0 U235) 3.325 2.902 2,901 3.393 3.217 3.399 3.589 3.605 3.203 ASSEMBLY TYPE 15X15 15X15 15X15 15X15 15X15 15X15 15X15 15X15 15X15 NUMBER OF ASSEMBLIES 22 II 10 6

12*

33 36 24 2

FUEL RODS PER ASSEMBLY 2011 2011 204 204 2011 204 204 204 204 INITIAL ENRICHMENT (W/0 U235)

ASSEMBLY IDENTIFICATION

. D02,D05 J05,J19 OA1,0A5 2A3,2A7 OBll,OB5 OC3,0C5 OD1*0D9 3D7-3D9 W14,W15 006,010 J33,J37 OA6,0A8 3A1,3AII OB7,0B8 OC6,0C7 100-109 400-409 D12,D13 J38,J39 1A2, 1A3 3A6,5A2 089, 1BO 1C0, 1C2 200-209 5D0-5D9 014,D22 J45,J51 1A5,1A6 182,183 1C4*1C7 3D0-306 6DO D25,D27 1A7,1A9 1811, 186 1C8,2C2 031-033 187,189 2C3,2C8 D39,U41 3C1,3C3 D42-0li4 3C4,3C8 046,048 3C9,IICO 049,050 4C2,IIC4 4C5,IIC7 IICll,4C9

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SECTION 2 BURNUP FOLLOW The burnup history for the Surry Unit 1, Cycle 7 core is graphically depicted in Figure 2.1.

The Surry 1, Cycle 7 core achieved a burnup of 11984 MWD/MTU. As shown in Figure 2.2, the average load factor for Cycle 7 was 72. 6% when referenced to rated thermal power (2441 MW(t)).

Radial (X-Y) burnup distribution maps show how the core burnup is shared among the various fuel assemblies, and thereby allow a detailed burnup distribution analysis. The NEWTOTE 3 computer code is used to calculate these assemblywise burnups. Figure 2. 3 is a radial burnup

• distribution map in which the assemblywise burnup accumulation of the core at the end of Cycle 7 operation is given. For comparison purposes, the design values are also given. Figure 2.4 is a radial burnup distribution map in which the percentage difference comparison of measured and predicted assemblywise burnup accumulation at the end of Cycle 7 operation is given. As can be seen from this figure, the accumulated assembly burnups were generally within +/-3.1% of the predicted values. In addition, deviation from quadrant symmetry in the core, as indicated by the burnup tilt factors, was no greater than +/-0.52%.

The burnup sharing on a sub-batch basis is monitored to verify that the core is operating as designed and to. enable accurate end-of-cycle sub-batch burnup predictions to be made for use in reload fuel design studies.

Sub-batch. definitions are given in Figure 1.1. As seen in Figures 2.5, 2.6, and 2.7, the sub-batch burnup sharing for Surry Unit 1, 7

Cycle 7 followed design* predictions very closely with each sub-batch deviating by no more than 2. 3%

from design.. Symmetric burn up in conjunction with good agreement between actual and predicted assemblywise burnups and sub-batch burnup sharing indicate that the Cycle 7 core did deplete as designed.

8

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PERCENT 100 90 60 70 60 50 40 30 20 10 0

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4 MONTH THERMAL ENERGY GENERATION !N MONTHIMWHTl LOAD FACTOR:---------------------------------------------

AUTHORIZED POWER LEVEL IMWTI X HOURS !N MONTH

!EXCLUDES REFUELING OUTAGES!

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L K

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---

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8.931 13.88132.59115.531 36.701 15.581 34.191 33.611 34.191 15.581 36.701 15.531 32.591 13.881 8.931 I 11.351 37.131 32.991 34.401 15.661 43.311 33.511 24.941 34.051 43.241 15.231 33.321 32.551 37.471 11.091 I 11.231 37.261 32.811 33.931 15.391 43.T11 33.941 24.841 33.941 43.111 15.391 33.931 32.811 37.261 11.231 1 9.061 14.011 32.331 15.701 37.071 15.531 33.811 33.421 33.811 15.501 36.751 15.151 32.641 14.031 8.931 I

8.931 13.881 32.591 15.531 36.701 15.581 34.191 33.611 34.191 15.581 36.701 15.531 32.591 13.881 8.931 R

I 28.311 14.351 39.021 30.611 28.821 15.321 42.491 15.461 28.911 30.00I 38.931 14.591 29.131 I 28.401 14.261 38.711 30.251 29.021 15.661 42.581 15.661 29.021 30.251 38.711 14.261 28.401 I 30.981 13.041 15.41 I 35. 751 30.00I 36.451 15. 141 36.451 30.261 35.591 15.421 13.21 I 30.401 I 30.631 13.00l 15.381 35.351 30.411 36.621 15.4.41 36.621 30.411 35.351 15.381 13.00I 30.631 1 29.971 37.441 15.641 38.531 15.351 33.431 15.181 38.521 15.311 37.931 29.801 I 29.761 37.581 15.401 38.701 15.561 33.761 15.561 38.701 15.401 37.581 29.761 I 31.431 13.501 14.501 32.421 32.371 32.211 14.141 12.921 29.741 I 30.331 12.981 14.251 32.571 32.811 32.571 14.2'.il 12.981 '30.331 1 30.911 28.521 13.921 37.231 13.631 28.041 30.831 I 30.701 28.061 13.831 37.231 13.831 28.061 30.701 I

9.111 11.231 8.791 I

8.891 11.111 8.891 p

N M

L K

J H

G F

E D

C B

11 A

2 3

4 5

6 7

8 9

10 11 12 13 14 15

2 3

II 5

6 7

9 10 11 12 13 111 15 BATCH 4C3 6C3 7A2 7B2 BA BB 9A 9B It Figure 2.4

• SURRY 1 s CYCLE 7 ASSEMBL YWISE ACCUMULATED BURN UP COMPARISON OF MEASURED' AND PREDICTED (1000 MWD/MTU)

L II J

G F

£ 0

e I

A I

a.IOI 11.021 8.831 I -o.961 *1.361 -o.671 I

MEASURED I

I M/P I DIFF I I 30.331 21.2111 13.641 37.021 13,511 27.741 30.151 I *1.201 0.641 *1.361 *0,561 *1,791 *1,141 0.511 I 30.471 12.221 111.041 32.961 32.671 31.981 111.331 13.061 29.811 I

0.11111 -5.911 *1.471 1.201 -0.11111 -1.121 0.5111 0.571 *1.731 I 29.751 37.1121 111.651 38.IIOI 15,1101 33.63115.1191 39.011 15.221 37.551 29.771 I -0.031 *0.1131 *11.871 *0.751 *1.001 *0.381 *0.1101 0.811 *1,181 *0.091 0.0111 I 30.711 12.1111 15.021 35.301 29.971 36.1171 15.1131 36.721 30.651 34.9111 111.671 12.591 30.721 I o.261 -1.911 -2.351 -0.141 -1.1161 -0.1111 -0.081 0.211 o.781 -1.191 -11.651 -s.131 o.301 I 28.311 111.301 38.391 29.171 28.48115.521112,601 15.581 29.321 30.081 38.321 13.701 27.671 I *0.311 0.301 *0.831 *1,271 *1.861 *0.891 0,041 *0.501 1.031 *0.571 -1.011 *3,191 *2.571 2

3 6

I 9.061 13,981 33.231 15,621 36.1191 15.1101 311.1161 33.791 34,581 15.591 36.631 15.0111 32,281 13.211 a.731 7

I 1,531 0.741 1.981 0.561 *0.561 *1,171 0.771 0.531 1.131 O.ll<!I *0.191 *3,1111 -0,921 -4.281 *2,221 I 11.351 37.131 32.991 34.1101 15.661 113.311 33.511 211.9111 311.051 43,2111 15.231 33.321 32,551 37.1171 11.091 a

1 1. t1 I -o.351 0.5111 1.401

1. 1111 o.1181 -1.261 o.381 o.331 o.321 -1.0111 -1.911 -0.801 o.571 -1.231 I 9.061 111.011 32.331 15.101 37.071 15.531 33.111 33.421 33.111 15.501 36.751 15,151 32.6111 111.031 8.931 9

I 1.521 0.931 *0.781 1.101 1.001 *0,361 *1.121 *0.571 *1.111 *0.511 0.141 *2.1121 0.171 1.111 0.081 1 28.31 I 14.351 39.021 30.611 28.821 15.321 112.1191 15.1161 28.i,*11 30.00I 38.931 111.591 29.131 10 I -o.331 0,681 0.121 1.111 -0.101 *2.151 -0.211 -1.:s11 -o.401 -0.821 o.581 2,351 2.561 I 30.981 13.0111 15.1111 35.751 30.001 36.1151 15.1111 36.1151 30.261 35.591 15,421 u.211 30.401 I

1.1111 0.311 0.111 1.111 *1.361 *0.451 -1,951 *0.471 *0.521 0.661 o.261 1,611 *0.731 I 29.971 37.1141 15.641 38.531 15,351 33.431 15.181 38.521 15.311 37.931 29.101 I

0.711 *0.371 1.551 *0.1131 *1.351 *0.961 -2.391 *0.1151 *0.561 0.921 0.151 11 12 I 31.1131 13.501 14.501 32.421 32,371 32.211 14.141 12.921 29. 7111 I

3.621 11.001 1,121 -0.1111 -1.331 -1.121 -0.801 -0.521 -1.9111 13 I STANDARD DEV I I

"'0.99 I

It M

I 30.911 21.521 13.921 37.231 13.631 21.0111 30.831 I

o.691 1,651 o.631 -0.011 *1.461 -0.051 0.1131 I

9,111 11.231 8,791 I

2.521 o.5111 *1.131 L

K J

G F

[

BURNUP SHARING (MWD/MTU)

CYCLE 2 I CYCLE 3 I CYCLE 4 I CYCLE 5 I CYCLE 6 I CYCLE 7 I TOTAL 5.72 6.54 14.36 8.81 35.43 8.33 16.48 11.82 36.63 17.08 7.34 7.62 32.04 12.66

• 16.84 11.54 41.04 20.33 9.73 30.06 18.06 13.58 31.64 12.70

12. 70 14.95 14.95 D

C I ARITHMETIC AVG I IPCT DlfF c *0.321 14 I AVG ABS PCT I I DIFF = 1.09 I a

A BURNUP TILT 15 S2/6B4 S2/10. 96 15.38 4.61 30.95 NW= -0.15 NE= -0.36 SW=

0.52 SE= -0.01 S2/9B 14.27 14.27 CORE AVERAGE 11.98 12

44000 40000 36000 s u 32000 B

B

[3C1 A 28000 T

C

,r-H 24000 B

u, R

t,J 20000 ~

u p

M 16000 w

D I M 12000 T

u 8000 4000 0 -~

I 0

Figure 2.5 SURRY UN IT 1 - CYCLE 7 SUB-BATCH BURNUP SHARING SUB-BATCH SYMBOL -

~-

i..e--

A__.~

~

• I

~

A.

4C3 DIAMOND

. ~

762 SQUARE

-c:I'"""

-- r"

~

A--

~

_-a-i,..,---

I SA TRIANGLE

_.r

~

~

~

A-I 2000 4000 6000 8000 10000 CYCLE BURNUP MWD/MTU 13 9A STAR

~

~

I 12000

~

~

~

I 14000

Figure 2.6 SURRY UNIT 1 - CYCLE 7 SUB-BATCH BURNUP SHARING SUB-BATCH

• sYMBOL 6C3 DIAMOND 7A2 SQUARE 98 TRIANGLE 44000-+---+--+-----t----+---+---+---+---+---+---+--+--+-----t---+

40000-t-----+---+----!----+----+---+---+---+---+---+--+---+----!-----t-36000-t---+---+--t---+--+--+---+---+---+--+----!::,eoo=__..::;.....+ _ __..;t----+

s u 32 000-t---+---+--t---+--+--+--""P!J"--~f-7 --+--+---+-=::::ool~-;...._t-,.....--+

B

-~

_e,..-<::I B

Ar

--r=

A 2 BOO O-t---+---+---21?""""'-A!l"::;;...;;..+--=t--:=::-et"""F--a--;;......;--+----+---+--+---+--t--t----+

T C

~

~_...

H B u R

24000~~;;;_-+---+---t---+--+--+---t---+---+--+---+---I--I----+

N 2 0000-+---+-----1....---+--+----+-----+------+----+-----..----+

u p

M 16000-t---+---+-----tt---+--+---+---+---+---+--+---+---+----::_--::>"""=t-......---+

w D

I

~-

M 12000-t---+---+-----tf---+--+---+---+---+---+~~a~'+--+---+---+---+

T u 8000-+---+---+-----tf---+--+--+,~./"~-~--+---+---+---+--t-----,t----+

-~

4000-+---+---+--__.t.""'P"~_.....;...._+--+---+--+---+---+--+---+---+-----tf----+

o-~6........,....,.+......... ~. -~ *..,....,..-~.~-~--~---~-+ *.,........,,......;...........,....,..j""T"'e'.,....,...;'-:--,-~;.,.-.""T"'e'.;...,-,~......-.,...,...,...;.......,......,........,.""T"'e'...+-~...-+-

• . . I

'. I

'.' I

.'.'I 0

2000 4000 6000 8000 10000 12000 14000 CYCLE BURNUP MWD/MTU 14

.~

SURRY UN IT 1 - CYCLE 7 SUB-BATCH BURNUP SHARING SUB-BATCH SYMBOL 8B DIAMOND 52/664 SQUARE Figure 2. 7 52/96 TRIANGLE 36000-+----+----+-----+----+----+----+----+----+-----t-----t----+----+----+-----+

s U 32000 B

B A 28000 T

C H

B u R

N 2 00 00-+----+.k'"~....::J-.,.....---+-----+-----+---4---4---4---1-----1----I----I----I----+

u p

M 1 so o 0-+----+---+---+-----+-----+---4---4-----+---1-----1----1----1--_,,,---,,i..,.:::~;.......+

w D

I M 1 2 o o o-+----+---+---+-----+---4---4---4-----+---1-_,,,---,,,~..,~-=--+---+---+---+

T u 0-~

I I

, *** ',. I I

I

. .'I '

o 2000 4000 6000 8000 10000 12000 14000 CYCLE BURNUP MWO/MTU 15 r

SECTION 3 REACTIVITY DEPLETION FOLLOW The primary coolant critical boron concentration is monitored for the purposes of following core reactivity and to identify any anomalous reactivity behavior. The FOLLOW4 computer code was used to normal fa~

"actual" critical boron concentration measurements to design conditions*

taking into consideration control rod position, xenon and.samarium concentrations, moderator temperature, and power level. The normalized critical boron concentration versus burnup curve for the Surry 1, Cycle 7 core is shown in Figure 3.1. It can be seen the.t the measured data compare to within 20 ppm of the design prediction.

This corresponds to approximately +/-0.19% AK/K which is within the +/-1% AK/K criterion for reactivity anomalies set forth in Section 4.10 of the Technical Specifications 2

• In conclusion, the trend indicated by the critical boron concentration verifies that the Cycle 7 core depleted as expected.

16

C R

1 T

1400 1200 1 I 000 C

A L

B 0

R 600 0

N C

0 N

C 600 E

N T

R A

T 1

400 0

N p

p M

200 0 -

I 0

\\.,

)Q<

Figure 3. 1 SURRY UNIT 1 - CYCLE 7 CRITICAL BORON CONCENTRATION vs. BURNUP HFP, ARO X

MEASURED PREDICTED

-*~...

..,~

~

!If(

' ~

~

~'-

~

'~.

~ r.,.,

~.

~

• ,,. I

...... I 2000 4000

,.. I

.. I 6000 6000 CYCLE BURNUP !MWO/MTUl 17

~ ~

~

• I I

• * * * * *

• I 10000 12000 14000

SECTION 4 POWER DISTRIBUTION FOLLOW Analysis of core power distribution data on a routine basis is necessary*to verify that the hot channel factors are within the Technical Specifications limits and to ensure that the reactor is oper'ating without any abnormal conditions which could cause an "uneven" burnup distribution. Three-dimensional core power distributions are determined from movable detector flux map measurements using the INCORE 5 computer program. A summary of all full-power flux maps taken since the completion of startup physics testing for Surry 1, Cycle 7 is given in Table 4.1.

Power distribution maps were generally taken at monthly intervals with additional maps taken as needed.

During an approach to criticality in June, 1984, control rod B-6 became stuck at approximately 58-60 steps (1/4 withdrawn).

When all attempts to move the control rod fa;i.led,. Vepco conducted a safety evaluation to determine whether power operation could resume with the stuck control rod.

After all relevant safety analyses were found to be satisfactory for power operation up to 88% of full power, the reactor was slowly taken to approximately 79% power (this additional power reduction provided operational flexibility with respect to the control rod insertion limits). Several full-core flux maps were acquired during the power escalation to monitor the power distribution.

The first full-core map at 79% power (map number 30 of Table 4.1) indicated a 3% quadrant power tilt ratio.

The tilt ratios continued to decrease during cycle operation to an EOC value of less than 2%, as indicated in Table 4.1. All 18

hot channel factors were determined to be within their Technical Specifiaction limit during reduced power operation.

Radial (X-Y) core power distributions for a representative series of incore flux maps are given in Figures 4.1 through 4.3. Figure 4.1 shows a power distribution map that was taken early in cycle life. Figure 4. 2 shows a power distribution map that was taken near mid-cycle burnup.

Figure 4.3 shows a map that was taken late in Cycle 7 lif~. The radial power distributions were taken under equilibrium operating conditions with the unit at approximately full power until June, 1984, and at approximately 79% power thereafter.

In each case, the measured relative assembly powers were generally within 4.3% of the predicted values with an average percent difference of no greater than 1.6% which is considered good agreement. In addition, as indicated by the INCORE tilt factors, the power distributions were essentially symmetric for all cases prior to the misalignment of control rod B-6.

An important aspect of core power distribution follow is the monitoring of nuclear hot channel factors. Verification that these factors are within Technical Specifications limits ensures that linear power density and departure from nucleate boiling ratio (DNBR) limits will not be violated, thereby providing adequate thermal margins and maintaining fuel cladding integrity. The Technical Specifications Limit on the axially dependent heat flux hot channel factor FQ(Z) was (2.18/P) x K(Z), where K(Z) is the hot channel factor normalized operating envelope and P is the fraction of rated thermal power. Figure 4. 4 is a plot of the K(Z) curve associated with the 2.18 FQ(Z) limit. The axially dependent heat flux hot channel factors, FQ(Z), for a representative set of flux maps are given in Figures 4.5 through 4.7. It should be noted that the FQ(Z) measured values and FQ(Z) x K(Z) limits depicted in Figure 4. 7 have 19

been normalized from 79% power to 100% power for comparison with Figures 4.5 and 4.6. Throughout Cycle 7, the measured values of FQ(Z) were within the Technical Specifications limit. A summary of the maximum values of axially-dependent heat flux hot channel factors measured during Cycle 7 is given in Figure 4.8.

Figure 4.9 shows the maximum values for the heat flux hot channel factors measured as a function of burnup during Cycle 7.

As can be seen from this figure, there was approximately 15% margin to the limit at the beginning of the cycle, which remained fairly constant throughout cycle operation.

The increase in the limit is due to the power dependence of the Technical Specifications limit and the reduced-power operation with stuck control rod B-6.

The value of the enthalpy rise hot channel factor, F-~H, which is the ratio of the integral of the power along the rod with the highest integrated power to that of the average rod, is routinely ~allowed. The Technical Specifications limit for this parameter is set such that the departure from1 nucleate boiling ratio limit will not be violated.

Additionally, the F-delta H limit ensures that the value of this parameter used in the LOCA-ECCS analysis is not exceeded during normal operation.

The Cycle 7 limit for the enthalpy rise hot channel factor was 1. 55 x

[l+0.3(1-P)], where Pis the fractional power level. The maximum values of F-del ta H versus burnup are shown in *Figure 4.10. Again, the increase in the limit is due to reduced power operation with stuck control rod B-6.

The Technical Specifications require that target delta flux* values be determined periodically. The target delta flux is the delta flux which would occur at conditions of steady-state operation (i.e., full power Pt-Pb

,':Delta Flux =

2441 X 100 where Pt= power in top of core (MW(t))

Pb= power in bottom of core (MW(t))

20

prior to June, 1984 and 79%

power thereafter), all rods out, and equilibrium xenon. Therefore, the delta flux is measured with the core at or near these conditions and the target delta flux is established at this measured point. Since the target delta flux varies as a function of burnup, the target value is updated monthly. Operational delta flux-limits are then established about this target value. By maintaining the value of delta flux relatively const~nt, adverse axial power shapes due to xenon redistribution are avoided. The plot of the target delta flux versus burnup, given in Figure 4.11, shows the value of this parameter to have been approximately -0. 5% at the beginning of Cycle 7. By the middle of the cycle the value of delta flux had shifted to approximately -3% where it remained until control rod B-6 became inoperable.

During stuck rod operation, the value of delta flux initially increased to +2% before decreasing to 0% at EOG.

This power history can also be observed in the corresponding core average axial power d~stribution for a representative series of maps given in Figures 4.12 through 4.14. In Map Sl-7-11 (Figure 4.12) taken at approximately 1612 MWD/MTU, the axial power distribution had a slightly peaked cosine shape with a peaking factor of 1.16. In Map Sl-7-19 (Figure 4.13) taken at approximately 6,639 MWD/MTU, the axial power distribution had flattened somewhat with an axial peaking factor of 1.14. Finally, in Map Sl-7-62 (Figure 4.14) taken at approximately 10,602 MWD/MTU, the axial power distribution remained fairly flat with the axial peaking factor of 1.12. The history of F-Z during the cycle can be seen more clearly in a plot of F-Z versus burnup given in Figure 4.15.

In conclusion, the Surry 1,

Cycle 7

core performed very satisfactorily with power distribution analyses verifying that design predictions were accurate and that the values of the FQ(Z) and F-delta H 21

hot channel factors were within the limits of the Technical Specifications.

22

I I

I

~

w TABLE 4.1 SURRY UNIT 1 - CYCLE 7

SUMMARY

OF INCORE FLUX MAPS FOR ROUTINE OPERATION I

1 2

I I

I I

BURNI I

F-Q(T) HOT F-DH(N) HOT ICORE F(Z)

I 4

I I

I UP I IBANK CHANNEL FACTOR CHNL. FACTOR I

MAX I

3 I QPTR I AXIAL NO. I MAP NO.

DATE MWD/IPWRI D

I I

I IF(XY)I I

OFF OF I 8

11 12 13 14 18 19 22 23 24 MTU 1(%)1STEPSIASSYIPINIAXIALI ASSYIPINIF-DH(N)IAXIAL F(Z) I IMAX ILOCI SET THIMI I

I I

I I

I POINT F-Q(T)

I I

IPOINT I

I I

I

(%)

BLESI I __ I_I __ I_I_I __

I_I I ____ I __ I __ I_I

_I 7-19-83 807 100 223 ROB Ml 35 1.811 L13 LM 1.473 34 1.159 1.422 1.008 SWl-0.200 49 (5)

I 8-11-83 1612 100 227 L13 LM 44 1.803 L13 LM l.476 34

1. 152 1.429 1.008 SWl-0.384 49 I

9-16-83 2815 100 227 L13 LM 45 1.807 L13 LM 1.473 34 1.144 1.416 1.012 SEl-1.308 46 I

10-27-83 3650 100 228 L13 LM 45 1.812 L13 LM 1.479 45 1.138 1.427 1.006 SWl-1.593 46 I

11-16-83 4329 100 228 L12 LD 44 1.828 L12 LO

1. 491 45 1.147 1.436 1.013 SWl-2.525 45 (6,7)

I 1-05-84 5730 100 227 L12 LD 45

1. 831 Ll2 LD 1.485 45 1.151 1.429 1.006 SEl-3.338 39 I

2-03-84 6639 100 228 L12 LO 45

1. 813 L12 LD 1.486 45

1. 144 1.424 1.006 SWl-3. 152 38 (8) j 1. 143 I

3-20-84 7630 100 226 L12 LO 45

1. 787 LOB GH 1.479 45 1.441 1.006 SEl-2.941 39 I

I 5-09-84 8782 100 227 L12 LD 45 1.797 LOB GH 1.480 46 11. 148 1.442 1.006 SEl-4.012 39 I

I 6-12-84 9593 100 227 L12 LO 46

1. 785 LOB GH 1.492 46 11. 141 1. 450 1.005 SEl-3.844 39 I

r I

NOTES: HOT SPOT LOCATIONS ARE SPECIFIED BY GIVING A~SEMBLY LOCATIONS (E.G. H-8 IS THE CENTER-OF-CORE ASSEMBLY),

FOLLOWED BY THE PIN LOCATION ~DENOTED BY THE "Y" COORDii-iA1*i: WITH THE FIFTEEN ROWS OF FUEL RODS LETTERED A THROUGH RAND THE 'X" COORDINATE DESIGNATED IN A SIMILAR MANNER),

IN THE 112 11 DIRECTION THE CORE IS DIVIDED INTO 61 AXIAL POINTS STARTING FROM THE TOP OF THE CORE.

1. F-Q(T) INCLUDES A TOTAL UNCERTAINTY OF 1.08.

2, F-DH(N) INCLUDES A MEASUREMENT UNCERTAINTY OF 1.04.

3. F(XY) IS EVALUATED AT THE MIDPLANE OF THE CORE.

4. QPTR -

QUADRANT POWER TILT RATIO.

5. MAPS 9 AND 10 WERE QUARTER CORE MAPS USED FOR CALIBRATION OF THE EXCORE DETECTORS.

6. MAPS 15 AND 16 WERE QUARTER CORE MAPS USED FOR CALIBRATION OF THE EXCORE DETECTORS.

7. MAP 17 WAS NOT USED DUE TO INVALID DATA.

8. MAPS 20 AND 21 WERE QUARTER CORE MAPS USED FOR CALIBRATION OF THE EXCORE DETECTORS.

MAP NO.

(9, 10) 30 (11,12) 62

( 13) 66 TABLE 4.1 (CONT'D)

I 1

2 I

I I

I BURNI I

F-Q(T) HOT F-DH(N) HOT ICORE F(Z)

I 4

I I

UP I IBANK CHANNEL FACTOR*

CHNL.FACTOR I

MAX I

3 QPTR I AXIALI NO. I DATE MWD/IPWRI D I I

I F(XV)

I OFF I OF I MTU 1(%)1STEPSIASSVIPINIAXIALI ASSVIPINIF~DH(N)IAXIALI F(Z)I MAX IL I SET ITHIMI I

I I

I I

IPOINTIF-Q(T)

I I

I POINTI I

10 I

(%) I BLESI 1--1-1--1-1-1--1

_I_I 1--1--1--

IC_I I_I I

I I

I I

I 6-22-841 96451 791 223 I L08 I EHi 12 I 1.869 L08 I EHi 1.545 I 12 11

• 158 I 1

• 507 1.0313ISWI 2.3611 39 I I

I I

I I

I I

I I

I I

I I

I I

I 8-06-841106021 791 225 I J061 DGI 12 I 1. 819 L081 DGI 1.527 I 12 11.12411.481 1.0226ISWl+0.2591 39 I I

I I

I I

I I

I I

I I

I I

I I

I 9-10-841115571 791 228 I J061 DGI 12 I 1.807 L081 DGI 1.522 I 12 I 1.12111.488 1.0195ISWl+0.089I 39 I I

I I

I I

I I

I I

I I

I I

I I

I

9.

MAP 25 WAS A QUARTER CORE MAP.

10. MAPS 26 THROUGH 29 WERE POWER ASCENSION MAPS FOR OPERATION WITH CONTROL ROD B-6 STUCK AT 58 STEPS.

11. MAPS 31 THROUGH 58 WERE DAILY SURVEILLANCE MAPS TAKEN TO MONITOR QUADRANT POWER TILT WITH INOPERABLE EXCORE POWER RANGE NUCLEAR INSTRUMENTATION CHANNEL N-43;

12. MAP 59 WAS A 49% POWER FULL-CORE MAP USED FOR CALIBRATION OF THE EXCORE DETECTORS.

MAPS 60 AND 61 WERE QUARTER-CORE MAPS USED FOR CALIBRATION OF THE EXCORE DETECTORS.

13. MAPS 63 THROUGH 65 WERE WEEKLY POWER TILT SURVEILLANCE MAPS WITH CONTROL ROD B-6 STUCK AT 58 STEPS.

R Figure 4. 1 p

H PREDICTED ttEASURED SURRY UNIT 1 - CYCLE 7 ASSEMBLYWISE POWER DISTRIBUTION Sl-7-11 L

K J

H G

F E

D

• 0.77. 0.98. 0.77.

• 0.77. 0.98. 0.76.

0.38 0.66 1.17 0.97 1.17 0.66 0.38 0.35 0.65. 1.15. 0.95. 1.15. 0.65 0.38

• 0.37. 1.07. 1.15. 1.13. 1.12. 1.13. 1.15. 1.07. 0.37 *

• 0.35. 1.01. 1.12. 1.12. 1.10. 1.11. 1.14. 1.07. 0.38.

C

8.

PREDICTED ttEASURED

• 0.38. 0.71. 1.29. 1.04. 1.25. 1.10. 1.25. 1.04. 1.29. 0.71. 0.38 *

• 0.36. 0.68. 1.22. 1.01. 1.22. 1.07. 1.23. 1.04. 1.28. 0.70. 0.38.

. 0.38. 1.08. 1.28. 1.01. 1.19. 0.97. 1.22. 0.97. 1.19. 1.01. 1.28. 1.08. 0.38 *

• 0.38. 1.07. 1.27. 1.00. 1.16. 0.95. 1.20. 0.97. 1.19. 1.02. 1.25. 1.08. 0.39.

• 0.66. 1.15. 1.04. 1.18 1.26. 1.26. 0.94. 1.26. 1.26. 1.18. 1.04. 1.15. 0.66 *

• 0.67. 1.16. 1.04. 1.17. 1.24. 1.25. 0.95. 1.28. 1.26. 1.18. 1.03. 1.16. 0.68.

A 0.78 1.17 1.13 1.25 0.96. 1.25 ** 1.13 1.13 1.13 1.25 0.96. 1.25 1.13 1.17 0.78 l

2 3

4 5

6

• 0.81 1.22 1.17 1.26 0.95. 1.23. 1.14 1.15 1.16 1.27 0.97. 1.25 1.15 1.20 D.80 7

• 0.99. 0.98. 1.12. 1.09. 1.21. 0.93. 1.11. 1.23. 1.11. 0.93. 1.21. 1.09. 1.12. 0.98. 0.99 *

• 1.04. 1.02. 1.16. 1.11. 1.19. 0.93. 1.08. 1.26. 1.13. 0.95. 1.22. 1.10. 1.13. 1.02. 1.02.

8

• 0.78. 1.17. 1.13. 1.25. 0.96. 1.25. 1.13. 1.13. 1.13. 1.25. 0.96. 1.25. 1.13. 1.17. 0.78 *

• 0.81. 1.21. 1.16. 1.21. o.99. 1.26. 1.12. 1.14. 1.15. 1.25. o.96. 1.23. 1.15. 1.21. 0.80.

9

• 0.66. 1.15. 1.04. 1.18. 1.26. 1.26. D.94. 1.26. 1.26. 1.18. 1.04. 1.15. 0.66 *

• 0.68. 1.17. 1.05. 1.19. 1.26. 1.25. 0.94. 1.26. 1.26. 1.17. 1.02. 1.16. 0.69.

• 0.38. 1.08. 1.28. 1.01 1.19. 0.97. 1.22. 0.97. 1.19. 1.01. 1.28. 1.08 0.38 *

• 0.38. 1.09. 1.29. 1.00. 1.18. 0.96. 1.20. D.96. 1.17. 1.01. 1.28. 1.08. 0.39.

0.38 0.71 1.29. 1.04 1.25 1.10 1.25 1.04 1.29 D.71 0.38 0.38 0.71 1.28. 1.03 1.23 1.08 1.22 1.03. 1.27 0.70 0.38 o.37. 1.01. 1.15. 1.13. 1.12. 1.13. 1.15. 1.01. o.37 0.38. 1.10. 1.16. 1.11. 1.10. 1.11. 1.12. 1.05. D.37

• 0.38. 0.66. 1.17. 0.97. 1.17. 0.66

• 0.38 *

• 0.39. 0.68. 1.19. 0.97. 1.16

• 0.65. 0.37.

0.77 0.98 0.77 -

• 0.80. 1.01. 0.77.

Standard Deviation= 1.348 Average% Diff. = 1.6

SUMMARY

MAP NO: Sl-7-11 DATE:

8/11/83 CONTROL ROD POSITIONS:

F-Q(T) = 1.803 D BANK AT 227 STEPS F-DH ( N ) = 1. 4 7 6 F( Z)

= 1. 152 F(XY)

= 1.429 BURNUP = 1612 MWD/MTU 25 POWER:

100%

QPTR:

NW 0.9909 I NE 1.0041

---

1----------

sw 1.0076 I SE 0.9973 A.O =

-0.38(%)

10 11 12 13 14 15

R Figure 4.2 p

N SURRY UN IT 1 - CYCLE 7 ASSEMBLYWISE POWER DISTRIBU_TION Sl-7-19 11 L

K H

6 F

E D

C B

PREDICTED HEASURED

, 0,72, 0,90. 0,72.

. 0.72. 0.89. 0.72,

PREDICTED 11EASURED

• 0,40. 0.67. 1.13. 0.92. 1.13. 0,67. 0.40 *

• 0,38. 0.67. 1.13. 0,91

• 1.12

• 0.68. 0.41 *

• 0.39. 1.08. 1.20. 1.11. 1.09. 1.11, 1.20. 1.08. 0.39,

• 0.37. 1,03. 1.18. 1.11, 1.07. 1.10

• 1.22

• 1,10. 0,40 *

• 0.39. 0,72. 1.27. 1,04. 1.31. 1.11. 1.31. 1.04. 1.27. 0.72. 0.39 *

• 0.39. 0.70. 1.22. 1.02. 1.30. 1.09. 1.30. 1,05. 1.28. 0.72. 0.40.

• 0.40

• 1.08

• 1.27

• 0.98

• 1.15

• 1.00

• 1.31

• 1.00

• 1.15 * ;0.98

• 1.27

• 1.08

• 0.40 *

• 0.39. 1.06, 1.25. 0.98. 1.13. 0.99. 1.30. 1.00. 1.17 *. 0.99. 1.23. 1.08. 0.41.

• 0,67. 1.20. 1.03. 1,14. 1.22. 1.32. 0.9~. 1.32. 1,22. 1.14. 1.83. 1.20. 0.67 *

. 0.67. 1.19, 1,03. 1.13. 1.21. 1,31. 0.97. 1.32. 1,24 *. l.13. 1.02. 1.19. 0.68.

A

• 0.72. 1.14. 1.11. 1.31

• 0.99. 1.31. 1.11. 1.09. 1.11. 1.31. 0.99. 1.31. 1.11. 1.14. 0.72

• l 2*

3 4

5 6

• 0.73. 1.15. 1.12. 1.31. 0.98. 1.29. 1.11. 1.09. 1.12. 1.32. 0.98. 1.29. 1.11. 1.13. 0.73.

7

• 0.90. 0.93. 1.09. 1.11. 1.31. 0.97. 1.08. 1.16. 1.08. 0.97. 1.31. 1.11. 1.09. 0.93. 0.90 *

• 0.91. 0.93. 1.09. 1.12. 1.33. 0.98. 1.03. 1.17. 1.08. 0.97. 1.30. 1.09. 1.09. 0.94. 0.91.

8

• 0.72. 1.14. 1.11. 1.31. 0.99. 1.31. 1.11. 1.09. 1.11, 1.31. 0.99. 1,31. 1.11. 1.14. 0.72 *

• 0.73. 1.14. 1.11. 1.32. 1.01. 1.31. 1.07. 1.07. 1.11. 1.31. 0.99. 1.29. 1.14. 1.16. 0.73.

9

• 0.67. 1.20. 1.03. 1.14. 1.22. 1.32. 0.97. 1.32. 1.22. 1.14. 1.03. 1.20. 0.67 *

• 0.67. 1.20. 1.04. 1.16. 1.20. 1.27. 0.95. 1.30. 1,22. 1.15. 1.06. 1.25. 0.70.

• 0.40. 1.08. 1.27. 0.98. 1.15. 1.00. 1.31. 1.00. 1.15. 0,98. 1.27. 1.08. 0.40 *

• 0.40. 1,09. 1.28. 1.00. 1.11. 0.97. l.~8. 0.98. 1.14. 1.00. 1.29. 1.11. 0.41.

• 0.39. 0.72. 1.27. 1.04. 1.31. 1.11. 1.31. 1.04. 1.27. 0.72. 0.39 *

• 0.40. 0.73. 1.34. 1.02. 1.28. 1.08. 1.28. 1.04. 1.28. 0.72. 0.40.

• 0.39. 1.08. 1.20. 1.11. 1.09. 1,11. 1.20. 1.08. 0.39 *

• 0.41. 1.13, 1.22, 1.09. 1,07, 1,10. 1,19, 1.08. 0.40.

• 0.40. 0.67. 1.13. 0.92. 1.13. 0,67. 0.40 *

• 0,42. 0.70. 1.18. 0.93. 1.13. 0.67. 0.40.

• 0.72

• 0.90

• 0.72 *

• 0.75. 0.92. 0.72.

Standard Deviation = 1.285 Avg.% diff. = 1.5

SUMMARY

MAP NO: 51-7-19 QATE:

2/ 3/84 CONTROL*ROD POSITIONS:

F-Q(T) = 1. 813 D BANK AT 228 STEPS F-DH( N) = 1.486 F(Z)

= 1.144 F(XY)

= 1.424 BURNUP = 6639 26..

MWD/MTU POWER:

100%

QPTR:

NW 0.9915 I NE 1.0003

---

1----------

sw 1.0058 I SE 1.0025 A.O =

-3.15(%)

10 11 12 13 14 15

Figure 4.3.

II p

N PREDICTED HEASURED SURRY UN IT 1 - CYCLE 7 ASSEMBLYWISE POWER DISTRIBUTION Sl-7-62 L

K J

H 6

• 0.70

• 0,86

• 0.70 *

, 0.72. 0.89. 0.72.

F E

0,41 0.69 1.13 0.90 1,13 0.68 0.41 0.40 0.70. 1.15. 0.92. 1.14. 0.69 0.42 D

• 0.41. 1.09. 1.25. 1.11. l,08. 1.11. 1.23. 1.07. 0.40 *

• 0.39. 1.06. 1.24. 1.13. 1,08. 1.11. 1.25. 1.10. 0.41.

C B

PREDICTED HEASURED

• 0,41. 0.73. 1.28. 1.05. 1.36, 1.13. 1.36. 1.03. 1.25. 0.70. 0.37 *

• o.41. 0.12. 1.2s. 1.04. 1.37. 1.13. 1.36. 1.05. 1.25. o.69. o.37.

• 0.42. 1.10. 1.28. 0,98. 1.14. 1.03. 1.38. 1.02. 1.12. 0.95. 1.20. 0.96. 0.32 *

• 0.41. 1.08. 1.27. 0.99. 1.13. 1.03. 1.38. 1,03. 1.13. 0.95. 1.15. 0.95. 0.33.

. 0.69. 1.25 1.05 1.14 1.21. 1.37 1.00 1.36 1.19. 1.09 0.97 1.06 0.40

• 0.69. 1.25. 1.05. 1.13. 1.21. 1.38. 1.01. 1.37. 1.20. 1.07. 0.94. 1.03. 0.40.

A 0.71 1.14 1.12 1.37 1.03 1.37 1.12 1.08 1.11 1.35 0.99 1.29 1.00 0.96 0.61 l

2 3

4 5

6 0.73 1.16 1,13 1.37 1.01 1.35 1.12 1.08 1.11 1.35 0.96 1.22 0.96 0.93 0.60 7

• 0.87. 0.92. 1,09. 1.13. 1.39. 1.00. 1.07. 1.13. 1.06. 0.99. 1.35. 1.09. 1.02. 0.84. 0.79 *

• 0.89. 0.93. 1.09. 1.15. 1.41. 1.02. 1.04. 1.13. 1.06. 0.99. 1.32. 1.03. 0.99. 0.84. 0.79.

8

• • • • • • • • • • • • • • • • • • • • • • • • 0.................................................................................

0.71 1.14 1.12 1.37 1.03 1.37 1.12 1.08 1.11 1.35 1.01 1.33 1.07. 1.08 0.66

• 0.73. 1.15. 1.12. 1.38. 1.04. 1.37. 1.08. 1.06. 1.11. 1.35. 1.01. 1.26. 1.10. 1.10. 0.67.

9

. 0.69. 1.25. 1.05. 1.14. 1.21. 1.37. 1.00. 1.37. 1.20. 1.12. 1.03. 1.22. 0.67 *

• 0.69. 1.25. 1.06. 1.15. 1.20. 1.33. 0.98. 1.35. 1.20. 1.13. 1.05. 1.27. 0.69.

0.42. 1.10. 1.28. 0.98. 1.14. 1.03. 1.39. 1.03. 1.13. 0.97. 1.26. 1.08. 0.41

• 0.42. 1.12. 1.30. 1.00. 1.11. 1.01. 1.35. 1.00. 1.12. 0.99. 1.29. 1.11. 0.42.

0.41. 0.73. 1.28. 1.05. 1.37. 1.13. 1.36. 1.05. 1.27. 0.73. 0.40

• o.42. o.76. 1.34. 1,03. 1.35. 1.11. 1.34. 1.04. 1.28. o.74. o.42.

0.41. 1.10. 1.25. 1.12. 1.08. 1.11. 1.24. 1.09. 0.41

• 0.43. 1.15. 1.27. 1.10. 1.07. 1.10. 1,24. 1.11. 0.42.

0.42 0.69. 1.14. 0.91. 1.13. 0.69 0.41

• 0.44. 0.72. 1.15. 0.91. 1.13. 0.68. 0.41.

0.71 0.86 0.70

• 0.73. 0.88. 0.70.

Standard Deviation== l.303 Avg.% diff. = 1.6

SUMMARY

MAP NO: 51-7-62 DATE:

8/ 6/84.

CONTROL ROD POSITIONS:

F-Q(T) = 1.819 D BANK AT 225 STEPS F-DH(N) = 1.527 F(Z)

= 1. 124 F(XY)

= 1. 481 BURNUP = 10602 MWD/MTU 27 POWER:

79%

QPTR:

NW 1.0176 I NE 0.9544

. ---------1----------

sw 1.0226 I SE 1.0054 A.O =

0.26(%)

10 11 12 13 14 15

1.0 SURRY UN IT 1 - CYCLE 7 HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE Figure 4.4 1-}-* -

-.-~-- U__jl---W---l---1--l-+--1--J--j~+-l-++1-H++++t-H+t-++t+t--t-1++-++t+t--t-1++--ttrl I

-f- -

.. 1--+--+-4--'--....i...,~~

r H--+-t-++-H- - - f-t- --1-4-1--i-+-l--l

++-++-Jc+++ *!-f-. 1-(6.1, 1.0)

. i*+-t-* *- - -. -t-t-t- -

t 1-..-. '.

-~-,- --.,-,-,.--, ri-r-rr+-+t+-t-+t-t-H--H--t-

!-!-1-'-t-+*

+-

t-(10.9, 0.94)

• - i l

! ~ -

I rJ+--!-+-++-++--J - -t--cr-l-t---1--+-1-+!-+-L+--. H-'-++-++-++-H--t--1--+-l--l---l--++++-1--H+!-++++H---l+t-t-\\--t+-tt--lrtttttt-tt f-

.J.,--

ff

+-

-, -, r-+--H-++-H-+-+-'1--f--+--t--++ *~H--++++++++-+++++--i---:H--t+t--t+t+-lrH-+++t-~H-i~I

.--i-----t-,--1--+--IH--!--++++-+--H--l-+-++-++-t-t -t-2 4

6 8

10 12 CORE HEIGHT (FI,)

28

N -

2.5 +

2.0 +

~ O'

~

SURRY UN IT 1 - C°VCLE 7 HEAT FLUX HOT CHANNEL FACTOR, F6(Z)

Sl-7-11 xxxx xx X

X XX XXXX Figure 4.5 XX X

X X

xxxxxxxx X

XX XX X X

X X

XX X

X

, p::

0

~ 1.5 +

xxxx X

X X

~

I 1.0 +

~

~

X X

X X

X

~

-x X

~

i_o.5:

-o.o +

I..... I.

61 55 BOTTOH OF CORE

• I

• so X

X X

X X

X X

.I.... I... I.... I.... I.

~

~

~

~

~

• I

• 20

• I

• 15

• I *

.

• I.** I 10 S

l TOP OF CORE AXIAL POSITION (NODES)

z.s +

.-, z.o +

N -

H O' Ii<

SURRY UN IT 1 - CYCLE 7 HEAT FLUX HOT CHANNEL FACTOR, F6(Z)

Sl-7-19 XX XX X X

X X

X xxxxxxxx xxxxxx

~

0 XX XX X

X X

XX X

X X

H u 1.5 +

X pi;..

X

~

~

~

X u

H X

0 1.0 +

I::

-x

~

B X

pi;..

H <

~

I:: 0.5 +

-:o.o +

I..**. I.

61 55 BOTTON OF CORE X

X X

.I.*. I.**. I ***. I **** I.**. I.

• I *

~

~

~

~

~

~

20 AXIAL POSITION (NODES) 30 Figure 4.6 X X X X X

XX X

X X

X X

• I *

. I

.. I... I 15 10 5

l TOP OF CORE

Z.5 +

z.o +

~

0 1.5 +

~

u xx SURRY UN IT 1 - CYCLE 7 HEAT FLUX HOT CHANNEL FACTOR, F6(Z)

S1-7-62 XX XX XX x.x X X X X X X X X X

~

XX X xxxxxxxx xx X

X

~

ra::I !

c..,

1.0 +

~

~

X X

X X

~

-x X

~ 0.5 +

r,::i.

c:

.o.o +

I..*** I 61 55 BOTTot1 OF CORE X

X X

X X

• I *

. I...* I.... I.*.* I...* I.

• I

• 50 45 40 35 30 25 20 AXIAL POSITION (NODES) 31 Figure 4. 7 X X X XX X XX X

X X

X

• I *

. I

.. I... I 15 10 5

1 TOP OF CORE

2.2 2.0 l.8 1. 6 1. 4 F

l.2 Q

1. a p

o.e 0.6 0.4 0.2

~

o.o,_

I 61 Figure 4.8 SURRY UNIT 1 - CYCLE 7 MAXIMUM HEAT FLUX HOT CHANNEL FACTOR, FQ*P, vs.

l l

55 50 45 AXIAL POSITION FQ

• P LIMIT

• MAXIMUM FQ

• P r-----

• • *~.

l

~~

~

40 35 30 25 20 15 10 AXIAL POSITION lNOOEI 32 l

\\

\\

• \\

• \\

• \\. \\

\\

~ I.

5

M Figure 4.9 SURRY UN IT 1 - CYCLE 7 MAXIMUM HEAT FLUX HOT CHANNEL FACTOR, FQ(z),

vs. BURNUP TECH SPEC LIMIT X MEASURED VALUE R

2.6-1----t----+----+----t----t--._...--i----t---t-.......... -t-----t-----t---1------ X I

M U 2-5-+---+---+---+---+---+---+--+---+---+--+-+--+--+--+----+

M H

E 2,4-+---+---+---+---+----+---+---+---+---+--+--+----+---+--+----+

A T

F 2.3-+---+----t----+---+---+----t--._...--i----+--t--+---t---t-----t----,...

L u X

H 0

T C

H A

N 2. G-1---+----t----+----+----+----t--._...--i----t---t---t---t---t-----,...

N E

L F

1. g-+---+---+---+---+---+---+---+--+---+---+---+---+--+----+

A X

C X

X X

xx T

l.B-+--.:..:.X+--~v......+--A+-v---'-'X-+---+---+--X-+--X-+-...,..~-+---+---+-....,v-+--+----+

Q X

R 1.7-+---+---+---+---+---+---+---+---+---+---+---+---+--+----+

1.s-~,....,....,....... _

...... _.,..._...,....,......,............-+-,....,....,.........+-,~.--!-r

.................. !-r,....,._..,.._.,.......,....,....,...,....,....,.....,....,-+-,-.,..

0 2000 4000 6000 BODO

~0000 12000 14000 CYCLE BURNUP !MWO/MTUJ 33

Figure 4.10 SURRY UN IT 1 - CYCLE 7 MAXIMUM ENTHALPY RISE HOT CHANNEL FACTOR, F delta-H~

vs. BURNUP TECH SPEC LIMIT X MEASURED VALUE E 1.60-l---1---1---1---+---+---i---+---+--+--+-+--+--+--+----r N

T H

A x

L I.55....:J----4----4----1---.+----+-....... -+---+---+---+-~+x:-:---+--t---t---t p

y V

R I* 1. 50..:l---+---+---+---+---l--+--+--+--+--+--+--t--+--+

5 X

X X

E l-~X.......1,~vi<-4-.......1,-* X0-1---+--+--+-_,a.X+___.:X+--t--+--t---t---t X

H 0

T C

H A

N N

E L

F*

A C

T 0

R I.JO-l---1---1---1---+---+---l---+---+--+---+--+--+--+---r I,Z0-~1..,...,....,....,.....!..........,....,.....!........--......l-..,...,....,,...+.,..............-l-,........,...,..+,.............4-,........ ~ 11-,-,-...,......-1h...;,-,-.~1..............-...,...,........-f.......--.......+..,...,..........-f" 0

2000 4000 6000 6000 10000 12000 14000 CYCLE BURNUP !MWD/MTUl 34

T R

R G

E T

'o E

L T

q F L u

X I N p

E R

C E

N T

SURRY UN IT 1 - CYCLE 7 TARGET DEL TA FLUX vs. BURN UP Figure 4.11-a--------------+---+---+---+-------------

6 4

2 0

t:.

t:.

t:.

t:.

-2 t:.

t:.

-4

-6 +---+----+----+----+---+---+---+---+---+--+--+--+--+----t- ~.................+.................,......... l"'T"'t-......-l"'T"'t-.............+-.......... -..+-.............+-,......,......+-,......,....-!-,....,.........+-,,...,..,-,-,.,..._...,._...,..,.,....,.-I-T..,....,.,....-....,........,,-+-

0 2000 4000 6000 8000 10000 12000 14000 CYCLE BURNUP !MWO/MTUl 35

.::i

~

N H g

~

0 z --

N -N i:,...

1.5 +

1.2 +

0.9 +

SURRY UN IT 1 - CYCLE 7 CORE AVERAGE AXIAL POWER DISTRIBUTION Sl-7-11 F2 (z) = 1.152 Axial Offset= -0.384 xxxxx xxxxxxxx xxxxxx XX X

X X

XX Figure 4. 12 X

X X

XXXXX XX X X

X X

X X

X X

X X

X X

X*

X X

X 0.6 +

X X

-x 0.3 +

o.o +

I..... I

• 61 55 BOTTOM OF CORE I

• 50 I

• 45

• I ***,

I **** I **** I

• 110 35 30 25 AXIAL POSITION (NODES) I

• I

• I

• 20 15 10 xx X

X X

X I *** I 5

1 TOP OF CORE

SURRY UNIT 1 - CYCLE 7 CORE AVERAGE AXIAL POWER DISTRIBUTION Sl-7-19 1.5 +

Fz(z) = 1.144 Axial Offset= -3.152 1.2 +

x*x XX XX X X

xxxxxxx X

XX X

X X

X X

X X X X X X X X X

X 0.9 +

X X

X X

-x 0.3 +

o.o +

I...,

I.

61 55 BOTTOM OF CORE I

• 50 X

X X

X X

I *, *. I,

I.... I. *

• I I

• 45 40 35 30 25 20 AXIAL POSITION (NODES)

• 37 X

Figure 4.13 X X X X X X X

X X

X X

XX X

X X

X I

• I

• I... I 15 10 5

1 TOP OF CORE

0 w N

H

- i 0 z --

N -N J:z4 1,5 +

Fz (z) = 1.124 SURRY UN IT 1 - CYCLE 7 CORE AVERAGE AXIAL POWER DISTRIBUTION Sl-7-62 Axial Offset= +o.259 1.2 +

X X X

X X

X X

0.9 +

X X

0.6 +

-x -

0.3 +

0.0 +

X I,, *.. I.

61 55 BOTTOM Of CORE X X X X X X X XX XX X

X XX XX X X X X X X X XX X

X X

X X

X I '

I.*.,

I..,

I,

I,

I

• I

• 50 45 40 35 30 25 20 AXIAL POSITION (NODES) 38 Figure 4.14 xxxxxx X

X X

X I

• I

• 15 10 X

X X

X X

X I *** I 5

1 TOP OF CORE

Figure 4.15 SURRY UN IT 1 - CYCLE 7 CORE AVERAGE AXIAL PEAKING FACTOR, F-z, vs. BURNUP 1. 3 A

X I

A L

p E

A K

1.2 I

N G

F A A A

A C

A

/:,.

A

'1 T

A A

A 0

A R

A 1. 1 1.o-~ *.,....,....,,......+

..,.._........,..+,.............+-.,...,...-,,+,...,.,....,...,...,+-,-,_...,.

...... t-,,...,..,...,..-!-,..........,.,..,....,.......,...h..,...,......+,...,..,....,...h...............h...............!-....,..,..~

I I

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2000 4000 6000 8000 10000 12000 14000 CYCLE BURNUP IMHO/MTUI 39

SECTION 5 PRIMARY COOLANT ACTIVITY FOLLOW Activity levels of iodine-131 and 133 in the primary coolant are important in core performance follow analysis because they.are used as indicators of defective fuel. Additionally, they are also important with respect to the offsite dose calculation values associated with accident analyses. Both I-131 and I-133 can leak into the primary coolant system through a breach in the cladding. As indicated in the Surry Power Station Technical Specifications, the dose equivalent I-131 concentration in the primary coolant is limited to 1.0 µCi/gm for normal steady state operation. Figure 5.1 shows the dose equivalent I-131 activity level history for the Surry 1, Cycle 7 core. The demineralizer flow rate averaged 106 gpm during power operation.

The data shows that during Cycle 7, the core operated substantially below the 1. 0 µCi/ gram limit during steady state operation (the spike data is associated with power transients and unit shutdown). Specifically,.the cycle average value of dose equivalent I-131 was 0.0788 µCi/gram which is approximately 8% of the Technical Specifications limit. This level of coolant activity resulted from two apparent fuel defect formation events that occurred early in Cycle 7.

The ratio of the specific activities of I-131 to I-133 is used to characterize the type of fuel failure which may have occurred in the reactor core. Use of the ratio for this determination is feasible because I-133 has a short half-life (approximately 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />) compared to that of 40

I-131 (approximately eight days).

For pinhole defects, where the diffusion time through the defect is on the order of days, the I -133 decays out leaving the I-131 dominant in activity, thereby causing the

• ratio to be 0.5 or more. In the case of large leaks, uranium particles in the coolant, and/or "tramp" uranium*, where the diffusion mechanism is negligible, the I-131/I-133 ratio will generally be less than 0.1. Figure 5.2 shows the I-131/I-133 ratio data for the Surry 1, Cycle.7 core: The I-131/I-133 ratio generally remained around 0.5, which indicates possible pinhole defects in the fuel cladding.

• "Tramp" uranium consists of small particles of uranium which adhere to the outside of the fuel during the manufacturing process.

41

Figure. 5.1 SURRY UN IT 1 - CYCLE 7 DOSE EQUIVALENT 1-131 vs. TIME

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SECTION 6 CONCLUSIONS The Surry 1, Cycle 7 core has completed operation. Throughout this cycle, all core performance indicators compared favorably with the design predictions and all core related Technical Specifications limits were met with significant margin.

No abnormalities in reactivity, power distribution, or burnup accumulation were detected except those anticipated "during operation with stuck control rod B-6. In addition, the mechanical integrity of the fuel has not changed significantly since the beginning of Cycle 7, as indicated by the radioiodine analysis.

44

SECTION 7 REFERENCES

1)

Mr. T. C. Hartsfield, "Surry Unit 1, Cycle 7 Startup Physics Test Report," VEP-NOS-5, July, 1983.

2)

Surry* Power Station Unit 1 and 2 Technical Specifications, S_ections 3.1.D, 3.12.B, and 4.10.

3)

'Mr. T. K. Ross, "NEWTOTE Code", VEPCO NFO-CCR-6 Rev-8, April, 1984.

4)

Mr. R. D. Klatt, Mr. W. D. Leggett, III, and Mr. L. D. Eisenhart, "FOLLOW Code, 11 WCAP-7482, February, 1970.

5)

Mr.

W.

D.

Leggett, III and Mr.

L. D. Eisenhart, "INCORE Code,"

WCAP-7149, December, 1967.

45