ML19320C689

Supplemental Reload Licensing Submittal for Browns Ferry Nuclear Power Station,Unit 2,Reload 3.

ML19320C689
Person / Time
Site:	Browns Ferry
Issue date:	06/30/1980
From:	Engel R, Hilf C GENERAL ELECTRIC CO.
To:
Shared Package
ML19320C684	List: ML19320C683 ML19320C687 ML19320C689
References
Y1003J01A12, Y1003J1A12, NUDOCS 8007170574
Download: ML19320C689 (20)

Text

.

O Y1003J01A12 Class I June 1980 o

SUPPLEMENTAL RELOAD LICENSING S"MITTAL FOR BROWNS FERRY NUCLEAR POWER STATION UNIT 2 RELOAD NO. 3 Prepared: [

C.L. Ililf Reload Fuel L censing Approved: ,

R.E. Engel(/ Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS OlVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORM A 95125 (ooll7o6W' GEN ER AL $ ELECTRIC

Y1003J01A12 IMPORTANT NOTICS REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report uas prepared by General Electric solely for The Tennessee Valley Authority (TVA) for TVA's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending TVA's operating License of the Brcuns Ferry Nuclear Unit 2. The infomation contained in this report is believed by General Electric to be an accurate and true representa-tion of the facta knoun, obtained or provided to Generat Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting information in this document are contained in the contract betuean The Tennessee Valtey Authority and General Electric Company for nuclear fuel and related services for the nuclear system for Brouna Ferry Nuclear Plant Unita 1 and 2, dated June 17, 1966, and nothing 1

contained in this, document shall be ccnetrued as changing said con-

, tracc. The use of this information except as defined by said con-l tract, or for any purpose other tha. that for uhich it is intended, is not authorized; and with r oc to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes many representation or warranty (e.xpress or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately ovned rights; nor do they assume any respcnsibility for liability or damage of any kind ,

uhich may result from such use of such information.

b t

4

Y1003J01Al2 CONTENTS Page

1. PLANT-UNIQUE ITEMS 1

2. RELOAD FUEL BUNDLES 1

~

3. REFERENCE CORE LOADING PATTERN 1

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CORE SYSTEM WORTH - NO VOIDS, 20*C 2

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY 2

6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS 2

7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS 3

8. SELECTED MARGIN IMPROVEMENT OPTIONS 3

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS 3

10. LOCAL ROD WITHDRAWAL ERROR (WII11 LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

4

11. OPERATING MCPR LIMIT 4

12. OVERPRESSURIZATION ANALYSIS STRE!ARY 4

13. STABILITY ANALYSIS RESULTS 5 ,

14. LOSS-0F-COOLANT ACCIDENT RESULTS 5

12. LOADING ERROR RESULTS 5

16. CONTROL ROD DROP ANALYSIS RESULTS 5

17. REFERENCES 5 APPENDIX A 19 4

i l

iii/iv

Y1003J01A12 ILLUSTRATIONS Figure Title Page 1 Reference Core Loading Pattern 6 2 Scram Reactivity and Control Rod Drive Specifications 7 3 Plant Response to Generator Load Rejection without Bypass 8 4 Plant Response to Loss of 100*F Feedwater Heating 9 5 Plant Response to Feedwater Controller Failure, Maximum Demand 10 6 Limiting RWE Rod Pattern 11 7 Plant Response to MSIV Clocure 12 8 Decay Ratio 13 9 Doppler Reactivity Coefficient Comparison for RDA 14 10 Accident Reactivity Shape Function at 20'C 15 11 Accident Reac?tvity Shape Function at 286*C 16 12 Scram Reactivity Function at 20*C 17 13 Scram Reactivity Function at 286*C 18

/

v/vi

Y1003J01Al2

1. PLANT-UNIQUE ITEMS (1.0)*

Iten. different from or not included in Reference 1:

o

Fuel Loading Error LHGR

Appendix A Number of dperable Safety / Relief Valves: Appendix A Safety / Relief Valve Capacity: Appendix A Spring Safety / Relief Valve Capacity: Appendix A Separate MCPR and ACPR limits reported for 8x8 and 8x8R fuels 4

2. RELOAD FUEL BUNDLES (1.0, 3.3.1 and 4.0)

Fuel Type Number Number Drilled Irradiated, Initial Core 7D250 Type 2 12 12 Irradiated, Initial Core 7D250 Type 3 112 112 Irradiated. Reload 1 8DB274H 24 24 Irradiated, Reload 1 8DB274L 108 108 Irradiated, Reload 2 8DRB284 232 232 Irradiated, Reload 2 89B274L 36 36 New P8DRB284 240 240 Total 764 764

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 16,052 mwd /t Assumed reload cycle core average exposure at end of cycle: 17,582 mwd /t Core loading pattern: Figure 1 l

t I

l

• 0 ) refers to areas of discussion in " Generic Reload Fuel Application", i NEDE-240ll-P-A-1, August 1979, l 1

Y1003J01Al2

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CORE SYSTEM l WORTH - NO VOIDS, 20*C (3.3.2.1.1 and 3.3.2.1.2) i BOC k,gg Uncontrolled 1.120 Fully Controlled 0.957 Strongest Control Rod Out 0.984 R, Maximum Increase in Cold Core Reactivity with

• Exposure Into Cycle, ok 0.0009

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) 600 ppm Shutdown Margin (ak) 0.026 t

(20*C, Xenon Free)

6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)

Void Coefficient N/A* (c/% Rg) -7.53/-9.41 Void Fraction (%) 40.18 Doppler Coefficient N/A (c/*F) -0.219/-0.208 Average Fuel Temperature (*F) 1383 Scram Worth N/A ($) -37.22/-29.78 Scram Reactivity vs Time Figure 2 4

l

• N = Nuclear Input Data A = Used in Transient Analysis 2

Y1003J01A12

7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 7x7 8x8/8x8R/P8x8R Exposure EOC 4 EOC 4 Peaking factors 1.24 1.20 (local, radial -

1.30 1.52 and axial) 1.40 1.40 R-Factor 1.10 1.05 Bundle Power 5.489 6.424

(MWt)

Bundle Flow 119.8 109.4 (103 lb/hr)

Initial MCPR 1.21 1.27

8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

Recirculation Pump Trip 9.

CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1) i I

l I

• LCPR Core ,

p p

, Power Flow $ Q/A si v P8x8R Plant Transient Exposure (%) (%) (% NBR) (% NBR) (psig) (psig) 7x7 8x8 8x8R Response

Load Rejection 30C4-EOC4 104.5 160 263.4 112.5 1230* 1253 0.14 0.19 0.20 Figure 3 l without Bypass Loss of 100*F -- 104.5 100 12 3.8 12 3. ) 1011 1069 0.09 0.12 0.12 Figure a Feedwater i Heating Feedveter 50C4-EOC4 104 5 100 183.6 114.2 1156 1191 0.09 0.14 0.14 Figure 5 Controller Failure i

• Less than 25 psi margin. (As indicated in Appendix A, only 10 of the 11 safety /

relief valves are assumed to be operating.)

3 l

Y1003J01A12

10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(5.2.1)

Rod ACPR MLHGR Position Limiting Rod Block (Feet P8x8R P8x8R Rod Reading Withdrawn) 8x8 8x8R 7x7 8x8 8x8R 7x7 Pattern 104 3.0 0.15 0.14 0.18 12.0 15.0 14.7 Figure 6 105 3.5 0.18 0.17 0.22 12.5 16.3 15.7 Figure 6 106* 4.0 0.20 0.20 0.25 12.9 17.3 16.4 Figure 6 107 4.0 0.20 0.20 0.25 12.9 17.3 16.4 Figure 6 108 4.5 0.23 0.22 0.28 13.2 17.9 16.9 Figure 6 109 5.0 0.25 0.25 0.30 13.4 18.3 l'.2 Figure 6 110 5.0 0.25 0.25 0.30 13.4 18.3 17.2 Figure 6

• Indicates setpoint selected

11. OPERATING MCPR LIMIT (5.2)

BOC4 to EOC4 1.27 (8x8 fuel) 1.27 (8x8R and P8x8R fuel) 4 1.32 (7x7 fuel) 1

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3)

Power Core Flow sl v Plant Transient (%) (%) (psig) (psig) Response MSIV Closure 104.5 100 1266 1301 Figure 7 (Flux Scram) 4

Y1003J01A12

13. STABILITY ANALYSIS RESULTS (5.4)

Decay Ratio: Figure 8 Reactor Core Stability:

Decay Ratio, x2 /*0 -

0.88

, (105% Rod Line - Natural Circulation Power)

Channel Hydrodynamic Performance Decay Ratio, x2 /*0 (105% Rod Line - Natural Circulation Power) 8x8/8x8R/P8x8R chantiel 0.38 7x7 channel 0.23

14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2) (8DRB284)

MAPLHGR results for the P8DRB284 type fuel are conservative compared to the 8DRB284 type fuel previously reported.

15. LOADING ERROR RESULTS (5.5.4)

Limiting Event: Rotated Bundle P8DRB284 MCPR: 1.07

16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

. Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 Plant specific analysis results Parameter not bounded: Scram Reactivity Shape Function at 20 C Resultant peak enthalpy: 131 cal /gm l

17. REFERENCES l

1. " Generic Reload Fuel Application," August 1979 (NEDE-2iO11-P-A-1) .

5

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l 1 3 5 7 9il13i5171921'325272331232537394: 43 45 47 49 51 53 55 57 59 FUEL TYPE

• A = IC 7D250 (Type 2) E = R? 8DRB284 B = IC 7D250 (Type 3) F = R2 8DB274L C = R1 8DB274L G = R3 P8DRB284 D = R1 8DB274H

• IC = Initial Core, R1 = Reload 1, R2 = Reload 2, etc.

Figure 1. Reference Core Loading Pattern 6

- _ _ . -. = - - .

Y1003J01A12 100 r 45 CONTROL ROD CRIVE VS TIME SCRAM REACTIVITY VS TIME C-678 CAD IN PERCENT 1-NOMINAL SCRAM CURVE IN (-51 90- 2-SCRAM CURVE USED IN ANALYSIS

-40 80-

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E 50- L;

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o_

-20 40-

-15 30-

-10 20-10- -5 0 --

i i i 0

, 0 1 2 3 4 I TIME (SECONOS) l Figure 2. Scram Reactivity and Control Rod Drive Specifications 7

1' I NEUTRON F* LUX l VESSEL PFES RISE (PSI) 2 RVE StiffCE E AT FLUX 2 SAFETT VfLVE FLOW 150* 3 CORE INLf I FLOW 3 ELIEF VfLVE FLOW 300*

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TIME (SEC1 O.4 0.8 1.2 1.6 TIE ISECl Figure 3. Plant Response to Generator Load Rejection Without Bypass

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TIME (SEC) 120. 160.

llME LSECl Figtire 4. Plant Response to Loss of 100*F Feedwater IIcating 4

9 i NEUTRON FLUX i VESSEL PFES RISE (PSil f 2 RVE SURFfCE EAT Flut 2 SAFETT YFLVE FLOW 150" 3 CORE IM.E T FLOW 3_ RELIEF VFLVE FLOW I

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TIME ISECl TIE ISEC)

Figure 5. Plant Response to Feedwater Controller Failure, Maximum Demand 9 6 . .

Y1003J01A12 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 32 32 32 32 55 4 8 6 8 4 51 32 30 30 30 30 32

. 47 6 4 0 4 0 4 6 43 32 30 46 46 30 32 39 4 2 8 2 8 2 4 35 30 28 30 30 28 30 31 0 12 0 10 0 12 0 27 30 28 30 30 28 30 23 4 2 8 2 8 2 4 19 32 30 46 46 30 32 15 6 4 0 4 0 4 6 11 32 30 30 30 30 32 7 4 8 6 8 4 3 32 32 32 32 Notes: 1. Rod pattern is 1/4 core mirror symmetric (full core shown).

2. Number indicates number of notches withdrawn out of 48. Blank is a withdrawn rod.

3. Error rod is (22,31) .

Figure 6. Limiting RWE Rod Pattern 11

I ,

1 EUTR(N FLUX 1 VESSEL PFES RISE (PS!!

2 AVE SURFFCE HERT FLUX 2 SAFETT VfLVE FLOW 3 CORE Ittf i FLOW 150. 300. 3

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8 U

t:

O V

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Figure 7. Plant Response to MSIV Closure

i Y1003J01A12 1

i 1.2

. ULTIMATE STABILITY LIMIT

, 1.0 ---- --____ _ _ _ _ _ _ _ _

0.8 -

NATURAL CIRCULATION

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0 20 40 60 80 100 PERCENT POWER Figure 8. Decay Ratio 13

Y1003J01A12

0. 0 ,

A CALCULATED VALUE-COLD B CALCULATED VALUE-HSB C BOUNDING VALUE FOR 280 CAL /G COLD .

-5.O D B0UNDING VALUE FOR 280 CAL /G HSB A

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b -20.0 C

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-30.0 l' /

4

-35.0

0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE (DEG C.)

I ;

Figure 9. Doppler Reactivity Coefficient Comparison for RDA 14

4" Y1003J01A12 20.0 , , .

A CALCULATED VALUE B BOUNDING VALUE FOR 280 CAL /G 15.0 G

{- gcc c E

e b 10.0 <

4 t;

C 3

5

" 5.0 ,

0. 0 ,_

O. 0 5.0 10.0 15.0 20.O ROD POSITION, FEET GUT Figure 10. Accident Reactivity Shape Function at 20*C 15

Y1003J01A12 20.O A CALCULATED VALUE B B0UNDING VALUE FOR 280 CAL /G 15.0 G

g g-0 e a

E D 10.0 N "

E t3 5

5.0 ~

c -

0. 0 0.0 5.0 10.0 15.0 20.0 .

ROD POSITION, FEET GUT Figure 11. Accident Reactivity Shape Function at 286*C 16 l

F

' ~

Y1003J01A12 40.O A CALCULATED VALUE B B0UNDING VALUE FOR 280 CAL /G 30.0 2=

bi .

E E

9 A

)

a 20.0 ^

?

J l-E t3 10.0 )

0. 0 _

O. 0 1. 0 2.0 3. 0 4.0 5.0 6. 0 ELAPSED TIME, SECONDS Figure 12. Scram Reactivity Function at 20*C 17

Y1003J01A12 75.O A CALCULATED VALUE p B B0UNDING VALUE FOR 2 a cav:: .

G

=

a g S0.0 --

9 i;

4i C

E

& 25.0 /

U

.I O.0 #

/ '

0. 0 1. 0 2.0 3. 0 4.0 5. 0 6. 0 ELAPSED TIME, SECONDS

! Figure 13. Scram Reactivity Function at 286*C l

18 i

=--

i .. .

3 Y1003J01A12 4

APPENDIX A Fuel Loading Error LHGR: 17.1 kW/ft i

Safety / Relief Valve Capacity at Setpoint (No./%): 10/63.6 l Spring Safety Valve Capacity at Setpoint (No./%): 2/14.2 4

0 t .

4 i.

I i

1 k

e i

i i

i j

4 1

4 1

4 C

1 i

e t

I i

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