ML20198H241

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Rev 0 to Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plant Unit 2,Reload 6
ML20198H241
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 03/31/1986
From: Charnley J, Elliott P, Casey Smith
GENERAL ELECTRIC CO.
To:
Shared Package
ML20198H231 List:
References
23A4748, 23A4748-R, 23A4748-R00, NUDOCS 8605300189
Download: ML20198H241 (26)
v • d • e


Text

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    • 0L" MARCH 1986 m -

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SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2, RELOAD 6

[88"2868: 8888ljg4 GEN ER AL $ ELECTRIC

23A4748 Revision 0 Class I March 1986 l

l SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PIANT UNIT 2, RELOAD 6 Prepared:

O P. E. Elli.ott Fuel Licen. sing Verified: .

C. W. Smith' Suel L

,-. s Appr ed: _, ,

' . -> . Chaftfle y Manager, Fuel Licens i

NUCLEAR ENERGY BUSINESS OPEHATIONS . GENERAL E' ECTRIC COMPANY SAN JOSE. CALIFORNIA 95 t25 GENERAL $ ELECTRIC I/2 u .

23A4748 Rzv. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Carolina Power and Light Company (CP&L) for CP&L's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending CP&L's operating license of the Brunswick Steam Electric Plant Unit 2. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General 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 Supplemental Agreement to the Contract between Carolina Power and Light Company and General Electric Company for Reload Fuel Supply and Related Services for Brunswick Steam Electric Plant Unit 2, effective June 25, 1980, and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in 1 this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

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23A4748 Rev. 0

1. PLANT UNIQUE ITEM (1.0)*

Information in Section 4 and Appendix A provided by Carolina Power & Light Co. Appendix A Plant. Parameter Changes Appendix B Turbine Control Valve Configuration Appendix C

2. . RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)

Cycle Loaded Number Irradiated 8DRB265H 3 20 P8DRB265H 4 48 P8DRB265H 5 136 CT P8DRB284H 5 24 BP8DRB299 6 184 New BP8DRB299 7 148 TOTAL 560

(

3. REFERENCE CORE LOADING PATTERN (3.3.1) l .

Nominal previous cycle core average exposure at end of cycle: 16040 mwd /MT.

Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 16040 mwd /MT Assumed reload cycle core average exposure at end of cycle: 18742 mwd /MT Core loading pattern: Figure 1

  • () Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-7, August 1985. A letter "S" preceding the number refers to the appropriate section in the United States Supplement, NEDE-24011-P-A-7-US, August 1985.

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23A4748 Rev. 0

4. CALCUIATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTd - NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)*

Beginning of Cycle, K,ff Uncontrolled 1.127 Fully Controlled 0.966 Strongest Control Rod out 0.989 R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, Ak 0.003

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (a) j3na (20*C, Xenon Free) 600 0.035

6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)

(COLD WATER INJECTION EVENTS ONLY)

Void Fraction (%) 41.3 Average Fuel Temperature (*F) 1285 Void Coefficient N/A** (d/% Rg) -6.97/-8.71 Doppler Coefficient N/A (4/*F) -0.177/-0.168 Scram Worth N/A ($)

  • See Appendix A.
    • N = Nuclear Input Data, A = Used in Transient Analysis.
      • Generic exposure independent values are used as given in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-7-US, August 1985.

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7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS

( S.2.2)

Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR Exposuret h0C to EOC-2000 mwd /ST BP/P8x8R 1.20 1.51 1.40 1.051 6.422 114.1 1.26 8x8R 1.20 1.53 1.40 1.051 6.516 112.0 1.24 Exposure: EOC-2000 mwd /ST to EOC BP/P8x8R 1.20 1.41 1.40 1.051 6.002 116.5 1.35 8x8R 1.20 1.44 1.40 1.051 6. 13 3 114.2 1.32

8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: Yes Improved Scram Time: No

\ Exposure Dependent limits: Yes Exposure Points Analyzed: EOC and E0C-2000 mwd /ST

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3 )

I Single Loop Operation: Yes Load Line Limit: Yes Extended Load Line Limit: No Increased Core Flow: No Flow Point Analyzed: N/A Feedwater Temperature Reduction: No 7

23A4748 Rov. 0

10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)

Exposure Range: BOC to EOC-2000 mwd /ST Flux Q/A Transient (% NBR) (% NBR) BP/P8x8R 8x8R Figure Load Rejection Without Bypass 541 122 0.19 0.17 2 Loss of Feedwater Heater 126 124 0.16 0.16 3 Feedwater Controller Failure 164 112 0.07 0.06 4 Exposure Range: EOC-2000 mwd /ST to EOC A

Flux Q/A Transient (% NBR) (% NBR) BP/P8x8R 8x8R Figure Load Rejection Without Bypass 521 130 0.28 0.25 5 Loss of Feedwater Heater 126 124 0.16 0.16 3 Feedwater Controller Failure 164 112 0.07 0.06 4

11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(S.2.2.1)

Limiting Rod Pattern: Figure 6 Rod Block Rod Position Reading (feet withdrawn) BP/P8x8R 8x8R BP/P/8x8R f 104 3.5 0.11 0.11 13.70 105 3.5 0.11 0.11 13.70 /

106 4.0 0.12 0.12 14.19 107 4.5 0.13 0.13 14.64 f 108 5.0 0.14 0.14 14.9 3 109 5.5 0.15 0.15 15.16 110 6.0 0.17 0.17 15.44 Setpoint Selected: 107 8

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23A4748 Ray. 0

12. CYCLE MCPR VALUES (S.2.2)

Non-Pressurized Events Exposure Range: BOC to EOC BP/P8x8R 8x8R

{

Loss of Feedwater Heater 1.23 1.23 Fuel Loading Error 1.20 Rod Withdrawal Error 1.20 1.20 Pressurization Events l Option A Option B BP/P8x8R 8x8R BP/P8x8R 8x8R Exposure Range:

BOC to EOC-2000 mwd /ST Load Rejection Without Bypass 1,32 1.29 1.12 1.10 Feedwater Controller Failure 1.19 1.18 1.13 1.12 Exposure Range:

E0C-2000 mwd /ST to EOC Load Rejection Without Bypass 1.41 1.38 1.29 1.26 Feedwater Controller Failure 1.19 1.18 1.13 1.12

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) l 1

's1 v Plant Transient (psig) (psig) Response MSIV Closure (Flux Scram) 1221 1257 Figure 7 l

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I 23A4748 R2v. 0

14. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed: Extrapolated APRM Decay Ratio Figure 8 Reactor Core Stability Decay Ratio, x2 /*0: 0.78 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 Channel Type BP/P8x8R 0.30 8x8R 0.30

15. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Misoriented Bundle Analysis Yes*

Event Initial MCPR Resulting MCPR Misoriented 1.18 1.07

16. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)

Bounding Analysis Results:

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 (s) not Bounded, Cold: Scram Reactivity Resultant Peak Enthalpy, Cold: 171.6 Parameter (s) not Bounded, HSB: Scram Reactivity /

Accident Reactivity Resultant Peak Enthalpy, IISB: 229.2

  • ACPR penalty of 0.02 for the tilted misoriented bundle is applied to the cycle MCPR value reported in Section 12.

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23A4748 Rsv. 0

17. LOSS-OF-COOIANT ACCIDENT RESULT (S.2.5.2)

See " Loss-of-Coolant Analysis Report for Brunswick Steam Electric Plant Unit No. 2," September 1977 (NEDO-24053, as amended).

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23A4748 Rav. 0

' .MMMMM.
oMMMMMMMMMo
- MMMMMMMMMMM
.MMMMMMMMMMM.

CMMMMMMMMMMMMM CMMMMMMMMMMMMM IMMMMMMMMMMMMM CMMMMMMMMMMMMM IMMMMMMMMMMMMM

"MMMMMMMMMMM" MMMMMMMMMMM
"MMMMMMMMM"
"MMMMM" iIIIIIIIII 1 3 5 7 9111315171921232527293133353733414345474951 FUEL TYPE B = P8DR 6 511 E = BP8DRB29 C = P8DRB26511 F = BP8DRB299 Figure 1. Reference Core Loading Pattern 12

23A4748 Ray. O I NEufRON FLUE 1 VESSEL PRESS RISELPSI) 2 AVE SURF ACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET ELOW 3 RELIEF VALVE FLOW 130.0 388.0 '. eva.a ss u.*LuE eter b 180. 8 'M M 1 200.0 m v- -

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0. 0 2. 0 4.0 6.0 8.0 2. 0 4.0 6.0 f!ME (SECONOS) TIME (SECONOS)

LEVEL (INCH-REF.SEP.SKRT) V  ! VO) REACTIVITY 2 VESSEL STEAiFLOW 2 00P ER REACT!VIly 3 TURBINE STELMFLOW 200.0 m rEgg. pen r_cu 1.4 3i SCR.

rg u_A REACT egicr I.VITY verv

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e. o 2. s 4.s s.e e.o 2. s 4. e s.e TIME (SECONCS) f!ME (SECCND$1 Figure 2. Plant Response to Generator Load Rejection, Without Bypass (EOC-2000) 13

23A4748 R2v. O 150.0 1 NEUIRON FLUX 1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE K Af FLUX 2 REL IEF VALVE FLOW 3 CORE INLET FL0n 3 BYPiSS VALVE FLOW 13 0.0 ' r emr '= rt ste L

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100.0 E

W 10 0. 0 ' -

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0. 0 0.0 100.0 200.0 0. 0 100.0 200.0 flME (SECONOSI f!ME (SECONOS1 1 LEV EL(INCH-REF-SEP.SKRT) 1 VOI ) REACTIVITY 2 VES5EL STEAMFLOV 2 00P)LER REACTIVITY 3 TUR)1NE STE ANFLOV 3 SCRnN REAcilVITY 150.0 ' rre ,warro eteu 1.0 ' tot u_ eEict'y'Yv

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0.0 100.0 200.0 0. 0 100.0 200.0 f!ME (SECOND$1 flME (SECOND$1 Figure 3. Plant Response to Loss of 100*F Feedwater lleating 14

23A4748 Rsv. 0 150.0 1 NEUTRON FLU ( l VESSEL PRESS RISE (PSI) 2 AV : SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CO PE INLET rLOW 3 RELIEF VALVE FLOW 13 0.0 '"O t' "

p' RET 4 BYPASS VALVE FLOW

& 100.0 g a; ~ n 'Ig.

- 100.0 ' .' '. - - -

th y 50.0 w

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C. 0 L 0. 0 20.0 30.0 0. 0 10.0 20.0 30.0 TIME ($[CONDS) TIME (SECONOSI

! LEVEL (INCH-REF.$EP.SM R T) 1V010RIACTIVITY 2 VESSEL STEA1 FLOW 2 00PP Ef REACTIVITY 3 TURBINE STE AMFLOW 3 SCR A f EACT I 135,0 i rggpitro r nu 1.0 a vnri_ c i r.VliY virv 4 4 i i 4 l

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l 5 l i 100.0 L s ' 4' ;

.0 tj C w 88 A9 - q

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53.0

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0. 0 18.0 20.0 30.0 0. 0 10.0 20.0 30.0 TIPE ($ ECON 01) TIME (SECOND$1 Figure 4 Plant Response to Feedwater Controller Failure 15

23A4748 Rsv. O l

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l NEUTRON FLLE 1 YESSEL PRESS RISE (PSI) 2 AVE SURF ACE HE AT FLUX 2 SAFETY VALVE FLOW 3 C RE !RU rLOW ,,,. ggg 3At{jFt0$

is .. .

N 100.0 ' M g 200.0 hN E

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ILEVEL(INCH-REF-SEP.SKRT) 1' 10 REACTIVITY 2 VESSEL STEA1 FLOW 2 PLER REAtilv!TY 2....  ? M__ E_ L S'! ?_?"

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!!PE (SECONOS) f!Mt (SECONOS)

Figure 5. Plant Response to Generator Load Rejection, Without Bypass (EOC) 16

r 23A4748 Rsv. O l

l NOTES: 1. ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC.

2. NUMBER INDICATES NUMBER OF NOTCIIES WITHDRAWN OUT OF 48. BLANK IS A WITilDRAWN ROD.
3. ERROR ROD IS (18,31).

2 6 10 14 18 22 26 51 40 47 14 10 14 43 40 40 40 39 6 6 6 35 40 36 44 31 10 6 0 6 l 27 40 44 36 Figure 6. Limiting Rod Withdrawal Error Rod Pattern I

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23A4748 Rev. O I

I hEUTRON F UX 1 VESSEL PRESS RISE (PSI) 2 AVE SURFA:E HEAT FLUX 2 SAFETY VA VE FLOW 3 EORE INLET FLOW 3 RELIEF VALVE FLOW 15 0.0 300.0 2 ovonSS vi uE etow E l td 10s.0 1 200.8 -

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0. 0 5.0 0.0 5.0 TIME (SECOND$1 TIME (SECONDS) i ILEVEL(INC4 REF-SEP-SKRT)  ! VO!D REACTIVITY 2 VESSEL STEAMFLOW OOPPLER PEACTIVITY 3 TURetNE STE AMFLOW u REACTIVITY 2 eggguarto _rtgu 1.0 . 2 rgrit egirtivit<

200.4 b .

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140.0 2,0

c. 9 S.0 8.0 5.O TIME ($(CON 0$1 TIME (SECONOS)

Figure 7. Plant Response to MSIV Closure (Flux Scram) 18 i

1

i 23A4748 Rev. O Ab ATURAL C :RCULATIO 4 B1 05 PERCENT ROD LI VE CL LTIMATE STABILITY LINE 1.00 (:  ::

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.75 /

x N

X d

r

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0. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure 8. Reactor Core Decay Ratio 19

23A4748 R2v. 0

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-15.0

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-35.0 -A4;ALCUL nE0 VALU -COLO B CALCUL ATED VALui-HSB C BOUND VAL 280 C \L/G COLD D BOUND VAL 280 CAL /G HSB

-40.0

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

,1.... . ... m .,g.,c.. m e.. o . m c  ;

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23A4748 Rsv. 0 20.O A ACCIDENT FUNCTION 8 BOUNDING VALUE 280 CAL /G 17.5 15.0 2

U 12.5 w&UU g ,

4 A A A o

10.O  !

2 7.5 l--

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'0. 0 5.0 10.0 15.0 20.O ROD POSITION, FEET GUT Figure 10. Accident Reactivity Shape Function Cold Startup 21

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23A4748 R2v. 0 20.O A ACCIDENT FUNC TION 8 BOUNDING VALU E 280 CAL /G 17.5 15.O m

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0. 0 5.0 10.0 15.0 20.0 ROD POSITION, FEET GUT l no... u. m u m a _ u m ,s,.. er - u a se m e, j

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l 23A4748 Rev. 0 30.O i A SCRAM FUNCTION 8 BOUNDI NG VALUE 280 CAL /G 25.0 m

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,, 15.0 to us z

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0. 0 1. 0 2. 0 3.0 4.0 5. 0 6. 0 ELAPSED TIME, SECONOS Figure 12. Scram Reactivity Function Cold Startup 23

23A4748 ,

R2v. 0 50.O A SCRAM FUNCTION B BOUNDIflG VALUE 280 CAL /G 40.0 ,

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g 30.0

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0. 0 1. 0 2. 0 3.0 4.0 5.0 6.0 ELAPSED TIME, SECONDS Figure 13. Scram Reactivity Function Hot Startup 24

l 23A4748 Roy. O APPENDIX A This cold shutdown margin design evaluation was performed by Carolina Power & Light Company personnel using NRC approved methodology

  • applied in conformance with Carolina Power & Light Company's Quality Assurance Program.

This evaluation provides a high degree of confidence that greater than the Technical Specification requirement of 0.38% AK/K margin to critical will be maintained throughout the cycle, and that 0.38% AK/K plus "R" will be measured during beginning of cycle shutdown margin demonstration startup physics testing.

I I

  • Letter f rom Domenic B. Vassallo to E. E. Utley, " Brunswick Reload Licensing Methodologies", Docket Nos. 50-325/324, May 18, 1984, 25/26

l 23A4748 Rsv. O I

APPENDIX B Plant Parameter Changes:

Pressure Relief Systems (Table S-2-4.1, NEDE-24011-P-A-7-US)

Safety / Relief Valve Type E (i.e. , capacity at reference Pressures of 1080 psig +3% is 789,000 lb/hr for each S/RV) 27/28

r 1

23A4748 Ray. O APPENDIX C TURBINE CONTROL VALVE CONFIGURATION The transient and GETAB analyses presented in the body of this report are based on turbine control valves in a full-arc configuration and on the power supply to the recirculation Motor-Generator Sets from offsite power.

29/30 (FINAL)

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