ML20053C186

Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plant Unit 2,Reload 4 (W/Recirculation Pump Trip Feature), Revision 1

ML20053C186
Person / Time
Site:	Brunswick
Issue date:	03/31/1982
From:	Charnley J, Engel R, Zarbis W GENERAL ELECTRIC CO.
To:
Shared Package
ML20053C160	List: ML20053C159 ML20053C164 ML20053C174 ML20053C186
References
DRF-L12-00306-1, DRF-L12-306-1, Y1003J01A37, Y1003J01A37-R01, Y1003J1A37, Y1003J1A37-R1, NUDOCS 8206010557
Download: ML20053C186 (33)

Text

.... _. - -

l Y1003J01A37 DRF L12-00306-1 REVISION ~1 CLASS 1 MARCH 1982 i

Il1 l,'

I i

l SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2, RELOAD 4

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82060105~57 GEN ER AL h ELECTRIC

Y1003J01A37 DRF L12-00306-1 Revision 1 g

Class I March 1982 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSUICK STEAM ELECTRIC PLANT UNIT 2, RELOAD 4 (UITil RECIRCULATION PUMP TRIP FEATURE)

Prepared:

h W. A. Za i Licensing Engineer Reload F icensing Verified

• J/ S.' Charnle'y,

//

Senior Licensing Engifeer Reload Fuel Licensing d

Approved

. -E. $ngel,' Manager 9/

~

Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNI A 95125 e

GEN ER AL $ ELECTRIC i

Y1003J01A37 Rev. 1 e

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 Fuel Contract Supplemental Agreement between Carolina Power and Light Company and General Electric Company for Brunswick 1

& 2 dated January 28, 1974, 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 complete-ness, accuracy or usefulness of the information ccntained in 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.

9 11

1 Y1003J01A37 Rev. 1 1.

PLANT-UNIQUE ITEMS (1.0)

Local Rod Withdrawal Error Appendix A Confirmation of Single Loop Operation Appendix B Safety Analysis Results for 7x7 Fuel Appendix C Additional LOCA Results Appendix D 2.

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

Fuel Type Cycle Loaded Number Number Drilled Irradiated 8DB274L 2

100 100 Irradiated 8DB274H 2

36 36 Irradiated 8DRB265H 3

64 64 Irradiated 8DRB283 3

68 68 Irradiated P8DRB265H 4

132 132 New P8DRB265H 5

136 136 New P8DRB284H 5

24 24 Total:

560 560 3.

REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle exposure:

15,356 mwd /ST Minimum previous cycle exposure:

14,956 mwd /ST Assumed reload cycle exposure:

16,609 mwd /ST Core loading pattern:

Figure 1 4.

CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)

BOC k,gg Uncontrolled 1.112 Fully Controlled 0.954 Strongest Control Rod out 0.989 R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, ak 0.0 e

• ( ) refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-4, January 1982.

1

Y1003J01A37 Rev. 1 5.

STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (Ak) ppm (20*C, Xenon Free) a 600 0.036 6.

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

EOC Void Coefficient N/A* (C/% Rg)

-7.86/-9.83 Void Fraction (%)

41.8 Doppler Coefficient N/A (C/% *F)

-0.219/-0.208 Average Fuel Temperature (

• F) 1312 Scram Worth N/A ($)**

7.

RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.?.2)

Peaking Factors Fuel (Local, Radial, Bundle Power Bundle Flow Initial Design Exposure Axial)

R-Factor (MWt)

(103 lb/hr) MCPR 8x8 EOC5 (1.22, 1.44, 1.40) 1.098 6.152 110.5 1.19 8x8R EOCS (1.20, 1.58, 1.40) 1.051 6.723 111.0 1.20 P8x8R EOC5 (1.20, 1.56, 1.40) 1.051 6.658 112.1 1.21 8.

SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization:

No Recirculation Pump Trip: Yes Rod Withdrawal Limiter: No Thermal Power Monitor: Yes Measured Scram Time: No f

Exposure Dependent Limits: No l

f

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

• • Ceneric exposure independent values are used as given in " General. Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-4, January 1982.

2 I

Y1003J01A37 Rev. 1 9.

CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1) 0 Q/A (Uncorrected)

(g (g

py, Transient Exposure NBR)

NBR) 8x8 8x8R P8x8R

Response

Turbine BOC-EOC 330 116 0.12 0.13 0.14 Figure 2 Trip Without Hypass Loss of BOC-EOC 124 122 0.13 0.13 0.13 Figure 3 100*F Feedwater Heating Feedwater BOC-EOC 113 109 0.04 0.04 0.05 Figure 4 Controller Failure 10.

LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(S.2.2.1)

See Appendix A.

11.

OPERATING CYCLE MCPR VALUES (S.2.2)

Nonpressurization Events:

Exposure Range:

BOC to EOC 8x8 8x8R P8x8R Loss of Feedwater Heating 1.20 1.20 1.20 Fuel Loading Error 1.22 Rod Withdrawal Error 1.29 1.21 1.21 Minimum Required by LOCA 1.20 1.20 1.20 Pressurization Events:

Exposure Range:

BOC to EOC Option A Option B Transient 8x8 8x8R P8x8R 8x8 8x8R P8x8R Turbine Trip 1.24 1.25 1.26 1.16 1.17 1.18 Without Bypass I

Feedwater Controller 1.16 1.16 1.17 1.13 1.13 1.14 Failure j

D 3

1

Y1003J01A37 Fev. I 12.

OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3)

Psi P

Plant y

Transient (psig)

(psig)

Response

MSIV Closure (Flux Scram) 1208 1244 Figure 5 13.

STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed:

105%

Figure 6 Decay Ratio:

i x2 X0 0.73

/

Reactor Core Stability Decay Ratio, Channel Hydrodynamic Performance Decay Ratio, x2/X0 Channel Type 8x8R/P8x8R 0.30 8x8 0.36 14.

LOADING ERROR RESULTS (S.2.5.4)

Variable Water Cap Misoriented Bundle Analysis:

Yes Includes 2.2% Power Spiking Penalty:

Yes l

Initial Resulting MCPR MCPR 1.20 1.07 15.

CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)

Doppler Reactivity Coefficient :

Figure 7 Accident Reactivity Shape Functions:

Figures 8 and 9 Scram Reactivity Functions:

Figures 10 and 11 Plant Specific Analysis Results:

Parameters not bounded: Accident Reactivity and Scram Reactivity, cold j

Resultant Peak Enthalpy = 224.8 cal /gm l

h l

4

7 Y1003J01A37 Rev. 1 16.

LOSS-0F-COOLA.'1T ACCIDENT RESULTS (S.2.2.2)

" Loss of Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit No.

2," General Electric Company, September 1977 (NEDO-24053, as amended).

9 Fuel Type:

P8DRB284H Exposure MAPLHGR PCT 0xidation (mwd /ST)

(kW/ft)

(*F)

Fraction 200 11.2 2101 0.025 1000 11.2 2099 0.025 5000 11.7 2146 0.028 10000 12.0 2180 0.031 15000 12.0 2184 0.031 20000 11.7 2158 0.029 25000 11.0 2061 0.021 30000 10.3 1962 0.015 35000 9.7 1863 0.010 40000 9.0 1775 0.007 0

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Y1003J01A37 Rev. 1

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l MMMMMMMMMMM "MMMMMMMMM" "MMMMM" l IIIliIIIi 1 35 7 9111315171921232527293133353739414345474951 FUEL TYPE A = 8DB274L E = P8DRB265H B = 8DB274H F = P8DRB265H C = 8DRB265H G = P8DRB284H D = 8DRB283 H = 8DRB265H i

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Y1003J01A37 Rev. 1 A NATURAL CIRCULATIC N B 105 PERCENT ROD LI NE C ULT. PERFORMANCE L IMIT 1.00

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Reactor Core Decay Ratio 11

Y1003J01A37 Rev. 1 O. 0

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

Figure 7.

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Y1003J01A37 Rev. 1 20.0 17.5 15.0 0

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Accident Reactivity Shape Function Cold Startup 13

Y1003J01A37 Rev. 1 20.0 17.5 15.0 0

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Accident Reactivity Shape Function Hot Startup 14

r.

Y1003J01A37 Rev. 1 30.0 A SCRAM FUNCTION 8 BOUNDING VALUE 2 30 CAL /G 25.0 9

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Scram Reactivity Function Cold Startup 15

Y1003J01A37 Rev. 1 70.0 A SCRAM FUNCTION B BOUNDING VALUE 2 30 CAL /G 60.0 0

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Figure 11.

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16

Y1003J01A37 Rev. 1 APPENDIX A LOCAL ROD WITHDRAWAL ERROR For 8x8 fuel, local Rod Withdrawal Error results are reported in accordance with Reference A-1.

For 8x8R and P8x8R fuel, these results are reported in accordance with Reference A-2.

ACPR Rod Block Reading (%)

8x8 8x8R/P8x8R 104 0.13 0.11 105 0.16 0.11 106 0.19 0.13 107*

0.22 0.14 108 0.28 0.16 109 0.32 0.16 110 0.36 0.17 REFERENCES A-1.

Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), " Change in General Electric Methods for Analysis of Control Rod Withdrawal Error,"

May 18, 1981.

A-2.

" General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-4,

" January 1982.

• 1ndicates set point selected 17/18

r-Y1003J01A37 Rsv. 1

.e APPENDIX B 4

CONFIRMATION OF SINGLE LOOP OPERATION The previous Single Loop Operation analysis performed for Brunswick 2 (Reference B-1) has been verified to be applicable for Cycle 5.

REFERENCES B-1.

" Brunswick Steam Electric Plant Units 1 and 2 Single-Loop Operation,"

General Electric Company, September 1981 (NED0-24344).

i

+

.O I

19/20.

. ~..

j Y1003J01A37 Rev. 1 APPENDIX C SAFETY ANALYSIS RESULTS FOR 7x7 EUEL e

There is a possibility that 7x7 fuel may be used by Brunswick 2 during Cycle 5.

Transient analysis results for 7x7 fuel are presen' ed in Table C-1 to provide t

for such an occurrence. Should 7x7 fuel be loaded into the core, the procedures outlined in Reference C-1 will be followed to assure that the final core loading pattern is an acceptable deviation from the reference core loading pattern presented in item 3 of this submittal. A comparison of the data for 7x7 fuel with the data presented in items 9 and 11 for the other fuel types shows that 7x7 fuel, if loaded, will have no effect on Cycle 5 operating limits.

REFERENCE C-1.

" General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-4, January 1982, a

1 21

Y1003J01A37 Rav. 1 Table C-1

SUMMARY

OF TRANSIENT ANALYSIS FOR 7x7 FUEL 1.

RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS Peaking Factors Bundle Bundle Fuel (Local, Radial, Power Flow Initial Design Exposure Axial)

R-Factor (MWt)

(103 lb/hr)

MCPR 7x7 EOC 5 (1.24, 1.33, 1.40) 1.100 5.672 122.1 1.15 2.

CORE-WIDE TRANSIENT ANALYSIS RESULTS ACPR Q/A Uncorrectedg Transient Exposure

(% NBR)

(% NBR) 7x7

/

Turbine Trip Without Bypass EOC 5 330 116 0.08 Loss of 100* Feedwater lleating E0C 5 124 122 0.11 Feedwater Controller Failure EOC 5 113 109 0.03 3.

LOCAL ROD WITilDRAWAL ERROR If 7x7 fuel is used, it will be located on or near the periphery of the core, far enough away from the error rod that the Rod Withdrawal Error will have no impact.

4.

OPERATING CYCLE MCPR VALUES Nonpressurization Events:

MCPR Transient Exposure (7x7)

Loss of Feedwater Heating EOC 5 1.18 Fuel Loading Error EOC 5 N/A Rod Withdrawal Error EOC 5 N/A Minimum Required by LOCA 1.20*

• For 7x7 fuel, the flow factor, K, is based on the 112% flow curve line rather g

than the 102.5% line, a conservatism that makes the requirement that the operating limit MCPR be greater than 1.23 not applicable in this case.

22 a

Y1003J01A37 Rev. 1 Table C-1 SUMS \\RY OF TRANSIENT ANALYSIS FOR 7x7 FUEL (Continued)

Pressurization Events:

MCPR (7x7)

Transient Exposure Option A Option B Turbine Trip Without Bypass EOC 5 1.20 1.12 Feedwater Controller Failure E0C 5 1.15 1.12 5.

STABILITY ANALYSIS RESULTS Channel Hydrodynamic Performance Decay Ratio, x /*0 2

7x7 Channel 0.22 6.

LOSS-OF-COOLANT ACCIDENT RESULTS LOCA results for exposures up to 30,000 mwd /ST were previously reported for the following fuel type in Reference C-1.

Additional LOCA results, denoted by a bar in the right hand margin, are presented here for exposures up to 40,000 mwd /ST.

Fuel Type:

7D230 Exposure MAPLHGR PCT 0xidation (mwd /ST)

(kW/ft)

(*F)

Fraction 200 14.9 2198 0.031 1000 15.0 2197 0.031 5000 15.1 2197 0.030 10000 14.6 2195 0.030 15000 14.1 2199 0.073 20000 13.8 2198 0.073 25000 13.7 2199 0.072 30000 13.8 2197 0.071 35000 12.8 2082 0.049 40000 11.5 1898 0.024 23

Y1003J01A37 Rev. 1 REFEPINCE C-1.

" Loss of Coolant Accident Report for Brunswick Steam Electric Plant Unit No.

2," Ceneral Electric Company, idepterier 1977 (NED0-24053, as amended).

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Y1003J01A37 Rav. 1 APPENDIX D ADDITIONAL LOCA RESULTS Loss-of-coolant accident (LOCA) results were previously reported (References D-1, D-2 and D-3) for exposures up to 30000 mwd /ST for the fuel types presented in the following tables. Additional LOCA results are presented here for exposures up to 40,000 mwd /ST for these fuel types. These additional results are denoted by a bar in the right-hand margin.

REFERENCES D-1.

" Loss of Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit No.

2," General Electric Company, September 1977 (NED0-24053, as amended).

D-2.

" Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plant Unit 2 Reload 2," Ceneral Electric Company, January 1979 (NEDO-24587).

D-3.

" Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plant Unit 2 Reload 3," General Electric Company, March 1980 (NEDO-24235).

+

O 25

Y1003J01A37 Rev. 1 Fuel Type:

8D274L Exposure MAPLHGR PCT 0xidation (mwd /ST)

(kW/ft)

(*F)

Fraction 200 11.2 2064 0.020 1000 11.3 2069 0.020 5000 11.9 2144 0.026 10000 12.1 2148 0.025 15000 12.2 2178 0.028 20000 12.1 2185 0.029 25000 11.6 2136 0.025 30000 10.9 2046 0.019 35000 10.2 1959 0.014

~

40000 9.6 1872 0.010 l

l a

26

r Y1003J01A37 R:v. 1 Fuel Type:

8D274H Exposure MAPLHGR PCT 0xidation o

(mwd /ST)

(kW/ft)

(*F)

Fraction 200 11.1 2056 0.020 1000 11.2 2055 0.019 5000 11.8 2126 0.024 10000 12.1 2150 0.026 15000 12.2 2181 0.028 20000 12.0 2182 0.029 25000 11.5 2130 0.025 30000 10.9 2047 0.019 35000 10.2 1961 0.014

~

40000 9.6 1875 0.010 k

27

Y1003J01A37 Rev. 1 Fuel Type:

8DRB265H Exposure MAPLHGR PCT 0xidation iMWd/ST)

(kW/ft)

(*F)

Fraction 200 11.5 2154 0.030 1000 11.6 2156 0.029 5000 11.9 2192 0.032 10000 12.0 2196 0.032 15000 12.0 2200 0.033 20000 11.8 2197 0.033 25000 11.3 2138 0.027 30000 10.7 2056 0.021 35000 10.1 1970 0.015 40000 9.4 1883 0.011 4

28

Y1003J01A37 Rev. 1 Fuel Type:

8DRB283 Exposure MAPLHGR PCT 0xidation g

(mwd /ST)

(kW/ft)

(*F)

Fraction 200 11.2 2122 0.027 1000 11.2 2117 0.026 5000 11.8 2184 0.032 10000 12.0 2197 0.033 15000 11.9 2194 0.032 20000 11.8 2197 0.033 25000 11.3 2132 0.027 30000 11.1 2106 0.025 35000 10.4 2021 0.019

~

40000 9.8 1938 0.014 O

~

29

Y1003J01A37 R:v. 1 Fuel Type:

P8DRB265H Exposure MAPLHGR PCT 0xidation (mwd /ST)

(kW/ft)

(*F)

Fraction 200 11.5 2138 0.028 1000 11.6 2146 0.028 5000 11.9 2174 0.030 10000 12.1 2187 0.031 15000 12.1 2196 0.032 20000 11.9 2177 0.030 25000 11.3 2101 0.024 30000 10.7 2016 0.018 35000 10.1 1916 0.012

~

40000 9.4 1823 0.009 M

30

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