ML20053C174
| ML20053C174 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 03/31/1982 |
| From: | Charnley J, Engel R, Zarbis W GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20053C160 | List: |
| References | |
| DRF-L12-00306-1, DRF-L12-306-1, Y1003J01A37, Y1003J1A37, NUDOCS 8206010547 | |
| Download: ML20053C174 (37) | |
Text
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{ MAR 1982 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2, RELOAD 4 8206010 5'/7 GENER AL h ELECTRIC
. . . . - . = . .
4 4
Y1003J01A37 DRF L12-00306-1
- Revision 0
- . Class I March 1982 i
4 i l J
I SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2, RELOAD 4 (WITHOUT RECIRCULATION PUMP TRIP FEATURE) t i
Prepared: M
- W.A.[aris Licensin Engineer q
] Reload Fuel Licensing f,
Verified: # .
- J.ps.Mharnley e Senior Licensing Eng neer Reload Fuel Licensing i Approved
- .
- h. E. Engel, Manage 8
!$t Reload Fuel Licensing i l
{
- NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE CALIFORNI A 95125 G E N ER AL $ ELECTRIC i
Y1003J01A37 Rev. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT s
PLEASE READ CAREFULLY
%is report was prepared by General Electric solely for Carolina Power and Libat Company (CP&L) for CP&L's use with the U.S. Nuclear Regulatory Com-mission (USNRC) for amending CP&L's opt. rating 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 completeness, accuracy or usefulness of the information contained 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.
I 11
Y1003J01A37 Rev. 0
- 1. H AiT UNIOUE ITEMS (1.0)*
Local Rod Withdrawal Error Appendix A e
Confirmation of Single Loop Operation Appendix B Sa fety 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 Typee Cycle Loaded Number Number Drilled Irradiated 8DB274L 2 100 100 Irradiated 8DB27411 2 36 36 Irradiated 8DRB265H 3 64 64 Irradiated 8DRB283 3 68 68 Irradiated P8DRB263H 4 132 132 New P80RB26511 5 136 136 New P8DRB28411 5 24 24 Total: 560 560
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Neminal previous cycle exposure: 15,356 mwd /ST Minimum previous cycle exposure: 14,936 mwd /ST Auumed reload cycle exposure: 16,609 mwd /ST Core loading pattern: Figure 1
- ( ) reftrs to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-4, Jaauary 1982.
1
Y1003J01A37 Rev. 0
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH -
NO VOIDS, 20*C (3.3.2.1.1 and 3.3.2.1.2) e EOC k eff ,
Uncontrolled 1.112 Fully Coatrolled 0.954 Strongest Control Rod Out 0.989 R, Maximum Increase in Cold Core 0.0 Reactivity with Exposure into Cycle, Ak
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) pgm (20*C, Xenon Free) 600 0.036
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 and S.2.2)
EOC EOC -2000 mwd /ST vo id Coefficient N/A* (c/T Rg) -7.86/-9.83 -7.97/-9.97 Void Fraction (%) 41.76 41.76 Doppler Coefficient N/A (c/% *F) -0.219/-0.208 -0.208/-0.198 Average Fuel Temperature (*F) 1312 1312 Scram Worth N/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-4, January 1982.
2
ll -
Y1003J01A37 Rev. 0 l
- 7. RELOAD-UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)
Peaking Factors Fuel Exposure (Local, Radial, Bundle Power Bundle Flow Initial Design (mwd /ST) Axial) R-Factor (HWt) (103 lb/hr) MCPR 8x8 EOC (1.22,1.35,1.40) 1.098 5.744 113.1 1.28 EOC-2000 (1.22,1.46,1.40) 1.098 6.239 109.9 1.17 8x8R EOC (1.20,1.47,1.40) 1.051 6.267 113.7 1.29 EOC-2000 (1.20,1.60,1.40) 1.051 6.815 110.5 1.18 P8k8R EOC (1.20,1.44,1.40) 1.051 6.144 115.2 1.32 EOC-2000 (1.20,1.58,1.40) 1.051 6.741 111.7 1.19
- 8. SELECTED _ MARGIN IMPROVEMENT OPTIONS (S.2.2.2) r -
' Transient Recategorization: No
! Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: Yes ,
. Measured Scram Time: No I
' Exposure Dependent LJmits: Yes Ex,posures Analyzed: EOC and EOC-2000 mwd /SI
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1) f 4
ACPR Exposure h Q/A (Une trected) py Transient (mwd /ST) (% NBR) (% NBR) 8x8 8x8R P8x8R Response
/ Load EOC 553 127 0.21 0.22 0.25 Figure 2a
< Rejection ,
I Without E0C-2000 436 117 0.10 0.11 0.12 Figure 2b Bypass Loss of BOC-EOC 124 122 0.13 0.13 0.13 Figure 3 100*F /
Heating /
Feedwater ' EOC; 122 109 0.04 0.04 0.05 Figure 4a Contro11er l
- Failure - E0C-2000 114 109 0.04 0.04 0.04 Figure 4b ef>
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Y1003J01A37 Rev. 0
- 10. LOCAL R0D WITilDRAWAL ERROR (WITil 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 1.20 1.20 1.20 lleating Fuel Loading Error -- --
1.22 Rod Withdrawal Error 1.29 1.22 1.25
! Minimum Required by 1.20 1.20 1.20 LOCA Pressurization Events:
p n ption B Exposure Transient (mwd /ST) 8x8 8x8R P8x8R 8x8 8x8R P8x8R Load EOC 1.34 1.35 1.38 1.22 1.23 1.26 Rejection Without E0C-2000 1.22 1.23 1.24 <l.08 <l.08 <l.08 Bypass Feedwater EOC 1.16 1.16 1.17 1.10 1.10 1.11 Controller Failure E0C-2000 1.16 1.16 1.16 1.10 1.10 1.10
- 12. OVERPRESSURIZATION ANALYSIS SDDIARY (S.2.3) si v Plant Transient (psiQ (psig) Response MSIV Closure 1208 1244 Figure 5 (Flux Scram)
~
4
F Y1003J01A37 Rev. 0
- 13. STABILITY ANALYSIS RESULTS (S.2.4)
Rod Line Analyzed: 105%
, Decay Ratio: Figure 6 Reactor Core Stability Decay Ratio, x2 /*0 *
/
Channel liydrodynamic Performance Decay Ratio, x2 XO Channel Type 8x8R/P8x8R 0.30 8x8 0.36
- 14. LOADING ERROR RESULTS (S.2.2.4)
Variable Water Gap Misoriented Bundle Analysis: Yes Includes 2.2% Power Spiking Penalty: Yes Initial Resulting MCPR MCPR 1.20 1.07
- 15. CONTROL R0D DROP ANALYSIS RESULTS (S.2.2.1)
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 Resultant Peak Enthalpy = 224.8 cal /gm 4
5
Y1003J01A37 Rev. 0
- 16. LOSS-OF-COOLANT ACCIDENT RESULTS (5.2.5.2)
" Loss of Coolant Accident Analysis Report for Brunswick Steam Electric ,
Plant Unit No. 2," General Electric Company, September 1977 (NEDO-24053, as amended). .
Fuel Type: P8DRB28411 Exposure '
MAPLilGR 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 4
6 l
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Y1003J01A37 Rev. 0
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- sMMMMMMMMMMMs
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- "MMMMMMMMM" "MMMMM" l IIIIIIIIi 1 3 57 9111315171921232527293133353739414345474951 FUEL TYPE A = 8DB274L E = P8DRB26511 B = 8DB274}i F = P8DRB265H C = 8DRB26511 G = P8DRB28411 D = 8DRB283 11 = 8DRB26511 Figure 1. Reference Core Loading Pattern 7
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Figure 4b. Plant Response to Feedwater Controller Failure, E0C-2000 mwd /t o m - -. _
h I of uTroJ FLlix J agt *lefD F HFAT FLUX 1 vfSSEL PfES RISE IPSin
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Y1003J01A37 Rev. 0 A NATURAL IRCULATIC N B 105 PERC NT ROD LI NE -
C ULT. PER ORMANCE L IMIT 1.00 (: :
A x .75 N
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Figure 6. Reactor Core Decay Ratio 14 I
t.
- Y1003J01A37 Rev. O l
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- 0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE DEG C.
1 Figure 7. Doppler Coefficient in 1/A*C 15
Y1003J01A37 Rev. 0 20.0 17.5 15.O m
o
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A ACCIDENT FUNCTION 8 BOUNDING val.UE 280.C AL/G
- 0. 0
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Figure 8. Accident Reactivity Shape Function Cold Startup 16
Y1003J01A37 Rev. 0 I
20.0 i
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A ACCIDENT FullCTION 8 BOUNDING val.UE 280 CAL /G
- 0. O _
m
[ 0.0 5.0 10.0 15.0 20.0 ROD POSITION, FEET OUT Figure 9. Accident Reactivity Shape Function Hot Startup 17
Y1003J01A37 Rev. 0 30.O A SCRAM FUNCTION -
B BOUNDING VALUE 2 30 CAL /G 25.0 c
O 1
w 20.0 -
l t--
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a
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- 0. 0 1.0 2. 0 3. 0 4.0 5.0 6. 0 ELAPSED TIME, SECONDS I e Figure 10. Scram Reactivity Function Cold Startup 18
Y1003J01A37 Rev. 0
~
1 70.0 A SCRAM FUNCTION 4 B BOUNDING VALUE 2 30 CAL /G 60.O e,
O w
i 50.0 x
C 40.0 w f a
to w
5 30.0 20.0 -
o w
a:
10.O !
- 0. 0 m
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- 0. 0 1.0 2. 0 3. 0 4.0 5.0 6. 0 ELAPSED TIME, SECONDS Figure 11. Scram Reactivity Function Hot Startup 19/20
Y1003J01A37 Rev. O 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 0.12 0.15 105 0.16 0.12 0.15 106 0.19 0.14 0.17 107* 0.22 0.15 0.18 108 0.28 0.17 0.20 109 0.32 0.17 0.20 110 0.36 0.18 0.21 REFERENCES ,
i A-1. Letter, R. E. Engel (CE) to T. A. Ippolito (NRC), "Chaage in General Electric Methods for Analysis of Control Rod t.'ithdrawal Error."
May 18, 1981.
A-2. " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-4, January 1982. <
1 l
1
- Indicates set point selected 21/22 l
l _ _ .
i I
Y1003J01A37 Rev. O l 1
APPENDIX B 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,"
NEDO-24344, September 1981.
23/24
Y1003J01A37 Rev. O APPENDIX C SAFETY ANALYSIS RESULTS FOR 7x7 FUEL There is a possibility that 7x7 fuel may be used by Brunswick 2 during
. Cycle 5. Transient analysis results for 7x7 fuel are presented in Table C-1 to provide 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. "Ceneral Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-4, January 1982.
j Q
25
Y1003J01A37 Rev. O Table C-1
SUMMARY
OF TRANSIENT ANALYSIS FOR 7x7 FUEL
- 1. RELOAD-UNIQUE TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS Peaking Factors Bundle Bundle .
Fuel Exposure (Local, Radial, Power Fuel Initial Design (mwd /ST) Axial) R-Factor (MWt) (103 lb/hr) MCPR 7x7 EOC (1.24, 1.25, 1.40) 1.100 5.344 124.3 1.23 EOC-2000 (1.24, 1.36, 1.40) 1.100 5.819 121.2 1.12
- 2. CORE-WIDE TRANSIENT ANALYSIS RESULTS
. . ACPR Exposure $ Q/A Uncorrected Transient (mwd /ST) (% NBR) (% NBR) 7x7 Load Rejection Without Bypass EOC 5 553 127 0.15 EOC 5-2000 436 117 0.05 Loss of 100*F Feedwater EOC 5 124 122 0.11 Heating Feedwater Controller Failure EOC 5 122 109 0.03 EOC 5-2000 114 109 0.03
- 3. LOCAL ROD WITHDRAWAL 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:
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, Kg, is based on the 112% flow curve line ,
rather 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.
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Y1003J01A37 Rev. O Table C-1
SUMMARY
OF TRANSIENT ANALYSIS FOR 7x7 FUEL (Continued)
Pressurization Events:
MCPR (7x7)
Transient Exposure Option A Option B Load Rejection Without Bypass EOC 5 1.27 1.17 EOC 5-2000 1.17 <1.08 Feedwater Controller Failure EOC 5 1.15 1.09 EOC 5-2000 1.15 -
1.09
- 5. STABILITY ANALYSIS RESULTS Channel Hydrodynamic Performance Decay Ratio, x2 !*0 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 0.049 35000 12.8 2082 40000 11.5 1898 0.024 27
Y1003J01A37 Rev. O
"'7ERENCE C-1. " Loss of Coolant Accident Report for Brunswick Steam Electric Plant Unit No. 2," General Electric Company, September 1977 (NEDO-24053, as amended) .
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Y1003J01A37 Rev. O APPENDIX D ADDITIONAL LOCA RESULTS Loss-of-coolant accident (LOCA) results were previously reported (References D-1, a D-2 and D-3) for exposures up to 30,000 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 (NEDO-24053, as amended).
D-2. " Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plant Unit 2 Reload 2," General 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 (NED0-24235).
29
Y1003J01A37 Rev. O 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 _
f 30
Y1003J01A37 Rev. O I
Fuel Type: 8D274H Exposure MAPLHGR PCT 0xidation (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 _
o 31
Y1003J01A37 Ikev. O P
Fuel Type: 8DRB265H Exposure MAPLHCR PCT 0xidation -
(mwd /ST) (kW/ft). (*F) Fraction 0.030 200 11.5 2154 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 _
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Y1003J01A37 Rev. O Fuel Type: 8DRB283 Exposure MAPLHGR PCT 0xidation (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 10.4 35000 2021 0.019 40000 9.8 1938 0.014 _
e 33
Y1003J01A37 Rev. O 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
- r 25000 11.3 2101 0.024 30000 10.7 2016 0.018 35000 10.1 1916 0.012
- \ 40000 9.4 1823 0.009 _
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