ML20092N840
| ML20092N840 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/31/1984 |
| From: | Friday J, Gridley R, Lambert P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20092N825 | List: |
| References | |
| 23A1765, 23A1765-R, 23A1765-R00, NUDOCS 8407050114 | |
| Download: ML20092N840 (24) | |
Text
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23A1765 CLASSI M AY 1984 1
SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2, RELOAD 5 J .-
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23A1765 Revision 0 Class I May 1984 SUPPLEMENTAL REIDAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2. RELOAD 5
&,v Prepared: ~ P. A. Lambert Fuel Licensing Verified: - -
V_
( J. D. Frfday
/ Fuel Licensing Approved: dI E R. L. Gridley, ManagS /
Fuel & Services Licensing ECLEAR ENERGY BUSINESS OPERATIONS
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENER AL h ELECTRIC 1/2 I
23A1765 Rev. 0 1MPORTANT 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 informa-tion in this document are contained in the Supplemental Agreement to the Con-tract 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 Coepany 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 contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liabillcy or damage of any kind which may result from such use of such information. __
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23A1765 Rev. 0
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- 1. 7LANT UNIQUE ITEM (1.0)*
m .
N Plant Parameter Changes Appendix h
- 2. % REIDAD FUEL BUNDLES (1.0, 2. 0, 3. 3.1 AND 4. 0) ,
f
-~
- ,j'A " Cycle-Loaded Number Number Drilled s **\
Irrarlipred - r 8DB274L._ 2 18 8 8DRB265H 3 , 40 40 8DRB283 ,
, 3 36 36 P8DRB265H - % 4 132 132 P8DRE284H
% '3 ,, 24 24 s , ,
P8DRB265}i C. - --
. 5 136 136
~
New -
BP8DRB99 cx 6 184 184
~
TOTAL -
560 560
- 3. , REFERENCE CORE IDADING ATTERN (3.3.1)
L. '^
Nominal,' previous cycle core average exposur,e at,
.o end of cycle: N' '
15680 mwd /ST
,s.
Minimum previous cycle cord' average expdsuge at
'end of cycle f rom cold shutdown considerations: 156b0 mwd /ST Assumed re16ad cycle core average exposure at end of cycle: - ,
16336 mwd /ST
- . . s.
Core loading pattern: ,. ; Figure 1
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- ( N Refers to area of discussion in " General Electric Standard Application for ReactnFiuel]" .,NEDE-24011-P-A-6. April 1983. A letter "S" preceding the number o refers to i eh appropriate section in the United States Supplement, NEDE-24011-P-A-6 US, April'1983. *
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l l 23A1765 Rev. 0 1
- 4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONIROL SYSTEM WORTH -
NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)
Beginning of Cycle, k ff Uncontrolled 1.110 Fully Controlled 0.955 Strongest Control Rod Out 0.986 R, Maximum Increase in Cold Core Reactivity with Exposurc into Cycle. Ak 0.0
- 5. STANDBY LIQUID CONTROL SYSTEM SHUIDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (A)
]3ng (20*C, Xenon Free) 600 0.036
- 6. REIDAD 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) 1292 Void Coefficient N/A* (c/% Rg) -7.00/-8.75 Doppler Coefficient N/A (c/*F) -0.175/0.166 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-6-US, April 1983.
6
23A1765 Rev. 0
- 7. REIDAD UNIQUE GETAB TRANSIENT ANALYSIS INIT!AL CONDITION PARAMETERS (S.2.2) ea ing actors Fuel Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (1000 lb/hr) MCPR
__ (MWt)
Exposure: BOC to E0C-2000 mwd /ST BP/P8x8R 1.20 1.55 1.40 1.051 6.583 113.5 1.23 8x8R 1.20 1.54 1.40 1.051 6.545 112.3 1.23 8x8 1.22 1.40 1.40 1.098 5.975 112.0 1.23 Exposure: E0C-2000 mwd /ST to EOC BP/P8x8R 1.20 1.42 1.40 1.051 6.062 116.5 1.34 8x8R 1.20 1.46 1.40 1.051 6.199 114.2 1.31 8x8 1.22 1.33 1.40 1.098 5.674 113.9 1.30
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization: No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: Yes l
Improved Scram Time: No Exposure Dependent Limits: Yes Exposure Points Analyzed: EOC and EOC-2000 mwd /ST
- 9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)
Single Loop Operation: No Load Line Limit: No Extended Load Line Limit: No Increased Core Flow: No Flow Point Analyzed: N/A Feedwater Temperature Reduction: No 7
23A1765 Rev. O I
- 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)
Exposure Range: BOC to E0C-2000 mwd /ST ACPR Flux Q/A Transient (% NBR) (% NBR) BP/P8x8R 8x8R 8x8 Figure Load Rejection Without Bypass 469 119 0.16 0.14 0.13 2a Loss of Feedwater Heater 126 124 0.16 0.16 0.16 3 Feedwater Controller Failure 115 110 0.05 0.05 0.05 4a Exposure Range: EOC-2000 mwd /ST to E0C Flux Q/A ACPR Transient (% NBR) (% NBR) BP/P8x8R 8x8R 8x8 Figure Load Rejection without Bypass 516 128 0.27 0.24 0.23 2b Loss of Feedwater Heater 126 124 0.16 0.16 0.16 3 Feedwater Controller Failure 141 111 0.05 0.05 0.05 4b
- 11. IDCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(S.2.2,1)
Limiting Rod Pattern: Figure 5 Includes 2.2% Power Spiking Penalty: Yes O ( ! )
Rod Block Rod Position Reading (feet withdrawn) BP/P8x8R 8x8R 8x8* BP/P/8x8R 8x8*
104 3.5 0.12 0.12 17.68 105 4.0 0.15 0.15 18.23 106 4.0 0.15 0.15 18.23 107 4.5 0.17 0.17 18.40 108 5.0 0.19 0.19 18.40 109 5.5 0.20 0.20 18.40 110 6.5 0.23 0.23 18.40 Setpoint Selected: 107
- 0n periphery of core (low power region) and not limiting 8
23A1765 Rev. 0
- 12. CYCLE MCPR VALUES (S.2.2)
Non-Pressurization Events Exposure Range: BOC to EOC BP/P8x8R 8x8R 8x8 Loss of Feedwater Heater 1.23 1.23 1.23 Fuel Loading Error 1.20 Rod Withdrawal Error 1.24 1.24 Pressurization Events Option A Option B BP/P8x8R 8x8R 8x8 BP/P8x8R 8x8R 8x8 Exposure Range:
BOC to EOC-2000 mwd /ST Load Rejection 1.28 1.26 1.25 1.09 1.08 1.08 Without Bypass l
Feedwater Controller 1.17 1.17 1.17 1.11 1.11 1.11 Failure Exposure Range:
EOC-2000 mwd /ST to EOC Load Rejection 1.40 1.37 1.36 1.28 1.25 1.24 Without Bypass Feedwater Controller 1.17 1.17 1.17 1.11 1.11 1.11 l Failure
- 13. OVERPRESSURIZATION ANALYSIS SlNMARY (S.2.3) sl v Plant Transient (psig) (psig) Response MSIV Closure 1218 1255 Figure 6 (Flux Scram) 9
23A1765 Rev. 0
- 14. STABILITY ANALYSIS RESULTS (S.2.4)
Rod Line Analyzed: 105%
Decay Ratio: Figure 7 Reactor Core Stability Decay Ratio, x2 /*0: 0.67 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 Channel Type BP/P8x8R 0.20 8x8R O.20 8x8 0.28
- 15. IDADING 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 R0D DROP ANALYSIS RESULTS (S.2.5.1)
Bounding Analysis Results:
Doppler Reactivity Coefficient: Figure 8 Accident Reactivity Shape Functione: Figures 9 and 10 Scram Reactivity Functions: Figures 11 and 12 Plant Specific Analysis Results:
Parameter (s) not Bounded Cold: None Resultant Peak Enthalpy, Cold: N/A Parameter (s) not Bounded HSB: Accident Reactivity Resultant Peak Enthalpy HSB: 197.6
- ACPR penalty of 0.02 for the tilted misoriented bundle is applied to the cycle MCPR value reported in Section 12.
10
23A1765 Rev. 0
- 17. IDSS-OF-COOLANT 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).
Fuel Type: BP8DRB299 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 10.9 2061 0.022 1000 11.0 2063 0.022 5000 11.5 2108 0.025 10000 12.2 2194 0.032 15000 12.1 2198 0.032 20000 12.0 2197 0.032 25000 11.5 2141 0.027 30000 11.0 2048 0.020 35000 10.3 1960 0.029 40000 9.7 1846 0.009 45000 9.0 1772 0.007 11
23A1765 Rev. O 9
- ' eMMMMM.
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- MMMMMMMMMMMMM
- MMMMMMMMMMMMM
'::MEMMMMMMMMMMM
- "MMMMMMMMMMM" MMMMMMMMMMM
- "MMMMMMMMM"
- "MMMMM" lIIIIIIIIi 1 3 57 9111315171921232527293133353739414345474951 FUEL TYPE A = 8DB274L E = P8DRB265H B = 8DRB265H F = P8DRB284H C = 8DRB283 H = BP8DRB299 D = P8DRB265H Figure 1. Reference Core Loading Pattern 12
23A1765 Rav. 0 i NEUTRON FLU ( l VESSEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 15 0. 0 300.0 i evo AS? "* L'-E rLCu 100.0 1
% 200.0 (r
50.0 100.0 3
f
- 0. 0 . . 0.3 _ , , f.
0.0 2.0 4.0 6.0 0.0 2.0 4.0 6.0 TIME (SECONOS) TIME (SECONOS)
I LEVEL (INCH-REF-SEP-SKRT) i VO! REACTIVITY 2 VESSEL STEA1 FLOW 2 DOP ER REACTIVITY 200.0 .___. . e_ . h -_ . .
1.0 ! hk_0.), N)h ._
O 3
100.0 -
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-1.0 .
100.0 2.0
- 0. 0 2.0 4.0 6.0 0.0 2.0 4.0 6.0 f!ME (SECONOS) flME (SECONOS)
Figure 2a. Plant Response to Generator Load Rejection Without Bypass (E0C-2000) 13
\
f 23A1765 Rev. 0 1 NEUTRON FLUk 1 VESSEL PRE h RISE (PSI) 2 AVE SURF ACE HEAT FLUX 2 SAFETY VAltt FLCW 3 CORE INLET PLOW 3 RELIEF VALVE FLOW 130.0 300.0 _ svor:s " a.L- rLcu 100.0 h- --
a-200.0 f
~
50.0 \ 300.0 w- : - ;-- ;_
- - ; , =;: L : = :: :
4.0 6.0 0.0 2.0 4.0 6.0
- 0. 0 2.0 TIME (SECONDS) TIME (SECONOS) 1 LEVEL (INCH- REF-SEP-SKRT) 1 10 REACTIVITY 2 VESSEL STEA1 FLOW 2D LER REACTIVITY 3 TURBINE STE -FFLOW 3 SCR IVITY 200.G e rEEtwatEn r_gu 1.0 a ver ni RE A._CT cr ctruerv
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100.0 , , Aka E 0. 0 . ] /
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--100.0 -2.0 I 4,0 6.0 0. 0 2.0 4.0 6.0
- 0. 0 2.0 TIME (SECONDS) TIME (SECONOS)
Figure 2b. Plant Response to Generator Load Rejection Without Bypass (EOC) l 14
23A1765 Rev. 0 1 NEUTRON FLUX IVESLEL PRESS RISE (PSI) 2 AVE SURF ACE FC.i.' FLUX 2 RELlEF VALVE FLOW 3 CDRE INLET FLOW 3 BfFASS VALVE FLOW 150.3 *cy r "tET s '_*
T
~
', T>' 100.0 100.0 0 0 0 0 'O O O O I
50.0 1
50.0 0.0 ;3 3 0; 2 ;0 3 3;I
- 0. 0 0.0 100.0 200.0 0. 0 100.0 200.0 TIME (SECONOSI TIME (SECONOS)
ILEVEL(INCH REF-SEP-SKRT) 1 VOI ) REACTIVITY 2 VESSEL STEAMFLOW 2 DOP'LER REACTIVITY 3 TUR31NE STEAMFLOW 3 SCR CT IS O. 0 ' egg,un tro ergu 3, o a 797 .tLMogREA.cr!v!TY tug 7v
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50.0 W.3 ,o e
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-2.0
- 0. 0 100.0 200.0 0. 0 100.0 200.0 flME (SECONOS) TIME (SECONOS)
Figure 3. Plant Response to Loss of 100*F Feedwater IIcating 15
23A1765 Rev. 0
,_ 150.0 1NEU. CNFluk 1 VESSEL PREEh RISE (PS!J 2 AVE _ FACE HEAT FLUX 2 SAFETY V ALVE FLOW 3 MLET rLOW 3 RELIEF VALVE FLOW 130.0 N_rDRE tecE 'LET ' '_? 4 BYPASS VALVE FLOW 100.0 g a g, g 4
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- 50.0
{W r~
50.0 i
,d tb 4
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- 0. 0 .
p
- 0. 0 10.0 20.0 30.0 0. 0 10.0 20.0 30.r IIME (SECONOS) ilME (SECONOS) 1 LEVEL (INCH-7EF-SEP-SKRT) i V01 hEACTIVITY 2 VESSEL STEAMFLOW 2 PLFR REACTIVITY 3 TURBINE STEAMFLOW 3, SC AM RE ACT!vlTY 150.0 e FEEgunTEc r_gu 1.0 gug7y
, , , , , 79l4 - f E N
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50.0 / h . .0 W
. -2.0 u
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0.0 10.0 20.0 30.0 0.0 10.0 20.0 30.0 flME ttECON05) flME (SECONOS)
Figure 4a. Plant response to Feedwater Controller Failure. Maximum Demand (EOC-2000) 16
23A1765 Rev. 0 s 150.0 1 NI.UTR$4 Flut i VESSEL PRES 3 RISE (PSI) l 2 A1'E SL?F/CE HEAT FLUX 2 SAFETY V ALVE FLOW 3 FLOW 3 RELIEF VALVE FLOW 150.0 ,
3 100.0 . : : y 50.0 50.0 0.0 x= :=, =
- 0. 0 .
- 0. 0 10.0 20.0 30.0 0.0 10.0 20.0 30.0 TIME (CECONOS) TIME (SECONOS) l I LEVEL (INCH- 4EF-SEP-SMRT) 1 VO!D dEACTIVITY
- 2 VESSEL STEA1 FLOW 1
2 DOP E 7 REACTIVITY
, 3 TURBINE STEWLOW REACTIV!fY i 150.0 'Jsscwnsa s _gu 1.0 SCR _ g3crlyugtv 3, 191 ;
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[ 100.0 .
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0.0 ~/' % r
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- . - *1.0 0.0 10.0 20.0 30.0 0.0 10.0 20.0 30.0 flME ($tCONoll flME (SECON05)
Figure 4b. Plant Response to Feedwater Controller Failure, Maximum Demand (EOC) 17
23A1765 Rev. O NOTES: 1. ROD PATIERN IS 1/4 CORE MIRROR SYMMETRIC.
- 2. NO. INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN ROD.
- 3. ERROR ROD IS (22, 31) .
2 6 10 14 18 22 26 51 32 32 47 2 6 43 32 36 36 39 2 6 14 35 32 36 44 44 31 6 14 0 27 32 36 44 44 Figure 5. Limiting Rod Withdrawal Error Rod Pattern 18 t
23A1765 Rev. O I NEUTRON F_UX l vre T . rn' .. . ,c SI) 2 AVE SURFr:E HEAT FLUX 2 SA~C' f VALVE FLOW 3 CORE INLET FLOW 3 RELIEF VALVE FLOW 150.0 300.0 i nPa 2 va'yE ettu a
W 100.0 ~# 200.0 r
l ~ 30 0 100.0 i
i
- 0. 0 ._ . , 0.0
- 0. 0 5.0 0.0 5.0 I!ME (SECONOSI TIME (SECONOS) 1 LEVEL (INCr4-REF-SEP SKRT) 1 VOIO REACTIVITY 2 VESSEL ST[AMFLOW PPLER F[ ACTIVITY 3 IURBINE STEAMFLOW 3, S REAC11VITY 200.0 m rEEgym.vgo eggu 1.0 .
vg7.t eg.71;y37y b .
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- t 3 0.0 100.0 0,W -- %:,> 4 '
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} , , 3 g 500.0 -2.0 C. 0 5.0 8.0 5.O ilME t%COND$1 ilME (SECOM)$1 Figure 6. Plant Response to MSIV Closure (Flux Scram) 19 L.
23A1765 Rev. O Ab ATURAL C :RCULATIO 4 81 05 PERCENT ROD LI 4E CL LTIMATE STABILITY LINE 1.00 C c:
G .75 A s
(N X
b s
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0.00
- 0. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure 7. Reactor Core Decay Ratio 20 h
23A1765 Rev. 0
- 0. 0
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.0 4 ggLggL 37gg y3Lg 7 ggLg B CALCUL ATED VALUli-HSB C BOUND VAL 280 CAL /G COLD D BOUNO VAL 280 C AL/G HSB
-40.0
- 0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE DEG C.
Figurc 8. Fuel Doppler Coefficient in 1/A*C 21
23A1765 Rev. 0 20.O I A ACCIDENT FUNC TION 8 BOUNDING VALU E 280 CAL /G 17.5 l
15.0 m
o I
0 00 0 W 12.5 x
W Ej 10.0 a
g A A s
- 7.5 s
o
$ 5. 0 u
- 2. 5
- 0. 0
- 0. 0 5.0 10.0 15.0 20.0 R0D POSITION, FEET OUT Figure 9. Accident Reactivity Shape Function Cold Startup 22
23A1765 Rev. 0 20.O A ACC1DEilT FUNC TION E EOUNDING VALU E 280 CAL /G 17.5 15.0
,2 i
W 12.5 ~@M^ 0 z /
l$
lgj 10.O o
b 7.5 E
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- 0. 0 5.0 10.0 15.0 20.0 ROD POSITION, FEET OUT Figure 10. Accident Reactivity Shapo Function llot Startup 23
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23A1765 Rev. 0 4'.O A SCRAtt FUt,CTION B BOUT:DI NG VALIJE 280 CAL /G 35.0 m
o 30.0 I
w 25.0 F-
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Q m 20.0 !
4 e
w z
v g 15.0 _.
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cr 5.0 1 A0 /
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- 0. 0 1.0 2. 0 3. 0 4.0 5.0 f;.0 ELAPSED TIME, SECONDS Figura 11. Scram Reactivity Funetton Cold cartup 24 i
23A1765 Rev. 0 60.0 A SCRAM FUNCTION 8 BOUNDING VALUE 280 CAL,0 40 0
/
l 30 0 a
.2 a
s N
3 20.0 9 E
10 0 -
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/
/
/
00 #'ii A.1
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00 1.0 20 30 40 60 00 EL APSED TIMt. SECONDS Figurn 12. Scram Renetivity Function llot Startup 25/26 L _ _ _ _ _ _
23A1765 Rev. O APPENDIX A Plant Parameter Changes:
l Pressure Relief Systems (Table S.2-4.1, pg. US.2-93 NEDO-24011) l Safety / Relief Valve Type: E l (i.e., capacity at reference pressure of 1080 +3% psig is 789,000 lb/hr for each S/RV) l 27/28 (FINAL)
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