ML20116K659

From kanterella
Revision as of 03:02, 16 May 2020 by StriderTol (talk | contribs) (StriderTol Bot insert)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
Rev 0 to Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plant,Unit 1,Reload 4 (W/O Recirculation Pump Trip)
ML20116K659
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 04/18/1985
From: Charnley J, Lambert P, Zarbis W
GENERAL ELECTRIC CO.
To:
Shared Package
ML20116K608 List:
References
23A4663, 23A4663-R, 23A4663-R00, NUDOCS 8505030349
Download: ML20116K659 (32)
v • d • e


Text

- - - -

23A4663 APRIL 985 l

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 1, RELOAD 4 (WITHOUT RECIRCULATION PUMP TRIP)

[ GENER AL h ELECTRIC

k-23A4663 Revision 0 Class I April 1985 l

l SUPPLEMENTAL REIDAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PIANT UNIT 1, REIDAD 4 (WITHOUT RECIRCUIATION PUMP TRIP)

Prepared: 44, P. A. Lambert Verified:

W. A,. @ bis m

/,

A ro . /.

M S.' Chaprfey, M6fidgel "

Fuel Licensing NUCLEAR ENERGY BUSINESS OPERATIONS

  • GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 95125 GENER AL $ ELECTRIC 1/2

23A4663 Rev. 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 United States Nuclear '

Regulatory Commission (USNRC) for amending CP&L's operating license of the Brunswick Steam Electric Plant Unit 1. 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 infor-mation in this document are contained in the Fuel Contract Supplemental

' Agreement between Carolina Power and Light Company and General Electric Company for Brunswick Steam Electric Plant Units 1 and 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, 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.

3/4

23A4663 Rev. 0

1. PLANT UNIQUE ITEM (1.0)* j Transient Analysis Assumptions: Appendix A Fuel Mechanical Design Methods: Appendix A
2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)

Fuel Type Cycle Loaded Number Number Drilled Irradiated 8DRB265L 2 20 20 8DRB283 2 20 20 P8DRB265H 3 16 16 P8DRB285 3 140 140 P8DRB265H 4 72 72 P8DRB284H 4 72 72 P8DRB299 4 36 36 New BP8DRB299 5 184 184 Total 560 560

  • ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel", NEDE-24011-P-A-6, dated April 1983. A letter "S" preced-ing the number refers to the appropriate country-specific supplement.

5

23A4663 Rev. O l

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 17,518 mwd /ST Miniaua previous cycle core average exposure at end of cycle from cold shutdown considerations: 17,118 mwd /ST Assumed reload cycle core average exposure at end of cycle: 16,764 mwd /ST Core loading pattern: Figure 1

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

Beginning of Cycle, k,gg Uncontrolled 1.110 Fully Controlled 0.957 Strongest Control Rod Out 0.983 R, Maximum Increase in Cold Core Reactivity 0.000 with Exposure into Cycle, &

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin ( &)

m (20*C, Xenon Free) 600 0.033 1

l l

6

23A4663 Rev. 0

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

E0C 5-2000 mwd /ST E0C 5 Void Fraction (%) 41.3 41.3 Average Fuel Temperature (*F) 1279 1279 Void Coefficient N/A* (d/% Rg) -8.45/-10.57 -8.43/-10.54 Doppler Coefficient N/A (d/*F) -0.211/-0.200 -0.221/-0.210 Scram Worth N/A* ($) ** **

7. REIDAD 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 l Exposure: BOC 5 to E0C 5-2000 mwd /ST B P/P8x8R 1.20 1.52 1.40 1.051 6.488 110.9 1.24 8x8R 1.20 1.52 1.40 1.051 6.470 109.9 1.24 Exposure: EOC 5-2000 mwd /ST to E0C 5 B P/P8x8R 1.20 1.43 1.40 1.051 6.083 113.9 1.33 8x8R 1.20 1.46 1.40 1.051 6.198 111.9 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 Measured Scram Time: No Exposure Points Analyzed: 2

  • N = Nuclear Input Data, A = Used in Transient Analysis
    • Generic exposure independent values are used as given in *Ueneral Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-6, dated April 1983.

7

23A4663 Rev. 0

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

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

Exposure Range: B0C 5 to E0C 5-2000 mwd /ST EPR Flux Q/A Transient (% NBR) (% NBR) BP/P8x8R 8x8R Figur e Load Rejection Without Bypass 365 119 0.15 0.13 2a Loss of Feedwater Heater 127 125 0.17 0.17 3 Feedwater Controller Failure 223 116 0.09 0.09 4a Exposure Range: E0C 5-2000 mwd /ST to EOC 5 ACPR Flux Q/A Transient (% NBR) (% NBR) BP/P8x8R 8x8R Figure Load Rejection Without Bypass 526 127 0.26 0.23 2b Loss of Feedwater Heater 127 125 0.17 0.17 3 Feedwater Controller Failure 350 125 0.21 0.19 4b 8

23A4663 Rev. 0

11. LOCAL 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 ACPR MLHGR (kW/ft)

Rod Block Rod Position R ead,ing_ _

(feet withdrawn) BP/P8x8R 8x8R BP/P8x8E and 8x8R 104 3.5 0.12 0.12 17.8 105 4.0 0.13 0.13 18.4 106 4.0 0.13 0.13 18.4 107 4.5 0.15 0.15 18.6 108 5.0 0.16 0.16 18.6 109 6.0 0.19 0.19 18.6 110 10.0 0.23 0.23 18.6 Setpoint Selected: 107

12. CYCLE MCPR VALUES (S.2.2)

Non-Pressurization Events Exposure Range: BOC to E0C BP/P8x8R 8x8R Loss of Feedwater Heater 1.24 1.24 Fuel Loading Error 1.20 Rod Withdrawal Error 1.22 1.22 9

l 23A4663 Rev. O Pressurization Events Option A Option B BP/P8x8R 8x8R BP/P8x8R 8x8R Exposure Range:

BOC 5 to E0C 5-2000 mwd /ST Load Rejection Without Bypass 1.27 1.25 1.08 1.08 Feedwater Controller Failure 1.21 1.21 1.15 1.15 Exposure Range:

E0C 5-2000 mwd /ST to EOC 5 Load Rejection Without Bypass 1.39 1.36 1.27 1.24 Feedwater Controller Failure 1.34 1.32 1.27 1.25

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) s1 y Transient (psig) (psig) Plant Response MSIV Closure 1214 1248 Figure 6 (Flux Scram)

14. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed: 105%

Decay Ratio: Figure 7 Reactor Core Stability Decay Ratio, x2 /*V: 0.73 Channel Hydrodynamic Performance Decay Ratio, x2/ *0 Channel Type BP/P8x8R and 8x8R 0.48 10

23A6663 Rev. 0

15. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes Includes 2.2% Power Spiking Penalty: Yes Event Initial MCPR Resulting MCPR f 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 Functions: 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: 220 cal /gm 11

23A4663 Rev. 0

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

" Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit 1," General Electric Company, November 1978, (NED0-24165, as amended).

Fuel Type: BP8DRB299/P8DRB299 Exposure MAPLHGR PCT Local Oxidation (mwd /ST) (kW/ft) (*F) (Fractions) 200 10.9 2029 0.019 1,000 11.0 2029 0.018 5,000 11.5 2071 0.021 10,000 12.2 2155 0.027 15,000 12.3 2178 0.029 20,000 12.1 2170 0.028 25,000 11.5 2104 0.023 30,000 11.0 2005 0.016 35,000 10.3 1900 0.011 40,000 9.7 1820 0.008 45,000 9.0 1745 0.006 Fuel Type P8DRB285 Exposure MAPLHGR PCT Local Oxidation (mwd /ST) (kW/ft) (*F) (Fractions) 200 10.9 2038 0.019 1,000 11.0 2048 0.020 5,000 11.8 2141 0.026 10,000 12.3 2177 0.029 15,000 12.2 2174 0.028 20,000 11.8 2131 0.025 I

25,000 11.0 2031 0.018 30,000 10.4 1928 0.012

)

35,000 9.8 1844 0.009 40,000 9.2 1761 0.007 12

23A4663 Rev. O Fuel Type: P8DRB265H ,

l

)

Exposure MAPLHGR PCT Local Oxidation (mwd /ST) (kW/ft) (*F) (Fractions )

200 11.5 2103 0.024 1,000 11.6 2111 0.024 5,000 11.9 2135 0.025 10,000 12.1 2147 0.026 15,000 12.1 2157 0.027 20,000 11.9 2138 0.025 25,000 11.3 2063 0.020 30,000 10.7 1977 0.015 35,000 10.3 1891 0.011 40,000 9.6 1801 0.008 13

23A4663 Rev. 0 sMMMMEs

sMMMMMMMMMs
MMMMMMMMMMM
eMMMMMMMMMMMs
M E M M M M M M M M M M M

':MEMMMMMMMMMMM

MMMMMMMMMMMMM
MMMMMMMMMMMMM
MMMMMMMMMMMMM

': "MMMMMMMMMMM" l' MMMMMMMMMMM

"MMMMMMMMM"
"MMMMM" l IIIIIIIIi 1 3 5 7 9111315171921232527293133353739414345474951 PUEL TYPE A = 8DRB265L E = P8DRB265H B = 8DRB283 F = P8DRB284H C = P8DRB265H G = P8DRB299 D = P8DRB285 H = BP8DRB299 Figure 1. Reference Core leading Pattern 14

23A4663 Rev. 0 1 NEU*.RON FLUK 1 VESSEL PRESS RISE (PSI) 2 AV'. SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 C";RE INLET TLOW 3 RELIEF VALVE FLOW 15 0.0 . 300.0 = eve *S? untuE etnu l J 0 0. 0 N 200.0 j -  %

b 5

v t' -

" 50.0 100.0

-- 0 0  ; _:

0. 0 . 0.0 , , , , , ,, ,, ,, ,, ,
0. 0 2.0 4.0 6.0 0.0 2.0 4. 0 6. 0 TIME (SECOWS1 TIME (SECONOS) 1 LEVEL (INCH-REF-SEP-SMRT)  ! VO! REACTIVITY 2 VESSEL STEA1 FLOW 2 00P ER REACTIVITY 3 TURBINE STEW LOW 3 SCRA REACTIVITY 200.0 2 rEE5mTEo e_ rw I.0 e n v i_i_ acar ryugev f

M E f~

,00.0 . . . . <h= #

a 0.0 . /

8 C

x.__ * .

F G y O 00 <

- . } ., . !d -1.0 g

l

-100.0 2.0

0. 0 2.0 4.0 8.0 0.0 2. 0 4.0 8.0 TIME (SECONOSI TIME (SECONOSI Figure 2a. Plant Response to Generator Imad Rejection Without Bypass (E0C 5-2000 mwd /ST) 15

23A4663 Rev. 0 1 NEUTRON FLU ( l VESSEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 150.0 300.0 t eyo_SSE uituE rLey 91,0.0 .

~

A,N 200.0 I

}[ h g W N m

!E 50.0 100.0

-P- 0- 0 0 0

0. 0 0.0

~ . . . _ L. _.

0. 0 2. 0 40 6.0 0.0 2. 0 4.0 6.0 TIPE (SECONDS) TIME (SECONOS) 1 LEVEL (INCH-REF-SEP-SMRT) i VO REACTIVITY 2 VESSEL STEA1 FLOW 2 00P ER REACTIVITY l0W 200.0 E nu l.0 3 h$)L k)h Euf 5

~ .

A E

) /

7-

,00.0 .: . Li_ .e .

E 0.0 . .

I .

N NM l

, ; - b.

0. 0 i .l.0 l 1

-100.0 2.0

0. 0 2.0 4.0 6.0 0.0 2.0 4. 0 6. 0 1 TIME (SECOND$1 TIME (SECONO$1 Figure 2b. Plant Response to Generator Inad Rejection Without Bypass (EOC 5) 16 l

l

I l

23A4663 Rev. 0 150.0 1 NEUIRON FLUX 1 VESBEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 RELIEF VALVE FLOW 3 CORE INLET FLOW 3 BYPLSS VALVE FLOW 150.0 ' C0= r  !_E' ST

- ~ 7;_%%% 100.0 B .

r t" 200.0 '  ;  ;  ;  ;  ;  ;  ;  ; '

& l b

.- 50.0 5

W t

50.0 0.0 ,_

__ _2 I_ -

0. 0 L 0.0 100.0 200.0 0. 0 100.0 200.0 TIME (SECONOS) f!ME (SECONOS)

ILEV EL(INCH-REF-SEP.SKRT) i VOI ) REACTIVITY 2 VES sEL STEAMFLOW 2 DOP'LER REACTIVITY 3 IUR llNE STEAMFLOW 3 SCRLM REACTIVITY 136.0 ' rer?u nTEn e' Ou 1.0 m egtig neartivity i

2, ___A .*ANA AA $

.0 j yf% v, ,

- i~= W ' iia V 'vyy y w

,, ,-_; ,- n , _

, u.. ,,-_u._A. .___L. ,,

- 2 - _ g _ -A C

E e-50.0 W .l.0 W o

-- ; - -- : -  ;. _@ f 0= 0 .n , 2.0 0.0 100.0 200.0 0. 0 100.0 200.0 TIME (SECON0$1 f!ME (SECONOSI Figure 3. Plant Response to Loss of 100*F Feedwater Heating 17

23A4663 Rev. 0 150.0 1 N1 TR FLU ( l VESSEL press RISE (PSI) 2 A' S ACE HEAT FLUX 2 SAFETY VALVE FLOW 3 Ci l EI ET TLOW 3 REL EF VALVE FLOW 150.0 'E aE 'u Et '=

4 BYP S VALVE Flow l

M N 100.0 1

100.0 -

l s

50.0 w

1 w

r 4

" 50.0 y 4M i+ 4 0.0 ., , , _

m.,,, , . , , m . . . . -

0. 0 .
0. 0 10.0 20.0 30.0 0. 0 10.0 20.0 30.0 f!ME (SECONOS) TIME (SECONDS)

ILEVEL(!NCH-DEF-SEP-SKRT) 1Vd!O aEACTIVITY 2 VESSEL STEA9 FLOW 2 00PPL R REACTivtTY 3 TURBINE STEW LOW 3, S, CRAM,REACT!vlTY 150.0 'reenwaTen r_gu 1.0 ,

iv _ _ e g _a_ c truity 0 0 4 0  ? .

/

s

. ~  :;; ,f r 0 a 0.0

  • 100.0 _

\ 8 J\ E l 50.0 ,/ ( l ,

l \ C h

y

.l.0 0.0 ,,, , 2,g

0. 0 10.0 20 . 0 30.0 0. 0 10.0 20.0 30.0 flME (SECONOS) f!ME (SECON05)

Figure 4a. Plant Response to Feedwater Controller Failure (EOC 5-2000 mwd /ST) 18

23A4663 Rev. 0 150.0 1 NElITRON LU( l VESSEL PRESS RISE (PSI) 2 AVl. SURF CE HEAT FLUX 2 SAFETY VALVE FLOW 3 C0 t : IPL T TLOW 3 RELI F VALVE FLOW 150.0 ' C0 ':

'"__. E 'M 4 BYP/b VALVE FLOW M 1 100.0

, - - > Y 100.0  ;  ;  ;  ;

a:

b I

}

g 50.0 ~ r U

M E

50.0 <  !  ;- 0 0.0

,-a, a . , - a ...,

0. 0 ,
0. 0 10.0 20.0 30.0 0. 0 10.0 20.0 30.0 TIME (SECONOSI TIME (SECONOS) 1 LEVEL (INCH-REF-SEP-SKRT) i VO ACT!v!TY 2 VESSEL STEA1 FLOW 2 D0f REACTIVITY 3 TURBINE STERMFLOV 3, S,CI EACT.IVITY 130.0
  • rEE09atEo e_Cu 1.0 g f,_ eige3uerv 0 i i ?  ? 0 k

- -  :: t@ E .

. ~. /

,00.0 t- -

5 0.0 E. -

p'm g m.- ,'

31

% \p s 8 'l

\ E 50.0

/  :

I -1.0 w

0. 0 - -2.0 0.0 10.0 20.0 30.0 0. 0 10.0 20.0 30.0 TIME (SECONOS) TIME ISECONOS)

Figure 4b. Plant Response to Feedwater Controller Failure (EOC 5) 19

23A4663 Rev. 0 2 6 10 14 18 22 26 30 34 38 42 46 53 51 16 16 47 46 46 43 16 12 6 6 12 16 39 35 12 6 20 20 6 12 31 40 27 6 20 0 0 20 6 23 40 19 12 6 20 20 6 12 15 11 16 12 6 6 12 16 7 46 46 3 16 16 NOTES: 1. NUMBER INDICATES NUMBER OF NOICHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN ROD.

2. ERROR ROD IS (22,27).

s Vigure 5. Limiting Rod Withdrawal Error Rod Pattern i' 20 l

l 23A4663 Rev. O I NEUTRON F_U X 1 VESSEL PRESS RISE (PSI) 2 AVE SURFA:E HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET FLOW 3 RELIEF VALVE FLOW 150.0 300.0 ' eye

  • SS u *_uE rLeu a ,

id 100.0 er

~

200.0 E h b

s.

5 W

W 50.0 100.0 '

0 0 O  ; -

0. 0 . 0.0 , , , ,, ,, _ , ,
0. 0 5.0 0.0 5.0 TIME (SECOE S1 TIME (SECOWS) iLEVEL(INC4-REF-SEP-SKRT 3 REACTIVITY 2 VESSEL STEAMFLOW 2 DOPP RREACTIVITY 3 TURBINE STEAMFLOW 3 SCR AM EA CT

'rEE0uivEn rLOy 1.0 * ' vge _it ri 7,IVITY 3ugev 200.0 3

3 1 2 - ^- ' ^

i-' 5- 00 m 100.0

-p-- ~ ; '7 ;____; 5 k

~ '

8 C

w C 08

. h,-1.0

-100.0 -2.0

0. 0 5.0 0.0 5.0 TIME (SECONOSI TIME (SECONOSI Figure 6. Plant Response to MSIV Closure (Flux Scram) 21

/

23A4663 Rev. O

\

AN ATURAL CI RCULATION B1 05 PERCEb T ROD LI NE CU _T. PERFC RMANCE L IMIT 1.00 (: c A

.75

/

I 3

.50 ,

3 t

7 r

.25 i

F4 0 0.00

0. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure 7. Reactor Core Decay Ratio i

22 i

L_

23A4663 Rev. 0

0. 0 .

-5.0

-10.0 sH i ff s

-15.O f Z

r <

w g -20.0

- (

it w

8 -25.O c::

u 8

o

-30.0 l -35.0 ,c,,c,,, ,r en u ,, ,,e cn, n l

BdhLdUL ki5 VAE 5- 5 5~

C BOUND /AL 280 cal./G COLD

! D BOUND /AL 280 cal /G HSB l -40.0

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

Figure 8. Puel Doppler Coefficient in 1/A*C 23

e 23A4663 Rev. 0 20.0 17.5 15.0 0

12.5 #U U U r

?

gj -10.0 ,

a e .

s t- 7.5 E

w a

W 5.0 w

! 2.5 A ACCIDENT FUNCTION B BOUNDING VALUE 280 CAL /G

0. 0
0. 0 5.0 10.0 15.0 20.0 ROD POSITION, FEET OUT Figure 9. Accident Reactivity Shape Function, Cold Startup 24 j

L

23A4663 Rev. 0 20.0 17.5 15.0 A a

Il 4%

Y 12.5 Y?U U r

?

Gj 10.0 ca C 7.5 t

r O

E 5.0 2.5 A ACCIDENT FUNCTION B BOUNDING VALUE 280 CAL /G O. 0 l 0. 0 5.0 10.0 15.0 20.O ROD POSITION, FEET OUT Figure 10. Accident Reactivity Shape Function, Hot Standby 25

r 23A4663 Rev. 0 40.O A SCRAM F JNCTION B BOUNDIN 3 VALUE 280 CAL /G 35.0  !

g 30.0 b

25.0

= l e

a n

n 20.0 8

5 3

o 15.0 -

/

c u 10.0 j ,

x 5.0 ,

0. 0 .. A

. . , w -

0. 0 1.0 2. 0 3.0 4.0 5.0 6. 0 ELAPSED TIME, SECONDS 1

l

! Figure 11. Scram Reactivity Function, Cold Startup 26

23A4663 Rev. 0 50.O A SCRAM F JNCTION 8 BOUNDIN 3 VALUE 280 CAL /G 40.0 C

x ,

< 30.0 /

d a

G E

.20.0 n C

E r

M E 10.0 a

0. 0 m,a 1

,, m

0. 0 1.0 2. 0 3.0 4.0 5.0 6. 0 ELAPSED TIME, SECONDS

}

Figure 12. Scram Reactivity Function, Hot Standby 27/28

23A4663 Rsv. O APPENDIX A ADDITIONAL INFORMATION

1. 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.
2. General Electric's approved fuel thermal mechanical design model, TEXICO (documented in Revision 6 to NEDE-24011-P-A), was used in the analysis of Brunswick 1, Reload 4.

i l

l l

l 29/30 (FINAL) i

a m , u,,- s -,---, *--.- - -- & _ .--m. - - --- ------ s- .m - . 1= ,

[

1 I

i i

i 5

4 I

1, I

4 1 l

1 l

l l

l t

GENER AL h ELECTRIC ,

j l

, i

-- . - - . . . . . . , . _ _ . - . - -,._-_n..,_ __ -_-_. - _ - _ _ ___.,-.,_,nn, _ _ ,,,-~,,,-,--,- ,n nna, ., ,

Retrieved from "https://kanterella.com/w/index.php?title=ML20116K659&oldid=2807323"