ML20114D355

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Rev 0 to 23A1813, Supplemental Reload Licensing Submittal for Cooper Nuclear Power Station Unit 1,Reload 9
ML20114D355
Person / Time
Site: Cooper Entergy icon.png
Issue date: 11/30/1984
From: Charnley J, Lambert P, Zarbis W
GENERAL ELECTRIC CO.
To:
Shared Package
ML20114D343 List:
References
23A1813, 23A1813-R, 23A1813-R00, NUDOCS 8501310230
Download: ML20114D355 (74)
v • d • e


Text

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' d's's A NOVEMBER 1984

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SUPPLEMENTAL RELOAD
LICENSidG SUBMITTAL FOR COOPER NUCLEAR POWER STATION j UNIT 1, RELOAD.9 i

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[BA'ih8ER S885$! GEN ER AL h ELECTRIC

23A1813 Revision 0 Class I November 1984 P SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR COOPER NUCLEAR POWER STATION, UNIT 1 RELOAD 9 Prepared:

  • P. A. Lambert Fuel Licensing Verified: (

W. A. bis Fuel nsing Approv d- - #

,4 S. Charnley -

uel Licensing Ma er I

l NUCLEAR ENERGY BUSINESS OPERATIONS

  • GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 95125 GENERAL $ ELECTRIC 1/2 k _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___

h 23A1813 R:v. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY P

l This report was prepared by General Electric solely for Nebraska Public a Power District (NPPD) for NPPD's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending NPPD's operating license of the Cooper I Nuclear Station. The information contained in this report is believed by i

General Electric to be an accurate and true representation of the facts known, obtaincd 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 contract between Nebraska Public Power District and General Electric Company for nuclear fuel and related services for the nuclear system for Cooper Nuclear Station and nothing con-tained in this document shall be construed as changing said contract. The use of this information except as defined by said contract, or for any pur-pose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric 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.

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> 3/4 b . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . . . _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _

23A1813 Rev. 0

1. PLANT UNIQUE ITEM (1.0)*

Not Applicable f

L

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

!~

> Fuel Type Cycle Loaded Number Number Drilled Irradiated 8DRB283 5 12 12 P8DRB265L 6 8 8 P8DRB283 6 72 72 P8DRB265L 7 36 36 P8DRB283 7 76 76 P8DRB265L 8 56 56 P8DRB283 8 56 56 P8DRB265L 9 56 56 P8DRB283 9 60 60 New P8DRB265L 10 88 88 P8DRB283 10 28 28 I

Total 548 548

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end -

of cycle: 18291 Wd/ST Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 17883 Wd/ST Assumed reload cycle core average exposure at end l of cycle: 17356 Wd/ST f

Core loading pattern: Figure 1

  • ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-6, dated April 1983. A letter "S" pre- l ceding the number refers to the appropriate country-specific supplement.

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4. CALCULATED 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 1.109 Fully Controlled 0.955 Strongest Control Rod Out 0.989 R, Maximum Increase in Cold Core Reactivity 0.000 with Exposure into Cycle, Ak

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Sr.utdown Margin (ak) ,

E2m (20*C, Xenon Free) '

600 0.038 4

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

(COLD WATER INJECTION EVENTS ONLY)

Void Fraction (%) 39.6 Average Fuel Temperature (*F) 1286 Void Coefficient N/A* (C/% Rg) -6.36/-8.20

. Doppler Coefficient N/A (C/*F) >

-0.197/-0.187

. Scram Worth N/A ($) **

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l

  • 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-240ll-P-A-ti, dated April 1983..

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23A1813 Rev. 0 p f7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS s

em (S.2.2)

Peaking Factors Bundle Power Bundle Flow Initial Design' Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR Exposure: BOC.to E0C-1000 mwd /ST P8x8R Mt .E0 "1.50 1.40 1.051 6.383 110.3 1.28

-8x8R. 1.20 - 1.53 1.40- 1.051 6.486 108.4 1.25 t

1 . Exposure: EOC-1000 mwd /ST to EOC P8x8R' 1.20 1.48 1.40 1.051 6.301 110.8 1.30 8x8R 1.20 1.51 1.40 1.051 6.397 109.0 1.27

8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No

t. Recirculation Pump Trip: No i Rod Withdrawal' Limiter: No Thermal Power Monitor:' No Improved Scram Time: Yes (Option B)
Exposure Dependent _ Limits: 'Yes .

-Exposure. Points Analyzed:. EOC, EOC-1000 mwd /ST

9. OPERATING FLEXIBILITY-OPTIONS (S.2.2.3)

' Single Loop Operation: .Yes l Load Line Limit: Yes

[ Extended Load Line Limit: No

! Increased Core Flow: .No Flow Point Analyzed: N/A Feedwater Temperature. Reduction: No e

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[_ _ - - _ _ _ _ _ - - - _ - - _ - _ - _ _ _ - - _ _ _ _ _ _ .

I 23A1813 Rav. 0

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

Exposure Range: BOC to EOC-1000 mwd /ST ACPR Flux- Q/A Transient (% NBR) (% NBR) P8x8R 8x8R Figure t Load Rejection Without Bypass 484 124 0.21 0.18 2a Loss of Feedwater Heater 123 122' O.14 0.14 3 Feedwater Controller Failure 268 119 0.13 0.11 4a Exposure Range: EOC-1000 mwd /ST to EOC Flux Q/A Transient (% NBR) (% NBR) P8x8R 8x8R Figure Load Rejection Without Bypass 477 125 0.23 0.20 2b

, Loss of Feedwater Heater 123 122 0.14 0.14 3 Feedwater Controller Failure 276 121 0.15 0.13 4b i

11. IACAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

. TRANSIENT

SUMMARY

(S.2.2.1)

Limiting Rod Pattein:' Figure 5 Includes 2.2% Power Spiking Penalty: Yes

,_ Rod. Block Rod Position 0

? Reading , (feet withdrawn) P8x8R 8x8R P8x8R 8x8R .a b 3.5' 104 O.11 0.11 17.79 -17.79 105~ 4.0 0.13 0.13 18.04 18. 04 ,

106 4.5 0.14 0.14 18.06 18.06 107 5.0 0.15 0.15 18.06 18.06' 108 5.0 0.15 0.15 18.06- 18.06 109 5.5 0.16 0.16 18.06 18.06

~110- 7.0 0.19 0.19 18.06 18.06 -

Setpoint Selected: 108 .

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23A1813 Rsv. O i

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12. CYCLE MCPR VALUES (S.2.2)

Nonpressurization Eveats Exposure Range: BOC to EOC I P8x8R 8x8R Loss of Feedwater Heater 1.21 1.21 Fuel Loading Error 1.22 Rod Withdrawal Error 1.22 1.22 Pressurization Events Option A Option B P8x8R 8x8R P8x8R 8x8R Expoeure Range: EOC-1000 mwd /ST Load Rejection Without Bypass 1.34 1.31 1.13 1.11 Feedwater Controller Failure 1.25 1.23 1.19 1.17 Exposure Range: E0C-1000 mwd /ST to EOC Load Rejection Without Bypass 1.36 1.33 1.24 1.21 Feedvater Controller Failure 1.27 1.25 1.21 1.19

-13. OVERPRESSURI4ATION ANALYSIS

SUMMARY

(S.2.3)

P,1 Py Transient (psig) (psig) Plant Response MSIV Closure 1220 1252- Figure 6 (Flux Scram) r h

9

[ 23A1813 Rsv. O ji 14. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed: Extrapolated Rod Block Line Decay Ratio: Figure 7 Reactor Core Stability Decay Ratio, x2 /*0: 0.86 Channel Hydrodynamic Performance Decay Ratio, x2 /*0*

Channel Type {

8x8R/P8x8R 0.43 ,

15. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial CPR Resulting CPR Misoriented 1.20 1.07

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

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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-1 Parameter (s) not Bounded, Cold: Scram Reactivity.

Accident Reactivity Resultant Peak Enthalpy, Cold: 246.0 Parameter (s) not Bounded, HSB: Accident Reactivity Resultant Peak Enthalpy. HSB: 215.0 h

17. IDSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2)_ -

I See " Loss-of-Coolent Accident Analysis Report for Cooper Nuclear  !

Power Station,", NEDO-24045 August 1977 (as amended).-

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1.1 I I l 1 1 I 1 3 5 7 S111315171921232527293133353733414345474951 FUEL TYPE _

A = 80RB283 6 = P8DRB283

B = P80RB265L H = P8DRB265L C = P80RB283 I = P8DRB283

! D =-P80RB265L J = P80RB265L (Cycle 10)

E= K = P80RB283 (Cycle '10) i Figure-1. Reference Core Loading Pattern 11

23A1813 Rev. 0 1 NEUTRON FLU ( .

1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE K AT FLUX 2 SAFETY VALVE FLOV 3 CORE IM.ET UW 3 RELIEF VALVE FLOW 150 8 30s.8 e ew=.*SS uAtur rLee

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0. 8 2.0 TIME (SEcolell TIME I E CONDS3 1 LEVEL (INCH-EF-SEP-SKRT) 1 V010 RF CT!VITY 2 VESSEL STEAWLOW 2 00PPLErt EACTIVITY
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Figure 2a. Plant Response to Generator Load Rejection Without Bypass (EOC-1000 mwd /ST) 12

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23A1813 Rev. 0 1 NEUTRON FLU ( 1 VESSEL PRESS R!SE(PSI) 2 AVE SURFACE MAT FLUX 2 SAFETY VALVE FLOW 3 CORE IM.ET LOW 3 RELIEF VALVE FLOJ '

158.0 . 300.8 ' eyn

  • SS u*Lur etey

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50.0 180.0 l " ^ ^

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8. 0 . . 0.0 _ _ _ ,, ,, , _ , ,,

, 0. 0 2.5 4.0 8.0 0. 8 2.0 4. 0 0.8 TIME (SEC0005) TIME (SEcoleS) 1 LEVEL (INCH-REF-SEP-SKRT) I votD R T!VITY 2 YESSEL STEAFLOW 2 DOPPLER EACTIVITY 3 TUR9INE STEUFLOW 3 SCR AM R CTIVITY 200.0 '

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s.e 2.s 4.s s.s TIME (SEcopes) TIME (SECONDS)

Figure 2b. Plant Response to Generator Load Rejection Without Bypass (EOC)

! 13

23A1813 Rev. 0 1 peEurRON FLUX 1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE lEAT FLUX 2 RELlEF VALVE FLOV 3 CORE IfLET FLOW 3 BYPLSS VALVE FLOW 150.0 ' raae ' =_" ??

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l l 1 LEV EL(INCH-REF.SEP-SKRT) 1 VOI 3 REACTIVITY 2 VES iEL STEAFLOW 2 00P'LER REACTIVITY 3 TUR II E STEAM LOW 3 SCRLM REACTIVITY l

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e. o tes.s 2es.o e. e toe.e 2oo.o TIME (SEcolell TIME (SEC080S1 r

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Plant Response to Loss of 100'F Feedwater Heating Figure 3.

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. . - - - . . _ _ . , - . , _w

23A1813 Rsv. 0 150.0 1 NEUTRON FLU .d iVESSEL PRESS RISE : PSI) 2 AVE SURFACE EJTl' LUX  ? SAFETY VALVE FLOW 3 CORE lit.ET LOh 3 RELIEF VALVE FLOW 150.0 ame

. , ,=," '? sBYPASS VALVE FLOW 100.0 0 100.0 :  ;  ;  ;  ?

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TIME (SECOW S1 TIME (SECOW S3 i

! LEVEL (INCH-REF-SEl'-SKRT) I VOID REACTI'h k / '

2 VESSEL STEAMFLOW 2 DOPPLER REAi TIMJ 3 TURBIE STEAMFLOW 3 SCRAM REACT :VITY 150. e

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0. 9 20.e ..0 40 8. 8 20.0 40.0 TIME (SECONOS) TIME (SECOND$1 Figure 4a. Plant Response to Feedwater Controller Failure (E0C-1000 mwd /ST) i 15
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23A1813 Rev. 0 150.0 I VESSEL PRESS RISE ; PSI) 1AVE 2 NEUTRON SURFACE FLU).f FLUX 2 SAFETY VALVE Flow 3 CORE INLET I .0W 3 RELIEF VALVE FLOW 158 0 ';=r, tu rv ' -

? 4 BYPASS VALVE FLOW it... S g ' *f '. - "_

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0. 0 20.0 40.0 8. 0 20.0 46.0 TIME (SEC088)S) TIME (SECOW S) l 1 LEVELLINCH-REF-SEP-SMRT) I YOID REACTIV T" 1 2 VESSEL STEAMFLOW 2 DOPPLER REAC 1 rITT 3 TURBINE STEAMFLOW I sse.e esr= >vE= et- s.o 3 S,CR,

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Figure 4b. Plant Response to Feedwater Controller Failure (EOC) 16 i

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51 38 47 12 2 6 43 38 40 39 2 8 12 35 38 40 l

l 31 10 6 12 0 27 38 40 NOTES: 1. Rod pattern 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 (26, 31) .

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Figure 5. Limiting Roc' Withdrawal Error Rod Pattern 17 i -

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23A1813 Rev. O I 1 NEUTRON F.UX 1 VESSEL PRESS RISECPSI) 2 AVE SURFA2 MAT FLUX 2 SAFETY VALVE FLOV d 3 CORE IM.ET FLOW 3 RELIEF VALVE FLOW 150.0 350.0 ' ero nes v.*LuE eteu

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200.0 h-n 5  ;

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0. 0 5.0 0.0 5. 0 TIME (SECONDS) TIME (SECOMS) 1 LEVEL (INC4-REF-SEP-SMRT) 010 REkflVITY 2 VESSEL STEAMFLOW DOPPLER REACTIVITY TEANFLOW 200.0- 31 re_

TURBINE,c_S 1.0 3. SCR.A.M oRE AE T, I, V vI, ,TY en_ u m . o. e, no.

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0. 0 5.0 0.0 5.0 l TIME (SECONDS) TIME (SECONDS) l l

Figure 6. Plant Response to MSIV Closure (Flux Scram) l 18

23A1813 Rzv. 0

.Ab ATURAL C'. RCULATIO i B1 05 PERCENT ROD LI 4E CL LTIMATE STABILITY LINE 1.00 C  ::

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s

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0. 0 20.0 40.0 80.0 80.0 100.0 120.0 1 PERCENT POWER Figure 7. Reactor Core Decay Ratio 19

23A1813 Rev. 0

0. 0

-5.0 m -

g -10. 0 , MM

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=

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-25.O CALCUL ATED VALUE-COLD CALCUL ATED VALUE-HSB C BOUND VAL 280 CAL /G COLD D BOUND VAL 280 CAL /G HSB

-30.0

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

Figure 8. Fuel Doppler Coefficient in 1/A*C [

20

1 I 23A1813 Rtv. 0 20.O A ACCIDENT FUNC TION B BOUNDING VALU E 280 CAL /G

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, 17.5 l

4 15.0 h

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W 12.5 1 x I

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( 0. 0 5. 0 10.0 15.0 20.O ROD POSITION, FEET OUT Figure 9. Accident Reactivity Shape Function (Cold Startup) 21

23A1813 Rev. 0 20.O A ACCIDENT FUNC TION 8 BOUNDING VALUE 280 CAL /G 17.5 15.0 T 12.5 M

U M

W gj 10.0 O

C 7.5 o

s 5. 0 2.5

0. 0
0. 0 5. 0 10.0 15.0 20.0 ROD POSITION, FEET OUT  !

Figure 10. Accident Reactivity Shape Function (Hot Startup) 22

23A1813 R:v. 0 30.O A SCRAM FUNCTION B BOUNDI NG VALUE 280 CAL /G

.25.0 a

20.0 m M -

W I

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m 15.0 Ez v

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t 10.0 8]

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0. 0 ,

u f 0. 0 1.0 2. 0 3. 0 4.0 5.0 6.0

. ELAPSED TIME, SECONDS Figure 11. Scram Reactivity Function (Cold Startup) 23

23A1813 Rev. 0 3 4

50.O A SCRAM FUNCTION B BOUNDI NG VALUE 280 CAL /G 1 40.0 m

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ELAPSED TIME, SECONDS Figure 12. Scram Reactivity Function (llot Startup) 24

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