ML20206R835

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Rev 0 to Supplemental Reload Licensing Submittal for Cooper Nuclear Power Station Unit 1,Reload 10
ML20206R835
Person / Time
Site: Cooper Entergy icon.png
Issue date: 05/31/1986
From: Charnley J, Casey Smith, Zarbis W
GENERAL ELECTRIC CO.
To:
Shared Package
ML20206R794 List:
References
23A4781, 23A4781-R, 23A4781-R00, NUDOCS 8609190216
Download: ML20206R835 (25)
v • d • e


Text

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.p CLASSI 2g MAY 1986 t

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Fa i SUPPLEMENTAL RELOAD

@ LICENSING SUBMITTAL FOR g COOPER NUCLEAR POWER STATION s UNIT 1, RELOAD 10 m

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i GENER AL h ELECTRIC

23A4781 Revision 0

( Class I

) May 1986 h

.O SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR COOPER NUCLEAR POWER STATION, UNIT 1 RELOAD 10 Prepared!

W.A.[arpis Fuel L%(nsing s =

Verified: U4 L/ . 3-C. W. SmitK P Fuel Licensing Approved .-

. CTUir'fiiley, Manage [

F L Licensing i

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l NUCLEAR ENERGY BUSINESS OPERATIONS

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

4-23A4781 R2v. O f

IMPORTANT NOTICE RECARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Nebraska Public f Power District (NPPD) for NPPD's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending NPPD's operating license of the Cooper Nuclear Station. 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 l 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 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 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 i .-. .

23A4781 R;v. O I

ACKNOWLEDCEMENT The engineering and reload licensing antlyses, which form the technical basis of this Supplemental Reload Licensing Submittal, were performed by C. C. Storey of the Nuclear Fuel Engineering Department.

l 5/6

23A4781 R;v. 0

1. PLANT UNIQUE ITEM (1.0)*

Appendix A: .CETAB Transient Analysis Initial Condition Parameter

2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0) l Fuel Type Cycle Loaded Number Number Drilled Irradiated P8DRB265L 7 20 20 P8DRB283 7 32 32 P8DRB265L 8 56 56 P8DRB283 8 56 56 P8DRD265L 9 56 56 P8DRB283 9 60 60 P8DRB265L 10 88 88 P8DRB283 10 28 28

)

New BP8DRB283 11 152 152

, Total 548 548 i

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 18032 mwd /ST

, Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 17632 mwd /ST Assumed reload cycle core average exposure at end of cycle: 17277 mwd /ST Core loading pattern: Figure 1 h

  • ( ) Refers to area of discussion in "Ceneral Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-7. dated August 1985. A letter "S" pre-ceding the number refers to the appropriate country-specific supplement.

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23A4781 R;v. 0

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL ~ SYSTEM WORTN - 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 0.955 4 Strongest Control Rod Out 0.985 R, Maximum Increase in Cold Core Reactivity 0.000 with Exposure into Cycle, Ak

5. STAND 8Y LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (Ak) g (20'C. Xenon Free) 600 0.036

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

(COLD WATER INJECTION EVENTS ONLY)

Vold Fraction (%) 40.0 Average Fuel Temperature (*F) 1285 Void Coefficient N/A* (4/% Rg) -6.46/-8.07 Doppler Coefficient N/A (4/*F) -0.199/-0.189 Scram Worth N/A ($)

1

  • N = Nuclear Input Data, A = Used in Transient Analysis
    • Ceneric exposure independent values are used as given in "Ceneral Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-7, dated August 1985.

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23A4781 R3v. 0

7. RELOAD UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)

Fuel Peaking Factors Bundle Power Bundle Flow Initial Desian Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR Exposure: SOC to EOC-1000 mwd /ST-BP8X8R/

P8x8R 1.20 1.54 1.40 1.051 6.528 109.3 1.25 h

Exposure: EOC-1000 mwd /ST to EOC BP8x8R/

P8x8R 1.20 1.51 1.40 1.051 6.406 110.0 1.27

8. SELECTED MARCIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: No Improved Scram Time: Yes (Option B) l Exposure Dependent Limits: Yes Exposure Points Analyzed: EOC, EOC-1000 mwd /ST

9. OPERATINC FLEXIBILITY OPTIONS (S.2.2.3)

Single-Loop Operation: Yes Load Line Limit: Yes Extended Load Line Limit: No Increased Core Flow: No Flow Point Analyzed: N/A Feedwater Temperature Reduction: No

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23A4781 R;v. 0 I

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

Exposure Ranget BOC to EOC-1000 mwd /ST t

ACPR Flux Q/A BP8x8R/

Transient (% NBR) (% NBR) P8x8R Figure Load Rejection Without Bypass 424 122 0.18 2a l Loss of Feedwater Heater 122 122 0.14 3 Feedwater Controller Failure 238 120 0.14 4a Exposure Ranget EOC-1000 mwd /ST to EOC I i ACPR Flux Q/A BP8x8R/

Transient (% NBR) (% NBR) P8x8R Figure Load Rejection Without Bypass 429 123 0.20 2b

, i Loss of Feedwater Heater 122 122 0.14 3 Feedwater Controller Failure 253 123 0.17 4b

11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) '

TRANSIENT

SUMMARY

(S.2.2.1) i Limiting Rod Patternt Figure 5 ACPR MLHCR (kW/Ft)

Rod Block Rod Position Reading (feet withdrawn) BP8x8R/P8x8R BP8x8R/P8x8R 104 4.0 0.12 17.69 105 4.5 0.13 18.01 106 5.0 0.15 18.07 107 5.5 0.16 18.07 108 6.5 0.19 18.07 109 9.5 0.23 18.07 110 10.0 0.24 18.07 Setpoint Selected: 106 1

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l 23A4781 R;v. 0

12. CYCLE MCPR VALUES (S.2.2)

Nonpressurization Events Exposure Ranget BOC to EOC BP8x8R/

P8x8R I

Loss of Feedwater Heater 1.21

, Fuel Loading Error 1.22 Rod Withdrawal Error 1.22 I

Pressurization Events I

Option A Option B BP8X8R/ BP8x8R/

t P8x8R P8x8R l

Exposure Ranget BOC to EOC-1000 mwd /ST I

l Load Rejection Without Bypass 1.30 1.11 Feedwater Controller Failure 1.27 1.20 I Exposure Ranges EOC-1000 mwd /ST to EOC i

Load Rejection Without Bypass 1.33 1.22 Feedwater Controller Failure 1.30 1.23 f 13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) f l

Psi Py

} Transient (psig) (psia) Plant Response HSIV Closure 1229 1250 Figure 6 (Flux Scram)

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23A4781 R^v. 0 4

14. STABILITY ANALYSIS RESULTS (S.2.4)

Not analyzed, in compliance with C. O. Thoma 2 (NRC) to H. C.,Pfefferten (CE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8, ' Thermal Hydraulic Stability Amendment to CESTAR II',"

April 24,~1985.

15. LOADINC ERROR RESULTS (S.2.5.4)

Variable Water Cap Misoriented Bundle Analysist Yes Event Initial CPR Resulting CPR Misoriented 1.20 1.07

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

Bounding Analysis Results:

Doppler Reactivity Coefficients Figure 7 Accident Reactivity Shape Functions: Figures 8 and 9 Scram Reactivity Functions: Figures 10 and 11 Plant Specific Analysis Results:

Parameter (s) not Bounded, Coldt Scram Reactivity, Accident Reactivity Resultant Peak Enthalpy, Coldt 158.7 cal /gm Parameter (s) not Bounded, HSBt Accident Reactivity Resultant Peak Enthalpy, HSB 244.0 cal /gm.

17. LOS3-OF-COOLANT ACCIDENT RESULT (S.2.5.2)

See " Loss-of-Coolant Accident Analysis Report for Cooper Nuclear Power Stttion," NEDO-24045, August 1977 (as amended).

12 1

23A4781 Rev. O l

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, B = P DRB283 C = P8DRB2 L D = P8DRB283 I = BP DRB 3 E = P8DRB265L Figure 1. Reference Core Loading Pattern 13

23A4781 R;v. O I NEUTRON FLU < l VESSEL PRES 3 RISE (PSI) 2 AVE SURFAtt HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE IM.ET : LOW 3 REllEF VALVE FLOW 150.0 300.0 ' ovoas? vaLuE Flee n

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I LEVEL (INCH-REF.SEP.SKRT) 2 VESSEL STEA1 FLOW V 1 VOID REACTIVtTY 2 DUPPLER REACTIVITY 3 IURBINE STE WLOW 3 SCRAM REACTIVITY 200.0 ' FEEgya rro r _nu 1.0 rnrat eracrrurry

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23A4781 R;v. O I NEUTRON FLU ( I VESSEL PRESS RISE (PSil 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 EDRE IM.El LOW 3 RELIEF valve FLOW 15.. . 3.... ' ava

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TIME (SECOND$1 TIME (SEC0fe93 Figure 2b. Plant Response to Generator Load Rejection Without Bypass (EOC) 15 i . , , ,

23A4781 R;v. O

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0.0 100.0 200.0 0. 0 100.0 200.0 TIME (SECOWS) TIME (SECONOS) 1 LEV IL(INCH.REF-SEP-SKRT) 1 VOI ) REACTIVITY 2 VESiEL STEAMFLOW 2 DOP'LER REACTIVITY 3 TURIINE STEAffLOW 3 SCRLM REACTIVITY 150.0 ' rEE M* ?EP [L'= 1.0 ' Tr" =.L = ^CT!w!TV g; , 2 O " O . I - 24 ^ I O 0 0 - '- ' ->

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Figure 3. Plant Response to Loss-of-Feedwater Heater 16

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23A4781 R;v. O 150.0 tNEUTRON FLU ( l VESSEL PRESS RISE (PSI) 2 AVE SURFACE HE I FLUX 2 SAFETY VALVE FLOW 3 CORE litET TLO' 3 RELIEF VALVE FLOW 150.0 e ce=E tu_E' "M 4 BYPASS VALVE FLOW k

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ILEVEL(INCH-4EF-SEP-SKRil i V010 REACTIVITI 2 VESSEL STEA1 FLOW 2 DOPPLER REACTIFI 'Y 3 TURBINE STE UFLOW 3, SCRAM RE/.CTIV! Yy 150.0 a FEEguATEo e_m 1. 0 797p_t eg3cvgyg

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Figure 4a. Plant Response to Feedwater Controller Failure (EOC-1000 mwd /ST) i 17

23A4781 R:v. 0 150.0 aNEUTRON FLU ( 1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE HE/ FLUX 2 SAFETY VALVE FLOW 3 CORE INLET TLD) 3 REllEF VALVE FLOW 130.0 ' Ca=E !"_E' E L , 4 BYPASS VALVE FLOW g i40.0 g , M. _

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I Figure 4b. Plant Response to Feedwater Controller Failure (EOC) 18

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3 10 NOTES: 1. NUMBER INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN ROD.

2. ERROR R0D IS (22,31) .

('I Figure 5. Limiting Rod Withdrawal Error Rod Pattern

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i NEUTRON F_UX 1 VESSEL PRESS RISE (PSI) 2 AVE SURFAI HEAT FLUX 2 SAFETY VA.VE FLOW 3 CORE IM.E T FLOW 3 RELIEF VA VE FLOW 15020 k 300.0

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ILEVEL(INC4-REF-SEP.SMRT ) i VOID REAtrlVITY 2 VESSEL SiEAMFLOW 2 DOPPLER REACilVITY 3 TURBINE STEAMFLOW 3, SCRAM :TIVIT v vm i, aciREAnervi,Y a crenwarro ei nu I.0 ft 200. t I

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A CALCUL ATED VALUE-COLD B CALCUL6TED VALUE- HSB C BOUNO VAL 280 CAL 1G COLO O BOUND VAL 280 CAL 1G HSB

-30.0 l 0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE DEG C.

Figure 7. Fuel Doppler Coefficient in 1/A*c t

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I Figure 9. Accident Reactivity Shape Function (Cold) 23 i . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ - . _ _ _ ___

23A4781 Rev. 0 60.O A SCRAM F JNCTION B BOUNDIN 3 VALUE 280 CAL /G 50.0 A O

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23A4781 Rsv. 0

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40.0 A SCRAM F JNCTION 8 BOUNDIN 3 VALUE 280 CAL /G 35.0 g 30.0 A i h 25.0 N

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23A4781 Rsv. O I'

APPENDIX A

-GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETER To more realistically represent actual plant data, a value of 520.4 Btu /lb was used as the inlet enthalpy instead of the value given in

" General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-7, dated August 1985.

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