ML20212Q230

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Rev 0 to Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station Unit 1,Reload 8 (Cycle 9)
ML20212Q230
Person / Time
Site: Quad Cities Constellation icon.png
Issue date: 08/31/1985
From: Charnley J, Plotycia G, Casey Smith
GENERAL ELECTRIC CO.
To:
Shared Package
ML20212Q208 List:
References
23A4691, 23A4691-R, 23A4691-R00, NUDOCS 8609040107
Download: ML20212Q230 (25)
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AUGUST 1985

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b SUPPLEMENTAL RELOAD L LICENSING SUBMITTAL FOR e- QUAD CITIES NUCLEAR POWER STATION T UNIT 1, RELOAD 8 (CYCLE 9)

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23A4691 Revision 0 Class I August 1985 1

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR QUAD CITIES NUCLEAR POWER STATION UNIT 1, RELOAD 8 (CYCLE 9)

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C. W. Smitii /

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A M harnley, Ma g el Licensing NUCLEAR ENERGY BUSINESS OPERATIONS . GENERAL ELECTRIC COMPANY SAN JOSE, CAltFORNIA 95125 GENER AL $ ELECTRIC 1/2

23A4691 R3v. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Commonwealth Edison Co. (Edison) for Edison's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending Edison's operating license of the Quad Cities Nuclear Power Station 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 information in this document are contained in the contract between Commonwealth Edison Co. and Iowa-Illinois Gas and Electric Co. and General Electric Company for fuel bundle fabrication .and services for Quad Cities Nuclear Power Station Units 1 and 2, dated December 1,1978, as amended 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 l

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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. 23A4691 Rev. 0

1. PLANT UNIQUE ITEMS (1.0)*

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Plant Parameters: See Appendix A R (Item 4): Value Shown Includes Effect of B 4C' Settling (0.0004 Ak)

Stability Analysis (Item 14): See Appendix B Control Rod Drop Analysis (Item 16): See Appendix C Fuel Mechanical Design Methods: See Appendix D

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

Fuel Type Cycle Loaded Number Number Drilled Irradiated BP8DRB265H 8 116 116 BP8DRB283H 8 80 80 P8DRB282 6 88 88 P8DRB265L 7 64 64 P8DRB265H 7 160 160 New BP8DRB299 9 144 144 BP8DRB282 9 72 72 TOTAL 724 724

  • ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel", NEDE-24011-P-A-6, April 1983; a letter "S" preced-ing the number refers to the appropriate section in the United States sup-piement, NEDE-24011-P-A-6-US, April 1983.

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23A4691

  • Rev. O l
3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 18,889 mwd /st Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 18,593 mwd /st Assumed reload cycle core average exposure at end of cycle: 18,683 mwd /st Core loading pattern: Figure 1

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 effective Uncontrolled 1.107 Fully Controlled 0.952 Strongest Control Rod Out 0.978 R, Maximum Increase in Cold Core Reactivity 0.009 with Exposure into Cycle, Ak

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (ak) yp3 (20*C, Xenon Free) 600 0.044 6

23A4691 Rev. 0

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

(REDY EVENT ONLY)

Void Fraction (%) 34.8 Average Fuel Temperature ('F) 1147 Void Coefficient N/A* (d/% Rg) -4.99/-6.24 Doppler ' Coefficient N/A (d/*F) -0.186/-0.177 Scram Worth N/A ($) **

7. RELOAD 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 Exposure: BOC9 to E0C9 BP/P8x8R 1.20 1.76 1.40 1.051 5.953 107.6 1.34

8. SELECTED MARGIN 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 (ODYN, Option B)

Number of Exposure Points: 1

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

7

1 23A4691 Rev. 0

9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3)

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

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

ACPR Flux Q/A Transient (% NBR) (% NBR) BP/P8x8R Figure Exposure: B0C9 to EOC9 576 118 0.27 2 Load Rejection Without Bypass Exposure: BOC9 to EOC9 120 120 0.17 3 Loss of 145'F Feedwater Heating Exposure: BOC9 to EOC9 309 117 0.18 4 Feedwater Controller Failure

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

SUMMARY

(S.2.2.1)

(Generic Bounding Analysis Results)

Rod Block Reading ACPR

% (All Fuel Types) 104 0.13 105 0.16 106 0.19 107 0.22 108 0.28 109 0.32 110 0.36 Setpoint Selected: 107%

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23A4691 Rev. 0

12. CYCLE MCPR VALUES (S.2.2)

Non-Pressurization Events Exposure Range: BOC9 to EOC9 BP/P8x8R Loss of Feedwater Heater 1.24 Fuel Loading Error 1.24 Rod Withdrawal Error 1.29 Pressurization Events Exposure Range: BOC9 to EOC9 Option A Option B BP/P8x8R BP/P8x8R Load Rejection w/o Bypass 1.40 1.35 Feedwater Controller Failure 1.31 1.23

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.2.3)

P'3 1 Py Transient (psig) (psig) Plant Response MSIV Closure 1313 1330 Figure 5 (Flux Scram)

14. STABILITY ANALYSIS RESULTS (S.2.4)

See Appendix B

15. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial MCPR Resulting MCPR Misoriented 1.22 1.07 9

23A4691 , Rev. 0

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

See Appendix C

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

" Loss-of-Coolant Accident Analysis Report for Dresden Units 2, 3, and Quad Cities Unit 1, 2 Nuclear Power Stations," NED0-24146A, April 1979 (as amended).

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= BP80 82 G= P80RB265H D= BP8DRB282 Figure 1. Reference Core Loading Pattern 11

23A4691

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Figure 2. Plant Response' to Generator Load Rejection, Without Bypass 12

23A4691 Rev. 0 150.0 1 NEU RON FLUX X VESSEL PRESS RISE (PSI) 2 8.vE StlPrtCE,9E e' FUW 2 RCL!Er VALVE FLOW 3NH i_t! tLGd 't 3 BYP'55 VALVE FLOW ~

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Figure 3. Plant Response to Loss of 145'F Feedwater Heating 13

23A4691 Rev. 0 150.0 1 NEUIRON ligUX 1 VES 2 AYE '"rr Lt ? ?.T ri Ux 2 SarhEL-TV YALVPRESS FLOW RISE (PSI) 3 COR 3 RELhCF V AL FLOW 150.0 e. ccc E"t' INLl*l.T

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Figure 4. Plant Response to Feedwater Centroller Failure 14

23A4691 Rev. 0 1 NEUTRON FLUX 1 VESEEL PRESS RISE (PSI) 2 7.YE Cler?.CE HEAT FLUX 2 REL IEF Vt.LVE FLOY 150.0 I 3 CORE IrtET FLOW 300.0 3 !*FITY VALVE Fl ?

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Figure 5. Plant Response to MSIV Closure, Flux Scram 15/16

23A4691 Rev. O APPENDIX A PLANT PARAMETER CHANGES Pressure Relief Systems (Table S2-4.1, pg. US2-93, NED0-24011)

Lowest Safety / Relief Valve setpoint (psig)* 1135 + 1%

Lowest Relief Valve setpoint (psig) 1115 Transient Operating Parameters (Table S.2-6, pg. US2-99, NED0-24011)

, Thermal Power, MWt 2511

Turbine Pressure (psig) 950 GETAB Initial Conditions (Table S.2-8, Pg. US2-102)

Reactor Core Pressure (psia) 1035 Inlet Enthalpy (Btu /lb) 523.7

  • This valve (i.e., the Target Rock Safety / Relief Valve) was considered to be out-of-service for the Core Wide Transient and Overpressurization analyses (Items 10 and 13) and, therefore, was not included in safety or relief valve flows.

17/18

23A4691 Rev. O APPENDIX B STABILITY ANALYSIS According to Reference B-1, Quad Cities Unit 1 is exempt from the current requirement to submit a cycle specific stability analysis to the NRC.

REFERENCES B-1. Letter, C.O. Thomas (NRC) to H. C. Pfefferien (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amend-ment 8, ' Thermal Hydraulic Stability Amendment to GESTAR II,

April 24, 1985.

19/20

23A4691 Rev. O APPENDIX C CONTROL ROD DROP ANALYSIS The cycle-specific control rod drop accident analysis has been discontinued for banked position withdrawal sequence (BPWS) plants based on the fact that in all cases the peak fuel enthalpy from a control rod drop accident would be much less than the 280 cal /gm limit. This change in procedures was reported and justified in Reference C-1. Reference C-2 indicates this change is acceptable to the NRC.

RETERENCES C-1. Letter, J. S. Charnley (GE) to C. O. Thomas (NRC), " Proposed Administrative Amendment to GE Licensing Topical Report NEDE-24011-P-A", January 25, 1984.

C-2. Letter, C. O. Thomas (NRC) to J. S. Charnley (GE), " Acceptance for Referencing of Licensing Topical Report Amendment 9 to NEDE-24031, Revision 6, 'GESTAR-II General Electric Standard Application for Reactor Fuel'", January 25, 1985.

21/22

23A4691 Rev. O APPENDIX D FUEL MECHANICAL DESIGN METHODS General Electric's approved fuel thermal mechanical design model, TEXICO (documented in Revision 6 to NEDC-24011-P-A), was used in the analysis of Quad Cities 1, Reload 8.

23/24 (FINAL)

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