ML20024C248

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Rev 0 to Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station,Unit 2,Reload 6 (Cycle 7).
ML20024C248
Person / Time
Site: Quad Cities Constellation icon.png
Issue date: 05/31/1983
From: Charnley J, Elliott P, Hill R
GENERAL ELECTRIC CO.
To:
Shared Package
ML20024C237 List:
References
22A8560, 22A8560-R, 22A8560-R00, NUDOCS 8307120446
Download: ML20024C248 (18)
v • d • e


Text

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22A8560

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j SUPPLEMENTAL RELOAD LICENSING l l l SUBMITTAL FOR QUAD C! TIES I NUCLEAR POWER STATION, UNIT E j l ll i

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  • 22A8560
Rev. O Class I May 1983 i

i SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR QUAD CITIES NUCLEAR POWER STATION UNIT 2, RELOAD 6 (CYCLE 7)

Prepared: - O P. E. Elliott Verified -

1. T. Hill' Approved: M E-ll-f3 J. S harnley Fuel censing Manager NUCLEAR POWER SYSTEMS DMSION + GENERAL ELECTRIC COMPANY SAN JOSE CALIFORNIA 95125
GENER AL h ELECTRIC 1

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22A8560 Rev. 0 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 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 information in this document are contained in the contract between Commonwealth Edison Co.

and Iowa-Illinois Gas & 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 docu-ment 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 unauthor-ized use, neither General Electric 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 liability or damage of any kind which may result from such use of such information.

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1. PIANT UNIQUE ITEMS (1.0)*

A. Plant Parameter Changes See Appendix A B. R (Item 4) Value Shown Includes Effect of B4 C Settling (0.0005 Ak)

C. Item 12 Includes ATWS RPT (1250 psig set point)

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

Fuel Type Number Number Drilled Irradiated:

Reload 2 8DB250 16 0 Reload 2 8D3262 20 0 Reload 3 8DB262 76 0 Reload 4 P8DRB26SL 180 180 Reload 5 P8DRB265L 80 80 P8DGB263L** 16 16 (Ramp) P8DGB263L**(Ramped 16 16 P8DGB298** 0C6) 32 32 P8DGB284** 68 68 P8DGB265L** 12 12 New:

Reload 6 BP8DRB265H 208 208 Total 724 612

  • ( ) refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A- (latest approved revision).
    • Barrier Fuel 1

-- - _ _ - - ~ - . . - - . . - , _ -.._-_ .

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i 22A8560 Rev. 0

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 19,274 mwd /t Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 19,074 mwd /t Assumed reload cycle core average exposure at end of cycle: 18,173 mwd /t 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)

Minimum Shutdown Margin, BOC, k effective Uncontrolled 1.102 Fully Controlled 0.946 Strongest Control Rod Out 0.982 R, Maximum Increase in Cold Core Reactivity 0.008 with Exposure into Cycle, Ak

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (ak)

EEE (20*C, Xenon Free) 600 0.045

6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2) j (REDY EVENTS ONLY)

Void Fraction 34.3 Average Fuel Temperature (*F) 1155. -

! Void Coefficient N/A* (C/% Rg) -5.12/-6.40 Doppler Coefficient N/A (C/*F) -0.191/-0.181 i Scram Worth N/A ($)

  • N = Nuclear Input Data A = Used in Transient Analysis
    • Generic e.*posure independent values are used as given in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A, (latest approved revision).

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22A8560 Rev. 0

7. RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)
Bundle Fuel Peaking Factors ~ Power Bundle Flow Initial Design Local Radial Axial R-Factor JMWt) (1000 lb/hr) MCPR Exposure
BOC7 to EOC7 P8x8R 1.20 1.79 1.40 1.051 6.072 105.2 1.31 8x8R 1.20 1.81 1.40 1.051 6.140 103.6 1.29 8x8 1.22 1.68 1.40 1.098 5.684 106.3 1.28
8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: No MeasurAd Scram Time: No Number of Exposure Points: 1

9. ODERATING 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 l

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22A8560 Rev. 0 l

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10.

CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1)

Flux 0 Q/A Transient (% NBR) (% NBR) P8x8R 8x8R 8x8 Figure Exposure: BOC7 to EOC7 451 116 0.24 0.22 0.21 2 Load Rejection Without Bypass Exposure: BOC7 to EOC7 121 120 0.17 0.17 0.17 3 Loss of 145'F Feedwater Heating Exposure: BOC7 to EOC7 252 114 0.15 0.14 0.14 4 Feedwater Controller Failure 11.

LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(S.2.2.1)

Limiting Rod Pattern: Figure 5 Rod Block Rod Position Reading (feet withdrawn) P8x8R/8x8R 8x8R/P8x8R 104 4.0 0.13 15.06 105 4.5 0.15' 15.50 106 5.0 0.18 15.75 107 5.5 0.20 15.92 108 6.5 0.23 16.00 109 9.0 0.28 16.00 110 9.5 0.29 16.00 Setpoint Selected: 107 4

22A8560 Rev. 0

12. CYCLE MCPR VALUES (S.2.2)

Non-Pressurization Events  !

Exposure Range: BOC to EOC P8x8R 8x8R 8x8 Loss of Feedwater Heater 1.24 1.24 1.24 Fuel Loading Error 1.21 Rod Withdrawal Error 1.27 1.27 Pressurization Events Exposure Range: BOC7 to EOC7 Option A Option B P8x8R 8x8R 8x8 P8x8R 8x8R 8x8 Load Rejection w/o Bypass 1.37 1.35 1.34 1.32 1.30 1.29 Feedwater Controller Failure 1.27 1.26 1.26 1.20 1.19 1.19

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.2.3)

PSL PV Transient (psig) (psig) Plant Response MSIV Closure 1298 1315 Figure 6 (Flux Scram)

14. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed: Extrapolated Rod Block Line -

Natural Circulation Decay Ratio: Figure 7

. Reactor Core Stability Decay Ratio, X /X 0.52 2 O Channel Hydrodynamic Performance Decay Ratio, X /X 2 O Channel Type 8x8R/P8x8R 0.18 8x8 0.28 -

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22A8560 Rev. 0

15. LGADING ERROR RESULTS (S.2.5.4) i Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial MCPR Resulting MCPR Misoriented 1.20 1.08
16. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)

L ximum Incremental Control Rod Worth: 0.36% Ak

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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APPENDIX A PLANT PARAMETER CHANGES Pressure Relief Systems (Table S2-4.1, pg. US2-93, NEDO-24011)

Safety / Relief Valve retpoint (psig) 1135 + 1%

Safety / Relief Valve capacity (% rated steam flow) 29 Safety Valve capacity (% rated steam flow) 59 Relief Valve setpoint (psig) 1115 Transient Operating Parameters (Table S.2-6, pg. US2-99, NEDO-24011)

Thermal Power, MWt 2511 Turbine Pressure (psig) 950 CETAB Initial Conditions (Table S.2-6, Pg. US2-102)

Reactor Core Pressure (psia) 1035 Inlet Enthalpy (Btu /lb) 523.7 NIWS - RPT initiated at 1250 psig 7

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B = 8DB262 (Reload 2) G = P8DGB263L (Reload 5)

C = 8DB250 (Peload 2) H = P8DGB298 (Reload 5)

D = 8DB262 (Reload 3) I = P8DGB784 (Reload 5) ,*

! E = P8DRB265L (Reload 4) J = BP8DRB265H (New)

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22A8560 Rev. 0 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, 35).

2 6 10 14 18 22 26 30 59 55 8 12 8 51 34 0 47 8 8 22 32 43 6 16 39 12 22 35 0 16 0 31 8 32 l

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22A8560 Rev. O Ab ATURAL C RCULATIO 4 81 00 PERCENT ROD LI 4E CL LTIMATE STABILITY LINE 1.00  :  ::

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NUCLEAR ENERGY BUSINESS OPERATIONS e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 96126

, . GEN ER AL $ ELECTRIC APPLICABLE TO:

NEDO-24146A

" ' " ^ " "" -

79NED273 ERRATA And ADDENDA T. I. E. NO-

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SHEET TITLE 10 NO.

DRESDEN UNITS 2, 3 AND QUAD DA E April 1983 CITIES 1.2 NUCLEAR POWER STATIONS April 1979 NOTE: CorrectsIIcopies of the applicable ISSUE DATE publication as specifiedbelow.

I EFERENCES INSTRUCTIONS ITE M PpRAG APH L NE) (CORRECTIONS AND ADDITIONS) 1

1. Page 4-11 Replace with new page 4-11.

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  • NEDO-24146A ATTMllA@T 3 Table 41 HAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Dresden 2,3/ Quad Citio 1,2 FUEL TYPE: P8DRB282 Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) __

(kW/ft) (*F) Fraction '

200 11.2 2132 0.029 1,000 11.2 2132 0.029 5,000 11.8 2183 0.033 10,000 12.0 2189 0.032 .

f 15,000 12.0 2199 0.033 20,000 11.8 2181 0.032 25,000 11.3 2110 0.025

  • 30,000 11.1 2075 0.043 35,000 10.4 1981 0.035 40,000 9.8 1886 0.021 NOTE: Credit taken for the effects of pre-pressurization of the fuel rods Table SJ HAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Dresden 2,3/ Quad Cities 1,2 FUEL TYPE: P8DRB265H/BPSDRB265H

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Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t) (kW/ft) (*F) Fraction 200 11.5 2163 0.032 1,000 11.6 2171 0.032 5,000 11.9 2192 0.033 10,000 12.1 2198 0.033 15,000 12.1 2200 0.033 20,000 11.9 2190 0.032 l 25,000 11.3 2116 0.026 30,000 10.7 2018 0.018 35,000 10.2 1934 0.022 I

40,000 9.6 1835 0.009 45,000 8.9 1755 0.007 NOTE: Same as Table 41 4-11

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