ML20235C433
| ML20235C433 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/30/1987 |
| From: | Charnley J, Elliott P, Lambert P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19304B493 | List: |
| References | |
| 23A5831, NUDOCS 8709240450 | |
| Download: ML20235C433 (18) | |
Text
{{#Wiki_filter:- _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ __ _ - - . 23A5831 Revision 0 Class I June 1967 SUPPLEMENTAL RELCAD LICENSING SUEMITIAL FOR QUAD CIIIFS NUCLEAR PCKER STATION UNIT 1, RELOAD 9 Preparedi~ F. A. Lambert Fuel Licensing Verified: YC P. E. Elliott h[ ~ Fuel Licensing Approved: TJ $7 J. f. Charnley, Man e Fuf1LicensinF NUCLE AR ENERGY BUSINESS OPERATIONS
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL h ELECTRIC 1/2 9709240450 870918 DR ADOCK 0500 4
23A5831 Rev. 0 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT F. EASE READ CAREFULLY This report was prepared by General Electric solely for Commonwealth Edison Company (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 informa-tion in this document are contained in the contract between Commonwealth Edison Company and Iowa-Illinois Gas and Electric Company and General Electric Company for fuel bundle fabrication and services for Quad Cities Nuclear Power Station Units 1 and 2, dated January 6,1986, and nothing contained in this document shall be construed as changing said contract. The use of this inf ormation 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 f rom such use of such inf ormation. 3/4
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i 23A5831 Rev. O ! ACFROWLEDGMENT The engineering and reload licensing analyses which form the technical basis of this Supplemental Reload Licensing Submittal, were performed by q T. P. Shannon of the Nuclear Fuel and Engineering Services Department. l t I j 5/6
23A5831 Rev. O l 1
- 1. PLANT-UNIQUE ITEMS (1.0)*
l GETAB and Transient Analysis Initial Conditions: Appendix A l i
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0) l Fuel Type Cycle Loaded Number l
Irradiated P8DRB265L 7 24 P8DRB265H 7 88 BP8DRB265H 8 116 BP8DRB283H 8 80 BP8DRB299 9 144 BP8DRB282 9 72 New BD300A 10 120 BD300B 10 80 Total 724
*( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-8, May 1986; a letter "S" preceding the number refers to the U.S. Supplement, NEDE-24011-P-A-8-US, May 1986.
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23A5831 Rev. 0
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposura 21,080 mwd /MT at end of cycle: Minimum previous cycle core average exposure at 20,860 mwd /MT end of cycle from cold shutdown considerations: Assumed reload cycle core average exposure at 22,176 mwd /MT end of cycle: 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, Keffective Uncontrolled 1.1052 Fully Controlled 0.9590 Strongest Control Rod Out 0.9868 R, l'eximum Increase in Cold Core Reactivity with 0.0 Exposure into Cycle, AK
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (AK) ppm (20*C, Xenon Free) 6!0 0.041 3 8
23A3831 Rev. 0
- 6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)
(Cold Water Injection Events Only) Void Fraction (%) 34.9 Average Fuel Temperature (*F) 982 Void Coefficient N/A* (d/1 Rg) -4.68/-5.85 Doppler Coefficient N/A (d/*F) -0.239/-0.227 i Scram Worth N/A ($) **
- 7. RELOAD UNIQUE GE1AB 1RANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)
Fuel. Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial .R-Factor (NWt) (1000 lb/hr) HCPR Exposure: ECC10 to EOC10 BP/P8x8R 1.20 1.85 1.40 1.051 6.280 104.2 1.26 GE8x8Eb- 1.20 1.87 3.40 1.051 6.311 106.0 1.26
- 8. SELFCIED MARGIN IMPROVEMENT OP110NS (S.2.2.2) 3ransient Recategorization: No Recirculation Pump Trip: No Rod kithdrawal Limiter: No l
Thermal Power Monitor: No Improved Scram idae: Yes (ODYN Option B) Exposure Dependent Limits: No Exposure Points Analyzed: 1 1 l l
*N = Nuclear Input Data, A = Used in Transient Analysis ** Generic exposure independent values are used as given in " General Electric Standard Application f or Reactor Fuel," NEDE-24011-P-A-8-US, dated Mey 1986.
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A
'Q 4
23A5831 i ' Rev. 0,t
\ , 'i ai
- 9. CPERATING FLEXIBILITY OPTIONS (S.2.2.3) ;,
-{
Single-Loop Operation: Yes Load Line Limit: - Yes ! Extended Load Line Limit: Yee Increased Core Flow: No Flow Io'nt Analyzed: Tid- , Feedwater Temperature Reduction: No ARTS Program:\ No' Maximum Extended Operating Domain: No
- 10. CORE-WIDE TRAh}iENT ANALYSIS REJLC.TS (5.2.2.1) !
Nethods Used: GEMINI Exposure Range: B0010 to EOC10 Flux 6 CFR Q/A ___
' , Transient (% NBR) (% NBR) PP/P8x8R GE_8x8EB Figure, Load Rejectign Without Bypass 505 120 0.19 0.19 -2 Loss of Feedwater Heating 121 119 '9.18 0.18 3 Feedwatey Controller Failure 230 116 0.34' O.14 4 5
3 Y t 4
23A5831 Rev. 0
- 11. . LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(S.2.2.1) Limiting Rod Pattern: Figure 5 Rod' Block Rod' Position ACPR MLHGR (kW/ft) Reading (ft withdrcus) BP/P8x8R GE8x8EB BP/ P8x8R GE8x8EB 104 3.5 0.11 0.11 16.37 17.37 101 3.5 0.11 0.11 16.37 17.37
,8 106 4.0 0.16 0.16 17.28 18.28 107 4.0 0.16 0.16 17.28 18.28 108 445 0.19 0.19 17.60 18.60 109 7.0 0.26 0.26 17.60 18.60 110 13.0 0.32 0.32 17.60 18.60 Setpoint Selected: 107
- 12. CYCLE MCPR VALUES (S.2.2)
Non-Pressurization Events Exposure Range: BOC10 to EOC10 BP/P8x8R GE8x8EB
.. . Loss of Feedwater Heating 1.25 1.25 Fuel. Loading Error' '
1.18 Rod W1 thdrawal Error 1.23 1. 23 Pressurization Events *
"xposure Ra nge: BOC10 to EOC10 Option A Option B BP / pbx 8R GE8x8EB BP /P8x8R GE8x8EB Lcad Rejection Without Bypass 1.33 1.33 1.28 1.28 Feedwater Contr oller Failure 1.27 1.27 1.22 1.22
*CpYN Adjustment Factors are documented in a letter, J. S. Charnley (GE) to G. C. Lainae (NRC), " GENIN 1/0 LYN Statistic al Adders f or BWR/2,3 Flant s (without RPT)-MOC/EOC," Kireb 13, 1987.
11 )
23A5831 Rev. 0
- 13. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3) sl v Transient (psig) (psig) Plant Response MSIV Closure 1295 1319 Figure 6 (Flux Scram)
- 14. SIABILITY ANALYSIS RESULTS (S.2.4)
Quad Cities Nuclear Power Station Unit 1 is exempt from the current requirement to submit a cycle-specific stability analysis as documented in the letter, C. O. Thomas (NRC) to H. C. Pfefferlen (GE), " Acceptance for Referencing of Licensing Iopical Report NEDE-24011, Rev. 6, Amendment 8, 'Ihermal Hydraulic Stability Amendment to GESTAR II'," April 24, 1985.
- 15. LOADING ERROR RESULIS (S.2.5.4)
Variable Kater Gap Disoriented Bundle Analysis: Yes Event p_CPR Disoriented 0.11
- 16. CC,NTROL ROD EROP ANALYSIS RESULIS (S.2.5.1)
Quad Cities Nuclear Power Station Unit 1 is a Banked Position Withdrawal Sequence plant, so the Control Rod Drop Accident Analysis is not required. NRC approval is documented in NECE-24011-P-A-8-US, May 1986. 12
23A5831 Rev. 0
- 17. LOSS-OF-COOLANT ACCIDENT RESULTS (S.2.5.2)
LOCA Method Used: SAFER /GESTR-LOCA Quad Cities Nuclear Power Station Units 1 & 2, " SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31345P, to be issued. 13
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23A5831 Rev. 0 8 8
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l l fMMMME"' l l l l l l l l l l l l 1 3 5 7 91113:51719212" " 'o313335373941434547,95153555753 FUEL TYPE A = BP8DRB265H E = BD300A B = BP8DRB283H F = BD300B C = BP8DRB299 G = P8DRB265L D = BP8DRB282 H = P8DRB265H Figure 1. Reference Core Loading Pattern 14
23A5831 Rev. O I 1 NEU RON FLUX 1 VESEEL PRESS RISEEPSI) 2 AVE SUPr ACE HE AT FLUX 2 S Ar :TY V/LVE FLCd 3 COR: INLET FLOW 3 REL IEF VALVE FLC.' 150.0 300.0 a 9:r*1: =_qr_- , 4
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Figure 2. Plant Response to Generator Load Rejection Without Bypass (E0C) 15
23A3831 Rev. 0 150.0 1 NEUTRCN FLUX 2 AVE SUFF ACE HE AT FLUX 21VE2hELPRE3SRISECPSI) PEL EF YALVE FLOW 3 COG E INLET Flaw 3 BtP$$5 HLVE FLCW l' J. ) + f.a 4 ::c : o n_ r + ef I 100.0 2 ,~ M - 3 : : 3: b t c o, a u p, , , Y y- s s s j E l 50.0 I f u 50.0
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'[DCFFLERPEACTIVITY I M PEACTIv!TY 33,9 i 7;r mi-r, <! na 1. *. b]SC;dgri:+:utrv i' .1 . ; w_ p Nr i
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t!"E ($ECCNDS! itME (SECONOS) Figure 3. Plant Response to Loss of Feedwater Heating (E0C) 16
23A5831 Rev. 0 1* 0. 0
) NEUTRON FLUX 1 VES EL PRESS RISE (PSI) 2 AVE SURF ACE Hi T FLUX 2 S AF TY VAL /E FLOW 3 CORE INLET FL. 3 REL EF V AL VE FLCW 150.0 ' CCr r INLE' ED" 4 BYP \55 VALVE FLCW t
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23A5831 Rev. 0 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 2 2
, 55 8 32 32 32 8 51 0 8 16 16 8 0 h
. 47 32 38 38 38 32 43 2 16 0 0 16 2 39 8 32 38 38 38 32 8 35 0 0 8 16 16 8 0 0 31 8 8 32 32 32 8 8 27 0 0 8 16 16 8 0 0 23 8 32 38 38 38 32 8 19 2 16 0 0 16 2 15 32 38 38 38 32 11 0 8 16 16 8 0 7 8 32 32 32 8 3 2 2 NOTES:
- 1. No. indicates number of notches withdrawn out of 48. Blank is a Withdrawn Rod.
- 2. Error Rod is (26, 43).
Figure 5. Limiting Rod Pattern 18
23A5831 Rev. O i g i NfukEON FLUX 1 VE5kEt. PREIS RISElFSI)
? AVE !UV ACE HE AT FLUX 2 57Frif VFL VE FLCW 3 COE i INLE FuCW 3 FELD Vf LVE FLCv 150,0 300 8 - ' E?I G L rLC'
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-i a i 100.0 -2.0 C. D 5.O 1 C. 0 C.D 5.0 10.O ilat (CE COND ) 11ML t SECONDS Figure 6. Plant Response to MSIV Closure, Flux Scram 19 -
23A5831 Rev. O APPENDIX A GETAB AND TRANSIENT ANALYSIS INITIAL CONDITIONS The values used in the GETAB and Transient Analysis for non-fuel power fraction and relief valve capacities and setpoints are given in Table A-1. The following values differ from the values reported in NEDE-24011-P-A-8-US, May 1986. Table A-1 PLANT PARAMEIER Parameter Analysis Value NEDE-24011 Value Non-Fuel Power Fraction 0.039 0.035 Relief Valve (RV): Capacity at Peference 558,000 645,000 Pressure (1bs/hr) 20 FINAL
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