ML20215A527

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Rev 0 to Supplemental Reload Licensing Submittal for Quad Cities Nuclear Power Station,Unit 2,Reload 8 (Cycle 9)
ML20215A527
Person / Time
Site: Quad Cities, 05000000
Issue date: 07/31/1986
From: Charnley J, Casey Smith, Zarbis W
GENERAL ELECTRIC CO.
To:
Shared Package
ML20215A501 List:
References
23A4758, 23A4758-R, 23A4758-R00, NUDOCS 8610060110
Download: ML20215A527 (16)
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Text

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JULY 1986 a

s SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR QUAD CITIES NUCLEAR POWER STATION UNIT 2, RELOAD 8 (CYCLE 9[

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23A4758 Revision 0 Class I July 1986 f

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR QUAD CITIES NUCLEAR P0'JER STATION UNIT 2, RELOAD 8 (CYCLE 9)

i Prepared
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C. W. Smith Verified:

W. A. is __,

Approved: - #- - /MI.e /

' J f~S'..CidrHisy,#Managef~'f /

Fuel Licensing / '

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NUCLEAR ENERGY BUSINESS OPERATIONS . GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENER AL $ ELECTRIC 1/2

23A4758 Rsv. O IMPORTANI NOTICE REGARDING CONIENIS OF THIS REPORT PLEASE 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 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 inf or-mation in this document are contained in the contract between Commonwealth Edison Company and General Electric Company for f uel bundle f abrication and services for Quad Cities Nuclear Power Station Units 1 and 2, dated December 1,1978, as amended and nothing contained in this doctanent shall be construed as changing said contract. The use of this information except as defined by said contract, or f or any purpose other than that f or which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or warranty (express or implied) as to the couplete-ness, accuracy or usefulness of the inf ormation contained in this document or that such use of such information may not inf ringe privately owned rights; nor do they assume any responsibility f or liability or damage of any kind which may result from such use of such information.

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23A4758 Rsv. 0

1. PLANT UNIQUE ITEMS (1.0)*

A. Plant Parameter Changes See Appendix A B. R (Item 4) Value Shown Includes Effect of B4 C Settling (G.0005 Ak)

C. Stability Analysis See Appendix B

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

Cycle Loaded Number Number Drilled Irradiated P8DRB265L 6 80 80 BP8DRB282** 8 72 72 BP8DRB283H** 8 104 104 P8DGB263L** 6 32 32 P8DGB298** 6 32 32 P8DGB284** 6 48 48 BP8DRB265H** 7 204 204 New BP8DRB299L*** 9 88 88 BP8DRB299** 9 64 64 Total 724 724

?

> *( ) refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A-7, dated August 1985.

    • Barrier Fuel
      • Barrier fuel included in Amendment 13 to GESTAR II. See Letter, J. S. Charnley (GE) to C. O. Thomas (NRC), " Proposed Administrative Amendment to GE Licensing Topical Report NEDE-24011-P-A",

September 24, 1985.

5

1 23A4758 Rev. 0

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end 20656 mwd /Mr of cycle:

Minimum previous cycle core average exposure at end 20329 mwd /Mr of cycle from cold shutdown considerations:

Assumed reload cycle core average exposure at end 21539 MW4/Mr of cycle:

Core loading pattern: Figure 1

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONIROL SYSTEM WORTH -

NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)

Minimum Shutdown Margin, B0C, k effective Uncontrolled 1.106 Fully Controlled 0.951 Strongest Control Rod Out 0.981 R, Maximum Increase in Cold Core Reactivity 0.005 with Exposure into Cycle, Ak

5. STANDT2Y LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin ( Ak) ppm (20*C, Xenon Free) 600 0.039 d

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23A4758 Rgv. 0

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

(REDY EVENTS ONLY)

Void Fraction (%) 34.4 Average Fuel Temperature (*F) 1147 Void Coefficient N/A* (d/% Rg) -5.23/-6.54 Doppler Coefficient N/A (d/*F) .19/ .18 Scram Worth N/A* ($) **

7. REIDAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)

Fuel Peaking Factors Bundle Power Bundle Flow Initial Destgn Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR Exposure: BOC9 to EOC9 P8x8R/ 1.20 1.73 1.40 1.051 5.870 108.1 1.36 BP8x8R 8x8R 1.20 1.75 1.40 1.051 5.938 105.8 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: No 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 Standard Application for Reactor Fuel," NEDE-24011-P-A-7, dated August 1985.

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23A4758 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 BP8x8R/

Transient (% NBR) (% NBR) ,P8x8R 8x8R Figure Exposure: BOC9 to E0C9 Load Rejection Without Bypass 619 120 0.29 0.27 2 Loss of 145*F Feedwater 121 120 0.17 0.17 3 Heating Feedwater Controller Failure 310 118 0.20 0.18 4

11. LOCAL R0D WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT

SUMMARY

(S.2.2.1)

(Generic Bounding Analysis Results)

Rod Block ACPR Reading BP8x8R/P8x8R/8x8R 104 0.13 105 0.16 -

106 0.19 107 0.22 108 0.28 109 0.32 j 110 0.36 Setpoint selected: 107 8

23A4758 Rty. 0

12. CYCLE MCPR VALUES (S.2.2)

Non-Pressurization Events Exposure Range: BOC9 to EOC9 BP8x8R/

P8x8R 8x8R Loss of Feedwater Heater 1.24 1.24 Fuel Ioading Error 1.20 -

Rod Withdrawal Error 1.29 1. 29 Pressurization Events Exposure Range: B0C9 to E0C9 Option A Option B BP8x8R/ BP8x8R/

P8x8R 8x8R P8x8R 8x8R Load Rejection w/o Bypass 1.42 1.40 1.37 1.35 i~

Feedwater Controller Failure 1.32 1.31 1.24 1.23

13. OVERPRESSUR.12ATION ANALYSIS

SUMMARY

(S.2.3)

P P s1 y Transient (psig) (psig) Plant Response MSIV Closure 1317 1334 Figure 5 (Flux Scram)

14. STABILITY ANALYSIS RESULTS (S.2.4)

See Appendix B.

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23A4758 Rev. 0

15. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes Resulting MCPR Event Initial MCPR Misoriented 1.18 1.07

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

Not analyzed for Banked Position Withdrawal Sequence plants.

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," NEDO-24146A, April 1979 (as amended).

6 i

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23A4758 R;v. 0

mMMMMMm
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mMMMMMMMMMMMm
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M M M M M M M M M M M M M M M
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MMMMMMMMMMMMM
  • MMMMMMMMMMM*
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MMMMMm i IIIIIIIIi 1 357 911131517192123252729313335373941434547495153555759 BUNDLE TYPE B = BP8DRB282 G = P8DGB263L 1:!"*ll*i" ":P8lil" J = BP8DRB265H E = BP8DRB299 Figure 1. Reference Core Loading Pattern 11

- _ _ _ _ _ - _ _ -_ _ _ l

mme .e l 1 NEUTRON FLUX l VESiEL PRESS RISEtPSI) 2 AVE SURF ACE HEAT FLUX 2 SAF ETY VALVE FLOW 3 CORE INLET FLOW iS .. . 2.... 3JAMssA.LVE es FLOW Le- i E

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0. 0 10.8 20.0 8. 0 10.0 20.0 TIME (SECCES) TIME (SECO@SI 1LEviL(INCH REF-SEP-SMRT) I 1 VOI ) REACilvlTY 2 VESSEL STEAMFLOW 2 00P'LER REACTIVITY 3 TUR31NE STE AMFLOW 3 SCRLM REACTIVITY

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... i e. 2.. ... i s. 28.o fine (SECONDS) IlPE (SECO251 Figure 2. Plant Response to Generator Load Rejection, without Bypass 12

23A4758 Rev. O ise.e 1NEUIRON FLUX X VESSEL PRESS RISEtPSI) 2 E {AT FLUX 2 RELIEF VALVE FLOW R; ]M_ET F .0V 3 BYP(SS VALVE FLOW 134.0 ' Emor ' RET E ?

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t. 6 100.0 240. 0 0. 0 104.0 200.0 flME (SECONOS) TIwE (SECONOS1 ILEVEL(INCH REF-SEP-SKRT) i VOI ) REACTiv i 2 VESiEL STEAMFLOW 2 DOP)LER REAC 'l ITY 3 TUR IINE STEAMFLOW y 1.0 3, S,CR,LM RE A.C,TI,r _,

150. 0 m rEE y a ?E n etgy g iL og g ,

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9. 0 ite.0 280.0 0. 0 100.9 200.0 TIME (SECON051 f!RE (SECONOS)

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

23A4758 Rsv. 0 150.0 1NEurRON LUX 1 YEsiEL PRESS RISEtPSI) 2 AVE SURF CE AT FLUX 2 SAF TY VAL FLOW h 3 EORE  : '"

!%:T F W 3 RELlEF VAL FLOW  ;

150. 0 ' tan ot SL 4 BYPLSS VA E FLOW o

+ 300.0 0

8 :fY 0 100.0 < #. ., .- '- -

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y 50.0 E [W r

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0. 0 l
0. 0 20.0 40.0 0. 0 20.0 40.0 TIME (SECONOS) f!ME (SECONOS1 1 LEY'L(INCH REF-SEP-SKRT)  ! VOI ) REA lotTY 2 VESSEL STEAMFLOW 2 DOP)LER Al: Y 3 TUR31NE STEANFLOW 1 SCRnM RE CT 150.0 3. 0 . vntu- e r_
  • r e. e_

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0. 0 20.0 40.0 0. 0 20.0 48 8 TIME ($ ECON 051 f!ME (SECONOS)

Figure 4. Plant Response to Feedwater Controller Failure 14

23A4758 Rav. O 1 VES iEL PRESS RISE (PSI) hI l 1 NEUTRON FLUX 2 AVE SURF ACE HE AT FLUX 2 REL IEF VALVE FLOW 3 co" "'"* 3.... 35". !,!!M_ _ i5: w is.. . _

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!!ME (SECONOS) f!ME (SECON05)

I LEY :L(INCH.REF-SEP.SKRT) I VOI ) EACTIVITY 2 VESLEL STEAMFLOW 2 00P) R REACTIVITY 3 TUR)!NE STE AMFLOW 3 SCRL REACTIVITY 200. 0 g rre matro r i_ gw l9 4 egtn_ erirteurtw 300.0 -

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0. 0 , -1.0 I y..

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9. 8 5.0 10.0 8.0 5.0 88.0 flME (SECONDS) f!ME (SECON051 Figure 5. Plant Response to MSIV Closure, Flux Scram 15/16 L .- _________ _ ___-________

23A4758 R:v. O APPENDIX A PLANT PARAMETER CHANGES Pressure Relief Systems (Table S2-4.1, pg. US2-92, NED0-24011)

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

! Lowest Relief Valve setpoint (psig) 1115 l Transient Operating Parameters (Table S.2-6, pg. US2-98, NED0-24011) l Thermal Power (MWt) 2511 Turbine Pressure (psig) 950 GETAB Initial Conditions (Table S.2-8 Pg. US2-102, NED0-24011)

Reactor Core Pressure (psia) 1035

{ Inlet Enthalpy (btu /lb) 523.7 ATWS - RPT initiated at 1250 psig

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

17/18

7.

23A4758 Rsv. O APPENDIX B STABILITY ANALYSIS i

According to Reference B-1, Quad Cities Unit 2 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. Pfefferlen (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8, ' Thermal Hydraulic Stability Amendment to GESTAR II'",

April 24, 1985.

19/20 (FINAL)

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