ML20154A214

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Rev 0 to Supplemental Reload Licensing Submittal for Clinton Power Station Unit 1 Reload 1,Cycle 2
ML20154A214
Person / Time
Site: Clinton Constellation icon.png
Issue date: 08/31/1988
From: Charnley J, Lambert P, Pentzien D
GENERAL ELECTRIC CO.
To:
Shared Package
ML19297G893 List:
References
23A591-R, 23A591-R00, 23A5921, NUDOCS 8809120111
Download: ML20154A214 (18)
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23A5921 REVISION 0 CLASS I AUGUST 1988 23A5921 REY. 0 SUPP11 MENTAL RELCAD LICENSING SUBMITTAL FOR CLINTON PokT.R STATION UNIT 1 RELOAD 1. ' CYCLE 2 Prepared by: -

P. A. Lambert Fuel Licensing

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Verified by:  !. -

D. C. Pentzien" Fuel Licensing I

Approved ya ,-- ,

. Y Charnley Manager, Fuel Licens ng i

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IMPORTAhT NOTICE REGARDING CONTERIS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by the General Electric Company (GE) solely for Illinois Power Company (IPC) for IPC's use with the United States Nuclear Regulatory Commission (USNRC) for amending IPC's operating license of the Clinton Power Station Unit 1. The information contained in this report is believed by GE to be an accurate and true representation of the

' facts known, obtained or provided to GE at the time this report was prepared.

The only undertakings of the GE respecting information in this document are contained in the contract between IPC and GE for nuclear fuel and related services for the nuclear system for Clinton 1, 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 GE nor any of the contributors to this document makes any representation or warranty (express l 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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ACKNOWLEDGDGNTS The engineering and reload licensing analyses,'which form the technical basis of this Supplemental Reload Licensing Submittal, were performed in the Fuel Engineering Section by T. C. Hoang.

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1. PLANT-UNIQUE ITEMS (l.0)*

I Appendix A GETAB and Transient Analysis Initial Conditions Appendix B: Basis for Analysis of Loss-of-Feedwater Heating Event Appendix C: Application of GEMINI Methods

2. RELOAD TUEL BUNDLES (1.0. 2.0. 3.3.1 AND 4.0)

Fuel Type Cycle Loaded Number Irradiated P85RB154 1 96 P8 SRB 200 1 360 New BP8 SRB 284LC** 2 88 BP8 SRB 284L 2 80 -

Total 624

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure at end of cycle: 9314 mwd /MT Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 8773 mwd /MT Assumed reload cycle core average exposure at and of cycle (all rods out, rated power): 12.393 HWd/HT Core loadir.g pattern: Figure 1

  • ( ) Refers to area of discussion in General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A-8 (dated May 1986); a letter "S" preceding the number refers to the United States Supplement.
    • Bundle-specific information for this bundle design was submitted to the U.S. NRC in a letter, J.S. Charnley (GE) to H.N. Berkov (NRC), "Proposed Amendment 16 to GE Licensing Topical Report NEDE-24011-P-A " August 8, 1986. U.S. NRC approval of this Amendment was documented in a letter.

A.C. Thadani (NRC) to J.S. Charnley (GE), "Acceptance for Referencing of Amendment 16 to GE Licensing Topical Report NEDE-24011-P-A, ' General Electric Standard Application for Reactor Fuel'," April 20. 1988.

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4. CALCULATED CORE EFTECTIVE MULTIPLICATION AND CONTROL SYSTEM '

WORTH- NO VOIDS. 20 DEG. C (3.3.2.1.1 AND 3.3.2.1.2)

Beginning of Cycle, k-effective Uncontrolled 1.108 Fully Controlled 0.934 Strongest Control Rod out 0.974 R. Maximum Increase in Cold Core Reactivity with Exposure into Cycle, Delta k 0.000

5. STANDBY LIQUID CONTROL SYSTE SHUTDOWN CAPABILITY (3.3.2.1.3) -

Shutdown Margin (Delta k)

E2m (20 den.C. Xenon Free) 660 0.065 ,

6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.S AND S.2.2) ,

Values normally reported in this section are REDY inputs. There were no transients aralyzed using REDY.

7. RELOAD-UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2 .2) l Exposure BOC2 TO EOC2 l

l Fuel Peakina Factors R- Bundle Bundle Flow Initial l

Design Local Radial Axial Factor Power (MWT) (1000 lb/hr) MCPR BP/P8x8R 1.20 1.55 1.40 1.051 7.054 107.1 1.12 l

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8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip Yes ,

Rod Withdrawal Limiter Yes Thermal Power Monitor Yes l Improved Scram Times No Exposure Dependent Limits: , No Exposure Pointa Analyzed: 1

9. OPERATING FLEXIBILTTY OPTIONS (S.2.2.3)

Single-Loor Operation: Yes ,

Load Line Limit No Extended Load Line Limit: No .

Increased Core Flows No -

Flow Point Analyzed: N/A Feedwater Temperature Reduction: No ARTS Program No Maaimum Extended Operating Domains No

10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1) i Methods Used: GEMINI Delta CPR '

Flux Q/A Transient (% NBR) (: NBR) BP/P8xBR Tirure Exposure Range BOC2 to EOC2 -

Pressure Regulator 141 103 0.04 2 Failure Downscale l

  • 0.11

Load Rejection Without 172 101 0.01 4 (

= Bypass .

  • See Appendix B '

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11. LOCAL ROD WITMDRAWAL ERROR (WITH LIMITING INSTRtMEb7 FAILURE)

TRAN$IENT

SUMMARY

(S.2t2.1)

( The generic bounding BWR/6 Rod Withdrawal Error analysis described in l NEDE-24011-P-A-8-US is applied the resulting delta CPR is 0.11.

12. CYCLE MCPR val.UES (S.2.2)*

l Non-Pressurization Events -

l l Exposure Ranges BOC2 to EOC2 BP/P8x8R Loss of 100*F Feedwater Heating 1.18 Rod Withdrawal Error 1.18 Pressurization Events 1

1 Exposure Range BOC2 to EOC2 Option A BP/P8x8R l

Pressure Regulator Failure Downscale 1.12 Feedwater Controller Failure 1.13 Load Rejection Without Bypass 1.09

  • GEMINI ODYN adjustment factors are provided in the letter from J.S.

Charnley (GE) to M.W. Hodges (NRC). "GEMINI ODYN Adjustment Factors for BWR/6." dated July 6. 1987. .

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13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) ,

Steam Line Vessel Pressure Pressure Plant r Transient (psin) (osin) Response MSIV Closure 1204 1247 Figure 5 (Flux Scram)

14. LOADING ERROR RESULTS (S.2.5.4)

Loading Error Results are not applicable for BWR/6 plants. NRC approval of the non-applicability of Loading Errors to BWR/6 plants is documented in Section S.2.5.4 of NIDE-24011-P-A-8-US, entitled "Loading Error Accident Calculational Hethods."

15. C0KrROL ROD DROP ANALYSIS RESULTS (S.2.5.1)

Banked Position Withdrawal Sequence is utilized at the Clinton Power Statio,n therefore, the bounding Control Rod Drop Analysis (CRDA) described in NEDE 24011-P-A-8-US is applied. NRC approval of the bounding analysis is given in the letter to J.S. Charnley (GE),

"Acceptance for Referencing of Licensing Topical Report NEDE-24011 Revision 6, Amendment 9 'GESTAR-II General Electric Standard Application for Reactor Fuel'," January 25, 1985.

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16. STABILITY ANALYSIS RESULTS (S.2.4)

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GE SIL-380 recommendations have been included in the Clinton Power  ;

Station operating procedures and Technical Specifications: therefore, ,

! no stability analysis is required. Furthermore, Clinton Power Station operating procedures require an immediate manual scram if both recirculation pumps trip.

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17. LOSS-OF-COOLANT ACCIDENT RESULTS (S.7.5.2)'

LOCA Method Used: SAFE /REFLOOD (see Clinton Power Station Final Safety Analysis Report)

Fuel Type BP8 SRB 284L Avera;;e Planar Exposure MAPLHGR 0xidation (CWd/MT) (kW/ft) PCT ('F1, Traction

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0.2 11.7 2018 0.016 .

1.1 11.8 2016 0.016 5.5 12.4 2053 0.017 11.0 12.6 2067 0.017 16.5 12.6 2075 0.018 22.0 12.5 2078 0.018 27.6 11.9 2007 0.014 '

38.6 10.8 1850 0.008 49.6 9.4 1681 0.004 r

Fuel Type- BP8 SRB 284LC Average Planar Exposure MAPLHGR Oxidation  !

(CWd/MT) (kW/ft) PCT ( F) Fraction 0.2 11.6 2007 0.016 1.1 11.8 2014 0.016 5.5 12.6 2071 0.018 11.0 12.6 2064 0.017 16.5 12.6 2073 0.018 22.0 12.6 2078 0.018 27.6 11.9 2008 0.014 38.6 10.3 1852 0.008 49.6 9.5 1680 0.004 t r

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  • APPENDIX A GETAD AND TRANSIENT ANALYSIS INITIAL CONDITIONS To accurately reflect actual plant pcrameters, the values shown in Table A-1 were used instead of the values reported in NEDE-24011.-P-A-8-US, May 1986.

1 Table A-1 ,

PLANT PARAMETER Parameter Analysis Value Thermal Power, MWt 2894 0

Rated Steamflow, 10 lb/hr 12.45 Dome Pressure, psig 1025 ,

Turbine Pressura, psig 980 .

Non-Fuel Power Fraction 0.038 Dual Mode Safety / Relief Valves Number of Valves 16 Relief Mode Low Setpoint, psig 1133 Safety Mode Low Setpoint, psig 1180 Capacity,1h/hr 924,933 (Ref. Pressure psig) (1190) t i

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  • 1 APPENDIX B BASIS FOR ANALYSIS OF LOSS-0F-FEEDWATER HEATER EVENT. I The Loss-of-Feedwater Heater event was analyzed with the 3D BWR Simulator code described in NEDE-24011-P-A-8-US. The transient analysis inputs normally reported in Section 6 of the licensing submittal are internally calculated in the 3D BWR Simulator code and in ODYN. The transient plots, flux, and Q/A normally reported in Section 10 are not outputs of the 3D BwR Simulator code; therefore, these items are not included in this doctunent.

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APPENDIX C APPLICATION OF GEMINI METHODS The GEMINI system of methods are used to perform the licensing analyses of clinton Power Station (CPS) Reload 1. The GEMINI system of rathods is described in Reference 1; NRC approval of these methods is documented in Reference 2. In Reference 3, the application of GEMINI methods in licensing analyses ir described. Pressurization events that could eatablish the Operating ',Jedt MCPR are analyzed at the 100% power level. Power level uncertainties specified in Regulatory Guide 1.49 are accounted for by adding adjustment factors to the calculated delta CPR.

NRC approval of this procedure is provided in Reference 4.

Rod Vithdrawal Error The NRC approved generic Rod Withdrawal Error analysis for PWR/61s described in Reference 5 is applied to CPS Reload 1. An evaluation of the impact of GEMINI methods on the generic analysis indicates that the results of the generic analysis continue to be conservative and bounding.

Overpressurization Analysis The MSIV Closura (Flux Scram) analysis is performed usin.c GEMINI methods at the 102% power level to account for the power level uncertainties specified in Regulatory Guide 1.49.

Control Rod Drop Accident The NRC approved boanding Control Rod Drop Accident analysis for Banked Position Withdrawal Sequence plants (such as CPS) described in Deference 1 is applied to CPS Reload 1. The impact of GEMINI methods on the results of the generic analysis is negligible.

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a Stability The NRC approved generic stability approach described in Reference 1 is applied to CPS Reload 1. The use of GEMINI methods does not impact the generic analysis.

References

1. Letter, J.S. Charnley (GE) to C.O. Thomas (NRC), "Amendment 11 to GE LTR NEDE-24011-P-A," February 27, 1985,
2. Letter, C.O. Thomas (NRC) to J.S. Charnley (GE), "Ac eptance for Referencing of Licensing Topical Report NEDE-24011-P-A, Rev. 6, Amendment 11, ' General Electric Standard Application for Reactor Fuel'," November 5, 1985. '

1

3. Letter, J.S. Charnley (GE) to H.N. Berkow (NRC), "Revised Supplementary Information Regarding Amendment 11 to GE Licensing l

Topical Report NEDE-24011-P-A," January 16, 1986.

4. Letter, G.C. Lainas (MRC) * , J.S. Charnley'(GE), "Acceptance for Referencing of Licensing Topical Report NEDE-24011-P-A, 'GE Generic Licensing Reload Report', Supplement to Amendment 11," Harch 22, 1986.
5. "GESSAR-II 238 BWR/6 Nuclear Island Design," NRC Docket No. STN50-447.

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24/25 (Final)

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