ML20134K424

Rev 0 to Supplemental Reload Licensing Submittal for Millstone Unit 1,Reload 10

ML20134K424
Person / Time
Site:	Millstone
Issue date:	08/31/1985
From:	Charnley J, Plotycia G, Zarbis W GENERAL ELECTRIC CO.
To:
Shared Package
ML20134K392	List: B11663, Proposed Tech Specs Changing Max Critical Power Ratio Operating Limits & MAPLHGR Curves for Cycle 10 & Correcting Typos ML20134K390 ML20134K424
References
23A4696, 23A4696-R, 23A4696-R00, NUDOCS 8508300193
Download: ML20134K424 (23)

Text

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23A4696 AUGU 1985 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR MILLSTONE UNIT 1 RELOAD 10

[BP R862 Bi88lip GEN ER AL h ELECTRIC

f 23A4696 Revision 0 Class I E

August 1985 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR MILLSTONE UNIT 1 RELOAD 10 Prepared:

co G. D. Plotycia'

(

Verified:

Nb W. A.

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Appro e-6 4S. Charnley, Manager [

uel Licensing NUCLEAR ENERGY BUSINESS OPERATIONS

• GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENER AL $ ELECTRIC 1/'

23A4696 Rev. 0 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY l

This report was prepared by General Electric solely for Northeast Utilities Service Company (NUSCo) for NUSCo's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending NUSco's operating license of the Millstone Nuclear Power Station. 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 infor-mation in this document are contained in the contract between Northeast Utilities Service Company and General Electric Company for nuclear fuel and related services for the nuclear system for Millstone Nuclear Power Station, dated April 14, 1967 and March 13, 1980 and nothing contained in this document shall be construed as changing said contracts. The use of this information except as defined by said contracts, 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 inf ringe 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.

3/4

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23A4696 Rsv. 0 1.

PLANT UNIQUE ITEMS (1.0)*

Control Rod Drop Analysis Appendix A GETAB and Transient Analysis Initial Conditions Appendix B Stability Analysis Appendix C Feedwater Temperature Reduction Analysis Appendix D Fuel Bundle Description Appendix E 2.

RELOAD FUEL BUNDLES (1,0. 2.0. 3.3.1 AND 4.0)

Fuel Type Cycle Loaded Number Number Drilled Irradiated P8DRB282 9

72 72 P8DRB283H 9

108 108

- I BP8DRB300**

10 200 200 V

6 New BP8DRB300**

11 200 200 580 580 r

3.

REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle core average exposure 17,533 mwd /ST

/

at end of cycle:

Minimum previous cycle core average exposure at 17,533 mwd /ST end of cycle from cold shutdown considerations:

Assumed reload cycle core average exposure at 18,256 mwd /ST end of cycle:

Core loading pattern:

Figure 1

• ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel", NEDE-24011-P-A-6, dated April 1983. A letter "S" preceding the number refers to the appropriate country-specific supplement.

• • See Appendix E.

5 l-

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23A4696 Rav. 0 I

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4.

CALCUIATED 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,gg Uncontrolled 1.106 Fully Controlled 0.954 Strongest Control Rod Out 0.979 R, Maximum Increase in Cold Core Reactivity 0.005 with Exposure into Cycle, Ak 5.

STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPMILITY (3.3.2.1.3)

Shutdown Margin (Ak) yJa (20*C, Xenon Free) 660 0.046 6.

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

(Cold Water Injection Events Only)

Void Fraction (%)

36.8 Average Fuel Temperature (*F) 1151 Void Coefficient N/A* (//% Rg)

-5.76/-7.20 Doppler Coefficient N/A (d/*F)

-0.183/-0.174 Scram Worth N/A ($)

• 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, dated April 1983.

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23A4696 Rev. 0 l

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RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2) 3 Peaking Factors Fuel Bundle Power Bundle Flow Initial Design Local. Radial, Axial' R-Fac to r (MWt)

(1000 lb/hr)

MCPR Exposure: BOC11 to EOC11 BP8x8R/ 1.20 1.62 1.40 1.051 5.4821 100.2 1.41 P8x8R 8.

SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2) f Transient Recategorization:

No Recirculation Pump Trip:

No Rod Withdrawal Limiter:

No Thermal Power Monitor:

No Improved Scram Time:

Yes (ODYN Option B)

Exposure-Dependent Limits:

No Exposure Points Analyzed:

1 9.

OPERATING FLEXIBILITY OPTIONS (S.2.2.3) l 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:

Yes 7

23A4696 Rev. 0*

10. CCRE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1) l Exposure: BOC11 to EOC11 I

ACPR Flux Q/A BP8x8R/

Transient

(% NBR)

(% NBR)

P8x8R Figure-Load Rejection Without Bypass 571 127 0.34 2

Loss of Feedwater Heater 116 115 0.14 3

Feedwater Controller Failure 109 108 0.07 4

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

SUMMARY

(S.2.2.1)

(Generic Bounding 'calysis Results)

ACPR Rod Block Reading (%)

(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: 108%

8

23A4696 Rev. 0

12. CYCLE MCPR VALUES (S.2.2)

Nonpressurization Events-Exposure Range: BOC11 to EOC11 BP8x8R P8x8R Loss of Feedwater Heater 1.21 1.21 Fuel Loading Error 1.26 Rod Withdrawal Error 1.35 1.35 Pressurization Events Exposure Range: BOC11 to EOC11 Option A Option B BP8x8R/P8x8R BP8x8R/P8x8R Load Rejection Without Bypass 1.45 1.42

- Feedwater Controller Failure 1.19 1.12

13. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3)

P,1 Py Transient (psig)

(psig)

Plant Response MSIV Closure 1254 1270 Figure 5 (Flux Scram)

14. STABILITY ANALYSIS RESULTS (S.2.4)

See Appendix C 9

T-23A4696 Rev. 0

15. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial MCPR Resulting MCPR Misoriented 1.24 1.07 16.

CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)

See Appendix A.

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

See " Loss-of-Coolant Accident Analysis Report for Millstone Unit 1 Nuclear Power Station", General Electric Company, July 1980 (NEDO-24085-1, as amended).

i I

10

23A4696 Rev. 0 BEM M M M M M B88+sB3Bi8888888895 B E B 8 M M B E N E+8 :+i ?+s M M

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=

'C B D D = BP8DRB300 Figure 1.

Reference Core Loading Pattern 11

23A4696 Rev. 0 I net??RON FLU (

1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLCW 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 13 0.0 300.0

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TIME (SECONOS) 1 LEVEL (INCH-REF-SEP.SKRT3

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TIME (SECONOS) l Figure _2.

Plant Response to Generator Load Rejection Without Bypass, E0011 12

i 23A4696 Rev. 0 1 NEUIRON FLUX 1 VESSEL PRESS RISE (PSI) 2 AYE SURFACE HEAT FLUX 2 RELIEF VALVE FLOW 3 CORi ULET FLOW 3 BYPiSS VALVE FLOW 13 0. 0 e eno r tu rr g -, eim,

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Plant Response to Loss of 100*F Feedwater Heating 13 r

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1 23A4696 Rev. O 150.0 1 NEUTRON FLUX 1 VESSEL PRESS RISE : PSI) i 2 AVE SURFACE HEAT l' LUX 2 SAFETY VALVE FLOW 3 CORE INLET FLOW 3 RELIEF VALVE FLOW 150.0

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TIME (SECONOS) 4 Figure 4.

Plant Response to Feedwater Controller Failure, EOC11 14

23A4696 R;v. 0 1 NEUTRON F.UX 1 VESSEL PRESS RISEfrSI) 2 AVE SURFA:E HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET FLOW 3 RELIEF VALVE FLCW d

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Plant Response to MSIV Closure (Flux Scram), EOC11 15/16

23A4696 Rsv. 0 APPENDIX A CONTROL ROD DROP ANALYSIS i

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 A-1.

Reference A-2 indicates that this change is acceptable to the NRC.

REFERENCES A-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.

A-2.

Letter, C. O. Thomas (NRC) to J. S. Charnley (GE), " Acceptance for Referencing of Licensing Topical Report Amendment 9 to NEDE-24011, Revision 6, 'GESTAR-II General Electric Standard Application for Reactor Fuel'", January 25, 1985.

)

17/18 z

23A4696 Rev. 0 APPENDIX B GETAB AND TRANSIENT ANALYSIS INITIAL CONDITIONS The values used in.the GETAB analysis for reactor core pressure and inlet enthalpy and in the transient analysis for rated steam flow are given in Table B-1.

.The following values are different from those reported in NEDE-24011-P-A-6-US, dated April 1983.

Table B-1 PLANT PARAMETER

' Parameter Analysis Value NEDE-740ll Value Reactor Core Pressure 1065 psia 1057 psia Inlet Enthalpy 526.0 Btu /lb 525.2 Btu /lb Ra.ed Steam Flow 7.99x10 lb/hr 7.94x106 + 0.2% lb/hr 6

Safety / Relief Valve (SRV)

Number of SRVs at:

Lowes t Setpoint Capacity (psig)

(lb/hr) 1095 791,000 0

4 1095 829,000 3

2 l

1125 791,000 3

0 a

19/20 t

1 23A4696 Rsv. 0 APPENDIX C STABILITY ANALYSIS According to Reference C-1, Millstone Unit 1 is exempt from the current requirement to submit a cycle specific stability analysis to the NRC.

REFERENCES C-1.

Letter, C. O. Thomas (NRC) to H. C. Pfefferien (GE), " Acceptance for Referencing of Licensing Topical Report NEDE-24011, Rev. 6, Amendment 8,

' Thermal Hydraulic Stability Amendment to GESTAR II'", April 24, 1985.

i l

21/22

23A4696 Rrv. 0 APPENDIX D FEEDWATER TEMPERATURE REDUCTION AT EOC11 Analyses were performed for end-of-cycle (EOC) 11 operation with the last-stage feedwater heaters valved out-of-service, in order to justify operation with feedwater temperature reduced by 75*F.

The prescurization events of Saction 12 were reanalyzed for operation at the reduced feedwater temperature. This appendix presents the results of these transient analyses.

The balance of the safety analysis required to justify operation at a reduced feedwater temperature (as defined in Reference D-1) will be provided by NUSCO.

REFERENCES:

D-1.

" General Electric Standard Application for Reactor Fuel",

NEDE-24011-P-A-6-US, dated April 1983.

D.1 CORE AVERAGE EXPOSURE Assumed reload core average exposure 18976 mwd /ST for Feedwater Temperature Reduction (FWTR) analysis (Extended EOC11)

D.2 RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS Peaking Factors Fuel Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt)

(1000 lb/hr)

MCPR Exposure: EOC11 to Extended EOC11 B P8x8R/ 1.20 1.67 1.40 1.051 5.640 99.3 1.39 P8x8R 23

)

23A4696 Rev. 0" D.3 CORE-WIDE TRANSIENT ANALYSIS RESULTS l

Exposure: EOC11 to Extended EOC11 ACPR Flux Q/A BP8x8R/

Transient

(% NBR)

(% NBR)

P8x8R Figure Load Rejection Without Bypass 515 125 0.32 D-1 Feedwater Controller Failure 121 114 0.12 D-2 D.4 CYCLE MCPR VALUES Exposure Range: EOC11 to Extended EOC11 Option A Option B BP8x8R/P8x8R BP8x8R/P8x8R Load Rejection Without Bypass 1.45 1.40 Feedwater Controller Failure 1.24 1.17 l

24

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23A4696 Rev. 0 1 NEUTRON FLU (

1 VESSEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CORE INLET TLOW 3 RELIEF VALVE FLOW 150.0 300.0 t ny==ss untuE eteu 100.0 200.0

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SCRA, 4 rnr.M REACTI, VI,TY 200.0 e errtuaTro r;cu 1.0 2

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Figure D-1.

Plant Response to Generator Load Rejection Without Bypass, FWIR 25 a

23A4696 Rev. 0,

150.0 1 NEUIRON F x 1 VESSEL PRESS RISE (PSI) 2 AT FLUX 2 SAFETY VALVE FLOW n! INLET FL W 3 RELIEF VALVE FLOW 150.0 C^*7

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Figure D-2.

Plant Response to Feedwater Controller Failure, FWTR 26

23A4696 Rev. 0 4

APPENDIX E l

FUEL BUNDLE DESCRIPTION The BP8DRB300 fuel bundle description will be provided in Amendment 13 of

" General Electric Standard Application for Reactor Fuel", NEDE-240ll-P-A-6, 1

dated April 1983. This information was previously provided in Reference E-1.

l

REFERENCE:

E-1.

Letter, W. G. Counsil (NUSCO) to D. M. Crutchfield (NRC), " Millstone Nuclear Power Station, Unit 1, Fuel Bundle Proprietary Information,"

March 27, 1984.

P k

27/28 (FINAL)

1 A

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GENER AL $ ELECTRIC 1

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