ML20072T049

Suppl Reload Licensing Submittal for Edwin I Hatch Nuclear Plant Unit 2,Reload 3 (Cycle 4)

ML20072T049
Person / Time
Site:	Hatch
Issue date:	01/31/1983
From:	Charnley J, Hilf G, Hill R GENERAL ELECTRIC CO.
To:
Shared Package
ML20072T041	List: ML20072T037 ML20072T045 ML20072T049
References
TAC-49989, TAC-51054, Y1003J01A57, Y1003J01A57-R00, Y1003J1A57, Y1003J1A57-R, NUDOCS 8304070422
Download: ML20072T049 (23)

Text

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V1003J01A57 CLASSI l JANUARY 1983 I I

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l SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR EDWIN 1. HATCH NUCLEAR PLANT UNIT 2, RELOAD 3 (CYCLE 4) l I

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Y1003J10A57 Revision 0 Class I January 1983 SUPPLEMENTAL RELOAD LICENSING SUBMITIAL FOR

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EDWIN I. BATCH NUCLEAR PLANT UNIT 2, RELOAD 3 (CYCLE 4)

Prepared *

. L. Hilf Verified

1. RM. Hill ' " /

Approved:

J.Mi. Charnley, Managlifr Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION + GENERAL ELECTRIC COMPANY SAN JOSE. CALIFORNIA 95125 GENER AL h ELECTRIC i

Y1003J01A57 Rav, 0 I

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by General Electric solely for Georgia Power Company ,

(GPC) for GPC's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending GPC's operating license of the Edwin I. Hatch Nuclear Plant, 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 Georgia Power Company and General Electric Company for nuclear fuel and related services for the nuclear system for Hatch 2 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 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 from such use of such information.

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Y1003J01A57 Rev. 0

1. PLANT UNIQUE ITEMS (1.0)*

GETAB Analysis Parameters Appendix A

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

Fuel Type Cycle Loaded Number Number Drilled Irradiated 8DRB221 (IC) 1 68 68 8DRB221 (IC) 1 92 92 P8DRB284LA 2 164 164 P8DRB283 3 120 120 New P8DRB26'H 4 116 116 Total' 560 560

3. REFERENCE CORE LOADING PATTERN (3.3.1)

., Nominal previous cycle core average exposure at end of cycle: 14015 mwd /ST Minimum previous cycle core average exposure at end of cycle from cold ~ shutdown considerations: 14015 mwd /ST Assumed reload cycle core average exposure at end of cycle: 15160 mwd /ST Core loading pattern: Figure 1

(

• ( ) refers to area of discussion in " General Electric Standard Application for Reactor Fuel," (NEDE-24011-P-A-4) January 1982.

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Y1003J01A57 Rev. 0

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20 DEGREE C (3.3.2.1.1 and 3.3.2.1.2)

Beginning of Cycle, k effective Uncontrolled 1.108 -

Fully Controlled 0.952 Strongest Control Rod Out 0.985 .

R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, Ak 0

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (aK) pgm (20 Degree C, Xenon Free) 660 0.050

6. REIDAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 and S.2.2)

(Loss of Feedwater Heating Event Only)

EOC 4 Void Fraction (%) 41.4 Average Fuel Temperature- 1301 (degree F)

Void Coefficient N/A* (cents /% Rg) - 8.28,

-10.35 Doppler Coefficient N/A (cents / degree F) -0.227/ -

-0.215

, Scram Worth N/A ($)* * ,

• N = Nuclear Input Data A = Used in Transient Analysis

• • Generic, exposure-independent values are used as indicated in NEDE-24011-P-A-4.

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Y1003J01A57 Rev. 0 4

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

Peaking Factors R- Pwr Bundle Flow Initial Fuel Design Local Radial Axial Factor (MWt) (1000 lb/hr) MCPR BOC 4 to EOC 4 P8x8R 1.20 1,48 1.40 1.051 6.294 113.7 1.28 8x8R 1.20 1.50 1.40 1.051 6.406 112.9 1.26

8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)

Transient Recategorization: No Recirculation Pump Trip: Yes Rod Withdrawal Limiter: No

'I Thermal Power Monitor: Yes Measured Scram Time: No Number of Exposure Points: 1

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1) 0 Flux Q/A Transient  % NBR  % NBR P8x8R 8x8R Figure Exposure: BOC 4 to EOC 4 Load Rejection w/o Bypass 510 122 0.22 0.19 2

. Exposure: -BOC 4 to EOC 4 Loss of Feedwater Heater 126 121 0.15 0.14 3 Exposure: BOC 4 to EOC 4 Feedwater Controller Failure 319 119 0.16 0.15 4 k

t 3

Y1003J01A57 Rev. 0

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

SUMMARY

(S.2.2.1)

Limiting Rod Pattern: Figure 5 Includes 2.2% Power Spiking Penalty: Yes MLHGR Pos t on

( *} -

0 Rod Block Feet 8x8R/

Reading Withdrawn) P8x8R 8x8R P8x8R t

104 3.5 0.08 0.08 17.1 105 3.5 0.08 0.08 17.1 3

106 4.0 0.09 0.09 17.4 107 4.5 0.10 0.10 17.5 108 5.0 0.11 0.11 17.5 109 6.5 0.12 0.12 17.5

- 110 8.0 0.14 0.14 17.5 Set Point Selected is: 110

11. CYCLE MCPR VALUES (S.2.2)

Nonpressurization Events Exposure Range: BOC 4 to EOC 4 P8x8R 8x8R Loss of Feedwater Heater 1.22 1.21 Fuel Loading Error 1.21 Rod Withdrawal Error 1.21 1.21 Pressurization Events Exposure Range: BOC 4 to EOC 4

• Option A Option B P8x8R 8x8R P8x8R 8x8R .

Load Rejection w/o Bypass 1.35 1.32 1.25 1.23 Feedwater Controller Failure 1.28 1.27 1.25 1.24 4

Y1003J01A57 Rev. 0

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(S.2.3) s1 y Plant Transient (psig) (psig) Response MSIV Closure 1189 1223 Figure 6 (Flux Scram)

13. STABILITY ANALYSIS RESULTS (S.2.4)

Rod Line Analyzed: Extrapolated Rod Block Line Decay Ratio: Figure 7 Reactor Core Stability Decay Ratio, X2/XO *0 Channel Hydrodynamic Performance Decay Ratio, X2/XO Channel Type 8x8R/P8x8R 0.65

14. LOADING ERROR RESULTS (S.2.5.4)

Variable Water Gap Misoriented Bundle Analysis: Yes Initial Resulting Event MCPR MCPR Misoriented 1.21 1.09

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

Bounding Analysis Results:

Doppler Reactivity Coefficient: Figure 8 Accident Reactivity Shape Functions: Figures 9 and 10 Scran Reactivity Functions: Figures 11 and 12 Plant Specific Analysis Results:

Parameter (s) Not Bounded, Cold: None Resultant Peak Enthalpy, Cold:

Parameter (s) Not Bounded, HSB: None Resultant Peak Enthalpy, HSB:

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Y1003J01A57 Rev. 0

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

MAPLHGR Table for Bundle Type 8DRB221 Exposure MAPLNGR PCT Local Oxidation (GWd/ST) (kW/ft) (Degree-F) (Fraction) 0.20 11.50 2162 0.031 1.0 11.50 2151 0.029 5.0 11.90 2187 0.033 10 12.10 2199 0.0 34 15 12.00 2200 0.034 20 11.90 2195 0.034 25 11. 30 2138 0.028 30 10.80 2061 0.022 35 10.20 1980 0.016 40 9.60 1897 0.012 MAPLHCR Table for Bundle Type P8DRB284LA Exposure MAPLHGR PCT .

Local Oxidation (GWd/ST) (kW/ft) (Dearee-F) (Fraction) 0.20 - 11.70 2189 0.034 1.0 11.80 2190 0.033 5.0 12.00 2198 0.033 to 12.10 2197 0.033 15 12.10 2198 0.033 20 12.00 2194 0.033 25 11.50 2130 0.026 30 10.80 2032 0.019 35 10.10 1934 0.014 40 9.50 1840 0.010 45 8.90 1758 0.007 MAPLHGR Table for Bundle Type: P8DRB283 Exposure MAPLHGR PCT Local Oxidation (GWd /ST) (kW/ft) (Dearee-F)- (Fraction) 0.20 11.30 2133 0.029 1.0 11.40 2134 0.028 5.0 11.90 2185 0.033 10 12.10 2195 0.033 15 12.10 2199 0.033 20 11.90 2184 ' O.032 25 11. 30 2112 0.025 30 11.10 2061 0.021 35 10.50 1981 0.030 40 9.80 1878 0.017 45 9.20 1788 0.008 MAPLHGR Table for Bundle Type: P8DRB265H Exposure MAPLHCR. PCT Local Oxidation (GWd /ST) (kW/ft) (Dearee-F) (Fraction) 0.20 11.50 2148 0.0 30 1.0 11.60 2157 0.030 .

5.0 11.90 2184 0.032 10 12.10 2198 0.033 15 12.10 2200 0.033 20 11.90 2188 0.032 25 11.30 2113 0.025 30 10.70 2027 0.019 35 10.20 1939 0.014 l 40 9.60 1840 0.009 45 8.90 1756 0.007 l l

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Y1003J01A57 Rev. O 2 6 10 14 18 22 26 l

51 8 47 44 38 38 43 6 18 6 39 38 38 44 35 8 6 0 31 44 38 44 38 27 6 18 12 NOTES: 1. ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC.

2. NUMBER INDICATES NUMBER OF NOTCJES WITHDRAWN OUT OF 48.

BLANK IS A WITHDRAWN ROD.

3. ERROR ROD IS (22,35).

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Y1003J01A57 Rev. O AF ATURAL C 'RCULATIO 1 81 05 PERCENT ROD LI VE CL LTIMATE STABILITY LINE 1.00 C  ::

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0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000~.C FUEL TEMPERATURE DEG C.

Figure 8. Fuel Doppler Coefficient in 1/4 Degree C 14

Y1003J01A57 Rev. 0 20.O A ACCIDENT FUNC TION 8 BOUNDING VALU E 280 CAL /G 17.5 15.0 1 m

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Y1003J01A57 Rav. 0 20.O A ACCIDENT FUNC TION 8 BOUNDING VALU E 280 CAL /G 17.5 .

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Y1003J01A57 Rev. 0 40.O A SCRAM FUNCTION 8 BOUNDI NG VALUE 280 CAL /G 35.0 .

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Y1003-101AS7 R1v. 0 80.O A SCRAM FUNCTION B BOUNDI NG VALUE 280 CAL /G 70.0 h E2 60.0 7

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Y1003J01A57 Rev. O APPENDIX A PLANT PARAMETERS GETAB Analysis Initial Conditions

• Core Power 2436 MWt 77.0 106 lb/hr Core Flow Reactor Pressure 1035 psia Inlet Enthalpy 526.9 Btu /lb Nonfuel Power Fraction 0.04

Safety Relief Valve Capacity 91.4%

• These values were inadvertently omitted from NEDE-24011-P-A-4. Tables S.2-4.1 and S.2-8.

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