ML20042A882
| ML20042A882 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/11/1982 |
| From: | Engel R, Galer R, Hilf C GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20042A874 | List: |
| References | |
| TAC-48110, Y1003J01A32, Y1003J1A32, NUDOCS 8203240232 | |
| Download: ML20042A882 (24) | |
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CLASSI DECEMBER 1981
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i SUPPLEMENTAL RELOAD l LICENSING SUBMITTAL FOR HATCH NUCLEAR POWER STATION UNIT 2 RELOAD 2
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GEN ER AL h ELECTRIC
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l Y1003J01A32 Revision 0 Class I December 1981
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SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR HATCH NUCLEAR PLANT UNIT 2, RELOAD 2 l
Prepared: U~A ~
C. L. Hilf Verified: dy R. R.' Galer Approved:[ ~2-/[ t-R. E. En g1, Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION
- GENER AL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENER AL h ELECTRIC i
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Y1003J01A32 Rev. O IMPORTANT NOTICE REGARDING CONTENTS OF TilIS 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 Hatch 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 i General Electric Company for nuclear fuel and related services for flatch 2, dated October 25, 1967, and nothing contained in this document shall be con- l strued as changing said contract. The use of this information except as de-fined 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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Y1003J01A32 -
Rev. O l
- 1. PLANT-UNIQUE ITEMS (1.0)*
Safety Relief Valve Capacity: Appendix A
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 and 4.0)
Fuel Cycle Type Loaded Number Number Drilled Irradiated 8DRB221(IC) 1 76 76 8DRB221(IC) 1 200 200 P8DRB284LA 2 164 164 New P8DRB283 3 120 120 Total 560 560
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end of cycle: 12047 mwd /t Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 12046 mwd /t Assumed reload cycle core average exposure at end of cycle: 14,621 mwd /t 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)
Minimum Shutdown Margin, BOC k eff Uncontrolled 1.112 Fully Controlled 0.955 Strongest Control Rod Out 0.986 R, Maximum Increase in Cold Core Reactivity with Exposure Into Cycle, ok 0.000
- ( ) Refers to area of discussion in " Generic Reload Fuel Application," NEDE- ,
240ll-P-A-2 and NEDO-240ll-A-2, July 1981. '
1 i
1 i
Y1003J01A32 -
Rev. 0
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) 4 l ppe (20 C, Xenon Free) 660 0.048
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 and 5.2)
(Loss of Feedwater Heating Event Only)
EOC-3 Void Fraction (%) 41.4 Average Fuel Temperature (OF) 1315.0 Void Coefficient N/A* (C/% Rg) -8.08/-10.10 Doppler Coefficient N/A (C/ F) -0.228/-0.217 l
-46.31/-37.05 Scram Worth N/A ($)
- 7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)
Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR BOC 3 to EOC 3 P8X8R I.20 1.48 1.40 1.051 6.298 113.7 1.28 q 8X8R 1.20 1.50 1.40 1.051 6.408 112.9 1.26
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2 2 Transient Recategorization: No Recirculation Pump Trip: Yes l Rod Withdrawal Limitier: No f Thermal Power Monitor: Yes Measured Scram Time: No ,
Number of Exposure Points 1
- N = Nuclear Input Data.
A = Used in Transient Analysis.
2
i l .
Y1003J01A32 ,
Rev. 0
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Flux Q/A 6CPR Transient (% NBR) (% NBR) P8X8R 8x8R Figure Exposure: BOC 3 to EOC 3 Load Rejection without Bypass 506 122 0.22 0.19 2 Exposure: BOC to EOC Loss of Feedwater 126 121 0.14 0.14 3 Heater Exposure: BOC 3 to EOC 3 Feedwater Controller Failure 295 121 0.18 0.17 4
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(5.2.1)
Limiting Rod Pattern: Figure 5 Includes 2.2% Power Spiking Penalty: Yes Rod Position MLHGR (kW/ft)
Rod Block (Feet dCPR 8X8R/
Reading Withdrawn) P8x8R/8x8R 7X7 8X8 P8X8R 104 3.0 0.12 0.11 16.91 105 3.5 0.13 0.12 17.48 106 4.0 0.15 0.14 17.82 107 4.0 0.15 0.14 17.82 108 4.5 0.16 0.15 17.85 109 5.0 0.17 0.16 17.85 110 5.0 0.17 0.16 17.85 Setpoint Selected is: 107
- 11. CYCLE MCPR VALUES (5.2)
Non-Pressurization Events Exposure Range: BOC to EOC P8X8R 8X8R Loss of Feedwater Heater 1.21 1.21 Fuel Loading Error 1.24 Rod Withdrawal Error 1.22 1.21 Pressurization Events Exposure Range: BOC 3 to EOC 3 Option A Option B P8X8R 8X8R P8X8R 8X8R Load Rejection without Bypass 1.35 1.32 1.25 1.23 Feedwater Controller Failure 1.30 1.29 1.27 1.26 3
\ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
Y1003J01A32 ,
Rev. O J
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
P Pv Plant g
Transient (psig) (psig) Response l
MSIV Closure 1205 1226 Figure 6 (Flux Scram)
- 13. STABILITY ANALYSIS RESULTS (5.4)
Rod Line Analyzed: Extrapolated Rod Block Line Decay Ratio: Figure 7 Reactor Core Stability Decay Ratio, x2/ x0: 0.83 Channel llydrodynamic Performance Decay Ratio, x2/ *0 Channel Type P8X8R 0.64
- 14. LOADING ERROR RESULTS (5.5.4)
Variable Water Cap Misoriented Bundle Analysis: Yes Includes 2.2% Power Spiking Penalty: Yes Initial Resulting Resulting*
Event MCPR MCPR LHCR Misoriented 1.22 1.07 17.66
- 15. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Bounding Analysis Results:
Doppler Reactivity Coef ficient: Figure 8 Accident Reactivity Shape Functions: Figures 9 and 10 Scram 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:
l
- To be eliminated after approval of event classification. i 4 1
Y1003J01A32 ,
Rev. 0
- 16. LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (5.5.2)
FUEL TYPE: P8DRB283 Exposure MAPLHGR PCT Local Oxidation (mwd /T) (kW/ft) ( F) (Fraction) 200 11.30 2133 0.029 1000 11.40 2134 0.028 5000 11.90 2185 0.033 10000 12.10 2195 0.033 15000 12.10 2199 0.033 20000 11.90 2184 0.032 f 25000 11.30 2112 0.025 30000 11.10 2061 0.021 35000 10.50 1981 0.030 40000 9.80 1878 0.017 5
Y1003J01A32 . Rev. 0
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. . j Y1003J01A32 ,
Rsv. 0 NOTES: 1. ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC.
- 2. NO. INDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDR AWN ROD.
- 3. ERROR ROD IS (26,31).
2 6 10 14 18 22 26 51 40 47 6 10 10 43 36 40 40 39 10 10 10 35 40 40 40 31 6 10 10 0 27 40 40 40 Y 1003J01 A32 Figure 5. Limiting RWE Rod Pattern 10
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Y1003J01A32 Rev. 0 1.25 ULTIMATE STABILITY LINE 1.00 N ATUR AL CIRCULATION 0.75 -
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8 0.50 -
105 PERCENT ROD LINE 0.25 -
0 ' ' ' ' I O 20 40 60 80 100 120 PERCENT POWER Y 1003J01 A32 Figure 7. Reactor Core Decay Ratio 12 L__
Rev. O Y1003J01A32 .
I i
0 -
-5 -
BOUND VALUE 280 cal /g COLD BOUND VALUE 280 cal /g HSB CALCULATED V ALUE - HSB CALCULATED VALUE - COLO 2
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Figure 8. Fuel Doppler Coefficient in 1/ A t 13
Y1003J01A32 Rsv. 0 20.0 -
17.5 -
15,0 -
BOUNDING VALUE 280 cal /g 12.5 -
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Y 1003JOI A32 Figure 9. Accident Reactivity Shape Function Cold Startup l
14
Y1003J01A32 ,
Rev. O I
20 0 -
17.5 -
15.0 -
BOUNDING VALUE 280 cal /g 12.5 -
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I-E ACCIDENT FUNCTION 7.5 -
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Figure 10. Accident Reactivity Shape Function Hot Startup 15 t -- - - - - - - -
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Y1003J01A32 Rev. 0 40 -
i 35 -
\
30 -
2 N
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- 20 -
Y> ,
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15 -
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t-jj/ j /' . bot >NCING VALUE 280 cal /g t
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. dLAPSED it1AE (sec)
Y 1003J01 A32 Figure 11. Scram Reactivity Function Cold Startup x
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. V1003J01A32 . Rev. 0
(
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BOUNDING VALUE 280 cal /g
' ' ' ' I 0
O 1 2 3 4 5 6 ELAPSED TIME (sec)
Figure 12. Scram Reactivity Function Hot Startup 17/18
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. Y1003J01A32 ,
R:.v. O APPENDIX A Safety Relief Valve Capacity 91.4%
4 19
Y1003J01A32 Rav. O LIST OF FIGURES
- 1. Reference Core Loading Pattern
- 2. Plant Response to Limiting Power and Pressure Increase Event
- 3. Plant Response to Limiting Coolant Temperature Decrease Event
- 4. Plant Response to Feedwater Controller Failure
- 5. Limiting Rod Withdrawal Error Rod Pattern
- 6. Plant Response to Overpressurization Event
- 7. Reactor Core Decay Ratio
- 8. Doppler Reactivity Coefficient Comparison for RDA
- 9. RDA Reactivity Shape Function at 20 deg. C
- 10. RDA Reactivity Shape Function at 286 deg. C
- 11. RDA Scram Reactivity Function at 20 deg. C
- 12. RDA Scram Reactivity Function at 286 deg. C l
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