ML20028A280

Supplemental Reload Licensing Submittal for Hatch Nuclear Power Station Unit 2 Reload 2.

ML20028A280
Person / Time
Site:	Hatch
Issue date:	09/30/1982
From:	Engel R, Galer R, Hilf C GENERAL ELECTRIC CO.
To:
Shared Package
ML20028A276	List: ML20028A275 ML20028A277 ML20028A280
References
Y1003J01A32, Y1003J01A32-R01, Y1003J1A32, Y1003J1A32-R1, NUDOCS 8211180034
Download: ML20028A280 (30)

Text

.

Y1003J01A32 Revision 1 Class I September 1982 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR HATCH NUCLEAR PLANT UNIT 2, RELOAD 2 Prepared:

C. L. Hilf Verified: )

R. R.' Galer Approved:[ 1/[ f's -

R. E. Eng21, Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION

• GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENERAL h ELECTRIC l 1

8211180034 821111 l PDR ADOCK 05000366 P PDR

Y1003J01A32 Rev. 1 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 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 General Electric Company for nuclear fuel and related services for Hatch 2, dated October 25, 1967, and nothing contained in this document shall be con-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 suca use of such information.

ii

, . ~

Y1003J01A32 Rev. 1

1. PLANT-UNIQUE ITDiS (1.0)*

Safety Relief Valve Capacity: Appendix A Feedwater Temperatura Reduction: Appendix B Error in ODYN Code: Appendix C

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. PEFERENCE CORE LOADING PATTERN (3.3.1) f 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,gg 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-24011-P-A-2 and NEDO-24011-A-2, July 1981.

1

Y1003J01A32 . Rev. 1

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (Ak) pga (200C, Xenon Free)

I 660 0.048

6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 and 5.2)

(Loss of Feedwater Heating Event only) l 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 Scram Worth N/A ($) -46.31/-37.05

7. REIDAD-UNIQUE CETAB 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 1.20 1,47 1.40 1.051 6.258 114.0 1.29 8X8R 1.20 1.50 1.40 1.051 6.373 113.1 1.27

8. SELECTED MARCIN IMPROVEMENT OPTIONS (5.2.2)

Transient Recategorization: No Recirculation Pump Trip: Yes Rod Withdrawal Limiter: No Thermal Power Monitor: Yes Measured Scram Time: No Number of Exposure Points 1 l  !

l 1

• N = Nuclear Input Data.

A = Used in Transient Analysis.

2

Y1003J01A32 Rev. 1

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Flux Q/A bCPR Transient (% NBR) (% NBR) P8X8R 8x8R Figure Exposure: BOC 3 to EOC 3 Load Rejection without Bypass 518 122 0.23 0.20 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 316 122 0.20 0.18 4

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

TRANSIENT

SUMMARY

(5.2.1)

Limiting Rod Pattern: Figure 5 Includes 2.2% Power Spiking Penalty: Yes Rod Position MLRGR (kW/ft)

Rod Block (Feet bCPR 8X8R/

Reading Withdrawn) P8x8R/8x8R 7X7 8X8 P8X8R i

104 3.0 0.14/0.12 16.91 105 3.5 0.15/0.13 17.48 106 4.0 0.17/0.15 17.82 107 4.0 0.17/0.15 17.82 108 4.5 0.18/0.16 17.85 109 5.0 0.19/0.17 17.85 110 5.0 0.19/0.17 17.85 Setpoint Selected is: 107

11. CYCLE MCPR VALUES (5.2) l l Non-Pressurization Events I Exposure Range: BOC to EOC P8X8R 8X8R Loss of Feedwater Heater 1.21 1.21 Fuel Loaling Error 1.24 Rod Withdrawal Error 1.24 1.22 Pressurization Events Exposure Range: BOC 3 to EOC 3 Option A Option B P8X8R 8X8R P8X8R 8X8R Load Rejection without Bypass 1.36 1.33 1.26 1.23 Feedwater Controller Failure 1.33 1.30 1.29 1.27 l

3

Y1003J01A32

• Rev. 1 i

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3)

P, Pv Plant Transient (pstg) (psig) Response 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/ *0: 0.83 Channel Hydrodynamic Performance Decay Ratio, x2/ *0 Channel Type P8X8R/8X8R 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 LRCR 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:

• To be eliminated af ter approval of event classification.

4

Y1003J01A32 Rev. 1

16. LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (5.5.2)

FUEL TYPE: P8DRB283 Exposure MAPLHGR PCT Local Oxidation (mwd /T) (kW/ft) (OF) (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 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  ;

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Y1003J01A32 Rev.,1

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NOTES: 1. ROD PATTERN IS 1/4 CORE MIRROR SYMMETRIC.

2. NO. lNDICATES NUMBER OF NOTCHES WITHDRAWN OUT OF 48. BLANK IS A WITHDRAWN 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 1G03J01 A32 Figure 5. Limiting RWE Rod Pattern 10

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12

l

• Y1003J01A32 Rev. 1 l

• 1

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BOUND VALUE 280 cal /g COLD BOUND VALUE 2M cal /g HSB i -10 -

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Y 1004J01 A32 Figure 8. Fuel Doppler Coefficient in 1/ A t 13

Y1003J01A32 . Rev.'l 20.0 -

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Y 1003JC . A32 Figure 9. Accident Reactivity Shape Function Cold Startup 14

Y1003J01A32 Rev. 1 20.0 -

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Y 1003J01 A32 Figure 10. Accident Reactivity Shape Function Hot Startup 15

Y1003J01A32

• Rev. 1 40 -

35 -

30 -

3 . SCRAM FUNCTION W

a

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OUNDING VALUE 280 cal /g 0 i I t t 0 1 2 3 4 5 ELAPSED TIME (sec) v.1003 J01 A32 Figure 11. Scram Reactivity Function Cold Startup 16

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Y1003J01A32 Rev. 1 l

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BOUNDING VALUE 280 cal /g 0

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Y-1003J01 A32 Figure 12. Scram Reactivity Function Hot Startup 17/18

Y1003J01A32 Rev. 1 APPENDIX A Safety Relief Valve Capacity 91.4%

t l

l 19/20

- Y1003J01A32 R:v. 1 t

' APPENDIX B FEEDWATER TEMPERATURE REDUCTION (FWTR)

Additional analyses were performed assuming a reduction of 63*F in feedwater temperature in order to determine operating limits in the event of a loss of feedwater heating.

Table B-1 provides the results of the transient analysis performed, and Table B-2 provides the resulting Minimum Critical Power Ratio (MCPR) values.-

The feedwater controller failure transient is the limiting event with MCPR operating limits higher than the limits without Feedwater Temperature Reduc-tion (FWTR).

Results of the stability analysis are presented in Table B-3 and in Figure B-3.

The analysis of the core response to a control rod drop accident showed all parameters to be within the bounding limits.

The ACPR of the rod withdrawal error is increased under FWTR conditions due to the higher initial CPR required by the pressurization transients with FWTR. However, in no case is RWE the limiting event.

As indicated in Section 5 of Reference B-1, the additional fatigue stresses are small compared to the allowable fatigue damage.

The loss-of-coolant accident analysis is relatively insensitive to feedwater temperature changes and is affected only slightly by a 63*F decrease. No change in MAPLHGR limits is required.

The operating limits for operation with FWTR at 100% power and 100% flow also bound operation within the region of the power / flow map described in Reference B-2.

21

Y1003J01A32 Rrv.' 1 -

i REFERENCES B-1. " Safety Review of Hatch Nuclear Power Station Unit No. 2 at Core Flow j Conditions Above Rated Flow Throughout Cycle 2," General Electric Company, October 1981 (NEDO-24292, Revision 2).

B-2. " General Electric Boiling Water Reactor Load Line Limit Analysis for Edwin I. Hatch Nuclear Plant Unit 2," General Electric Company, October 1980 (NEDO-24295) .

Table B-1 TRANSIENT ANALYSIS RESULTS WITH FWTR f Flux Q/A ACPR l P8X8R 8X8R Figure Transient (% NBR) (% NBR)

Exposure: BOC 3 to EOC 3 Load Rejection without Bypass 483 123 0.21 0.18 B-1 Feedwater Controller Failure 363 129 0.24 0.22 B-2 l -

Table B-2 MCPR VALUES WITH FWTR Exposure: BOC 3 to E0C 3

, Option A Option B P8X8R 8X8R P8X8R 8X8R Load Rejection without Bypass 1.34 1.30 1.24 1.22 Feedwater Controller Failure 1.37 1.35 1.34 1.32 22

I Y1003J01A32 Rev. I f

l Table B-3 STABILITY ANALYSIS RESULTS WITH FWTR Rod Line Analyzed: Extrapolated Rod Block Line ,

Decay Ratio: Figure B-3 Reactor Core Stability Decay Ratio, X /X : 0.91 2 o Channel llydrodynamic Performance Decay Ratio, X /X 2

Channel Type P8X8R/8X8R 0.63 23

E I 1 NEUTRON FLilX 1 VESSEL PflES RISE (PSI)

F 2 AVE SURFfCE HEAT FLUX 2 SAFETT VRLVE FLOW 3 COPE INLET FLOW 300* 3 PELIEF V4tVcyFLCW ,

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N O

> 1 VOID REACTIVITY  %

1 LEVEL (INCH-REF-SEP-SKIRT w 2 VESSEL S1EAMFLOW 2 00PPLER FEACTIVITY 200. 3 TURBINE 5TEAMFL.0_W 3*

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ir Figure B-1. Plant Response to Generator Load Rejection without Bypass with FWTR g

I NEUTRON FLUX 1 v553EL pries RISE (PSI) . l

. '. 2 SAFETY V VE FLC4 2 AVE SURFFCE HEAT FLUX l 150. J f - .- 3 3 .T W 125. ]h[h, [-

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y TOTAL REFCTIVIIT FEEDWRTU FL0ri 2 ,

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TIME (SEC) TIME ISEC) x m

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Figure B-2. Plant Response to Feedwater Controller Failure with FWTR g

Y1003J01A32 . Rev. 1 A NA1 URAL CIRCU .ATION B 10: PERCENT R )0 LINE C UL1 . PERFORMA JCE LIMIT 1.00 A

o R .75 0

p

> .50 -

8

.25 0.00

0. 0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POWER Figure B-3. Reactor Core Decay Ratio with FWTR 26

Y1003J01A32 Rev. 1 APPENDIX C An error was discovered in the ODYN code used to analyze the pressurization transients, and the analyses were redone (C-1).

REFERENCE C-1. Letter, H. C. Pfefferlen (GE) to D. G. Eisenhut (NRC), " Correction of ODYN Errors," June 8, 1982.

27/28

- Y1003J01A32 Rev. 1

~

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 Coef ficient 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 B-1. Plant Response to Generator Load Rejection without Bypass with FWTR B-2. Plant Response to Feedwater Controller Failure with FWTR B-3. Reactor Core Decay Ratio with'FWTR 29/30 (FINAL)  !

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