ML19263D349

Suppl Reload Licensing Submittal for Unit 1,Reload 3.

ML19263D349
Person / Time
Site:	Hatch
Issue date:	01/31/1979
From:	Engel R, Galer R GENERAL ELECTRIC CO.
To:
Shared Package
ML19263D335	List: ML19263D334 ML19263D349
References
79NED253, NEDO-24175, NUDOCS 7903270483
Download: ML19263D349 (26)

Text

NED0-24175 79NED253 Class I January 1979 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR EDWIN I. HATCH NUCLEAR PLANT UNIT 1 RELOAD 3 Prepared: *

• R.R. Galer, Program Engineer Operating Licenses I Approveo:

R. E. Eigel, Manager Operating Licenses I NUCLE AR ENERGY PROJECTS DivlSION

• GENER AL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 951:5 GEN ER AL h ELECTRIC7303270gg3

• ~-

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 1.

The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or pro-vided 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 Edwin I. Hatch Nuclear-Plant Unit 1, dated May 2,1968, and nothin5 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 Company nor any of the contributors to this document makes any representation or warranty (express er ic: plied) 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 in formation .

9

NEDO-24175 . .

6. RELOAD-UNIOUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)

EOC4 Void coefficient N/A* (C/% Rgo) -7.415/-9.269 Void Fraction (%) 40.12 Doppler Coefficient N/A (c/&*F) -0.231/-0.219 Average Fuel Temperature (*F) 1397 Scram Worth N/A (S) -39.79/-33.83 Scram Reactivity vs Time Figure 2

7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 7x7 8x8/8x8R Exposure EOC4 EOC4 Peaking factors (local, 1.27/1.36/1.40 1.17/1.59/1.40 radial and axial)

R-Factor 1.081 1.051 Bundle Power (MWt) 5.798 6.768 Bundle Flow 124.59 113.73 (103 lb/hr)

Initial MCPR 1.20 1.20

8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

(1) Recirculation Pump Trip (RPT)

(2) Thermal Power Monitor (TPM)

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Py Power Flow & Q/A Psi SCPR Plant Transient Exposure (t) (%) (%) (%) (psig) (psia) M 8x8/8x8R Response TT w/o BP BOC-EOC4 104.1 100 168.0 102.0 1149 1180 0.06 0.10 Fictre 3 Loss of BOC-E004 104.1 100 118.1 116.9 1014 1066 0.13 0.14 Figure 4 100* FV Heating Feedwater 30C-EOC4 104.1 100 135.4 106.8 1127 1161 0.06 0.07 Figure 5 Cent, aller Failure

• N = Nuclear Input Data A = Used in Transient Analysis 2

NED0-24175

1. PLANT-UNIQUE ITEMS (1.0)*

Revisions to Initial Conditions or Inputs: Appendix A Fuel Loading Error LHGR: Appendix B Densification Power Spiking: Appendix C New Methods - Fuel Loading Error: Appendix D

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

Fuel Type Number Number Drilled Irradiated Initial Core Type 1 68 68 Initial Core Type 2 36 36 Initial Core Type 3 32 32 8DB250 (Reload 1) 92 92 8DRB265H (Reload 2) 168 168 New 8DRB265H 164 164 Total 560 560

3. RFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle exposure: 13,056 mwd /t Assumed reload cycle exposure: 15,740 mwd /t Core loading pattern: Figure 1.

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WOR 1h - NO VOIDS, 20 C (3.3.2.1.1 and 3.3.2.1.2)

BOC k ggg Uncontrolled 1.110 Fully Controlled 0.950 Strongest Control Rod Out 0.983 R, Maximum increase in Cold Core Reactivity with Exposure Into Cycle, ak 0.0036

5. STANDBY LIOUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin (ak) ppm (20*C. Menon Free) 600 0.036

• ( ) refers to areas of discussion in " Generic Reload Fuel Application,"

NEDE-?'Oll-P-A, Revision 0, August 1978.

1

NED0-24175 . .

13. STABILITY ANALYSIS RESULTS (5.4)

Decay Ratio: Figure 8 Reactor Core Stability:

Decay Ratio, x2 /*0 *

(105% Rod Line -

Natural Circulation Power)

Channel Hydrodynamic Performance Decay Ratio, x /

2 0 (105% Rod Lite - Natural Circulation Power) 8x8/8x8R channel 0.39 7x7 channel 0.23

14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)

P3ference: 1*

15. LOADING ERROR RESULTS (5.5.4)

Loading Error Description MCPR Rotated Bundle 1.11 (8x3R)

Mislocated Bundle > 1.07 (8x8R into 8x8)

16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 'l Scrar. Reac;!/ity Functions: Figures 12 and 13 Plant-Specific Analysis Results Parameter not bounded: Accident Reactivity (cold)

Resultant peak enthalpies: 197.35 cal /gm

• Reference 1: " Gene-lu Fuel Application", NEDE-240ll-P-A, Subsection 5.5.2, August 1978.

4

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

SUMMARY

(5.2.1)

Rod 0 (

Rod Block (Feet Limiting Reading

• Withd rawn) 7x7 8x8 8x8R 7x7 8x8 8x8R Rod Pattern 105%* 3.5 0.19 0.17 0.12 14.0 14.0 14.6 Figure 6 106 4.0 0.22 0.20 0.14 14.8 14.9 15.8 Figure 6 107 4.5 0.26 0.22 0.16 15.3 15.2 16.4 Figure 6 108 4.5 0.26 0.22 0.10 15.3 15.2 16.4 Figure 6 109 5.0 0.28 0.25 0.18 15.6 15.5 16.9 Figure 6 110 5.5 0.31 0.27 0.20 16.0 15.8 17.4 Figure 6

11. OPERATING MCPR LIMIT (5.2)

BOC4 - EOC4 (7x7 fuel) 1.26 (8x8 fuel) 1.24 (8x8R fuel) 1.21

12. OVERPRESSURI7.ATION ANALYSIS SLW!ARY (5.3)

Power Core Flow sl v Plant Transient (%) (%) (psig) (psig) Response MSIV Closure 104.1 100 1194 1232 Figure 7 (Flux Scram)

Indicates setpoint selected

NED0-24175 .

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NED0-24175 52 @g@@@@@@@@@

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2. No. indicate number of notches withdrawn out of 48. Blank is a withdrawn rod.

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NED0-24175

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NED0-24175 70 60 -

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NEDO-24175 -

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NEDO-24175 100 so -

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-l NED0-24175 APPENDIX A CHANGES IN PARAMETERS FROM NEDE-24011 GETAB Transient Analysis Initial Condition Parameters:

Reactor Pressure, psia 1035.0 Inlet Enthalpy, Btu /lb 523.7 Plant Operating / Input Parameters:

S/RV Capacity, % 85.9 19/20

. NED0-24175 APPENDIX B FUEL LOADING ERROR LHCR'S Mislocated Bundle LHGR 8x8R into I.C. 7x7 18.84 Rotated Bundle 8x8R 15.18 21/22

NED0-24175 APPENDIX C DENSIFICATION POWER SPIKING Reference C-1 documents the NRC Staff position that " . . . it (is) acceptable to remove the 8x8 and 8x8R spiking penalty factor from the plant Technical Specification for these operating BWR's for which it can be shown that the predicted worst case maximum transient LHGR's , when augmented by the ecwer spike pens . ty, do not violate the exposure-dependent safety limit LHGR's".

The Hatch-1 Reload-3 license submittal contains the required information to remove the power spiking penalty from the Hatch-1 Technical Specifications.

All calculated maximum transient LEGR's, including those for Rod Wi thdrawal Error and Fuel Loading Error, are less than the 1% plastic diametral strain LHGR limits, as given in Reference C-2, by more than the power spike penalty of 2.2%.

REFERENCES C-1 " Safety Evaluation of the GE Methods for the Consideration of Power Spiking Due to Densification Ef fects in BWR 8x8 Fuel Design and Per-fo rmance", Reactor Safety Branch, DOR, Shy 1978.

C-2 Licensing Topical Report, "GE Boiling Water Reactor, Generic Reload Fuel Application", NEDE-240ll-P-A, May 1977.

23/24

NED0-24175 APPENDIX D NE'4 BC?DLE LOADING ERROR EVENT A'?ALYSES PROCEDURES The bundle loading error analyses results presented in Section 15 in this supple-cent are based on new analyses procedures for both the rotated bundle and the mislocated bundle loading error events. The use of these nee analyses procedures is discussed belev.

D.1 NE'4 ANALYSIS PROCEDURE FOR THE ROTATED 3C?DLE LOADING ERROR EVENT The rotated bundle loading error event analysis results presented in this supple-cent are based on the new analysis procedure described and approved in Reference D-1. This new tethod of performing the analysis is based on a = ore accurate detailed analytical codel.

The principle difference between the previous analysis procedure and the new anal-ysis procedure is the =odeling of the water gap along the axial length of the bundle. The previous analysis used a unifor= water gap, whereas the new analysis utilizes a variable water gap which is core representative of the actual condi-tion, since the interfacing between the top guide and the fuel spacer buttons, caused by tisorientation, causes the bundle to lean. The effect of the variable water gap is to reduce the power peaking and the R-factor in the upper regions of the liciting fuel rod. This results in the calculation of a reduced CPR for the rota ced bundle. The calculation was perforced using the same analytical

=cdels as were pre iously used. The only change is in the simulation of the water gap, which more accurately represents the actual geometry.

The nutber presented in Section 15 represents the sini=un CPR during Cycle 4 of the cost limiting rotated bundle.

D.2 NE'4 ANALYSIS PROCEDURE FOR THE MISLOCATED SC?DLE LOADING ERROR EVENT Ibe mislocated bundle loading error event analyses results presented in this sup-plement are based on the new analysis procedure described in Reference D-1. This 25

NEDO-24175 new method of performing the analysis employs a statistically corrected Haling procedure and analyzes every bundle in the core.

The use of the statistically corrected Haling analyses procedure indicates that the minimum CPR for mislocated 8DR265 bundles in the core is greater than the safety limit (1.0 7) for all exposures throughout Cycle 4 RE FERENCES D-1 Safety Evaluation Report (le tter), D. G. Eisenhut (NRC) to R. E. Engel (GE),

MEN-200-78, dated May 8, 1978.

26