ML19343C449

Supplemental Reload Licensing Submittal for Edwin I Hatch Nuclear Plant Unit 1,Reload 4.

ML19343C449
Person / Time
Site:	Hatch
Issue date:	07/31/1980
From:	Engel R, Kaminski G GENERAL ELECTRIC CO.
To:
Shared Package
ML19343C444	List: ML19343C443 ML19343C449
References
TAC-43589, TAC-48295, Y1003J01A13, Y1003J1A13, NUDOCS 8103240230
Download: ML19343C449 (25)

Text

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Y1003J01A13 JULY 1980 O

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SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR EDWIN 1. HATCH NUCLEAR PLANT UNIT 1 RELOAD 4 l '

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i 81032.4023o GENER AL h ELECTRIC

Y1003J01A13 July 1980 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR EDWIN I. HATCH NUCLEAR PLANT UNIT 1 RELOAD 4 Prepared: h G.I. Facinski Licensing Engineer Approved:k R. E. Engel, Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e eENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL $ ELECTRIC

Y1003J01A13 IMPORTANT NOTICE REGAPDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report axe prepared by General Electric solely for Georgia Power Company (GPC) for GPC's use uith the U.S. Naclear Regulatory Commission (USNRC) for amending GPC's opemting license of the Edvin I. Hatch Nuclear Plant Unit 1. The infomation contained in this report is believed by General Electric to be an accurate and true representation of the facte known, obtained or provided to General Electric at the time this report uas prepared.

The only undertakings of the Genent Electric Company respecting infomation in this document are contained in the contract between Georgia Pouer Company and General Electric Company for nuclear fuel and related services for the nuclear system for Edvin I.

Hatch Nuclear Plant Unit 1, dated May 2,1968, and nothing contained in this document shall be construed as changing said contract. The use of this ir.fomation except as defined by said contract, or for any purpose other than that for which it is intended, is not authorized; and uith respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any representation or a1rranty (express or implied) as to the completenese, accuracy or usefulnese of the information contained in this document or that auch use of such infomation may not infringe privately ovned rights; nor do they assume any responsibility for liability or damage of any kind which may result from auch use of such infomation.

Y1003J01A13

1. PLANT-UNIOUE ITEMS (1.0)*

Items different or not included in Reference 1**

Item 9. Transient Analysis Results: Appendix A Item 10. Rod Withdrawal Error: Appendix B ltem 12. Overpressurization Analysis: Appendix C Item 15. Loading Error: Appendix D

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

Fuel Tvpe Number Number Drilled Irradiated Initial Core Initial Type 1 60 60 Initial Core Initial Type 2 . 32 32 Reload 2 8DRB265H 168 168 Reload 3 8DRB265H 164 164 New Reload 4 P8DRB265H 136 136 Total 560 560

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle exposure: 15,695 mwd /t Assumed reload cycle exposure: 16,044 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)

BOC k,gg Uncontrolled 1.117 Fully Controlled 0.960 Strongest Control Rod Out 0.989 R, Max Increase in Cold Core Reactivity 0.0 with Exposure Into Cycle, Ak

• ( ) refers to areas of discussion in Reference 1.

• • Reference 1: " General Electric Boiling Water Reactor Geretic Reload Fuel Application," July 1979 (NEDE-24011-P-A-1).

1

Y1003J01A13

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3) ppm Shutdown Margin (Ak)

(20*C, Xenon Free) 600 0.033

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

EOC 5 Void Coefficient N/A*(c/% Rg) -6.97/-8.71 Void Fraction (%) 40.12 Doppler Coefficient N/A (c/*F) -0.216/-0.206 Average Fuel Temperature (*F) 1397 Scram Worth N/A ($) -38.85/-31.08 Scram Reactivity vs Time Figure 2-

'7. RELOAD-UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)

EOC 5 Exposure 7x7 8x8R P8x8R Peaking Factors 1.24 1.20 1.20 (local, radial 1.33 1.60 1.60 and axial) .1.40 1.40 1.40 R Factor ~1.100 1.052- 1.052 Bundle Power (MWt) 5.651 6.815 6.828 Bundle Flow'(103 lb/hr) !125.5: '112.2 113.0 Initial MCPR -1.19 1.21 1.21 I ,

i

• N = Nuclear Input Data A =_Used in Transient Analysis e

2

Y1003J01A13

8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

Recirculation Pump Trip Thermal Power Monitor Bundle Leading Error (New Method)

9. CORE WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Power Flow y

"' Plant

$ Q/A et Transient Exposure (1) (t) (2) (1) (psig) (psig) 7x7 8x8R P8x8R Response

!. css of 100'r -

104.1 100 123 122 1015 1066 0.12 0.14 0.14 Figure 3 Feedwater Heating

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

TRANSIENT

SUMMARY

(5.2.1)

Rod Block Rod Position " Limiting Reading (%) (Feet Withdrawn) 8x8R*/P8x8R 8x8R/P8x8R Rod Pattern 104 3.0 0.10 16.0 Figure 4 105** 3.5 0.11 16.7 Figure 4 106 4.0 0.13 17.1 Figure 4 107 4.5 0.14 17.2 Figure 4 108 4.5 0.14 17.2 Figure 4 109 5.0 0.15 17.2 Figure 4 110 5.5 0.16 17s2 Figure 4

• There are no 8x8 in core and no 7x7 in core interior - only on core periphery.

• • Rod Block Monitor Setpoint.

3

T1003J01A13

11. OPERATING MCPR LIMIT (5.2)

BOC5 to E005 1.21 (P8x8R fuel) 1.21 (8x8R fuel) 1.23 (7x7 fuel)

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3)

13. STABILITY ANALYSIS RESULTS (5.4)

Decay Ratio: Figure 5 Reactor Core Stability:

Decay Ratio, x2 /*0 0.72-(105% Rod Line - Natural Circulation Power).

~ Channel Hydrodynamic Performance Decay Ratio, x2/x0 (105% Rod Line - Natural Circulation Power)

P8x8R channel 0.29 8x8R channel 0.29 7x7 channel 0.23 1

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

Reference 2 Reference 2: " Loss-af-Coolant Accident Analysis for Edwin I. Hatch Nuclear Plant Unit 1", NEDO-24086, December 1977.

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Y1003J01A13

15. LOADING ERROR RESULTS (5.5.4)

Limiting Event: Rotated Bundle (8x8R)

MCPR: 1.07

16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Doppler Reactivity Coefficient: Figure 6 Accident Reactivity Shape Functions: Figures 7 and 8 Scram Reactivity Functions: Figures 9 and 10 All values are within bounding limits.

5

Y1003J01A13

eMMMMMs

sMMMMMMMMEs

MMMMMMMMMMM

eMMMMMMMMMMMs CMMMMMMMMMMMMM EMMMMMMMMMMMMM CMMMMMMMMMMMMM CMMMMMMMMMMMMM IMMMMMMMMMMMMM

"MMMMMMMMMMM" ll MMMMMMMMMMM

*HMMMMMMMM*

"MMMMM"

!IIililiIi 1 35 7 9111315171921232527293133353739414345474951 FUEL TYPE A = 7D234, INITIAL CORE TYPE 1 D = 8DRB265H, RELOAD 3 B = 7D234, INITIAL CORE TYPE 2 E = P8DRB265H, RELOAD 4 C = 8DRB265H, RELOAD 2 Figure 1. Reference Core Loading Pattere 6-

Y1003J01A13 100 ~ 45 C:>TPO. ROO DRIVE )s T!*I 5:A:= AE CTiv1TT VS T!*E C-679 CR0 IN FE9;ENT I-NNIN:. SCM" C 7.I IN t-si 93- 2-s:aA- Cts.1 usE Ik u.Tsts -40 80- __

-:a 70- 878C"D IN PERCENT

~

_ NOMINAL SCRAM CURVE IN {-8) d c

60- SCRAu Curve USED IN ANALYSl$

1 w >-

E -25 t z >

E 50- E'i

. c u

8

c. -20 C 40- -

-15 30-

~ 10' 20-10- -5 0 'L , , , ,0 0 1 2 3 4 TIME (SECONDS) rigure 2. Scram Reactivity and Control Rod Drive Specification,.EOC5 7

.UTRON FLUX 1 VESSEL PFES RISE (PSI) 2 A E SURFFCE KAT FLUX 2 RELIEF VfLVE FLOM 150* 3 C T Iq r FLOW 125* 3 DTfH55 VfLVE ft0H_

4 CLhE INLE T Sull 4 5 5 1, D V

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0. 40. 80. 120. 160. O. 40. 80. 120. 160.

TIE (SEC) TIME (SEC1 d

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c= 8 I LEVELLIPKH-REF-SEP-5KIRT 2 VESSEL 51EeNFLOW I v4 RETETIVITT 2D ER FEFCTIVITT ys 3 Tt0BINE TEAMFLOW

• 3V RFFCTIVITY w 4 I LLU@T FLOW 4 I J'4 Fd iCTI C 5

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D. = = * -2. * - $ *

0. 40. 80 20. 160. O. 40. g a 120. 160.

Figure 3. Plant Response to Loss of 100*F Feedwater lleating

Y1003J01A13 51 6 6 47 36 42 42 42 36 43 6 6 8 8 6 6 39 42 38 42 38 42 35 6 8 8 8 8 6 31 42 42 46 42 46 42 42 27 8 8 0 0 8 8 23 42 42 46 42 46 42 42 19 6 8 8 8 8 6 15 42 38 42 38 42 11 6 6 8 8 6 6 7 36 42 42 42 36 3 6 6 2 6 10 14 18 22 26 30 34 38 42 46 50 NOTES:

1. Nt.=ber indicates nu=ber of notches withdrawn out of 48. Blank is a withdrawn Rod.

2. Error Rod is (22,27) .

Figure 4. Limiting RWE Rod Pattern 9

Y1003J01A13 l 1

1.2 -

1.0 -- ----- -- ULT 6M ATE PER FORMANCE LIMIT 0.8 -

NATURAL CIRCULATION

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P 0.6 -

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0.4 -

0.2 - 105% ROD LINE 0 I IN O 20 40 60 80 100 PERCENT POWER Figure 5. Decay Ratio 10

._- . _ .- - _ - _ _ _ _ - _ . .. __ . . . ~ _ _ _ - - - . . - -

Y1003J01A13 0

IN ./ DELTA DEGREES C (MILL 1 0NTHS)

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+ BCUNDING VALUE FOR 2 0 CAL /G. COL 3

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@ X CALCULATED VALUE - C LD

. CALCULATc VALUE - H 9

-30 .

-35 ,

0 ' 00 4 800 1200 1600 2000 24:0 FUEL TEMP. DEG C Figure 6. Doppler Reactivity Coefficient 11

J Y1003J01A13 i

l COLD STAR rUP. BETA 0.005 PL 1.4 4

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Y1003J01A13 14 HOT STAR rUP. BETA 0.005 PL 1.5 12 en r

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Figure 9. Scram Reactivity Function, Cold Startup 14

Y1003J01A13

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+ BOUNDING VALUE FOR 21 s'0 CAL /G u CALCULATED VALUE en {

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--= - -.

1 Y1003J01A13 APPENDIX A CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

General Electric's one-dimensional core transient model ODYN computer code has been used for pressurization transient analysis (Reference A.1).

Power Flow a Q/A si y t.CPR Plant Transient Exposure (2) (2) (%) (1) (psig) (psig) 7x7 8x8R P8x8R Response Turbine Trip EOC5 104.1 100 360 114 1168 1188 0.08 0.11 0.12 Figure A-1 Without Bypass Feedwater EOC5 104.1 100 147 112 1137 1161 0.08 0.11 0.11 Figure A-2 Controller Failure REFERENCE A.1 NEDE-24154P, Volumes 1, 2, and 3, "One-Dicensional Core Transient Model", October 1978.

-A-1

I W UTRON dLUX l VESSEL IVES MISE IPS11 2 AVE SLMNE WRT FLUX 2 SoFE TT YrLvt FLOW 150. U O T INLt i FLCW 300. 3JttJfr y (OW 4 4 OffYlSS VI LVL LOW _

, 5 5 6

100. -

1 200, E

g M 1 1 e-

@ 50. 100, ,

9 O. 8- - L_ ' -

c. n ... . 2u 2u  : u .,

O. 1. 2 3. 4. O. 1. 2. 3. 4.

TIE (SECI TIE ISECl d

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b I LEVEttIM -REF-SEP-SMIRT I VO!D N ACTIVITT E 2 VESSEL S1EAMFLOW 7 00PPLf R rEncilv!TT >

200* 3 TUfoIE $ ifAMFLOW 4 iLLUHATET TIF~

g* M' 3 SCRAN RErCT!v!TT 4 TOTAL 7 UCTTVI C e-gg- i11 # _ a

0. I t

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L ._

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= 2",

.l.. .. ...... . .,. .........

O. 4. 2. 3. 4. O. 1. 2. 3. 4.

TIME ISECI 11MC iSECI Figure A-1. Plant Response to Turbine Trip Without Bypass, Trip Scram, EOCS

I WUTR(W dAJY g wESSEL FES CIT IPSil 2 RVE Sta rfE K RT Fltm 2 58W F TY VCLVE FLCW l'io. 3 COUNifp Ftru 125*

3 gr e S F(N_

$ $UO 4 lif fRMt L%t Obd

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100

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TIE ISECl TIE ISEC1 Y

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5- l VOID REACTivfTT E

- 2I WESSELLEVEll!%H-PEF-SEP-SMIRT 5 E8*FLtu 2 OOPPLER FEnCTIVITT -

w 3 TifnHIPT 9 TEswt rm 3 scnnM RErrigvgrY l'iO* 4 tIf.UETU TL UC ~

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=

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-0. 03.5 7. 10.5 14. -0. 03.5 7. 10.5 14.

TIE 15Ecl TIE 15ECl Figure A-2. Plant Response to Feedwater Controller Failure, Maximum llemand, EOCS

J Y1003J01A13 APPENDIX B DENSIFICATION POWER SPIKING Refsrence B-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 those operating BWR's for which it can be shown that the predicted worst case maximum transient LHGR's, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHGR's".

The Hatch-1 Reload-4 submittal contains the required information to demon-strate that the stated criterion is met for Hatch-1, Reload-4, Section 10 (Rod Withdrawal Error) and Appendix D (Linear Heat Generation Rate for Bundle Loading Error) include the densification effect in the calculated LEGR of the 8x8 fuels.

REFERENCE B.1 " Safety Evaluation of the General Electric Methods for the Consideration of Power Spiking Due to Densification Effects in BWR 8x8 Fuel Design and Performance," Reactor Safety Branch, DOR, May 1978.

3-1/B s

Y1003J01A13 APPENDIX C OVERPRESSURIZATION ANALYSIS SLHMARY (5.3)

I i

General Electric's one-dimensional core transient model ODYN computer code has been used for the overpressurization transient analysis (Reference A.1).

Power Core Flow st v Plant Transient (%) (%) (psig) (psig) Response MSIV Closure 104.1 100 1199 1232 Figure C-1 (Flux Scram)

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Y1003J01A13 APPENDIX D LOADING ERROR RESULTS (5.5.4)

Linear Heat Generation Rate (kW/ft): 16.8 D-1/D-2