ML20008D864

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Supplemental Reload Licensing Submittal for Hatch Nuclear Power Station Unit 2,Reload 1.
ML20008D864
Person / Time
Site: Hatch Southern Nuclear icon.png
Issue date: 06/30/1980
From: Engel R, Zanardi G
GENERAL ELECTRIC CO.
To:
Shared Package
ML20008D862 List:
References
Y1003J01A10, Y1003J1A10, NUDOCS 8010230376
Download: ML20008D864 (14)
v • d • e


Text

_ _ _ _ _ _ _ _ _ _ _ .- . -. --_ _

t Y1003J01A10 i

l June 1980 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR HATCH NUCLEAR POWER STATION UNIT 2 RELOAD 1 Prepared: ,

[

G. Zanardi Reload Fuel Licensing Approved:

l R. E. Engel a

Manager, Reload Fuel Licensing l

NUCLEAR POWER SYSTZMS DIVISION e GENER ' t. ELECTRIC COMPANY SAN JOSE, CALIFORNI A 951a5 GENER AL $ ELECTRIC se tevru. - . - _. _

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Y1003J01A10

'IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY I

j This report was prepared by General Electric solely for Georgia Power Company (GPC) for GPC's use uith the U.S. Nuclear Regulatory Commission l

'(USNRC) for amending GPC's operating ~ license of the Edvin I. Hatch

'e > lear Plant Unit 2. The infomation 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.

1 The only undertakings of the General. Electric Company respecting infor-mation 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 Edvin I. Hatch Nuclear Plant Unit 2, dated October 25, l967, and nothing contained in this document shall be con-

} strued as changing said contract. The use of this infomation 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 us_e, neither General Electric Company nor any of the con-tributors to this document makes any representation or varranty (express or implied) as to the comp; steness, accuracy or usefulness of the infor-

.mation contained in this document or that such use of such infomation I may not infringe privately owned rights; nor do they assume any respon-sibility for liability or damage of any kind which may result frcm such use of such information. l l

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Y1003J01A10

1. PLANT-UNIQUE ITEMS (1.0)*

Items different from or not included in Reference 1** - See Appendix A.

Item 2. Reload fuel bundles Appendix B Item 9. Transient analysis results Appendix C Item 10. Rod withdrawal error Appendix D Item 12. Overpressurization analysis Appendix E Item 15. Loading error Appendix F

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

Fuel Type Number Number Drilled Irradiated 8DRB176 Initial Core 84 84 Irradiated 8DRB219 Initial Core 3 12 312 New P8DRB284LA 164 164 Total 560 560

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle exposure: 8,468 mwd /t.

Assumed- reload cycle exposure: 12,150 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

eff Uncontrolled 1.109 Fully Controlled 0.954 Strongest Control Rod Out 0.983 R, Maximum Increase in Celd Core Reactivity 0.0 with Exposure Into Cycle, Ak l

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

, ** Reference 1: " Gen *ric Reload Fuel Application," NEDE-24011-P-A-1, July 1979.

1

Y1003J01A10 i

5. STANDBY LIQUD CONTROL SYSTEM SifUTDOWN CAPABILITY (3.3.2.I.3)

Shutdown Margin (Ak)

Iqmt (20*C, Xenon Free) 660 0.046

6. RFLOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3. 3.2.1. 5 AND 5.2) 5 EOC Void Coefficient N/A* (c/% Rg) -7.68/-10.00 i

Void Fraction (%) 41.82 Doppler Coefficient N/A (c/*F) -0.225/-0.214 Average Fuel Temperature (*F) 1357 Scrc, Worth N/A ($) -38.36/-30.69 Scram Reactivity v3 Time Figure 2

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

EOC 8x8R P8x8R Peaking. factors (local, radial and axial) 1.20 1.20 1.52 1.52 1.40 1.40 R-Factor 1.051 1.051 Bundle Power (MWt) 6.489 6.472 3

Bundle Flow (10 lb/hr) 112.14 112.35 Iaitial MCPR 1.24 1.24 1

4

8. GELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)  !

)

1 Thermal Power Monitor .

Recirculation Pump Trip

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

.Y1003J01A10 t

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

P P ACPR Power Flow $ Q/A s1 v Plant Transient Exposure (%) (%) (%) (Z) (psig) (psig) 8x8R P8x8R Response Loss of 100*F -

104.2 100 124.7 121.7 1016.1 1066 0.14 0.14 Figure 3 i Feedwater Heating f

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 3.5 0.12 15.02 Figure 4 105 4.0 0.14 15.33 Figure 4 106 4.5 0.15 , 15.33 Figure 4

  • 107 4.5 0.15 15.33 Figure.4 108 5.0 0.17 15.33 Figure 4 109 6.0 0.19 15.33 Figure 4 110 7.0 0.21 15.33 Figure 4 i

l 11. OPERATING MCPR LIMIT (5.2)

I l

8x8R/P8x8R BOC to EOC 1.24

  • Indicates setpoint selected.
    • Including 2 2% penalty due to fuel densification.

3

l Y1003J01A10 l

12. OVERPRESSURIZATION ANALYSIS

SUMMARY

(5.3) l

13. STABILITY ANALYSIS RESULTS (5.4)

Decay Ratio: Figure 5 Reactor Core Stability: 0.74 i

Decay Ratio, x /x0 (105% Rod Line - l Natural Circul t on Power)

Channel Hydrodynamic Performance ,

Decay Ratio, x2/X0 (100% Rod 4

Line - Natural Circulation Power) 8x8R/P8x8R channel 0.49

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

Fuel Type: P8DRB284LA i i

Exposure MAPLHGR PCT Oxidation (mwd /t) (kW/ft) (*F) Fraction j

200 11.7 2189 0.034 1,000 11.8 2190 0.033 5,000 12.0 2198 0.033 1

10,000 12.1 2197 0.033 '

15,000 12.1 2198 0.033 l 20,000 12.0 2194 0.033 l 25,000 11.5 2130 0.026 30,000 10.8 2032 0.019

15. LOADING ERROR RESULTS (5.5.4)

Limiting Event: Rotated Bundle P8DRB284LA MCPR: 1.07 4

Y1003J01A10

16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1) l j Parameter not bounded: Accident reactivity shape function, cold startup.

Plant-Specific Analysis Results i l Resultant peak enthalpies (cal /g): 244.7 i i

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4 1

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Y1003J01A10 1

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a 8 8'8 8 s B8 mi8 s @ @  ;

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88888JO5588E8888888888 42 -

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= a 8 e 8 8 8 8 8'a B s B 8 8 8 m 8 9 8 ss m8 8 i

=-8E B 8 @ 8 8 8 85 '"3 DDE 8 8 8 8 8 8 8 9 8 8 8E u -8'8 8 9 8 8 8 8 8'@ @bd 838 8 8 9 8 8 8 8 8 9 8 6'8 i

l 32-gggggggggggggggggggggggggs

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88&808858@88 a2 8888888888 IIIIIIIIII FUEL TYPE A = INITIAL CORE (8DR8176) 8 = INITIAL CORE 68DR8219)

C = RELOAD 1 (PSCR8204LA)

Figure 1. Reference Core Loading Pattern 6

Y1003J01A10 100 45 i

CONTROL ROD DRIVE VERSUS TIME SCRAM REACTIVITY VERSUS TIME

- e I

l 80 -

35 j

70 --

678 CRD IN PERCENT I 30 60 -

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NOMIN AL SCRAM CURVE IN (-3)

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p b

SCR AM CURVE USED IN ANALYSIS me 40 -

l 15 30 -

- 10 j m -

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10 -

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1 2 3 4 TIME (sec)

Figure 2. Scram Reactivity and Control Rod Drive Specifications, EOC 2 1

7  !

1

VI  ! EUTRON FLUX l VESSEL PFES FIISE (PSil 150* / A 11 2 AVE SURFFCE HEAT FLUX M ORE INLE T FLOW 125*

2 RELIEF VFLVE FLOW 3 BTPRSS VFLVE FLOW 4 c0HE INd T SUB 4 5 S 4

6

__ I g 100. 3 ^ I 75.

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50. as.

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ER A [ 2 t c1- b u o O.

0.

8- -

-25. 'w- *

40. 80. 120. 160. O. 40. 80. 120. 100.

TIME (SEC) TIME (SEC) b M

u o

ao l o 7M I LEVEL (INCH-KF-SEP-SNIflT 2 VESSEL SlEAMFLOW 1 YOID FIEACTIVITT 2 DOPPLER FEACTIVITT g

g l'iG* 3 TURBINE S TEAMFLOW J 3 SCRAM REFCTIVITT__ o 3*

4 t ttuWATEF FLOW 4 TOTAL BEFCTIVITT ll 5

.- D 100 m -

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- D.

,/ m A

U N 8 o

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50*

-1 W is 4 g

I g* ....n.... 1  ? -2. - '- I

  • l 0. 40. 80. 120. 160. O. 40. 80. 120. i'>') .

l TIE (SEC1 TIME (SEC1 Figure 3. 11atch 2 Reload 1, Loss of 100*F Feedwater lleating l

l

f.

l Y1003J01A10 1

51 8' 8 8 47 38 38 38 38 f_ 43 8 8 12 8 8 39 38 38 44 44 38 38 35 8 8 14 6 14 8 8 31 38 44 38 38 44 38 27 8 12 6 0 6 12 8 23 38 44 38 38 44 38 19 8 8 14 6 14 8 8 15 38 38 44 44 38 38 11 8 8 12 8 8 7 38 38 38 38 3 8 8 8 02 06 10 14 18 22 26 30 34- 38 42 46 50 NOTES:

1. Rod pattern is 1/4 core mirror symmetric.

i

2. Number indicates number of notches withdrawn out of 48. Blank is a withdrawn rod.
3. Error rod is (26,27).

Figure 4. hatch 2 Reload 1, Limiting Rod Pattern

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Y1003J01A10 1.2 I,

ULTIMATE PERFORMANCE LIMIT 1.0 - - --

t 0.8 -

-R w

5 9

NATURAL y CIRCULATION 6

8 105% ROD LINE 0.4 -

0.2 -

I I I I o ,

0 20 40 60 80 100 PERCENT POWER Figure 5. Hatch 2 Reload 1, Decay Ratio 10

Y1003J01A10 f

APPENDIX A f

This appendix contains all the Hatch Unit 2 specific plant data not included in " Generic Reload Fuel Application", NEDE-240ll-P-A-1, July 1979.

Table 1-1 Reactor Power - Number of- Fuel Length Power Density Plant (MWt) Fuel Bundles (in.) (kW/t)

BWR/4 8x8R and P8x8R 8x8R and P8x8R Hatch 2 2436 560 150 49.16 Table 4-1 Assemblies with Lower Plant Tie Plate Holes BWR/4 Hatch 2 All Fuel Table 5-3 Plants for Which the Bounding MCPR Safety Limit Applies Hatch 2 Table 5-4 SV RV or S/RV Number Number 1. oves t Opening of of Number of Set Capacity Lowest Capacity Time Time Safety Relief Safety / Relief Point at Setpoint Set Point at Setpoint Delay Constant Plant Valves Valven Valves (paig) (No./%) (psta) (No./%) (msec) (m s e cl _

BWR/4 Hatch 2 - -

11 - -

1090 + 1% 11/89.6 400 100 1

l I

l 11

Y1003J01A10 Table 5-6 k

Thermal Dome Turbine. Rated Steam . Core Power Pressure Pressure Flow x 106 plow x 10 6 (MWti0.2%) (psigt2 psi) (psig 2 psi) (1b'hr:0.2%) (1b/hr 0.2%)

BWR/4 Hatch 2 2536 1020 960 11 77 f

Table 5-8 6

Core Flow x 10 Reactor Core Pressure Iulet Enthalpy Non-Fuel Power (1b/hr) (psia) (Btu /lb) Fraction BWR/4 Hatch 2 77 1035 526.9 0.04 Table 5-13 Specific Plant Reference Lead Plant Plant LOCA Analysis Document LOCA Analysis Document Hatch 2 N/A* 5-31**

Table 5-17 Plants for which ATWS pump trip is' assumed in transient analysis Hatch 2

-*See "Edwin I. Hatch Nuclear Plant Unit Number 2, Final Safety Analysis Report",

Section 6.3.3

    • Reference in NEDE-24011-P-A.

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L Y1003J01A10 l

APPENDIX B RELOAD FUEL BUNDLES P8DRB284LA.

Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), General Electric Licensing Topical Report NEDE-240ll-P-A, " Generic Reload Fuel Application," Appendix D l Submittal, December 14, 1979.

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Y1003J01A10 l

, APPENDIX C CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1) s CE One Dimensional' Core Transient Model ODYN computer code has been used for 3 pressurization transients analysis (Reference C.1).

'?

Power Flow $ Q/A s1 y Plant Transient Exposure (%) (%) (%) (%) (psig) (psig) 8x8R P8x8R Response h

Load Rejection EOC 104.2 100_ 506.9 11b.4 1185.7 1208 0.17 0.17 Figure C.1 without Bypass Feedwater EOC 104.2 100 293.2 118.9 1150.7 1182 0.15 0.16 Figure C.2 Controller Failure REFERENCE C.1 NEDE-24154P, Volumes 1, 2, and 3, "One-Dimansional Core Transient Model,"

October 1978.

~

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'15

I 1 NEUTRON douX l 1 vi?fL rrrS RISC (PS!!

2 AYL SURFrCE HEAT FLUX 2 Sor E I f Yi 1.VE FLOW 150. 3_C0_f G NLI i FLOW 3m. 3 fJ t11r vi t.VE Ft rW

'l 'l litt4rf; V! !.VL ftrW 3 i

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l l 5 100. d - ' 200.

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0. 4. 8.  !?. 16. O. 4. 8. 12. 16.

TIME ISECl TIME ISEC) w o

o W

m o

1 LEVEL (IMH-fief-5EP-SMIRT 1 V010 REACTIVITT o 2 VESSEL SIEAMFLOW r  ? 00PPLFR ff ACTIVITT 200* 3 Tiin81 a ! IEnMFLOW y' / 3 srmM RFI f.flVITY FLEDWMU lTDW la IdfiOll i.TTVIC t-W

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TIME (SEC) TIME ISEC)

U Figure C.1 Hatch 2 Reload 1, Load Rejection w/o Bypass, EOC2 i

I R] 150.

LIUTfD4 FLtkl 2 .:VL StlRFf CE HEAT FLUX 4 lllE INL[ T FL(M 1 VFSTL PRES RI'iE IPSil 2 Mf f. I Y VI LVE FL(M (h  % l Ilit INLI I 'iUU 12'i. I if tlT VI ti-u :r3 t.Vt FLfM syL fluw h G 8 100. 3

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TIME (SEC)

TINE (SECl H

% b

% 8 t

1LEVElt!N(hi-REF-SEP-SK!R. I VOID REAR T!VITY O 2 YESSEL STEAMFLOW 2 .PFLER I EPCT[VITY >

l'iO' 3 TunRINE 5 TEAMFLOW A_ u y fiTEll%ATEF FLOW g* 3 .fWlM RFI CTIVITY Il ITICRC CTIVIIT $

a " A G 100.

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Figure C.2. Hatch 2 Reload 1, Feedwater Controller Failure, Maximum Demand, EOC2

Y1003J01A10 APPENDIX D DENSIFICATION POWER SPIKING Reference D-1 documents the NRC staff position that ". . . it (is) acceptable to remove the 8x8 and 8x3R spiking penalty factor from the plant Technical Specification for those operating BWRs for which it can be shown that the pre-dicted worst case maximum transient LHGRs, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHGRs".

The Hatch-2 Reloed-l submittal contains-the required information to demon-strate that the stated criterion is met for Hatch-2, Reload 1. Section 10, Rod Withdrawal Error, and Appendix F (Linear Heat Generation Rate for Bundle Loading Error) include the densification effect in the calculated LHGR of the 8x8 fuels.

REFERENCE D-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.

e 19/20

d

'Y1003J01A10 APPENDIX E OVERPRESSURIZATION ANALYSIS GE One-Dimensional Core Transient Model ODYN computer code has been used for overpressurization transient analysis (Reference E.1).

P,7 P Power Core Flow p, Transient (%)' (%) (psig) (psig) Response e MSIV Closure 104.2 100 1203.3 1234 Figure E.1 (Flux Scram) l 1

l REFERENCE

, E.1: NEDE-24154P, Volumes 1, 2, and 3, "One-Dimensional Core Transient Model", 0ctober 1978.

21 l I

L.

! EUTRON FLUX 1 VESSEL PfES RISE IPSI) 2 RVE SLFIFFCE TAT FLUX 2 SAFETT VrLVE FLOW 12- 3 CORE INLE T FLOW 3 REL((F V% VE FLOW _

4 300-y eTess5 vft.VE FLOW 5 5 6

100- A 200-8 n

x ~ 1

/

E b N

50. t00. 2 3 D. '-

O. 2E ' . .1 tu  : u .

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

TDE ISEC) TIME (SECl d

8 0 0 i LEVEL (INC H'-REF-SEP-SKIRT 2 VESSEL S1EAMFLOW I V010 REAC'TIVITY O g* 3 TUR8IPE ! TEAMFLOW n R DOPPLER fEFETIVITT FLOW g*

W

. / %SERAM REfCTIVITY 4 CTIVITY g FEEDHATEF 2

N

  • 7 ' "

100. -

h, 0 grw- - q m 1 1 1

0. '

y -

1- s t

- C -

w 300, . ..i. . 2. -- '- *

0. 1. 2. 3. 4. O. 1. 2. 3. 4.

TIME (SEcl TIME ISECl Figure E.1. Hatch 2 Reload 1, MSIV Closure, Flux Scram, EOC2

Y1003J01A10 APPENDIX F LOADING ERROR RESULTS Linear Heat Generation Rate (kW/ft): 17.7 i

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