NE DO-24204s

79NEU10CLA3S

JULY 197

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J-4

SUPPLEMENTAL RELOAD E

LICENSING SUBMITTAL FOR ,

PEACH BOTTOM ATOMIC POWER STATION ,.

UNIT 3, RELOAD NO. 3 iyVi

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/SO GENERAL h ELECTRIC4

NE:0-2 4 203A' 9N E D10 3Class !

July 1979

SUF?LEMENTAL RELCAD LICEN3ING SUEMITTALFOR

?EACH BOTTCM ATOMIC PCWER STATICNUNIT 3 RELOAD 3

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Prepared by 4[ ; Approved: '\ ', '- s, .7

A. M. Erv in R . O . E ru;;ge , "anagerOperating Licenses II

NUCLE AR E',E AGY 90;ECTS CIVISION * GENE R AL E LECT AIC CC*AP ANYSAN ;CSE, CA LIFO RNI A 95125

GENER AL $ ELECTRIC

/ 1

E D0-24 204 A

:.v.?O?'" ANT NO'":CE PEARD:NG

CONTEW'S CF TH:J REPORT

PLEASE READ CAPITUILY

This repor* was prepared by General Electric solely for PhiladelphiaElectric Company for use with ~he U.a. Nuclear Regulatory Commission(UadRC) !c: amending PECO's operating licen3e of the Peach 20::cm AtomicPower J ation, Uni '. The informetion ocntained in this report is

believed by General Electric to be an accurate and ::ue representationof the facts krcwn, obtained c: provid& to General Electric at the time

this repor: was prepared.

The only undertakings c! ~he Gem:31 ileccric Company respacting in!cima-tien in this c..cument a:e contained :n the contract between PhiladelphiaElectric Compang ani General Electric Company !c: nuclea: fuel and relatedservices !c: the nuclear system for Peach Bottom Atomic ?cwer acation,Uni 3, ani :vthirg ccn~ained in this document shall be c~nstrued as

u.gp.n, .i .v. , a3ia c., ". ~ .~a c ~. . '..'.a. ''=<=w-'.' ~. .k. i ~ .'~..''~.~a/~.~''.~. v. a"~. 3 o- a .' .' .~.a d- - -.. . a .'

-e - --

by said contract, c: ic: any 71:pse c~her than *ha t for which it is

intended, is nct aut.korized; and with respect :c any such unauchariceduse, neither Ge t ral Electric Company nor any of the contributors tothis document makes any representation c: warranty (express c: i= plied)as to the completeness, accuracy c: usefulness c! the information can-

tainel in this documer c: that such use c! such in!c:mation may not! :!:ime pri?a =ly cwnad ri;h~=I- mar do ~."p,. m :me any raspansibility

fc: liability or damage of any kin! which may result !::m such use of

such information.

,,..

NEDO-24204A

1. PLAff"-U'i!QUE ITEX3 ( 1.0 )'

Appendix A - Loading Error Limiting LHOR

Append Lx B - Pressarised Test Assembly

Appendix C - Fast Scram Control Ecd Drive

Appendix D - New ththods - Fuel Loading Errcr

Dundle PEDRB234H description is documented in non-approved submittal, Peference 2

2. RELOAD 7UEL EUNDLES C'.0, 3.3.1, and 4.0)

Fua l 'vra Number 'Jumba r Drilled

Ir rad ia ted 7D250 Type II 52 3

cDB274L 119 117

3DB274H 68 63

PTA 1 1

SDRB283 252 252

New P8DRB234H 272 272

.v 3, , ,e , ,.,, ,,

Ie c_.

, , e- . n - .L,c, uvRc ,nA ,,ilu- 0,a,i.cr.! ( .3.i)-m , -, , ,:.c .:. o . j.

Nominal previcus cycle expcsure: 3 3 6 3 M'4d/ t .

Assumed relcad c'/cle exposure: '7160 M'4d/t-

Co ra leading pattern : Figure 1.

. - . - - . . . - - _- ,,-....,,,,.,,.,,.n,In..., . . , q ,. .q ,*, Uw w . a . E~n ,-n.,- 2 .,01 1. -,,--y ,4 LnLvdLALT U LVNO " *'., r r L L i . d e am A o a v a r ov. .6uw

70 IDS, 200C (3.3 2.1.1 and 3.'.2.'.2)

B O C k o_ .c ,.

Uncontrolled 1.124

Fully Centrolled 0.9619

Strongest Control Rod Out 0.9869

R, Maximum Increase a. Cold Core Reactivity with 0.0000

Exposure into Cycle, ak

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

.

NE DC-2 4204 A

5. STANDBY LIOUID CONTROL SYSTEM SHUTDCWN CA?A3ILITY ( 3.3.2.1.3)

Shutdown Margin (ak)ppa (200C. Xenon Free)

660 0.032

6. RELOAD-UNIOUE TRANSIENT ANALYSIS IN?UTS ( 3 3 2.1.5 and 5.2 )

ECC4 ECC4-2000 mwd /t'

Void Coefficient N/ A'' (2/%Rg) -3.26/-10.33 -9 24/-11.307a id F rac t io n ( % ) 40.23 40.23Doppler 33 efficient N/A'' (t/ F) -0.2366/-0.2242 -0.2279/-0.21650

Average Fuel Temperature (cF) 1356 1356

Sc ram Worth N/ A** ( $ ) -35.48/-23.38 -33 93/-27.13Scra= Reactivity vs Time. Figure 2a Figure 2b

7. RELCAD-UNIOUE GETA3 TRANSIENT ANALYSIS INITIAL CCNDITION PARAMETERS (5.2)

7x7 3x3 3xBR PTA/?Sx3R

EOC4- ECC4- EOCu- 20CU-

2000 2000 '000 2000

_EO C4 mwd /t EOC4 mwd /t EOC4 mwd /t E0C4 mwd /t

Peaking factors

(local. rad ial

and axial) 1.24 1.23 1.22 1.22 1.20 1.20 1.20 1.20

1.23 1.30 1.37 1.43 1.39 1.56 1.47 1.56* 40 1.40 1.40 1.30 1.40 1.40 1.40 1.40.

R-Factor 1.100 1.100 1.098 1.098 1.C52 .052 1.052 1.052

Sundle Power (MWt) 5.41t 5.492 5 .77 7 6.02 6 6.237 6 .56 3 6.209 6.563Sundle 210w

(103 lbrne) 123.4 122 .3 111.1 109.3 111.1 109.3 111.3 109.5Initial M:?R 1.23 1.22 1.29 1.24 1.30 1.2u 1.32 1.24

'Mid Cycle Exposure Point* *N = 'hclear Input DstaA: Used in Transient Analysis

0- -

tie DO-24204 A

8. SELECTED MARGIN IMPROVEMENT OPTICNS (5.2.2)

Exposure Cependent Licits: From 20C4 to E0C4-2000 M'4d/t and from E004-2000

mwd /t to ECC4

Q. rn.D R -g'l u -97 * Q ' t4'0" T. O U* Q C.,0 v L A J' f c . c'_) . j 'Jd . t' A ..'./q T.Jt *T m off**CA..A. .1.w i- . . . \Ja,

A 4

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e-arn: - 'a:

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SU.P.A R Y ( 5 . 2 .1 )

Rod

Position ACPR* v;HCR f KW/ f t) * * * *

Rod Block (Feet 3x3R/ 3x3R/ isiting

Set Poin* With d rwn ) 7x7'' 3x3 P8x3R/?TA 7x7'' 3x3 ?3xiR/?TA Ro d P ' ttern

105 4.5 - 0.10 0.14 - 12.13 14.37 Tigure 61C6 5.0 - 0.10 0.16 - 12.09 14.41 rigure 6

107''' 6.0 - 0.11 0.20 - 12.01 14.53 Figure 6108 3.0 - 0.14 0.25 - 'i.70 16.39 rigure 6

109 9.0 - 0.15 0.27 - 12.06 17.99 rigure 5

110 12.0 - 0.16 0.33 - 12.56 13.09 rigure 6

' Based on an initial MCPR o f 1.43 ( 3x3) an d 1.34 (3x33 and P3x3R'**7x7 fuel is located Only en the core periphery and is not limiting; t here fo re

its response .2 not giten.

* ** Indicates setpoint selected.

"'* Includes the effects of densification power spiking.

tall aCFR ralues calculated from initial power of 104.5%.

7s

I

s

NEDO-24204A

11. CPERATING MCPR LIMIT (5.2)_

BOC4 to EOC4-2000 mwd /t

EOC4-2000 mwd /t to EOC4

1.23 (7x7 fuel) 1.23 (7x7 fuel)1.24 (3x3 fuel) 1.30 (3x3 fuel)1.27 ( 3 x3R fu el) 1.30 (3x3R fuel)1.27 (PTA/P8x3R fuel) 1.32 (?TA/P8x3R fuel)

12. O'IERPRESSURIZ ATICN ANALY3IS SUMMARY (5.3 )

ower Core Flow ?3L P Planty

(%) /%) fosi2) (osic) Rascensa

M3:7 Clocure 104.5 100 1271 1301 Tigure 7

(Flux Scram)

-

13 STABILITY ANALYSI3 RESULTS ( 5.4 )

Decay Ratio: Figure 3

Reactor Core S tability

Decay Ratio , x;/x . 0.9Ca

( 105 % Ro d L in e - Na tu ra l

Circulation Power)

Channel Hydrcdynacic Performan^a ecay Ratio, x2 Xn/ O

'105% Rod ine - NaturalCirculation Power)

3x3R/P3x33/?TA Channel 0.29

8x3 Channel 0.40

7x7 Channel <0.01.

dL. . ..

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F . *

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NEDO-24204A

14 LOSS-CF-COOLANT ACCIDENT RESULTS ( 5.5.2)

MAPLH OR PCT Local oxidation

Exposure (kW/ft) (cF) Fraction

(mwd /t) P SD H B28 4 H PSDR3234H PSDRB284H

200 11.3 1812 0.007

1000 11.3 1813 0.007

5000 11.7 1358 0.008

10000 12.1 1994 0.009

15000 12.0 1897 0.009

20000 11.6 1858 0.008

25C00 10.9 1777 0.006

30000 10.2 1700 0.004

15. LOADING ERRCR RESULTS (5.5.4)

Limiting Events: Mislocated Bundle P 3DRB2 34H MCP R 21.07

Ro tated Bundle ? 3DRB2 3uH MCPR ll 07

,g ,m,. (r. .1,.u v a . :.C,m . tD .-no ad. A L .r- eo.o, ...c:ob.; . orc u. o. . . e

Doppler Reactivity Coefficient: Figure 9

Ac c i den t Reactivity Shape Function: Figures 10 and 11

Scrim Reactivity Tunctions: Figures 12 ara 13

Plant 3pecific Analysis Results

Para eters No t Bounded:

3 c ~15 React ivity Functions: Cold and Hot 3tartup

Resu_ unt Peak Enthalpies (cal /g):

Cald Hot Startuo

,93 =ciIsi 4

a

'tEDO-24204A

17. REFERE! ICES

1. " General Electric Soiling '42.ter Reactor Generic Reload Fuel Applicaticn,"

August 1978, (!iEDE-24011-P-A).

2. " General Electric Boiling ' dater Reacur Gencric F.eloa d Fuel Application,"

May 1979, ( tIEEE-24011 -P- A , Amendment 3 ),

.

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01 02 05 07 09 11 13 15 17 13 21 23 25 27 29 3' 33 35 37 39 41 43 45 47 49 51 51 55 57 59

FUEL TYPE

A = 70 250 E SO R 9283=

8 = SO8274 L F = PS C A B 234 HC = 8DB274HO =PTA

Figure 1 Pe:'erence CSre Leading Pattern

7

I C

NE00-24204A

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C - 67A CRD IN PERCENT; I - NOMINAL SCR AM CURVE IN (-5) 40| 2 - SCRAM CURVE USED IN AN ALYSISI

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Figure 2a. Scram Reactivity and Centrol Rod DriveSpecifications frem E0C4-2000 mwd /t to EOC4

%D

NEDC-24204A

Ino 45C

C - 67A CAL IN PERCENT

30 1 - NOMINAL 2CR AM CURVE IN (-512 - SC9 AM CUR JE USED IN ANALYS;S 40

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Figure 25. Scram Beactivity and Centrol Rod DriveSpecifications frca 30C4 to E0C4-2000 T4d/t

9

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NEDO-24204A

APPENDIX A

LOADING ERROR LIMITING LHGR

This appendix provides the limiting linear heat generation rate (LHCR) resulting

from the bundle loading ermr (BLE) analysis.

Limiting Event LHGR (kW/f t)'

Misplaced Bundle 18,4

P8DRB284H

' Limiting bundle results including the effects of densification pcwer spiking( factor = 1.022).

A -1/ A-2 /r

|

NEDO-24204A

APPENDIX B

PRESSURIZED TEST ASSEMBLY

The pressurized test assembly (PTA) was described in NEDO-21363-1 Supplement 1,

dated November 1976 as amended by NEDO-21363 4 Supolement 4, dated January 1977.

In addition to describing the PTA, these licensing cccuments provided a safety

analysis for installation of the PTA in the Peach Bottom 3 reactor for cycle 2

and subsequent cycles. It is planned that the PTA will remain in the core during

cycle 4. The safety analysis performed for cycle 4 includes consideration of

the PTA as part of the reloaded core. Although the GETAB analysis was performed

for the PTA as a separate fuel type, its effect on the remaining safety analysia

is insignificant since the PTA configuration is basically the same as the reload

3 8x8R bundle.

B-1/B-2 -,

NEDO-24204A

APPENDIX C

FAST SCRAM CONTROL RCD DRIVE

The fast scram control rod drive (FSCRD) was described in NEDO-21363-2 Supple-ment 2. In addition to describing the FSCRD, the licensing document provided

the results of a safety review and evaluation which considered any effects the

presence of the FSCRD would have on the plant safety analysis. It was determined

that the inclusion of the FSCRD did not introduce an unreviewed safety question

and had no effect on parameters used in the plant safety analysis. For cycle

4 the FSCRD may be lef t installed for another cycle of operation.

NEDO-21363-2A, " General Electric Boiling Water Reactor Relcad 1 Licensing Amend-

ment for Peach Bottom Atomic Power Station Unit 3 Fast Scram Control Rod Drive,Second Supplement," July 1979, provides results of evaluation of the FSCRD which

was operated in Peach Bottom Unit 3 during cycle 2 and subsequently disassembled

and inspected. The report pro / ides performance results and a report of the

effects of the reactor environment on the drive mechanism. A safety evaluationis also provided which demonstrates that continued operation of the currentlyinstalled FSCRD, during cycle 4, does not introduce an unreviewed safety ques-tion and has no adverse effect on parameters used in the plant safety analysis.

C-1/C-2

'['

NEDO-24204A

APPENDIX D

NEW BUNDLE LOADING ERROR EVENT ANALYSES PROCEDURES

The bundle loading error analyses results presented in Section 15 in the supple-ment are based on new analyses procedures for both the eutated bundle and the

mislocated bundle loading error events. The use of these new analyses proce-dures is discussed below.

D.1 NEW ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOAD. Zi ERROR EVENT

The rotated bundle loading error event analysis results presented in this supple-ment are based on the new analysis procedure described and approved in ReferenceD-1. This new method of performing the analy313 is based on a more accuratedetailed analytical model.

The principal difference between the previous analysis procedure and the newanalysis procedure is the modeling of the water gap along the axial length ofthe bundle. The previous analysis used a uniform water gap, whereas the newanalysis utilizes a variable water gap which is more representative of theactual condition, since the interfacing between the top guide and the fuelspacer buttons, caused by misorientation, causes the bundle to lean. The effectof the variable water gap is to reduce the power peaking and the R-factor inthe upper regions of the limiting fuel rod. This results in the calculation of

a reduced CPR for the rotated bundle. The calculation was performed using thesame analytical models as were previously used. The only change is in the simu-lation of the cater gap, which more accurately represents the actual geometry.

The number presented in Section 15 represents the minimum CPR of the most limit-ing rotated bundle starting from an initial CPR of 1.22 which includes the 2%

allowance for uncertainties as required by the NRC.

D.2 NEW .tNALYSIS PROCEDURE FOR THE MISLOCATED BUNDLE LOADING ERROR EVENT

The misicaated bundle loading error event analyses results presented in thissupplement a based on the new analysis procedure described in Reference D-1.

This new method of performing the analysis employs a statistically corrected Halingprocedure and analyzes every bundle in the core.

D-1

,

NEDO-24204 A

The use of the statistically corrected Haling analyses pr ocedure indicates thatthe minimum CPR for the mislocated b 'ndle in the core is greater than the safety

limit.

REFERENCES

D-1 Safety Evaluation Report (letter), D. G. Eisenhut (NRC) to R. E. Engel(GE), F5N-200-78, dated May 8,1978.

D-2