ML19225D094
| ML19225D094 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/31/1979 |
| From: | Brugge R, Ervin A GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19225D095 | List: |
| References | |
| 79NED103, NEDO-24204-A, NEDO-24204A, NUDOCS 7908060407 | |
| Download: ML19225D094 (31) | |
Text
NE DO-24204s 79NEU10 CLA3S JULY 197
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-4 SUPPLEMENTAL RELOAD E LICENSING SUBMITTAL FOR ,
PEACH BOTTOM ATOMIC POWER STATION ,.
UNIT 3, RELOAD NO. 3 y i
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/SO GENERAL h ELECTRI 4
NE:0-2 4 203A
' 9N E D10 3 Class !
July 1979 SUF?LEMENTAL RELCAD LICEN3ING SUEMITTAL FOR
?EACH BOTTCM ATOMIC PCWER STATICN UNIT 3 RELOAD 3
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Prepared by 4[ - Approved: '\ s, ', .7 A. M. Erv in R . O . E ru;;ge , "anager Operating Licenses II NUCLE AR E',E AGY 90;ECTS CIVISION
/ 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 Philadelphia Electric Company for use with ~he U.a. Nuclear Regulatory Commission (UadRC) !c: amending PECO's operating licen3e of the Peach 20::cm Atomic Power J ation, Uni '. The informetion ocntained in this report is believed by General Electric to be an accurate and ::ue representation of 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 Philadelphia Electric Compang ani General Electric Company !c: nuclea: fuel and related services !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. - - -..
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- -- d 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 unauchariced use, neither Ge t ral Electric Company nor any of the contributors to this 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.
- 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
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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.
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., r r L L i . d e am w A o a v a r ov. .6u w w . a . E~n , .,01 1. - .*'. , q ,. .q ,*, U 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/-11.30 7a id F rac t io n ( % ) 40.23 40.23 Doppler 33 efficient N/A (t/ 0 F) -0.2366/-0.2242 -0.2279/-0.2165 Average Fuel Temperature (cF) 1356 1356 Sc ram Worth N/ A** ($) -35.48/-23.38 -33 93/-27.13 Scra= 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 6.563 Sundle 210w (103 lbrne) 123.4 122 .3 111.1 109.3 111.1 109.3 111.3 109.5 Initial 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 Dsta A: 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.
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-- r.7 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 6 1C6 5.0 -
0.10 0.16 -
12.09 14.41 rigure 6 107 6.0 -
0.11 0.20 -
12.01 14.53 Figure 6 108 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%.
7 s
I
- 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 Py Plant
(%) /%) 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;/xa. 0.9C
( 105 % Ro d L in e - Na tu ra l Circulation Power)
Channel Hydrcdynacic Performan^a n ecay Ratio, x2 /X O
'105% Rod ine - Natural Circulation Power) 3x3R/P3x33/?TA Channel 0.29 8x3 Channel 0.40 7x7 Channel <0.01 d
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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
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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 i4
=c iIs 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 ),
- I
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SO R 9283 8 = SO8274 L F = PS C A B 234 H C = 8DB274H O =PTA Figure 1 Pe:'erence CSre Leading Pattern 7
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NE00-24204A 100 -
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C - 67A CRD IN PERCENT
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0- l 0 a 1 2 3 4 5 6 TIME (sect Figure 2a. Scram Reactivity and Centrol Rod Drive Specifications frem E0C4-2000 mwd /t to EOC4
%D
NEDC-24204A Ino 45 C
C - 67A CAL IN PERCENT 30 1 - NOMINAL 2CR AM CURVE IN (-51 2 - SC9 AM CUR JE USED IN ANALYS;S 40 80 35 1
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Figure 25. Scram Beactivity and Centrol Rod Drive Specifications 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 /
l
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 evaluation is also provided which demonstrates that continued operation of the currently installed 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
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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 Reference D-1. This new method of performing the analy313 is based on a more accurate detailed analytical model.
The principal difference between the previous analysis procedure and the new analysis procedure is the modeling of the water gap along the axial length of the bundle. The previous analysis used a uniform water gap, whereas the new analysis utilizes a variable water gap which is more representative of the actual condition, since the interfacing between the top guide and the fuel spacer buttons, caused by misorientation, 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 limiting fuel rod. This results in the calculation of a reduced CPR for the rotated bundle. The calculation was performed using the same 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 this supplement a based on the new analysis procedure described in Reference D-1.
This new method of performing the analysis employs a statistically corrected Haling procedure and analyzes every bundle in the core.
D-1
NEDO-24204 A The use of the statistically corrected Haling analyses pr ocedure indicates that the 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