ML20003E155
| ML20003E155 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 02/28/1981 |
| From: | Engel R, Leaser J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20003E138 | List: |
| References | |
| Y1003J01A20, Y1003J1A20, NUDOCS 8104020358 | |
| Download: ML20003E155 (40) | |
Text
O' O :
l Y1003J01A20 Revision 0 Class I February 1981 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEAGI BOTTOM ATOMIC PO'4ER STATION UNIT 3, RELOAD NO. 4 Preparceby:_h J. D. Leaser Senior Licensing Engineer Approved:
R. E. EngeI, Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JO6d. CALIFORNI A 95125 GEN ER AL $ ELECTRIC wovo#353
Y1003J01A20 Rev. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT ,
PLEASE READ CAREFULLY This report was prepared by General Electric solely for Philadelphia Electric Company for use with the U.S. Nuclear Ragulatory Commission (USNRC) for amending PECO's operating license of the Peach Bottom Atomic Power Station, Unit 3. The information contained in this report is believed by General Electric to be an accurate and true representation of the facts (nown, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting informa-tion in this document are contained in the contract between Philadelphia Electric Company and General Electric Company for nuclear fuel and related services for the nuclear system for Peach Bottom Atomic Power Station, Unit 3, and nothing 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 or implied) as to the completeness, accuracy or usefulness of the information con-tained 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 information.
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Y1003J01A20 Rav. 0
- 1. PLANT-UNIQUE ITEMS (1.0)*
Appendix A - ODYN Transient Code Appendix B - Fast Scram Control Rod Drive Appendix C - New Bundle Loading Error Event Analyses Procedures Appandix D - GETAB Analysis Initial Conditions Appendix E - Margin to Spring Safety Valves Appendix F - Pressurized Test Assembly Fuel Rod Replacement Appendix G - Initial MCPR
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1, and 4.0)
Fuel Type Number Number Drilled Irrediated 8DB274H 23 23 PTA 1 1 8DRB283 252 252 PSDRB284H** 272 272 New P8DRB299*** 216 216 Total 764 764
- 3. REPERENCE CORE LOADING PATTERN (3.3.1) -
Nominal previous cycle exposure: 16205 mwd /t Assumed reload cycle exposure: 18208 mwd /t Core loading pattern: Figure 1.
- ( ) refers to areas of discussion in Reference 1.
- Described in Reference 2.
, ***Deteribed in Reference 3.
1
(_1 -
Y1003J01A20 Rsv. 0
- 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,f, Uncontrolled 1.115 Fully controlled 0.960 Strongest Control Rod Out 0.987 R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, ik 0.000
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (ak) gga (20*C. Xenon Free) 660 0.04
- 6. RELOAD-tWIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.3 and 5.2)
EOC5 EOC5-2000 mwd /t*
Void Coefficient N/A** (c/%Rg) -7.81/-9.77 -8.79/-10.99 Void Fraction (%) 40.22 40.22 Doppler Coefficient N/A** (c/*F) -0.221/-0.210 -0.212/-0.202
. Average Fu',1 Temperature (*F) 1391 1391 Scram Worth N/A** ($) -46.31/-37.05 -46.31/-37.05 Scram Reactivity vs Time Figure 2 Figure 2
- Mid Cycle Exposure Point
- N = Nuclear Input Data A = Used in' Transient Analysis e
L
Y1003JOL420 Rev. 0
- 7. RELOAD-UNIQUE GETAB TRANSIENT MIALYSIS INITIAL CCNDITION PARAMETERS (5.2) 8x8 8x8R PTA/P8x8R EOC5- EOC5- EOCS-2000 2000 2000 EOC5 Ed/t EOC5 Wd/t EOCS Wd / t Peaking factors (local, radiel and axial) 1.22 1.22 1.20 1.20 1.20 1.20 1.44 1.47 1.47 1.53 1.44 1.51 1.40 1.40 1.40 1.40 1.40 1.40 R-factor 1.098 1.098 1.051 1.051 1.051 1.051 Bundle Power (MWt) 6.007 6.202 6.184 6.455 6.081 6.376 Bundle Flow (103 lb/hr) 107.6 106.7 110.4 108.6 111.5 109.6 Inicial MCPR 1.33 1.27 1.33 1.27 1.36 1.29
- 8. SELECTED MARGIN UtPROVEMENT OPTIONS (5.2.2)
Exposure Dependent Limits: From BOCS to EOC5-2000 Wd/t and from EOC5-2000 Wd/ t to EOC5.
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
.wrn Feuer Cerg Flow -4 07A Fgt F, FTA & Plant Transiast _fposure (I) (%) (*185R) (NSal M (pett) M M F938R Ras m se Generator lead .
Rejecties- EOCS 104.5 100 no typees 903 129 1239 -1259 0.25 0.26 0.29 Fiswee la EDCS-2000 mwd /t 104.5 100 75 3 124 1223'- 1266 0.20 0.20 0.23 Figure 35 14ee of 100*F Feeeueter meeting toCS-EOCS 104.5 100 125 124 '1014 1069 0.15 0.16 0.16 Figure &
Feeeuster Centro 11er EOCS -104.3 . 100 390
- Feilet8 124 1161' 1202 0.21 0.21- 0.23 Figure Sa~
EOCS-200G mwd /t 106.3 100 33 7 121 1153 1196 0.13 0.13 0 17 Figure Sb-4 3
Y1003J0lA20 Rev. 0
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LD1ITING INSTRUMENT FAILURE) TRANSIENT SUICIARY (5.2.1)
Rod Position ACPR LHCR Rod Block (Feet) 8x8R/ 8x8R/ Limiting Set Point Withdrawn) 8x8* P8x8R/PTA 8x8* P8x8R/PTA Rod Pattern 104 3.5 -
0.08 -
16.0 Figure 6 105 4.0 -
0.09 -
16.3 106 4.0 -
0.09 -
16.3 197** 6.0 -
0.13 -
16.4 108 7.5 -
0.14 -
16.4 109 12.0 -
0.17 -
16.4 "
110 12.0 -
0.17 -
16.4
- 11. CYCLE MCPR VALUES (5.2) 3005 to
. EOC5-2000 mwd /t Option A Option B PTA/ PTA/
Pressurization Events 8x8 8x8R P8x8R 8x8 8xdR P8x8R Generator Load Rejection, 1.33 1.33 1.36 1.12 1.12 1.15 No Bypass Feedwater Controller 1.27 1.27 1.29 1.21 1.21 1.23 Failure Non-Pressurization Events Loss of 100*F Feedwater 1.22 1.23 1.23 1.22 1.23 1.23 Heating .
Fuel Loading Error 1.20 1.20 1.20 1.20 1.20 1.20 Rod Withdrawal Error
- 1.20 1.20
- 1.20 1.20 EOC5-2000 mwd /t to EOCS Option A Option L PTA/ PTA/
Pressurization Events 8x8 8x8R P8x8R 8x8 8x8R PSx8R iGenerator Load Rejection 1.39 1.39 1.42 1.27 1.27 1.30 No Bypass Feedwater' Controller 1.34 1.34 1.36 1.27 1.27 1.29 Failure Non-Pressurization Events Loss of 1000F Feedwater 1.22 1.23 1.23 1.22 1.23 1.23 Heating Fuel Loading Error 1.20 1.20 -1.20 1.20 1.20 1.20 Rod Withdrawal Error
- 1.20 1.20'
- 1.20 1.20
- 8x8. fuel.is not limiting; therafore its response to rod witharawal error is not.given.
- Indicates sacpoint selected.
'4
Y1003J01A20 Rev. 0
- 12. OVERPRESSURIZATION ANALYSIS SUPliARY (5.3)
Power Core Flow PSL Py Plant
(%) (%) (psig) (psig) Response t 4
MSIV Closure 104.5 100 1260 1292 Figure 7 (Flux Scram)
- 13. STABILITY ANALYSIS RESULTS (5.4)
Decay Ratio: Figure 8 Reactor Core Stability Decay Ratio, x2/ox: 0.87 (105% Rod.Line - Natural Circulation Power)
Channel Hydrodynamic Performance Decay Ratio, x2/x o (105% Rod Line - Natural '
Circulation Power) 8x8R/P8x8R/PTA Channel 0.31
-8x8 Channel 0.38
- 14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
See " Loss of Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 3," NEDO-24082, Addendus 3, February 1981.
- 15. LOADING IRROR.RESULTS (5.5.4)
Limiting Event: Rotated Bundle P8DRB299: MCPR > 1.07
, 5
i Y1003J01A20 Rsv. 0
- 16. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)
Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Function: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 '
Plant Specific Analysis Results Parameters Not Bounded:
Accident Reactivity Shape Functions: Cold and Hot Startup Scram Reactivity Functions: Cold and Hot Startup Resultant Peak Enthalpies (cal /g):
Cold Hot Startup 220 243
- 17. REFERENCES
- 1. " General Electric Belling Water Reactor Generic Reload Fuel Applicacion," July 1979, (NEDE-24011-P-A-1).
, 2. " General Electric Boiling Water Reactor Generic Reload Fuel Application," May 1979, (NEDE-24011-P-A, Amendment 3).
- 3. " General Electric Boiling Water Reactor Generic Reload Fuel Application," July 1980, (NEDE-24011-P-A, Amendment 8).
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Y1003J01A20 Rev. 0 100 50 90 - C - 679 CRD IN PERCENT 1 - NOMINAL SCRAM CURVE IN (-$)
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. . _ . . = _ _ - . . . ---
l Y1003J01A20 Rev. 0 i
02 06 10 14 18 22 26 30 i
l 59 12 55 12 31 4 32 47 12 8 43 4 32 36 39 12 8 12 35 4 32 36 0 31 12. 8 NOTES: 1.- ROD PATTERN'IS 1/4 CORE SYMMETRIC.
UPPER LEFT QUADRANT SHOWN ON MAP.-
- 2. NO. INDICATES NUMBER OF. NOTCHES' WITHDRAWN
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OUT 0F 48. BLANK IS A WITHDRAWN ROD.
- 3. .ERRORIROD IS-(30,35).
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Y1003J01A20 Rev. 0 1.2 ULTIMATE PERFORMANCE t 1.0 x== - ====m= - - - - - - - ==. i M IT ==---==============
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-35 ' ' ! ' I O 500 1000 1500 M 2500 3000 FUEL TEMPERATURE (*Cl l
l Figure 9. Doppler Reactivity Coefficient Comparison for RDA 1
17
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Y1003J01A20 Rev. 0 20 I
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O ACCIDENT FUNCTION e soVNOING VALUE 280 CAL /G I I g 0L to 16 20 ROD POSITION (ft CUT)
Figure 10. RDA Reactivity Shape Function at 20*C 18
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Y1003J01A20 Rev. O I
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Figure 11. RDA Reactivity Shape Function at 286*C 19
Y1003J01A20 Rev. 0 30 O SCRAM FUNCTION 9 BOUNDING VALUE 280 CAL /G 25 -
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Figure 12. RDA Scram Reactivity Function at 20*C 20
Y1003J01A20 Rev. O I
50 O SCRAM FUNCTION 4 BOUNDING VALL:1280 CAUG 40 -
l 3
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Figure 13. RDA Scram Reactivity Function at 286*C 21/22
Y1003J01A20 Rev. O APPENDIX A ODYN TRANSIENT CODE All rapid pressurization and over-pressure protection events have been analyzed using the ODYN transient code as specified in Reference A-1. Ccde over-pressure protection analysis results are deterministic as discussed in Reference A-2. The 4CPR values given for the pressurization events in Section 9 are the plant-specific deterministic values calculated by ODYN based on the initial MCPE given in Item 7 of this submittal. These aCPRs may be adjusted te reflect either Option A or Option B ACPRs by employing the con-version method described in Reference A-2. These adjustments are based on conservatism factors applied to the ratio ACPR/ICPR. The MCPR for the event is determined by adding the ACPR to the safety limit. Section 11 presents both the MCPRs for the non-pressurization events as well as the adjusted MCPRs (Option A and Option B) for the pressurization events.
The operating limit MCPR is the maximum MCPR of the following events:
- 1. Turbine trip ce load rejection without bypass based on ODYN.
- 2. Feedwater controller failure event based on ODYN.
- 3. Loss of feedwater heating event.
- 4. Rod withdrawal error event.
- 5. Bundle loading error accident.
- 6. Minimum required by LOCA.
- 7. Minimum required by Reference A-3, Appendix C, Page C-65.
23
Y1003J01A20 Rev. O where Items 3 thru 7 are calculated as described in Referenca A-3 but the MCPRs for the pressurization events analyzed with ODYN have been adjusted 4
as follows: .
I
- 1. MCPRs era adjusted for Option B for all plants choosing to operate under Option B which meet all scram specifications given in Reference A-4.
- 2. MCPRs are determined by a linear interpolation between the Option A MCPR and the Option 3 MCPR for all plants choosing to operate under Option B which do not meet the scram time specification. This interpolation is based on the tested measured scram time and is described in Reference A-4.
l REFERENCES l
A-1. Letter, R. P. Denise (NRC) to G. G. Sherwood (GE),
January 23, 1980.
A-2.- Letter, R. H. Buchholz (GE) to P. S. Check (NRC), "0DYN Adjustment
- Method for Determination of Operating Limits,"
January 19, 1981.
A-3.- " Generic Reload Fuel Application," NEDE-24011-P-A-1, August 1979.
A-4 Letter (with attacnment), R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request for Information on ODYN Computer Model," September 5,1980.
i 24
Y1003J01A20 Rev. O APPENDIX B FAST SCRAM CONTROL R05) DRIVE The fast scram control rod drive (FSCRD) was described in NED0-21363-2. In addition to describing the FSCRD, the licensing document provid(1 the results of a safaty review and evaluation which concluded that the inclusion of the FSCRD does not introduce an unreviewed safety question and has no effect on parameters used in the plant safety analysis. This safety review and evalu-acion is also applicable for operation of the currently installed FSCRD (S/N ,
7464) during cycle 5 operatien.
NED0-21363-2A, " General Electric Soiling Water Reactor Reload 1 Licensing Amend-ment for Peach Bottom Atomic Power Station Unit 3 Fast Scram Control Rod Drive, Second Supplement,d 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 provides performance results and a report of the effects of the reactor environment on the drive mechanism. A safety evaluation
-(Reference 3-1) is also provided which demonstrates that continued operation of the currently installed FSCRD, during cycle 5, does not introduce an-u.% reviewed safety question and has no adverse effect on parameters used in the plant safety analysis.
~
REFERENCES B-1.. General Electric Boiling Water Reactor Reload 1 Licensing Amendment for Peach Bottom Atomic Power Station Unic Number 3, Fast Scram Control Rod Drive Third Supplement, February 1981 (NEDO-21363-3).
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Y1003J01A20 Rev. 0 ,
1 APPENDIX C NEW BUNDLE LOADING ERROR EVELT 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 rotated bundle and the mislocated bundle loading error events. The use of these new analyses proce-dures is discussed below.
C.1 NEW ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT !
l The rotated bundle loading error event analysis results presented in this supplement are based on the new analysis procedure described and approved in Reference C-1. This new method of performing the analysis is based on a more accurate detailed analytical model.
The principal dif ference 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 utill:es 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 tha 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 model: as were previously used. The only change is in the simu-lation of the water gap, which more accurately represents the actual geometry. l The results of the analysis indicate for the P8x8R bundle a 15.3 kW/ft LHGR and a 0.13 ACPR'(includes a 0.02 penalty due to variable water gap
- R-factor uncertainty) with a minimum CPR of >1.07.
C.2 NEW ANALYSIS PROCEDURE FOR THE MISLOCN"ED BUNDLE LOADING ERROR EVENT The mislocated bundle loading error event analysis is no longer being reported, i as discuscad in Reference C-2.
1 27 f
_- _- . . . _ . _ . . . _ . _ _ . - - . - . . ~. . . -_. . -__ _ . -. _
- i Y1003J01A20 Rev. 0 REFERENCES 1
i C-1. Safety Evaluation Report (letter), D. G. Eisenhut (NRC) to R. E. Engel (GE), MFN-200-78, dated May 8,1978.
C-2. Letter, R. E. Engel (GE) to T. A. Ippolito (NRC), " Change
, in General Electric Methods for Analysis of Mislocated Bundle Accident," November 14, 1980.
I l
l l
4 4
4 4
f 28
r . .
I I
Y1003J01A20 Rsv. O i,-
r f -' APPENDIX D GETAB ANALYSIS INITIAL CONDITIONS i
j See'." General. Electric Boiling Itater Reactor Generic Reload Fuel Application,"
- j. ' July 1979, (NEDE-24011-P-A, Amendment 8). ,
4
'N
~
+
,, _29/30:
o E*... O+
....-w..,
Y1003J01A20 Rev. O APPENDIX E MARGIN TO SPRING SAFETY VALVES 1
The basis for providing pressure margin to the lowest setpoint of the unpiped spring safety valves conforms to the General Electric Company operational recommendation as described in Reference E-1. NRC acceptance of this criteria was documented in Reference E-2.
REFERENCES E-1. Letter, J. F. Quirk (GE) to Olan D. Parr (NRC), " General Electric Licensing Topical Report NEDE-24011-P-A, ' Generic Reload Fuel i
Application,' Appendix D, Second Submittal," dated February 28, 1979.
E-2. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 42 to Facility Operating License No. DPR-25, Commonwealth Edison Company, Dresden Nuclear Power Station, Unit No. 3, Docket No. 50-249.
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Y1003J01A20 Rev. O APPENDIX F PRESSURIZED TEST ASSEMBLY FUEL ROD REPLACEMENT F.1 INTRODUCTION The Pressurized Test Assembly (PTA) was described in NEDO-21363-4, Supplement 4 January 1977.
As a continuation of this program, during the Reload No. 4 refueling outage, up to six fuel rods will be removed from the PIA and replaced with fresh rods with a U-235 enrichment of 2.00 weight percent. The removed rods will be examined and punctured for fission gas pressure measurement. These rods vill not be used during future operation. The enrichment of the replacement rods is less than the initial enrichment of the rods they are replacing to compensate for fuel depletion and was selected to assure that the reactivity of the recon-stituted PTA will not exceed that of a non-reconstituted PTA. Therefore, the information contained in NEDO-21363-4, Supplement 4, is unaffected by the fuel rod replacements of the PTA since the nuclear characteristics of the reconsti-tuted bundle are essentially identical to a non-reconstituted bundle.
The purpose of this appendix is to report the results of the analyses and safety evaluation for operation of the reconstituted PTA during cycle 5.
F.2 EVALUTIONS AND ANALYSES F.2.1 Nuclear and Thermal Parameter Evaluations 9
Standard lattice physics calculations were made for the reconstituted PTA including simulation of the fresh rods. Cycle 5 operation was simulated by
" burning" the reconstituted PTA in 10 exposure steps.
Over the exposure range of interest, the computed lattice reactivity of the reconstituted PTA is within 0.4% AK of the non-reconstituted PTA reactivity.
The. fuel rod power _ peaking for the reconstituted PTA remains low, but is up to 2% greater than the value for the non-reconstituted PTA at the same exposure. Although the reduced gap conductance in the six fresh rods tends i
i I
33
Y1003J01A20 Rav. O to reduce the change in critical power ratio (aCPR) due to transients, which tends to reduce the MCPR operating limit, the effect of the small increase in fuel rod power peaking results in an overall slight decrease in the steady-state MCPR of the reconstituted PTA. To account for the slight decrease in steady-state MCPR, and the small increase in local peaking, the R-factors and local peaking fac ors for the PTA will be increased to account for the 2%
greater fuel rod power peaking factor. The process computer will compute the steady-state actual CPR and LHGR associated with the reconstituted PTA.
Using the BWR Simulator, power distribution analyses have been performed for Cycle 5 of Peach Bottom 3. The cote exposure distribution at beginning of Cycle 5 was obtained by simulating the estimated exposure accumulation to the end of Cycle 4 and representing the projected refueling and fuel rod replace-ment. These analyses show that the thermal limits for fuel assemblies su -
rounding the reconstituted PTA are significantly more limiting than for the reconstituted PTA. Estimated margins between the reconstituted PTA and the surrounding fuel assemblies are:
MCPR 3% to 12%
MAPLHGR 11% to 18%
Linear Heat Generation Rate 7% to 14%
Due to the relatively high exposure of the PTA, the neighboring fuel assemblies will operate closer to limits, and the PIA is predicted never to be limiting.
MCPR limits during transients are not affected because the PTA is predicted never to be limiting.
F.2.2 Mechanical Design Evaluation The six replacement- fuel rods in the PTA are mechanical?y sbnilar to the fuel rods which they are replacing and also to the standard fuel rods in the Reload 4 fuel bundles. The only mechanical difference is a longer upper end plug on each replacement rod to accommodate the irradiation growth of the rods in the PTA.
An analysis of differential roJ growth in the reconstituted PTA was performed.
34
YlD03J01A20 Rev. 0 The results of the analysis show that the replacement fuel rods are mechanically compatible with the irradiated rods and thus will have no adverse effect on the safety analyses for Cycle 3 or subsequent cycles for Peach Bottom 3.
The peak linear heat generation race of the reconstituted PTA is still within the operating limit of 13.4 kW/f t which was used in evaluating the mechanical perfctmance of the maximum duty fuel rod in Reload 4. Therefore, the results of the fuel rod thermal and mechanical design evaluations in NED0-21363-4, Supplement 4, are conservatively applicable to the reconstituted PTA.
F.2.3 Evaluation of the Effect of the ?resh Fuel Rods on PCT /MAPLHCR Reconstitution of the PTA will re.9 ult in a reduction of the planar average ex-posure of the assembly compared to that assumed in the analysis of the original PTA (NEDO-21363-4, Supplement 4, January 1977).
The following effects of the chaage in the exposure on LOCA analysis provide the basis for assessing the effect on Peak Clad Temperature:
- 1. The local power distribution in the center 16 rods is decreased.
This occurs because the fresh rods redistribute the power to the periphery of the bundle.
- 2. The calculated total stored energy is increased, This occurs because the calculated gap conductance at low exposures is generally mnailer, and since the reconstituted bundle has a lower effective exposure compared to the reconstituted PTA, the calculated stored energy would be higher compared to that calculated in the LOCA analysis for the non-reconstituted PTA. The maximum increase in total planar stored energy is approximately 1* for all planar exposures.
The resulting decrease in PCT due to the above effects is less than 200F.
Since the PCT of the non-reconstituted PIA is well below 22000F, the PCT of the reconstituted PTA will also remain below 22000F. Thus, the previously calculated MAPLHGRs for the PTA given in NED0-24082, " Loss-of-Cooiant Accident Analysis for Peach Bottom Atomic Power Station Unit 3, Addendum 2," June ;980, are valid for the reconstituted PTA.
35 9
Y1003J01A20 Rav. O F.2.4 Transient Analysis for Cycle 5 Based on the analysis results described in Section F.2.1 above, the transient analysis results contained in this submittal are unaffected by fuel rod replacement of the PTA.
F.3
SUMMARY
AND CONCLUSIONS It is concluded, based on the results of the evaluations and analysis described in Section F.2 that the accident and transient analyses of Cycle 5 are insigni-ficantly affected and the operating limits of Cycle 5 are unaffected by the intro-duction of the reconstituted PTA. The operating MCPR limit is given in Section 11 of this submittal.
s s
j 36 1
Y1003J01A20 Rsv. O APPENDIX G INITIAL MCPR 1te initial MCPR assumed in the calculation of the ACPR for the generator load rejection (no bypass) at EOC-2000 mwd /t for the PTA/P8x8R fuel is 0.01 below the operating limit MCPR. Justification for this is given in Referec.ce G-1.
Reference G-1: "Ge neral Electric Boiling Water Reactor Generic Reload Fuel Application," January 1979 (NEDE-24011-P-A, Amendment 2).
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! (FINAL) e s