ML20054M080
| ML20054M080 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 06/30/1982 |
| From: | Engel R, Leaser J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20054M057 | List: |
| References | |
| Y1003J01A20, Y1003J01A20-R01, Y1003J1A20, Y1003J1A20-R1, NUDOCS 8207090224 | |
| Download: ML20054M080 (41) | |
Text
-
Y1003J01 A20
""5'Jr"si JUNE 1982 l
SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM ATOMIC POWER STATION UNIT 3, RELOAD NO. 4 1
(i f
l "s?T88PSS85l2e GENER AL h ELECTRIC
I Y1003J01A20 Revision 1 Class I
- June 1982 F
SUPPLEMENTAL RELOAD LICENSING SUBMITTAL l FOR PEACH BOTTOM ATOMIC POWER STATION '
UNIT 3, RELOAD NO. 4 i
l 4
i i
i I
Prepared by: O AMV
, J. D. Leaser Senior Licensing Engineer i _
Approved: _ R. E. Enge[, Manager Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENER AL $ ELECTRIC i
L ____ ____
Y1003J01A20 Rev. 1 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 Regulatory 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 en accurate and true' representation j of the facts known, obtained or provided to' General Electric at the time i this report was prepared.
The only undertakings of the General Electric Company respecting informa-tion in this document are contained la 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 l
changing said contract. The use of this information except as defined by said contract, or for any purpose cther 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.
I I
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Y1003J01A20 R:v. 1
- 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 Appendix D - GETAB Analysis Initial Conditions Appendix E - Margin to Spring Safety Valves Appendix F - Pressurized Test Assembly Fuel Rod Replacement Appendix G - Pressurization Transient Reanalysis
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1, and 4.0)
Fuel Type Number Number Drilled Irradiated 8DB274H 23 23 PTA 1 1 8DRB283 252 252 P8DRB284H** 272 272 New P8DRB299*** 216 216 Total 764 764
- 3. REFERENCE 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.
- Described in Reference 3.
1
Y1003J01A20 Riv. 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 1.115 Fully Controlled 0.960 Strongest Control Rod Out 0.987 R, Maximum Increase in Cold Core Reactivity with Exposure into Cycle, Ak 0.000
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak)
EEm (20*C, Xenon Free) 660 0.04
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)
EOC5 E0C5-2000 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 Fuel Temperature (*F) 1391 1391 Scram Worth N/A** (S) -46.31/-37.05 -46.31/-37.05 Scram Reactivity vs Time Figure 2 Figure 2 i
- Mid Cycle Exposure Point
- N = Nuclear Input Data A = Used in Transient Analysis 2
~
Y1003J01A20 Rsv. 1
- 7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 8x8 8x8R PTA/P8x8R EOC5- EOC5- EOC5- EOC5- EOC5- EOC5-l 1000 2000 1000 2000 1000 2000
_EOC5 mwd /t mwd /t EOCS mwd /t mwd /t EOC5 mwd /t mwd /t Peaking factors 1.22 1.22 1.22 1.20 1.20 1.20 1.20 1.20 1.20 (local, radial 1.30 1.32 1.30 1.42 1.44 1,42 1.39 1.41 1.39 l and axial) 1.40 1.40 1.40 1.40 1.40 1.40 1.40 1.40 1.40 R-factor 1.098 1.098 1.098 1.051 1.051 1.051 1.051 1.051 1.051 Bundle Power 5.469 5.553 5.475 5.986 6.074 5.983 5.850 5.943 5.859 (MWt)
Bundle Flow 113.2 112.6 113.1 112.7 112.1 112.7 113.5 112.9 113.4 (103 lb/hr)
Initial MCPR 1.36 1.34 1.36 1.37 1.35 1.37 1.40 1.38 1.40
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)
Exposure Dependent Limits: From B0C5 to EOC5-2000 mwd /t From EOC5-2000 mwd /t to E0C 5-1000 mwd /t From E0C5-1000 mwd /t to E0C5
- 9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)
Core . .
p p Power Flow e Q/A si v PTA & Plant Transient Exposure (t) (t) (t NBR) (t NBR) (psig) (psig) 8x8 8x8R P8x8R Response Cenerator Load EOC5 104.5 100 831 130 1244 1265 0.29 0.29 0.32 Figure 3a Rejection EOC5-1000 mwd /t 104.5 100 878 128 1238 1258 0.27 0.27 0.31 Figure 3b No Bypass EOC5-2000 Wd/t 104.5 100 <781 <126 <1231 <1255 <0.22 <0.22 <0.25 Figure 3e
- toss of 100*F BOC5-EOC5 104.5 100 125 124 1014 1069 0.15 0.16 0.16 Figure 4 Feedwater Heating Feedwater EOCS 104.5 100 <418 <126 <1167 <1208 <0.24 <0.24 <0.26 Figure 5a
- Controller EOC5-1000 mwd /t 104.5 100 386 124 1160 1206 0.19 0.19 0.21 Figure 5b Failure EOC5-2000 mwd /t 104.5 100 358 121 1156 1197 0.16 0.16 0.18 Figure 5e
- See Appendix G.
3
t Y1003J01A20 Rgv. 1
- 10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
(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 -
16.3 107** 6.0 -
0.13 -
16.4 108 7.5 -
0.14 -
16.4 109 12.0 -
0.17 -
16.4 4
110 12.0 -
0.17 -
16.4
- 11. CYCLE MCPR VALUES (5.2)
BOC5 to E0C5-2000 mwd /t Option A Option B PTA/ PTA/
Pressurization Events 8x8 8x8R P8x8R 8x8 8x3R P8x8R Generator Load Rejection, 1.35 1.35 1.38 1.14 1.14 1.17 ***
No Bypass Feedwater Controller 1.28 1.28 1.30 1.22 1.22 1.24 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 EOC-1000 mwd /t Option A Option B PTA/ PTA/
Pressurization Events 8x8 8x8R P8x8R 8x8 8x8R P8x8R Generator Load Rejection, 1.40 1.40 1.44 1.18 1.18 1.21 No Bypass Feedwater Controller 1.32 1.32 1.34 1.25 1.25 1.27 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
- 8x8 fuel is not limiting; therefore its response to rod withdrawal error is not given.
- Indicates setpoint selected.
- See Appendix G. 4
. Y1003J01A20 Rzv. 1 E0C5-1000 mwd /t to EOC5 Option A Option B PTA/ PTA/
Pressurization Events 8x8 8x8R P8x8R 8x8 8x8R P8x8R Generator Load Rejection 1.42 1.42 1.45 1.30 1.30 1.32 No Bypass ,
Feedwater Controller 1.37 1.37 1.39 1.30 1.30- 1.31**_,___ _
Failure Non-Pressurization Events Loss of 1000F Feedwater 1.22 1.23 1.23 1.22 1.23 1.23 lleating 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
- 12. OVERPRESSURIZATION ANALYSIS
SUMMARY
(5.3)
Power Core Flow PSL Py Plant
(%) (%) (psig) (psig) Response MSlV Closure 104.5 100 <1265 <1297 Figure 7 **
(Flux Scram)
- 13. STABILITY ANALYSIS RESULTS (5.4)
Decay Ratio: Figure 8 Reactor Core Stability Decay Ratio, x2 X/ o: 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
- 8x8 fuel is not limiting; therefore its response to rod withdrawal error is not given.
- See Appendix G.
5
f Y1003J01A20 Rzv. I
- 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, Addendum 3, February 1981.
- 15. LOADING ERROR RESULTS (5.5.4)
Limiting Event: Rotated Bundle P8DRB299 MCPR > 1.07
- 16. CONTROL R0D 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 Boiling Water Reactor Generic Reload Fuel Application," July 1979, (NEDE-240ll-P-A-1).
- 2. " General Electric Boiling Water Reactor Generic Reload Fuel Application," May 1979, (NEDE-240ll-P-A, Amendment 3).
- 3. " General Electric Boiling Water Reactor Generic Reload Fuel Application," July 1980, (NEDE-240ll-P-A, Amendment 8).
o 6
Y1003J01A20 Rev. 1
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Y1003J01A20 Rev. 1 100 50 90 - C - 678 CRO IN PERCENT 1 - NOMIN AL SCRAM CURVE IN (-$)
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, ', Y1003J01A20 Rev. 1 w
02 06 10 14 18 22 26 30 59 12 55 12 51 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
_ Olif 0F 48. BLANK IS A WITHDRAWN ROD.
. 3. ERROR R0D IS (30,35).
3- Figure 6. Limiting RWE Rod Pattern 6 4
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Y1003301A20 Rev, 1 1.2 ULTIMATE PERFORMANCE LIMIT 1,0 - - . -mm - - - --=== - - - - - - - - - 0.8 - NATURAL CIRCU LATIO N x 5 o E 0.6 - 105% ROD LINE a S O 0.4 - 0.2 - i I I I o O 20 40 60 80 100 POWER (%) Figure 8. Decay Ratio 18
Y1003J01A20 Rev. 1 0
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l O CALCULATED VALUE - COLD 9 CALCULATED VALUE - HSB O BOUND VAL 280 CAL /G COLD
--30 8 BOUND VAL 280 CAL /G HSB I ! I I
-35 l 0 500 1000 1500 2000 2500 3000 FUEL TEMPERATURE ( C) l Figure 9. Doppler Reactivity Coefficient Comparison for RDA 19
Y1003J01A20 Rev. 1 20 15 - r -C D 2 6
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=
N 10 - d 5 G E a 5 - O ACCIDENT FUNCTION
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0u O 5 10 15 20 ROD POSITION (ft OUT) Figure'10. RDA Reactivity Shape Function at 20*C I 20
Y1003J01A20 Rev. 1 20 15 -
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t 2 0 5 e 5 - O ACCIDENT FUNCTION e BOUNDING VALUE 280 CAL /G l OL o 5 10 15 20 ROD POSITION (ft OUT) Figure 11. RDA Reactivity Shape Function at 286*C 21
Y1003J01A20 Rev. 1 30 O SCRAM FUNCTION e BOUNDING VALUE 280 CAL /G 25 - _ 20 - 5 E e w
*3 i
g 15 - t 2 0 5 m 10 - 5 - I ! I OC C T O 1 2 3 4 5 6 ELAPSED TIME (sec) Figure 12. RDA Scram Reactivity Function at 20*C 1 22
Y1003J01A20 Rev. 1 so O SCRAM FUNCTION
$ BOUNDING VALUE 280 CAL /G 40 -
' l; E 30 - 8 s V 4 1 5 b 4 y 20 - 10 - s l- e T I l 1 2 3 4 5 6 ELAPSED TIME (sec) Figure 13. RDA Scram Reactivity Function at 286 C 23
Y1003J01A20 R v. 1 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. Code over-pressure protection analysis results are deterministic as discussed in Reference A-2. The ACPR values given for the pressurization events in Section 9 are the plant-specific deterministic values calculated by ODYN based on the initial MCPR given in Item 7 of this submittal. These ACPRs may be adjusted to reflect either Option A or Option B ACPRs by employing the con-version method described in Reference A-2. These adjustments are based en 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 or load rejection without bypass based on ODYN.
- 2. Feedwater controller failure event based on ODYN.
- 3. Loss of feedwater hearing 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.
24
Y1003J01A20 Rev. 1 where Items 3 thru 7 are calculated as described in Reference A-3 but the MCPRs for the pressurization events analyzed with ODYN have been adjusted as follows:
- 1. MCPRs are adjusted for Option B for all plants choosing to operate under Option B which meet all scram specifications given in Ref erence A-4.
- 2. MCPRs are determined by a linear interpolation between the Option A MCPR and the Option B 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.
REFERENCES 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), "ODYN Adjttstment 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 attachment), R. H. Buchholz (GE) to P. S. Check (NRC), " Response to NRC Request for Information on ODYN Computer Model," September 5, 1980. l 25
Y1003J01A20 Rtv. l APPENDIX B FAST SCRAM CONTROL ROD DRIVE The fast scram control rod drive (FSCRD) was described in NED0-21363-2. In addition to describing the FSCRD, the licensing document provided the results of a safety 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-ation is also applicable for operation of the currently installed FSCRD (S/N 7464) during cycle 5 operation. NEDO-21363-2A, " General Electric Boiling Water Reactor Reload 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 provides performance results and a report of the effects of the reactor environment on the drive mechanism. A safety evaluation (Reference B-1) is also provided which demonstrates that continued operation of the currently installed FSCRD, during cycle 5, does not introduce an unreviewed 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 Unit Number 3, Fast Scram Control Rod Drive Third Supplement, February 1981 (NEDO-21363-3). 26
Y1003J01A20 Rev. 1 APPENDIX C 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 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 The rotated bundle loading error event analysis results pre ;ented 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 b,ned on a more accurate detailed analytical model. l 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 water gap, which more accurately represents the actual geometry. 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 MISLOCATED BUNDLE LOADING ERROR EVENT The mislocated bundle loading error event analysis is no longer being reported, as discusced in Reference C-2. 27
Y1003J01A20 R v. 1 REFERENCES 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," Noveraber 14, 1980. l 28
Y1003J01A20 Rev. 1 APPENDIX D GETAB ANALYSIS INITIAL CONDITIONS j See " General Electric Boiling Water Reactor Generic Reload Fuel Application," July 1979, (NEDE-24011-P-A, Amendment 8). 29/30
i . 1 . l Y1003J01A20 Rsv. 1 1 l l l APPENDIX E MARGIN TO SPRING SAFETY VALVES The basis for providing pressure margin to the lowest setpoint of the unpiped I 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-240ll-P-A, ' Generic Reload Fuel 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, 31/32 l
( .
- Y1003J01A20 Rzv. 1 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 PTA 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 will not be used during future operation. The enrichment of the replacement rods t 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 NED0-21363-4, 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 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 33
Y1003J01A20 . R v. 1 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 factors 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 i PTA. Using the BWR Simulator, power distribution analyses have been performed for Cycle 5 of Peach Bottom 3. The core 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 sur-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 PTA 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 mechanically similar 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 rod growth in the reconstituted PTA was performed. 34 ____- o
YJ003J01A20 Rev. 1 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 5 or subsequent cycles for Peach Bottom 3. The peak linear heat generation rate of the reconstituted PTA la still within the operating limit of 13.4 kW/ft which was used in evaluating che mechanical performance 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. l F.2.3 Evaluation of the Effect of the Fresh Fuel Rods on PCT /MAPLHCR Reconstitution of the PTA will result in a reduction of the planar average ex-posure of the assembly compared to that assumed in the analysis of the original PTA (NED0-21363-4, Supplement 4, January 1977). The following effects of the change 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.
l 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 smaller, 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 PTA 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 NEDO-24082, " Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 3, Addendum 2," June 1980, are valid for the reconstituted PTA. 35
R2v. 1 Y1003J01A20 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. 36
Y1003J01A20 Rev. 1 APPENDIX G PRESSURIZATION TRANSIENT REANALYSIS The transient results given in Sections 9 and 11 represent a reanalysis of the generator load rejection and the feedwater controller failure events. This reanalysis was performed to correct an error in the fuel length input to the transient code used to obtain the results given in Revision 0 of this document. The feedwater controller failure event was not reanalyzed for the EOC5 condi-tion since the original calculations demonstrate that this event is not limiting at that condition. Therefore, the small change in ACPR for this event that would result from a reanalysis would not affect the operating limit MCPR. The results shown for feedwater controller failure at EOCS were obtained conservatively by adding the differences between the original and revised analysis values (fiuxes, pressures, and ACPR's) for the EOC5 load rejection event. The ACPR resuits shown for the generator load rejection without bypass event at the EOC5-2000 mwd /t condition were obtained by a conservative extrapolation of the difference in the results between the original and the revised calcula-tions for this event at the E0C5 condition. The flux and pressure values shown were obtained conservatively by adding the differences between the original and revised analyses for the EOC5 load rejection event. The pressure values reported for the MSIV closure (flux scram) event in Section 12 were obtained by using a conservative pressure adder that was derived from a sensitivity study on fuel length variation. 37/38 (FINAL)
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