ML20100F045
| ML20100F045 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/31/1985 |
| From: | Charnley J, Dennison D, Lambert P GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20100E974 | List: |
| References | |
| 23A4660, 23A4660-R01, 23A4660-R1, NUDOCS 8504040193 | |
| Download: ML20100F045 (25) | |
Text
_ . ,
0 REVIS ON MARC 1985 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM l ATOMIC POWER STATION UNIT 3, RELOAD 6 pgg4oigas!
p 2888670 PDP GEN ER AL h ELECTRIC
23A4660 Revision 1 Class I March 1985 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM ATOMIC POWER STATION UNIT 3 RELOAD 6 1
Prepared: / [ -
A-o . Ms P. A. Lambert Fuel Licensing Verified: ~
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D. K. Dennison Fuel Licensing
//
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.-- j j S. Cli5rnley, Mana
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NUCLEAR ENERGY RUSINESS OPERATIONS
- GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 95125 GENERAL $ ELECTRIC 1/2
~ '
23A4660 Rev. 1 IMPORTANT NOTICE RECARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY 'g This report was prepared by General Electric solely for Philadelphia Electric Company (PECo) for PECo's use with the U.S. Nuclear Ragulatory Commissica (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 known, obtained or provided to General Electric at the time this report was prepared.
The only undertakings of the General Electric Company respecting infor-mation in this document are contained in " Contract between Philadelphia Electric Company and General Electric Company for Fuel Bundles and Services for Reload Fuel Supply for Peach Bottom Atomic Power Station Units 2 and 3, ,
dated November 16, 1979," 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 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 contained 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.
3/4
23A4660 Rev. I
- 1. PLANT UNIQUE ITEM (1.0)*
Appendix A: GETAB Analysis Initial Conditions
~
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.01 Fuel Type Cycle Loaded Number Number Drilled Irradiated P8DRB299 5 196 196 P8DRB28411 6 56 56 P8DRB299 6 224 224 PBLTAl 6 2 2 PBLTA2 6 2 2 New BP8DRB29911 7 140 140 BP8DRB299 7 144 144 Total 764 764
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end 18562 mwd /ST of cycle:
Minimum previous cycle core average exposure at end 18562 mwd /ST of cycle from cold shutdown considerations:
Assumed reload cycle core average exposure at end of 18089 mwd /ST cycle:
l l Core loading pattern: Figure 1 h s
- ( ) Refers to area of discussion in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-6. A letter "S" preceding the number refers to the U.S. supplement.
1 5
l 23A4660 -
Rdv. 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)
Beginning of Cycle, K-ef fective Uncontrolled 1.119 Fully Controlled 0.963 Strongest Control Rod Out 0.987 R, Maximum Increase in Cold Core Reactivity with 0.003 Exposure into Cycle, Delta K
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Ak) ppm (20*C, Xenon Free) 660 0.035
- 6. RELOAD UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)
(REDY EVENTS ONLY)
EOC EOC-2000 mwd /ST Void Fraction (%) 39.8 39.1 Average Fuel Temperature (*F) 1279 1281 Void Coefficient N/A* (c/% Rg) -8.50/-10.63 -9.59/-11.99 Doppler Coefficient N/A* (c/*F) -0.222/-0.211 -0.214/-0.203 Scram Worth N/A* ($) ** **
- N = Nuclear Input Data A = Used in Transient Analysis
- Generic exposure independent values are used as given in " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A-6, April 1983.
l 6
23A4660 Rev. 1
- 7. RELOAD UNIQUE CETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)
Fuel Peaking Factors Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR BOC 7 to EOC 7-2000 mwd /ST BP/P8x8R 1.20 1.58 1.40 1.051 6.641 108.4 1.22 PBLTA1 1.25 1.46 1.40 1.100 6.154 122.1 1.21 PBLTA2 1.25 1.44 1.40 1.100 6.041 113.6 1.21 EOC 7-2000 mwd /ST to EOC 7 BP/P8x8R 1.20 1.47 1.40 1.051 6.176 111.4 1.32 PBLTAl 1.25 1.33 1.40 1.100 5.590 125.7 1.33 PBLTA2 1.25 1.31 1.40 1.100 5.519 117.1 1.32
- 8. SELECTED MARGIN IMPROVEMENT OPTIONS (S.2.2.2)
Transient Recategorization: No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: No Improved Scram Time: Yes (ODYN Option B)
Exposure Dependent Limits: Yes Exposure Points Analyzed: EOC EOC-2000 mwd /ST
- 9. OPERATING FLEXIBILITY OPTIONS (S.2.2. 3)
Single Loop Operation: Yes Load Line Limit: No Extended Load Line Limit: No Increased Core Flow: No Flow Point Analyzed: N/A Feedwater Temperature Reduction: No 7
23A4660 -
Re'v. 1
(
- 10. CORE-WIDE TRANSIENT ANALYSIS RESULTS (S.2.2.1) 0 Flux Q/A Transient (%NBR) (%NBR) BP/P8x8R PBLTA1 PBLTA2 Figure Exposure: BOC to EOC-2000 mwd /ST Load Rejs etion w/o Bypass 542 121 0.19 0.18 0.18 2a Loss of 100*F Feedwater Heating 124 123 0.15 0.14 0.14 3 Feedwater Controller Failure 242 118 0.11 0.10 0.10 4a Exposure: EOC-2000 mwd /ST to EOC Load Rejection w/o Bypass 647 127 0.25 0.26 0.25 2b Loss of 100*F Feedwater Heating 124 123 0.15 0.14 0.14 3 Feedwater Controller Failure 320 124 0.18 0.18 0.17 4b
- 11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)
SUMMARY
(S.2.2.1)
Limiting Rod Pattern: Figure 5 Rod Block ( ee Reading withdrawn) BP/P8x8R PBLTA1 PTLBA2 BP/P8x8R PBLTA1 PTLBA2 104 4.5 0.16 0.16 0.16 14.99 14.99 14.99 105 5.0 0.18 0.18 0.18 14.99 14.99 14.99 106 5.5 0.19 0.19 0.19 14.99 14.99 14.99 107* 6.0 0.20 0.20 0.20 14.99 14.99 14.99 108 9.0 0.26 0.26 0.26 14.99 14.99 14.99 109 10.0 0.27 0.27 0.27 16.26 16.26 16.26 110 12.0 0.32 0.32 0.32 18.37 18.37 18.37
- Indicates Setpoint selected 8
L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
23A4660 Rev. 1
- 12. CYCLE MCPR VALUES (S.2.2)
Non-Pressurization Events BP/P8x8R PBLTA1 PBLTA2 Exposure Range: BOC to EOC Loss of 100*F Feedwater Heating 1.22 1.21 1.21 Fuel Loading Error 1.22 -- --
Rod Withdrawal Error 1.27 1.27 1.27 Pressurization Events Option A Option B BP/P8x8R PBLTAl PBLTA2 BP/P8x8R PBLTAl PBLTA2 Exposure Range:
BOC to EOC-2000 mwd /ST Load Rejection w/o Bypass 1.32 1.31 1.31 1.11 1.11 1.11 Feedwater Controller Failure 1.23 1.22 1.22 1.17 1.16 1.16 Exposure Range:
E0C-2000 mwd /ST to EOC Load Rejection w/o* Bypass 1.38 1.39 1.38 1.26 1.27 1.26 Feedwater Controller Failure 1.31 1.31 1.29 1.24 1.24 1.23
- 13. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3) sl v Transient (psig) (psig) Plant Response MSIV Closure 1242 1271 Figure 6 (Flux Scram) 9
23A4660 Rev. 1
- 14. STABILITY ANALYSIS RESULTS (S.2.4)
Rod Line Analyzed: 105% Rod Line Decay Ratio: Figure 7 Reactor Core Stability Decay Ratio, x2 /*0 0.86 Channel Hydrodynamic Performance Decay Ratio, x2 /*0 Channel Type BP/P8x8R 0.29 PBLTAl 0.13 PBLTA2 0.26
- 15. LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Misoriented Bundle Analysis: Yes Event Initial CPR Resulting CPR Rotated Bundle Error 1.20 1.07
- 16. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
Bounding Analysis Results:
Doppler Reactivity Coefficient: Figure 8 Accident Reactivity Shape Functions: Figures 9 and 10 Scram Reactivity Functions: Figures 11 and 12 Plant Specific Analysis Results:
Parameter (s) not Bounded, Cold: None Resultant Peak Enthalpy, Cold: N/A Parameter (s) not Bounded, HSB: Accident Reactivity Scram Reactivity Resultant Peak Enthalpy, HSB: 264.6 cal /gm
- 17. LOSS-OF-COOLANT ACCIDENT RESULT (S.2.5.2)
See " Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 3," General Electric Company, December 1977 (NED0-24082, as amended).
10
23A4660 Rev. 1
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- EMMMMMMMMME MMMMMMM" l IIIIIIIIIIIIi 1 357 911131517192123252729313335373941434547495153555753 FU EL TYPE l B = P8 2 CYCL 6) E = BPSDR (CYC C 7) 1 C = P80RB299 (CYCf.E 61 F = P8DHB299 ' CYCLE S)
Figure 1. Reference Core Loading Pattern l
11
1 23A4660 Rev. I 1 NEUTRON FLUk 1 VE!SEL PRES's RISE (PSI) 2 AVE SURF ACE HEAT FLUX 2 SAFETY VALVE FLOW 3 CCRE INLET rLOW 3* FELIEF 153.0 l 300.0 .
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Figure 2a. Plant Response to Generator Load Rejection Without Bypass, EOC7-2000 mwd /ST 12
i i
23A4660 Rev. 1 1NEUTRONFLUk I VESSEL FFE h RISECPSI) 2 AVE SURFACE HEAT FLUX 2 SAFETY VALNE FLOW 3 EORE ItLET ; LOW 3 RELIEF VALVE FLOW 150.0 '
300.0 : EveaI: 'it't e r l
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Figure 2b. Plant Response to Generator Load Rejection Without Bypass, EOC7 13 1
I
23A4660 -
Rei. 1 150.0 g
! NEU'RCN FLUX 1 VE5-EL FRESS RISE (PSI) 2 AVE SW ACE FE AT FLUX.' .
2 RELkEr VALVE FLOW y, g rM_ * ' 3 E!F $.55 VAL \E FLCW
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Figure 3. Plant Response to Loss of 100 F Feedwater Heating 14
Rev. 1 23A4660 150.0 1 NEUTRON dL 1 VESSEL FREEh RISE (PSI) 2 AVE SU E . AT FLUX 2 SAFETY VALVE FLOW
~ G,W 3 RELIEF VALVE FLOW 3ma - COREn_I .fT
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EOC7-2000 mwd /ST 15 m._ _
23A4660 -
. Rev. 1
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Figure 4b. Plant Response to Feedwater Controller Failure, EOC7 16
23A4660 Rev. 1 02 06 10 14 18 22 26 30 34 38 42 46 50 54 58 59 6 10 10 6 55 44 26 26 26 44 51 6 6 2 2 6 6 47 44 30 30 30 30 30 44 43 6 6 10 14 14 10 6 6 39 26 30 30 26 35 10 2 14 0 0 14 2 10 31 26 30 30 26 27 10 2 14 0 0 14 2 10 23 26 30 30 26 19 6 6 10 14 14 10 6 6 15 44 30 30 30 30 30 44 11 6 6 2 2 6 6 7 44 26 26 26 44 3 6 10 10 6 NOTES: 1. Number indicates number of notches withdrawn out of 48. Blank is a withdrawn rod.
- 2. Error rod is (26,35) .
Figure 5. Limiting Rod Pattern 17
23A4660 Rbv. 1 1NEUTRONFhUX 1VE5SELFFhSSRISE(PSI) 2 AVE SUCF?f E HE AT FLUX 2 SAFETY VALVE FLOW 3 CORE INLE7 FLOW 3 FELIEF VALVE FLOW 150.0 300.0
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Figure 6. Plant Response to MSIV Closure (Flux Scram) 18
23A4660 Rev. 1 Ab ATURAL C :RCULATIO V B1 05 PERCEl1T ROD LI VE CL LTIMATE STABILITY LINE _
1.00 (: ::
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PERCENT POWER Figure 7. Reactor Core Decay Ratio 19
23A4660 Rev. 1
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C BOUND VAL 280 CAL /G COLD D BOUND VAL 280 CAL /G HSB
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- 0. 0 500.0 1000.0 1500.0 2000.0 2500.0 3000.0 FUEL TEMPERATURE DEG C.
Figure 8. Fuel Doppler Coefficient in 1/A*C 20
23A4660 Rev. 1 20.0 .
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$ MP n ?
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@ 5.0 J 2.5 A ACCIDENT FUNCTION B BOUNDING VALUE 280 CAL /G
- 0. 0 -
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Figure 9. Accident Reactivity Shape Function, Cold Startup 21
23A4660 dev. 1 20.0 17.5 15.0
$ D W 12.5 x
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O E 5.0 2.5 A ACCIDENT FilNCTION B BOUNDING VALUE 280 CAL /G
- 0. 0 .,
- 0. 0 5.0 10.0 15.0 20.0 ROD POSITION, FEET OUT Figure 10. Accident Reactivity Shape Function, Hot Standby 22
23A4660 Rev. 1 30.O A SCRAM FL NCTION 8 BOUNDING VALUE 280 CAL /G .
25.0 / l m
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1.0 2.0 3.0 4.0 5.0 6.0 ELAPSED TIME, SECONDS Figure 11.
Scram Reactivity Function, Cold Startup 23
23A4660 Rev. 1 40.0 -
A SCRAM F JNCTION 8 BOUNDIN 3 VALUE 280 CAL /G 25.0 g 30.0 I
w 25.O f W
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D. 0 1.0 2. 0 3.0 4.0 5.0 .6. 0 ELAPSED TIME, SECONDS Figure 12. Scram Reactivity Function, llot Standby 24
, 23A4660 Rev. 1 APPENDIX A GETAB Analysis Initial Conditions:
Reactor Pressure, psia 1035.0 ,,
Inlet Enthalpy, BTU /lb 521.5 I
l l
i 25/26 (FINAL)
h GENER AL h ELECTRIC