ML20150E659
| ML20150E659 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 01/31/1988 |
| From: | Charnley J, Lambert P, Rash J GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19292J065 | List: |
| References | |
| 23A5889, 23A5889-R, 23A5889-R00, NUDOCS 8807150238 | |
| Download: ML20150E659 (27) | |
Text
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23A5889 Revision 0 CLASS I January 1988
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23A5889 REV. O SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM ATOMIC POWER STATION UNIT 3, RELOAD 7, CYCLE 8
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I Prepared by:
P. A. Lait 5ert Fuel Licensing Verified by: e p.~L. Rash Fuel Licensing Approved b
. ST Charnley j anager, Fuel Licensing 9
. I n m a m m ,, -~~.a-ar 175 Curw Aenue PDC SeJose CA 95125 1/2
23A5889 REV. 0 sgp n IMPCRTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ CAREFULLY This report was prepared by Gener -. Electric solely for Philadelphia Electric Company (PEco) for DSO,'s use with the U.S. Nuclear Regulatory Commission (USNRC) for tsending 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 information 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' and nothing contair.ed 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
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they assume any responsibility for liability or damage of any kind which may result from such use of such information.
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- L' 23A5889 REV. 0 ACKNOWLEDGEMENTS /
$ The engineering and reload licensing analyses, which form the
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K technical basis of this Supplemental Raload Licensing Submittal, were pe' formed in the Nuclear Fuel and Engineering Services Department by C. K. Fan and G. N. Marrotte.
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23A5889 REV. 0-Jk:
-1. ~ PLANT-UNIQUE ITEMS (1.0)a Increased Core Flow / Final Feedwater Temperature Reduction: Appendix A Extended Load Line Limit Analysis: Appendix B Analysis Conditions: Appendix C
- 2. RELOAD FUEL BUNDLES (1.0, 2.0, 3.3.1 AND 4.0)
Fuel Type Cycle Loaded Number Irradiated
. P8DRB299 (P8x8R) 5 4 P8DRB299 (P8x8R) 6 224 PBLTAl 6 2 PBLTA2 6 2 P8DRB284H (P8x8R) 6 56 BP8DRB299- (BP8x8R) 7 144 BP8DRB299H (BP8x8R) 7 140 >
(- New BD319Ana (GE8x8EB) 8 48 BD321Ana (GE8x8EB) 8 144
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. Total 764
- 3. REFERENCE CORE LOADING PATTERN (3.3.1)
Nominal previous cycle core average exposure at end of cycle: 15,858 NWd/ST Ninimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 15,858 NWd/ST.
i Assumed reload cycle core average exposure at end of cycle: '.8,861 NWd/ST Core loading pattern Figure 1 a( ) Refers to area of discussion in General Electric Standard Application for Reactor Fuel, NEDE-240ll-P-A-8 (dated Nay 1986); a letter "S"
( preceding the number refers to the United States Supplement.
as Bundle descriptions contained in the letter from J.S. Charnley (GE) to G.C. Lainas (NRC), "Proposed Amendment 18 to GE Licensing Topical Report NEDE-240ll-P-A (GE8x8NB Fue))," October 31, 1986.
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23A5889 REV. 0 4.- CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20 DEG. C (3.3.2.1.1 AND 3.3.2.1.2)
Beginning of Cycle, X-effective Uncontrolled 1.101 Fully Controlled 0.953 Strongest Control Rod-Out 0.978 R, Maximum Increase in Cold Core-Reactivity with Exposure into Cycle, Delth k O.009
- 5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (Delta k)
EEm (20 deg.C, Xenon Free) 660 0.042 8
23A5889 REV. 0
- 6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUT (3.3.2.1.5 AND S.2.2)
(COLD WATER INJECTION EVENTS ONLY)
Void' Fraction (%) 39.74 Average Fuel Temperature (degrees F) 1002 Void Coefficient N/Aa (cents /% Rg) -6.57/-8.22 Doppler Coefficient N/Aa (cents /deg. F) -0.202/-0.192 Scram Worth N/Aa ($) *e
- 7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (S.2.2)
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Fuel Peaking Factors R- Bundle Bundle Flow Initial Design . Local Radial- Axial Factor Power (NWT) (1000 lb/hr) MCPR Exposures EOC8-2000 mwd /ST ==a BP/P8X8R 1.20 1.62 1.40 1.051 6.844 112.3 1.19 GE8x8EB ...a 1.20 1.63 1.40 1.051 6.855 113.1 1.19 Exposure EOC8 BP/P8X8R i.20 1.52 1.40 1.051 6.404 109.1 1.27 GE8x8EB ==== 1.20 :.52 1.40 1.051 3.404 109.9 1.27
= N = Nuclear Input Data; A = Usto in Transient Analysis
- Generic exposure-independent values are used in General Electric Standard Application for Reactor Fuel, NEDE-240ll-P-A-8, May 1986 een These results are derived at increased core flow conditions, but may be
(- conservatively applied for operation at rated core flow.
==*a PBLTA1 and PBLTA2 fuel are bounded by the GE8x8EB values shown.
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23A5880 REV. 0
- 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: 2
- 9. OPERATING FLEXIBILITY OPTIONS (S.2.2.3) i Single-Loop Operation: Yes Load Line Limit No Extended Load Line Limit Yes (See Appendix B)
Increased Core Flow Yes (See Appendix A)
Flow Point Analyzed: 105% p, Feedwater Temperature Reduction: Yes (EOC8 only) (See Appendix A)
ARTS Program No Maximum Extended Operating Domain: No 10 l
I 23A5889 REV. 0
'10. CORE-WIDE TRANSIENT ANALYSl; htSVLTS (S.2.2.1)
- Methods Used: GEMINI Flux Q/A Delta CPR Transient (%NBR) (%NBR) BP/P8x8R GE8x8EBa* Figure Exposure Range BOC8 to EOC8 Loss of 100 degree F Feedwater Heating 121 119 0.14 0.14 2 Exposure Ranges BOC8 to EOC8-2000 mwd /ST **=
Load Resection Without Bypass 377 113 0.12 C.12 A-1
. Exposure Ranges EOC8-2000 mwd /ST to EOC8 Load Rejection Without Bypass 479 121 0.20 0.20 3 Feedwater controller Failure 272 116 0.14 0.14 4
- 11. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTR'JMENT FAILURE)
SUMMARY
(S.2.2.1)
Limiting Rod Pattern Figure 5 Rod Block Rod (Feet Delta CPR MLHGR ****
Reading % Withdrawn) BP/P8x8R/GE8x8EBee (kW/ft)_
104 4.0 0.13 15.26 15.48 105 4.5 0.14 106 5.0 0.15 15.70 107 5.5 0.17 15.91 108 6.0 0.18 16.13 109 8.5 0.21 17.81 110 9.5 0.22 18.01 4
Set Point Selected 107%
. These results are for operation at rated core flow conditions. For operation with increased core flow, with or without feedwater temperature reduction, see Appendix A.
- PBLTA1 and PBLTA2 fuels are bounded by the GE8x8EB values shown.
- . These results are derived at increased core flow conditions, but may be conservatively applied for operation at rated core flow.
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. ..** Results for all fuel types (see Sec. 2) are within the limits identified in Subsection 2.2.2.5.2 of NEDE-24011-P-A-8.
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23A5889 REV. 0
- 12. CYCLE MCPR VALUES (S.2.2) a Non-Pressurization Events BP/P8x8R GE8x8EB an Exposure Ranges BOC8 to EOC8 Loss of 100 degree F Feedwater Heating 1.18 1.18 Fuel Loading Error -- 1.18 Rod Withdrawal Error 1.21 1.21 Pressurization Events ===
Exposure Range BOC8 to EOC8-2000 mwd /ST aa**
Option A Option B BP/P8x8R GE8x8EBaa BP/P8x8R GE8x8EBan Load Rejection without Bypass 1.26 1.26 1.19 1.19
. Exposure Range EOC8-2000 mwd /ST to EOC8 Option A Option B BP/P8x8R GE8x8EBa= BP/P8x8R GE8x8EBan Feedwater Controller Failure 1.22 1.22 1.19 1.19 Load Rejection without Bypass 1.30 1.30 1.26 1.26 I
= These results are for operation at rated core flow conditions. For operation with increased core flow, with or without feedwater temperature reduction, see Appendix A.
== PBLTAl and PBLTA2 f uel are bounded by the GE8x8EB values shown.
a=a ODYN adjustment factors are documented in the letter from J. S. Charnley (GE) to H. N. Berkow (NRC), "Supplementary Information Regarding Amendment 11 to GE Licensing Topical Report NEDE-240ll-P-A," January 16, 1986.
- These results are derived at increased core flow conditions, but may be conservatively applied for operation at rated cere flow.
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23A5889 REV. O l 13. OVERPRESSURIZATION ANALYSIS
SUMMARY
(S.2.3) =
Steam Line Vessel l Pressure Pressure Plant Transient (psig) (psigl Response MSIV Closure '1232 1265 Figure A-4 (Flux Scram)
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- 14. LOADING ERROR RESULTS (S.2.5.4)
Variable Water Gap Misoriented Bundle Analysis: Yes Event Delta CPR Rotated Bundle Error 0.14
- 15. CONTROL ROD DROP ANALYSIS RESULTS (S.2.5.1)
Plant Specific Analysis Results:
Resultant Peak Enthalpy, Cold 107 '
Resultant Peak Enthalpy, HSB: 202.5
- 16. STABILITY ANALYSIS RE3ULTS (S.2.4)
GE SIL-390 recommendations have been included in the plant operating procedures and/or Technical Specifications; therefore, no stability analysis is required. NRC approval for deletion of a cycle-specific stability analysis is documented in NEDE-240ll-P-A-8-US. ,
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- 17. LOSS-OF-COOLANT ACCfDENT RESULTS LOCA Method Used: SAFE /REFLOOD/ CHASTE See "Loss-of-Coo. ant Accident Analysis for Peach Bottom Atomic Power Station r Unit 3," General Electric Compaa December 1977 (NEDO-24082, as amended).
- These results are derived at increased core flow conditions, but may be conservatively applied for operation at rated core flow.
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V 23A5889 Rev. 0
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FUEL TYPE B P H F 9DPB 99
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Figure 2. Plant Response to Loss of 100 F Feedwater Heating (BOC8 to EOC8) 15
23A5889 Rev. 0 I NEUTRO4 FLUr I YESSEL PRESS RIS((PSI) 2 AVE $4.M ACE W AT FLUX 2 SAFETY V ALY: FLOT no.e see. !M!E!^t!i !! !
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Figure 3. Plant Respot.ne to Load Rejection Without Bypass (EOC8) 16
23A5889 Rev. 0 150.0
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23A5889 Rev. 0 ,
2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 59 3.' 32 32 55 6 2 2 6 51 32 40 40 40 32 47 6 6 12 1 6 6 43 32 40 44 44 44 40 32 30 6 2 12 0 0 12 2 6 35 32 40 44 44 44 40 32 31 6 6 6 8 8 6 6 6 27 32 40 44 44 44 40 32 23 6 2 12 0 0 12 2 6 19 32 40 44 44 44 40 32 15 6 6 12 12 6 6 11 32 40 40 40 32 7 6 2 2 6 3 32 32 32 NOTE: No. indicates number of notches withdrawn out of 48. Blank is a withdrawn rod. Error rod is (26,39).
Figure 5. Limiting Rod Pattern for Rod Withdrawal Error Analysis 18
- 23A5889 Rsv. O APPENDIX A INCREASED CORE FLOW / FINAL FEEDWATER TEMPERATURE REDUCTION To provide'for improved operating flexibility and cycle extension for
- Peach Bottom 3, Cycle 8, analyses were performed for increased core flow (ICF) operation. The analyses cover operation with ICF at 100% power, 105%
core flow (100, 105) throughout the standard Cycle 8. After reaching end-of-cycle (EOC)* 8 exposure, operation can continue with ICF and/or last stage feedwater heaters valved out, followed by a natural reactivity coastdown to 70% power under conditions bounded by 110% core flow. Final feedwater temperature reduction (FFWTR) to approximately 328 F should occur only after the standard end-of-cycle.
This appendix presents the results of an evaluaiion of the limiting abnormal operational transients for Peach Bottom 3, Cycle 8, under ICF and/or FFWTR conditions. These transients were evaluated for the BOC8 to EOC8-2000 mwd /ST exposure range (ICF only) and for EOC8 at incres core flow with and without FFWTR.
Reference A-1 provides additional NSSS safety analyses of a generic nature (i.e., does not require further cycle specific analyses). These generic analyses remain valid for Cycle 8 and follow-on cycles.
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- EOC is defined as the core average exposure at which there is no longer sufficient reactivity to achieve rated thermal power with rated core flow, j
all control rods withdrawn (boyond rod position 24), all feedwater heaters f in service and equilibrium xenon.
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23A5889 R v. 0 A-1. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS Fuel Feaking Factors -
Bundle Power Bundle Flow Initial Design Local Radial Axial R-Factor (HWt) (1000 lb/hr) MCPR Exposure: E0C8-2000 mwd /ST (Increased Core Flow)
BP/P8x8R 1.20 1.62 1.40 1.051 6.844 112.3 1.19 GE8x8EB* 1.20 1.63 1.40 1.051 6.855 113.1 1.19 Exposure: E0C8 (Increased Core Flow)
BP/P8x8R 1.20 1.52 1.40 1.051 6.397 115.3 1.28 GE8x8EB* 1.20 1.52 1.40 1.051 6.383 116.1 1.29 Exposure: EOC8 (Increased Core Flow / Final Feedwater Temperature Reduction)
BP/P8x8R 1.20 1.57 1.40 1.051 6.613 113.8 1.25 GE8x8EB* 1.20 1.57 1.40 1.051 6.602 114.7 1.26 A-2. CORE-WIDE TRANSIENT ANALYSIS RESULTS**
Flux Q/A A CPR Transient (% NBR) (% NBR) BP/P8x8R GE8x8EB* Figure Exposure Range: BOC8 to E0C8-2000 mwd /ST (Increased Core Flow)
Load Rejection w/o Bypass 377 113 0.12 0.12 A-1 Exposure Range: EOC8-2000 mwd /ST to EOC8 (Increased Core Flow)***
Load Rejection w/o Bypass 538 123 0.21 0.21 A-2 1
Exposure Range: EOC8 (Increased Core Flow / Final Feedwater Temperature Reduction)
Feedwater Controller Failure 344 121 0.19 0.19 A-3
- PBLTA1 and PBLTA2 fuels are bounded by the GE8x8EB values shown.
- The results are derived at Increased Core Flow conditions, but may be !
conservatively applied for operation with ICF/FFWTR at E0C8. l l
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23A5889 Rsv.-0
,A-3. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT
SUMMARY
The rod' withdrawal error results reported in Section 11 are not aff,cted
% by ICF/FFWTR operation,' assuming that the rod block monitor is clipped at 107%.
A-4. CYCLE MCPR VALUES Pressurization Events:*
Option A Option B BP/P8x8R GE8x8EB** BP/P8x8R GE8x8EL**
. Exposure Range: BOC8 to EOC8-2000 mwd /ST (Increased Core Flow) 1.19 :
Load Kejection w/o Bypass 1.26 1.26 1.19 Exposure Rangc: EOC8-2000 mwd /ST to EOC8 (Increased Core Flow)
Load Rejection w/o Bypass 1.31 1.31 1.27 1.27 Exposure Range: EOC8 (Increaaed Core Flow / Final Feedwater Temperature Reduction) .
Feedwater Controller Failure 1.27 1.27 1.24 1.24 A-5. OVERPRESSURIZATION ANALYSIS
SUMMARY
Increased Core Flow sl v Transient (psig) (psia) Plant Response MSIV Closure 1232 1265 Figure A-4 (Flux Scram) l-A-6. LOADING ERROR RESULTS The fuel loading error analysis results reported in Section 14 ate unaffected by ICF/FFWTR operation.
I 1
- 0DYN Adjustment Factors are documented in the letter from J.E. Charnley (GE)
(
i to H.N. Berkow (NRC), "Supplementary Information Regarding Amendment 11 to GE Licensing Topical Report NEDE-24011-P-A." January 16, 1986.
- PBLTA1 and PBLTA2 fuels are bounded by the GE8x8EB values shown.
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23A5889 Rev. O A-7. CONTROL ROD DROP ANALYSIS RESULTS The control rod drop accident analysis results of Section 15 are unaffected by ICF/FFWTR operation.
A-8. STABILITY ANALYSIS RESULTS GE SIL-380 recommendations have been included in the plant operating procedures and/or Technical Specifications; therefore, no stability analysis is required. NRC approval for deletion of a cycle-specific stability analysis is documented in NEDE-24011-P-A-8-US.
A-9. LOSS-OF-COOLANT ACCIDENT RESULT The LGCA results and MAPLHGR limits are insensitive to operation with ICF and FFWTR; thus, the current limits documented in Reference A-2 remain valid for Peach Bottom 3.
REFERENCES:
A-1. "Safety Rev'.ew of Peach Bottom Atomic Power Station Unit No. 3 at Core Flow Conditions Above Rated Flow Throughout Cycle 6," NEDC-30519, March 1984.
A-2. "Losc-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 3," General Electric Company, December 1977 (NEDO-24082, as amended).
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Figure A-1. Plant Response to Loaa Rejection Wit}.out Bypass (E0C8-2000 mwd /ST Increased Core Flow) 23
Rev. 0 23A!889 .
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24 1
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23A5889 Rev. O
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Figure A-3. Plant Response to Feedwater Controller Failure (EOC8 Increased Core Flow / Final Feedwater Temperature Reduction) 25
23A5889 Rev. O' '
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TDE (KtDeCt) fine (ECORDS) l Figure A-4. Plant Response to MSIV Closure, Flux Scram l (Increased Core Flow) 26
N ..
- 23A5889 Rsv. O j APPENDIX B
{
EXTENDED LOAD LINE LIMIT ANALYSIS 1 ThisahpendixprovidesjustificationforoperationofPeachBottom3, I
\. Cyc'e 8, in the expanded operating domain bounded by the average power range monitot (APRM) rod block line, the rated power line and the rated load line.
4 The definition of the APRM rod block line (i.e., 0.58W + 50%, where W is 1 recirculation drive flow in percent of rated) given in Reference B-1, is applicable to D-ach Bottom 3, Cycle 8 operation. Similarly, the APRM scram line i
is given as 0.58W + 62%.
l Reference B-1 provides results of rnalyses which demonstrate that standard licensing basis results for pressurization transients bound the Extended Load Line Limit Analysis (ELLLA) point (i.e., 100% power /87% flow). To provide even !
further confirmation for Peach Bottom 3, Cycle 8, a feedwater controller failure (FWCF) analysis was performed at end-of-cycle 8 at 100% power /87% flow. Results are provided in Sections B-1 through B-3.
As described in Reference B-1, operation in the ELLLA region is within allowable design limits for the rod withdrawal error, slow flow runout events, overpressurization protection analysis, and the loss-of-coolant accident (LOCA) and containment analyses. Stability performance within the ELLLA region is also acceptable because the operating recommendations described in Section 16 have been incorporated into the Peach Bottom 3 operating procedures and/or Technical Specifications.
t 27
23A5889 Rev. 0
) .
B-1. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS Fuel Peaking Factors Bundle Power Bur.dle Flow Initial Design Local Radial Axial R-Factor (MWt) (1000 lb/hr) MCPR Exposures EOC8 1.40 1.051 6.659 92.1 1.17 BP/P8x8R 1.20 1.58 1.40 1.051 6.679 92.8 1.17 GE8x8EB* 1.20 1.59 B-2. CORE-WIDE TRANSIENT ANALYSIS RESULTS Flex Q/A aCPR
_BP/P8x8R GE8x8EB* Figure Transient (7. NBR)_ (% NBR)
Exposure Range: BOC8 to EOC8 231 114 0.10 0.11 B-1 Feedwater Controller Ftllure B-3. CYCLE MCPR VALUES Pressurization Events:**
Option A Option E BP/P8x8R GE8x8EB* BP/P8x8R GE8x8EB*
Exposure Range: BOC8 to EOC8 1.19 1.19 1.15 1.16 Feedwater Controller Failure
REFERENCE:
B-1. "General Electric Boiling Water Reactor Extended Load Line Limit Analysis for Peach Bottom Unit 2, Cycle 7, and Peach Bottom Unit 3, Cycle 7,"
NEDC-31298, May 1986.
- PBLTA1 and PBLTA2 fuels are bounded by the GE8x8EB values shown.
26 i
23A5889 Rev. 0 150..
8 KU'l#CN FM r i VESSEL PRCS5 RISE (PSI) 2 Avt'TuPFACEVAT FLUX 2 SAFETY WALY: Flow 3 CDR : IPLET rCOW 3 RELI VALv: FLOV I!!. 8 ^~ " "- 4 BYPA ALv: FLcv t ee.t b 186.t 3
~
'(
b N
g : : : :- + 1 St.e
'E S t.t '
CM4 i
1
. . . = ., = = =
.. e L O.0 18.0 29.0 S o.9 8.9 8. 9 29.. 3..t TIM sutnes) Tint istenes) 31.EVEL E INC)4-KF.StP.SKAT) I Vol viff 2 00Pg hE ACT!
- fiv!TV 2 VESSEL STEATLov
,,,., !Lu eisne ov i.. pmgtRREA _ysmm
. i h E e
L i s. . / t G .i.e
.. e 5 ..
... e... n. s e. . e i s. n.e m.
TasE REcemel T14 tette ms)
)
Figure B-1. Plant Response to Feedwater Controller Failare
)
(EOC8 Extended Load Line Limit Analysis) 29/30
)
Rev. O 23A5889 APPENDIX C ANALYSIS CONDITIONS To accurately reflect actual plant parameters, the values listed in Table C-1 were used instead of the values reported in NEDE-24011-P-A-B-US, May 1986.
l Table C-1 PLANT PARAMETERS Parameter Analysis Value NEDE-24011 Value Thermal Power NW/t 3293 3440 t_0.2%
Dome Pressure 1005 1020 t2 psi Turbine Pressure 951 960 +2 psi Non-Fuel Power Fraction 0.038 0.040
)
31/32
)
(Final)