ML20209E170
| ML20209E170 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 03/11/1996 |
| From: | Chin R, Shannon T, Touvannas G COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20209E127 | List: |
| References | |
| NDIT-960014, NDIT-960014-R01, NDIT-960014-R1, NUDOCS 9907140166 | |
| Download: ML20209E170 (8) | |
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NUCLEAR FUELSERVICES$EPARn!ENT NUC!.IAR DESIGN INFORMATION TRANSMTITAL M SATETY RI!ATED Originadng organization NDIT No. 960014 O NON-SATETY REIATED ' E NuclearFuelServices Rev. No. I O REOUIATORY RELATED D Other(specify) rage i or7 Stadon taSalle Unit ! Cycle t Ocneric To: L45 alls Central File Subject Final Neutronics Licensing Report (NLR) for 125alle 1 Cvele 8 neodore P. Shannon //
Preparer #
%'s Signature Dats Georre Touvannas -- ,_ ~M 3 Il!9I Renew er Reviewer's Signature Date Ronaid J. Chin 8& . x:a 7 J[8 #b f
NTS Supervisor NFS Supervisor's Signature Date Status ofinformation: M Verified 0 Unverified O Engineering Judgement Method and Schedule of Verification for Unverified NDITs:
l Description ofInformation: he NLR contains MCPR values for accident scenarios, the referense core loading. the standby liquid control syvem shutdown capability, maximum exposure limit compliance, and spent fuel pool and fresh fuel vault en6cality compliance. This is Rev. I of the NLR De Rev. 0 NLR was not issued as an NDIT.
Purpose ofInformation: Support Unit 1 Cycle 8 Startup Source ofInformation: BWR DesiEn cale note: IINPL:96414.DNDL:93 039 l
Supplemental Disvibution: David A. Ilenry, Deborah A. Worthington, NDIT File CilRON No:
9907140166 990714 PDR ADOCK 05000373 p FDR 1
r LaSalla 1 Neutronics Licensing Report
- Cycle 8 NDIT No. 960014 - Revision 1
+
Licensing Bas,is l
l This document,in conjunction with Reference 1, provides the licensing basis for LaSalle 1 Reload 7, Cycle 8. The analyses for this document are in References 2 - 9.
l Table of Contents
- ' l. Reference Core Loading
- 2. Calculated Core Effective Multiplication and Control System Worth -
No Volds, 20'C
- 3. Standby Liquid Control System Shutdown Capability
- 4. Core-wide AOO Analysis Results S. Local Rod Withdrawal Error With Limiting Instrument Failure
- 6. Loading Error Results
- 7. Cycle MCPR Values for Non-pressurization Events
- 8. Control Rod Drop Analysis Results
- 9. Maximum Exposure Limit Compliance
- 10. Spent Fuel Pool and Fresh Fuel Vault Criticality Compliance l
- 11. References l
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LaSalle 1 Neutronics Licensing Report i Cycle 8 )
NDIT No. 960014 - Revision 1 i 1. Reference Core Loading Cycle Initial Bundle Name Loaded # Bundles Enrichment GE98 P8CWB303-9CZ 100M 150-T 5 108 3.03 GE98 P8CWB314 9GZ-100M 150 CECO 6 72 3.14 GE9B-P8CWB313 9GZ-100M-150 CECO 6 128 3.13 GE98 P8CWB322-11GZ-100M-150 CECO 7 104 3.22 GE98-P8CWB320 9GZ-100M-150-CECO 7 104 3.20 GE98 P8CWB34312GZ 80M 150 CECO 8 104 3.43 GE98 P8CWB342-10GZ-80M 150 CECO 8 144 3.42 Cycle N 1 core average exposure at end of cycle: 25,791 mwd /ST Cycle N 1 core average exposure at end of cycle l for shutdown considerations: 25,291 mwd /ST Cycle N 1 core incremental exposure at end of cycle: 11,100 mwd /ST 1
Cycle N 1 core incremental exposure at end of cycle for shutdown considerations: 10,600 mwd /ST
)
l Cycle N core incremental exposure at end of cycle: 10,600 mwd /ST i
- 2. Calculated Core Effective Multiplication and Control System Worth -
No Voids,20*C Beginning of Cycle Keffective l
l Uncontrolled 1.120 Fully Controlled 0.966 Strongest control rod out 0.987 l
Beginning of Cycle Shutdown Margin, %AK 1.27 i Minimum Shutdown Magin, %AK 1.27 l R, Maximum increase in Cold Core Reactivity N with Exposure into Cycle, %AK 0.00 i Cycle incremental Exposure Corresponding to Minimum Shutdown Margin R-Value 0 mwd /ST Please note that these values may change slightly between this report and the Startup Report due to actual plant operation.
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LaSalle 1 Neutronics Licensing Report l Cycle 8 NDIT No. 960014 - Revision 1
- 3. Standby Liquid Control System ShutdowgCapability Boron Shutdown Margin (%AK) nom 20*C. Xenon Free 660 4.85.
The B-10 enrichment to be used in the SLCS for LaSalle 1, has been changed from
.18.9% to 45%. Conservatively, a concentration of 800 ppm of natural Baron was used for this reload licensing analysis because the actual Technical Specification minimum Boron Concentration (including allisotopes)is no less than 660 ppm.
The methodology documented in NFD-ND-900-1 Appendix B, Procedure 8.3 was J used.
- 4. Core-wide AOO Analysis Results Exposure range: BOC to EOC Flux Q/A ACPR Event (%NBR) (%NBR) GE91 Loss of Feedwater Heating 118 122 0.16 Loss of feedwater heating event analyzed at 100% Rated Power,87% Rated Flow, and an assumed inlet temperature decrease of 145 *F. The event was analyzed from BOC to EOC. -
- 5. Local Rod Withdrawal Error With Limiting Instrument Failure )
(
l Rod Block Rod Position ACPR Readina (ft. withdrawn) G193 104 4.5 0.17 l
105 5.0 0.20 106 6.0 0.23 107 8.5 0.28 .
108 9.5 0.28 109 10.0 0.28 110 10.5 0.28 UNBLOCKED 12.0 0.28 Setpoint recommended: UNBLOCKED Page 4 of 7 hd%
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LaSalle 1 Neutronics Licensing Report Cycle 8 NDIT No. 960014 - Revision 1
- 6. Loading Error Results q The fuelloading error includes both the mislocated bundle event and u,e misoriented bundle event. GE indicated in Reference 10 that if the operating limit MCPR will be >- 1.28 (as it will be for this cycle), then the mislocated bundle analysis is not required. Use of this bounding value of 1.28 implies a ACPR of {
0.21, which is reported here for the fuel loading error. The ACPR for the I misoriented bundle event is less than 0.21. Therefore, the fuel loading error l event is bounded by the 1.28 MCPR for the mislocated bundle event and its l associated 0.21 ACPR. j Event /ff_g Fuel Loading Error 0.21 {
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MCPR j
. \
Loss of Feedwater Heating 1,23 1 l
Rod Withdrawal Error 1.35 (for unblocked RBM setpoint)
Fuel Loading Error 1.28 These MCPR Operating Limits are based on a MCPR Safety Limit of 1.07.
- 8. Control Rod Drop Analysis Results LaSalle is a benked position withdrawal sequence plant, so the control rod drop accident (CRDA) analysis is not required for startups or normal operation. NRC approval is documented in References 11 and 12. However,in order to allow the i site the option of inserting control rods using the simplified centrol rod sequence shown in Table 1, a control rod drop accident analysis was performed for the simplified sequence. The results demonstrate that the 280 cal /gm Technical l Specification Limit is not exceeded. The simplified sequence is thus valid for LaSalle 1 Cycle 8.
j 9. Maximum Exposure Limit Compliance The maximum nodal exposure achieved during LaSalle 1 Cycle 8 is 45507 mwd /ST. This corresponds to 88% of the design limit. The maximum nodal exposure occurs in the bundle located at 09-42 in node 10. This value ensures that no exposure limits will be reached during normal cycle operation.
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LaSalle 1 Neutronics Licensing Reporz Cycle 8 NDIT No. 960014 - Revision 1 l
- 10. Spent Fuel Pool and Fresh Fuel Vault Criticality Compliance Fresh Fuel Type Peak Fuel Uncontrolled Storage Limit Reactivity
{
GE98 P8CWB343-12GZ-80M-150-CECO 1.218 1.29 1
GE98 P8CWB34210GZ-80M 150 CECO 1.245 1.29 l
3 The fresh fuel complies with the reactivity criteria for new fuel vault and spent fuel pool storage as described in Reference 13.
1 1. - References
'1. GE Proprietary Document 24A5180," Supplemental Reload Licensing Report for LaSalle County Station Unit 1 Reload 7, Cycle 8" draft.
I
- 2. Commonwealth Edison Calculation Note # BNDL:95-014. 1 l
- 3. Commonwealth Edison Calculation Note # BNDL:95 023. l
- 4. Cominonwealth Edison Calculation Note # BNDL:95-024.
- 5. Commonwealth Edison Calculation Note # BNDL:95 025.
- 6. Commonwealth Edison Calculation Note # BNDL:95 027,
- 7. Commonwealth Edison Calculation Note # BNDL:95 038.
- 8. Commonwealth Edison Calculation Note # BNDL:95-039.
- 9. Commonwealth Edison Calculation Note # BNDL:96-014.
- 10. GE Technical Design Procedure,"Mislocated Bundle CPR," TDP-0038, Rev. O, F. T. Bolger, issued February 1995.
- 11. GE Proprietary Document NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel", as amended and approved.
- 12. Commonwealth Edison Document NFSR 0085," Benchmark of BWR Nuclear Design Methods", as supplemented and approved.
- 13. " Spent Fuel Rack and New Fuel Vault Reactivity Limits for LaSalle County Station, Supplement 1", letter, Jack M. Dolter to Dr. Ron J. Chin, NFS:BSS:92-143, August 21,1992.
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LaSalla 1 Cycle 8 Neutronics Licensing Report NDIT No. 960014 - Revision 1
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Table 1 ,. l LSIC8 Shutdown Seouence Insertion BPWS Rod Grouc (Bank) Comments 10 or 9 48- 00 Either Group 10 or 9 may be inserted first.
8 48- 00 Groups 10 and 9 must be fully inserted prior to insertion of any group 8 rod.
7 48-12 All group 8 rods must be fully inserted prior to insertion of any group 7 rods.
7 12-00 All group 7 rods must be banked at 12 before continuing insertion to 00.
5 or 6 48-00 Groups 5 and 6 may be inserted without banking anytime after Groups 9 and 10 have been inserted and before Group 4.
4 48-00 All group 5-10 rods must be fully inserted prior to insertion of any group 4 rods.
3B 48-00 All group 4 rods must be fully inserted prior to insertion of any group 3B rods.
3A 48- 00 All group 3B rods m..a be fully inserted prior to insertion -T any group 3A rods, 2 48-00 Analyzed by Standard BPWS 1 48-00 Analyzed by Standard BPWS Page 7 of 7 ,
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Administrctiva Technical Requircmants - App:ndix A L1CS Reload Transient Analysis Results Attachment 2 LaSalle Unit 1 Cycle 8
. Supplemental Reload Licensing Report 1
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Lasalle Unit 1 cycle 8 May 1999
g i
GENuclear Energy 24A5180 Revision 1 Class I !
May 1998 -
l 24A5180, Rev.1 Supplemental Reload Licensing Report for LaSalle County Station Unit 1 Reload 7 Cycle 8 e
1 Approved - Approved '
G. A.) ationi, Manager W. H. Hetzel E*" #
Nuclear Fuel Engineering Fuel Project Manager fg1 Sd1
Important Notice Regarding Contents of This Report Please Read Carefully This report was prepared by General Electric Company (GE) solely for Commonwealth Edison i
Company (Comed) for Comed's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending Comed's operating license of the LaSalle County Stadon Unit 1. De infonnation
- contamed in this report is believed by GE to be an accurate and true representation of the facts known, obtained or provided to GE at the time this repon was prepared.
ne only undenakmgs of GE respecting infonnation in this document are contained in the con-tract between Commonwealth Edison Company and GE fornuclear fuel and related services for the nuclear system for LaSalle County Station Unit I and nothing contained in this document shall be construed as changing said contract. The use of this information except as defined by said con-tract, or ior any purpose other than that ior which it is intended,is not authorized; and with respec to any such unauthorized use, neither GE nor any of the contributors to this document makes any represeruation or warranty (expressed orimplied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not in-fringe privately owned rights; nor do they assume any responsibility forliability or damage o kind which may result from such use of such information.
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Acknowledgement The engineering and reload licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were perfonned by A. E Alzaben and E T. Bolger. The Supplemental Reload Licensing I Report Rev 1 wr.s prepared by E T. Bolger. This do Fuel Engineering.
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1 7he basis for this repon is GeneralElectric Standard Applicationfor ReactorFuel. NEDE-24011-P-A-11,
- - November 1995; and the U.S. Supplement. NEDE-24011-P-A-11-US November 1995. l l 1. Plant-unique items
, Appendix A: Analysis Condidons 3 Appendix B: Impact of Change to the APRM Flux Scram Serpoint
- 2. !
Reload Fuel Bundles '
Cycle Fuel Type Loaded Number IrradinteA-GE9B-P8 CWB 303-9GL100M-150-T (GE8x 8NB) 5 108 GE9B-P8 CWB 313-9GL100M-150-T (GE8x 8NB) 6 128 j
GE9B-P8CWB314-9GL100M-150-T (GE8x8NB) 6 72 GE9B-P8 CWB 322-11 GL100M-150-T (GE8 x8NB)
- 7 104 GE9B-P8CWB320-9GZ3-100M-150 -T (GE8x8NB) 7 104 Ncr GE9B-P8 CWB 342-10GL80M-150-T (GE8 x 8NB) 8 144 GE9B-P8CWB 343-12G2s80M-150-T (GE8x8NB) 8 104 Total 764
- 3. Reference Core Loading Pattern '
Nominal previous cycle core average exposure at end of cycle: This information will be provided by ComFA Minimum previous cycle core average exposure at end of cycle This information will be '
fmm cold shutdown considerations: provided by Comed Assumed reload cycle core average exposure at beginning of 15680 mwd /MT cycle:
( 14225 mwd /ST) l Assumed reload cycle core average exposure at end of cycle: 27365 MWdMr
( 24825 mwd /ST)
Reference core loading pattem:
Figure 1 Page 4
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- 4. Calculated Core Effective Multiplication and Control System Worth - No Voids,20*C This irdonnation will be pmvided by Comed.
- 5. Standby Liquid Control System Shutdown Capability This infonnation will be pmvided by Comed.
6.
Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis Initial Condition Parameters 12 Erposure: BOC8 to EOC8ICF Peaking Factors Fuel Bundle Bundle Initial Design IAcal Radial Axial R-Factor Power Flow MCPR (MWt) (1000 ib/hr) l GE8x8NB 1.20 1.65 1.40 1.000 7.021 112.8 1.27 Exposure: BOC8 to EOC8 ICF & FFWTR Peaking Factors Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)
GE8x8NB 1.20 1.70 1.40 1.000 7.220 111.0 1.25 Exposure: BOC8 to EOC8 ICF and RPT Out-of-Service Peaking Factors Fuel Bundle Dundle Initial Design Local Radial Arial R-Factor Power Flow MCPR (MWt) (1000 lb/hr)
GE8x8NB 1.20 1.60 1.40 1.000' 6.824 114.3 1.31 i
Erposure: BOC8 to EOC8 ICF and Turbine Bypass Valves Out-of-Service l
{
Peaking Factors I
Fuel {
Bundle Bundle Initial Design Local Radial Axial R-Factor l
Power Flow MCPR '
(MWt) (1000 lb/hr)
GE8xSNB 1.20 1.61 1.40 1.000 6.837 114.2 1.31
~~
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- 1. To provide 1Asalle Cammry Stanon Urus I with operstmg impervements, an mcnased are flow up to 105% analysis
- 1. Ar.alyses wen perf onned for opernuon war,h a reducson m the feedwater temperatet to 320 *F.
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Onv*v m
- Exposure: BOC8 to EOC8 ICF & FFWTR and Turbine Bypass Valves Out-of-Service Peaking Factors Fuel Bundle Bundle Initial Design IAcal Radial Axial R-Factor Power Flow MCPR (MWt) (1000Ib/hr)
GE8x8NB 1.20 1.65 1A0 1.000 6.992 112.7 1.30
- 7. Selected Margin Improvement Options Recirculation pump trip: Yes Rod withdrawallimiter. No Thermal power monitor.
Yes Impmved scram time:
Yes (ODYN Option B)
Measured scram time: No Exposure dependent limits:
No Exposure points analyzed:
1 (EOC)
- 8. Operating Flexibility Options Single-loop operation:
Yes Load line limit:
No Extended load line limit- Yes Maximum extended load line limit: No increased core flow thmughout cycle: Yes Flow point analyzed:
105.0 %
Inctrased core flow at EOC: Yes Feedwater temperature reduction throughout cycle: Yes Temperature tduction:
100.0"F Final feedwater temperature reduction:
Yes ARTS Pmgram:
No
! age 6
~'
_hcaUeU l awe ,
APRM Setdown Requirement Elimination Yes3 Maximum extended operating domain: No Moisture separator reheater OOS: No Turbine bypass system OOS: Yes Safetyhelief valves OOS: Yes (credit taken for 17 of I8 valves)
ADS valve OOS: Yes EOC RPT OOS: Yes Main steam isolation valves OOS: No TCV slow closure: Yes 4
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- 3. The standard APRM Sculown requirernenu for opersuan at off-4sted power # low condiaans are clinunated and the fuel thermal-<necharucal proiccuan is providal t'y the admuustrauan of power and flow dependent MCPR and MAPUlGR Limns.
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- 9. Core-wide AOO Analysis Results Methods used: GEMIN1; GEX1s-PLUS t
Exposure range: BOC8 to EOC8 ICF Event Flux Q/A GE8x8NB Fig. ;
(%NBR) (%NBR) l l Load Reject w/o Bypass 491 119 0.20 2' l Exposure range: BOC8 to EOC8 ICF & FFWTR Uncorrected ACPR Event Flux Q/A GE8x8NB Fig.
l
(%NBR) (%NBR)
FW Connoller Failure 345 121 l 0.18 3 Exposure range: BOC8'to EOC8 ICF and RPT Out-of-Service Uncorrected ACPR Event Flux Q/A GE8x8NB Fig.
(%NBR) (%NBR)
Load Reject w/o Bypass 595 124 0.24 4 Exposure range: BOC8 to EOC8 ICF and Turbine Bypass Valves Out-of-Service ;
Uncorrected ACPR Event Flux Q/A GE8x8NB Fig.
(%NBR) (%NBR)
FW Connoller Failure 529 125 0.24 5 i
Exposure range: BOC8 to EOC8 ICF & FFWTR and Turbine Bypass Valves Out-of-Service !
Uncorrected ACPR !
Event Flux Q/A GE8x8NB Fig.
(%NBR) (%NBR)
FW Controller Failure 467 126 023 6
- 10. Local Rod Withdrawal Errur (With Limiting Instrument Failure) AOO Summary This information will be provided by Corned.
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f IL Cycle MCPR Valuesd I
Safety limit: 1.07 Single loop operation safety limit: 1.08 Non-Pressurization events; This informa: ion will be provided by Comed.
Pressurization events:
Exposure range: BOC8 to EOC8ICF Exposure point: EOC8 Option A Option B GE8x8NB GE8x8NB '
l Load Reject w/o Bypass 3 1.29 Exposure range: BOC8 to EOC8 ICF & FFWTR Exposure point: EOC8 Option A- Option B GE8x8NB GE8x8NB FW ControllerFailure 1.29 1.27 Exposure range: BOC8 to EOC8 ICF and RPT Out-of-Service Exposure point: EOC8 Option A Option B GE8x8NB GE8x8NB Load Reject w/o Bypass 1.36 1.32 Exposure range: BOC8 to EOC8 ICF and 'Ibrbine Bypass Valves Out-of-Service Exposure point: EOCS Option A Option B GE8x8NB GE8x8NB FW Controller Failure 1.35 1.33
{
Exposure range: BOC8 to EOC8 ICF & FYWTR and Turbine Bypass Valves Out-of-Service kysure point: EOC8 l' Option A Option B GE8x8NB GE8x8NB FW Controller Failure 1.34 1.32 1
- 4. For a stngle-4oup opersnou, the MCPR operaung Imus is 0.01 treater than the two-locp value, i
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[~ ww- o . ..-...
- Reload 7 o Rev.1 I
- 12. Overpressurization Analysis Summary 5
' Psl Pv Plant Event (psig) (psig) Response l
l MSIV Oosure (Flux Scram) 1260 1296 l Figure 7 l
- 13. Loading Error Results This information will be provided by Comed. "
- 14. Contml Rod Drop Analysis Results This information will be pmvided by Comed.
- 15. Stability Analysis Results GE SII -380 recommendations have been included in the LaSalle County Station Unit 1 Technical Specifica-tions; therefore no stability analysis is required as documented in the leuer, C. O. Thomas (NRC) to H. C.
Pfefferlen (GE), Acceptancefor Referencing ofLicensing Topical Report NEDE-24011 Rev. 6, Amendment 8, Thermal Hydraulic Stability Amendment to GESTAR 11, April 24,1985.
LaSalle County Station Unit 1 recognizes the issuance of NRC Bulletin No. 88-07, Supplement 1. Power Oscillations in Boiling WaterReactors (BWRs), and will compiy with the recommendations contained there-in. LaSalle County Station Unit 1 also recognizes the issuance of NRC Generic Leuer 94-02, Long-term Solutions and Upgrade ofInterim Operating Recommendatiorafor Thermal-Hydraulic Instabilides in Boil-
{
ing Water Reactors, July ll,1994, and will comply with the recommendations contained in Letter BWROG-94-078 L. A. England (BWROG) to BWR Owners' Group Executive Committee and Primary Representatives. BWR Owners' Guidelinesfor Stability Interim Corrective Action, June 6,1994.
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- 5. The MsIV Dux serarn was performed at 1020 psig dame pressurt. AU other analyses were performed at 986 psig.
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- 16. Loss-of-Coolant Accident Results LOCA method used: SAFER /GESTR-LOCA Reference the LaSalle County Stadon Units 1 and 2 SAFERIGESTAR-LOCA Loss-of-Coolant Accident Analysis, NEDC-32258P, October 1993. The analysis in the LOCA report yielded alicensing basis peak clad ,
temperanne of 1260*F and a peak local oxidation fraction of <0.30%. The following table provides the most limiting and the least limiting MAPLHGRs for the new fuel design.
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- 16. Loss of-Coolant Accident Results (cont)
Bundle Type: GE9B-P8CWB343-120Z-80M-150L-T Average Planar Exposure MAPLHGR(kW#t)
(GWd/ST) (GWd/MT) Most Limiting Least Limiting 0.00 0.00 10.92 11.69 0.20 0.22 10.99 11.71 1.00 1.10 11.13 11.78 2.00 2.20 11.33 11.95 3.00 3.31 11.54 12.16 4.00 4.41 11.76 12.40 5.00 5.51 12.00 12.67 6.00 6.61 12.24 12.90 7.00 7.72 12.49 13.05 8.00 8.82 12.75 13.21 9.00 9.92 13.01 13.37 10.00 11.02 13.22 13.54 12.50 13.78 13.57 13.75 15.00 16.53 13.21 13.48 20.00 22.05 l
12.37 12.71 25.00 27.56 11.57 11.92 35.00 38.58 10.06 10.36 45.00- 49.60 8.64 8.95 51.27 56.52 5.64 6.14 ,
51.30 56.55 -
6.13 52.26 57.61 -
5.70 l l
)
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LAbALLb1 mio A ou Reload 7 Rev.1
- 16. Loss-of-Coolant Accident Results (cont)
Buridle Type. GE9B-P8CWB342-1002-80M-150-T Average Planar Exposure MAPLHGR(kW/ft)
(GWd/ST) (GWd/MT) Most Limiting Least Limiting 0.00 0.00 11.72 12.25 O.20 0.22 11.77 12.28 1.00 1.10 11.87 12.35 2.00 2.20 12.00 12.45 3.00 3.31 12.13 12.55 4.00 4.41 12.27 12.70 5.00 5.51 12.41 12.88 6.00 6.61 12.56 13.07 7.00 7.72 12.72 13.27 8.00 8.82 12.88 13.47 9.00 9.92 13.05- 13.65 10.00 11.02 13.21 13.76 12.50 13.78 13.31 13.82 15.00 16.53 13.05 13.51 20.00 22.05 12.45 I 12.79
. 25.00 27.56 11.63 11.95 35.00 38.58 10.04 10.37 l 45.00 49.60 8.63 8.96 51.07 56.29 5.69 6.22 51.18 )
56.42 -
6.16 {
52.16 57.49 -
5.72 52.16 57.50 -
5.72 Page 13
r w ey
- - m s+s s+8 M M M M s+s m
- E s+s s+s M M M M M M E s
- -mm M M M M M M M M E+s H M em
- r~s M M s+8 M M M M M M E+ss+8Ms+sm
- MMMMs+ss+sMMMMs+sMs+ss+ss+8
- s+8MM!+8s+ss+ss+ss+8Ms+8Ms+sMMM
- s+8MMMMEs+8Es+ss+ss+sMMEs+s 32-m CDogCDmog mm mamEmgmgmgmg d
30-de a m m fDe@CDs dm @ @+m de@esT[D CD to mm mm
- s+8 MMMMMMMMMMMMMM 24-massammEtDatDcamJt 22- m+in dtc m+m em de demFmdeomtummmtoetmn mm do m as ta m mE IDE mE mE EDs 20- EDE m m am (DE m 8 m 8[DJ mE mE am BE BE m o mE mE i s-Fm dm Fe em dm do mm mTm om Fe mm om do me F is i+ to mE ss (DE [DE EDE mE mE mE atD acD atn mE m e o 12 (DeFmtomememememmmFmFmFammdco 10 ID ED IDE mE 9E tDa ms tamTmm mm me m a mm m tD s- (De8mmmFmfmmTmEDmmTmFEDd8ftD s- CD B E m E (DE m TE m m m E a tn m 3 m E ID 4 CDtomemtomemCDraemFlomemmto 2- mmEEDEIDGmEmEmEmJCD tomtDmttTmFotDmomme 1 3 5 7 iillIIIIIIIIII 911 13151719 fj 23 25 27 29 3133 55 37 39.41 13 45 47 49 5153 55 57 59 Fuel Type A=GE9B-P8CWB322-11GL100M-15G-T (Cycle 7)
B-GE9B-P8CWB320-9GZ3-100M-150-T E=GE9B-PSCWB303-9GL100M-150-T (Cycle 5)
(Cycle 7) F=GE9B-P8CWB313-9GZ-100M-150-T C=GE9B-P8CWB343-12GL80M-150-T (Cycle 8) (Cycle 6)
D=GE9B-P8CWB342-10GL80M-150-T G=GE9B-P8CWB314-9GL100M-150-T (Cycle 6)
(Cycle 8)
Figure 1 Reference Core Loading Pattern I
Page 14
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I lt Neutron Flux Wssel Press Rise (psi)
Ave Surtaos Heat Rux - - - - - Safety Wlve Flow 150.0 - -- -- Core intot Flow 300.0 - --- Reisef Valw Flow
--- Bypass Wlve Flow .
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- =
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' I I 0.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Eme (sec) 2.0 Lewi(inch-REF-SEP-SKRT) Wid Reacevny Vasse! Steam Flow 200.0 - ---
- - - - - Do p r Reactivny 10 bine Steam Flow ---
--- eactmty Feedwater Flow
- - - T tal acevny g 1.0 -
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=
0.0 ___ g
.,.- o -1.0 -
., C h.
\'
-100.0 ' '
-2.0 '\ '
O.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec) .
E Figure 2 Plant Response to Load Reject w/o Bypass (BOC8 to EOC8 ICF)
I l
l i
Page 15 l I
r> ~
Keiosa /
RcV.1
. I l
i l J Neutron Flux f- Wssel Press Rise (psi) 150.0 - - - - Ave Surtmofest flux - - - - - Safety Wive Flow
. --- Coroplet Now 125.0 -
--- Reisef Valve Flow
-y Cdre Irdet Subcoolmg --- Bypess Wlve Flow
- a s t"., 100.0 - - - - - A_ - * *m .Et e 75.0 -
2en %,
2 c N re g -
C j
'N. , . g - 1 r- )
j 1 <
50.0 -
1 ,
25.0 - ~
~~
l )
I
-.I.....
l 1
0.0 I O.0
-25.0 '
7.0 14.0 0.0 7.0 14.0 i Time (sec)
Time (sec) i I
i 1
Lewi(inch-REF-SEP-SKRT) Wid Reacevity \
- - - - Vessel Steam Flow l 150.0 - --- Turbine Sesam Flow - - - - Doppier Reactvny '
1.0 --- Scram Reacovity li FeeewterJ.4ew -
, --- TotalReacevity
=
\ .
n g
\
t 100.0 - "
0.0 c3
& C, W-
- Tu . - - - - - -~-
T. -....4. .g l
- c ),
. E o 1.-
s n'o l *, **. O -
\
s l' l ' . . y
> )
50.0 - g*e ,.
g l' ,. - - co e, -1.0 l' . ' / -
. , C
,1::'
'; \ .
0.0 i b \
O.0
~2.0 i '$
7.0 14.0 0.0 7.0 14.0 Time (sec)
Time (sec)
E Figure 3 Plant Response to FW Contiviler Failure (BOC8 to EOC8 ICF & FFWTE) i Page 16
smou, j Neutron Flux Wssel Press Rise (psi)
Ave Surface Heat Rux - - - - - Safety Valve Flow 150.0 -
Corg inlet Flow 300.0 -
--- Relief Valve Flow
--- Bypass Wive Flow -
e *
,s #' %
g 100.0 ,.
- s -
,g 200.0 15 E
%.,,,,,* is
. E 50.0 - '*
100.0 -
f-~~___
/ \,
^
I
/ \
T j t--.
0.0 ' '
O.0 O.0 3.0 6.0 0.0 3.0 6.0 Time (Sec) Time (Sec) level (mch-REF-SEP-SKRT) id Re 'vity
- - - - - Vessel Sesam Flow - - - - - Doppier covrty 200.0 - --- Turbine Steem Row 1.0 -
-- Scram R ctmty
--- Feeerster Flow Total Rea -
Em ' ...
- C .-
t 100.0 %
e e y .*
$ ', g 0.0 , .
T ,
E '%.[**~~'*-.
g o
,\
g .,
j mo [.* * . *
---4--. O
'.c .'L-l .
.# \' .
. .* .2:
0.0 1- .rr------------ % D \
to -1.0 -
(
.. c) i
' c- -\ .
~
i
\
l.
-100.0 I
-2.0 ' I 0.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (sec) i Figure 4 Plant Response to Load Reject w/o Bypass (BOC8 to FOC8 ICF and RM' Out-of-Service)
Page 17 I
I%
E tscauau s 150.0 - ---
Neutron Rux
- - - - - Ave Surface Heat Flux j
\ .
Vessel Press Rise (psi) 175.0 - - - - - Sakty Vah Flow j
ore inist Flow --- Relief Va% Flow
~
' ore inlet Subcookng ,
--- Bypass Malve Flow -
, 125.0 -
v 100.0 "*
- -- p C* * ~ 'c, m., _q y
'. ve ~
j~ ~ '
C '{%A C
- l l I
'? ~, $ 75.0 I \
I
. I \
50.0 - '. *
{ l \
g _g 25.0 -
l '
~
l I
/
0.0 ' ' '
-25.0 l 0.0 8.5 17.0 0.0 8.5 17.0 Time (SeC) Time (SeC)
Lewi(inch-REF-SEP-SKRT) Void Reactmty f
- - - - - Vessel Swam Flow - - - - - Doppier Reactmty 150.0 - --- Turbine Smam Flow 1.0 - --- Scram Reactivity
- = " : r:- _ .c "k . .
--- Total Reacevity
. \ .
a m
. \
a C
- tw 1004)--- p\ ,#
E ~
A ,
R - - "" ~:-Em 2 -m ! -
C E t' 8 ~ ll \i .:*.' . ' . . y
- 1. ,
'l 1; '. :- ., g-
. . .2 50.0 - l
k -1.0 -
l.ll .l,'.: h.I E j . .. .*. \ **
L 0.0 ' I
I
-2.0 I 'I' -
0.0 8.5 17.0 0.0 8.5 17.0 Time (Sec) Time (seC) i Figure 5 Plant Response to FW Controller Failure (BOC8 to EOC8 ICF and 'Ibrbine Bypass Valves Out-of-Service) 1 Page 18 l
c nctona 1 -~ -
J Neutron Flux Vessel Press Rise (pri)
- - - - - Ave Surfacpes 175.0 - - - - - - Safety Valve Row 150.0
--- Core)1stFlow Relief Valve Row
-p CdRiinlet Subcookng --- Bypass Vane Flow
~
'liS.0 -
y 100,0 -- - - - M----
.h '.
1 y
. -s , . =
E \- E l~
E C 75.0 N' . Y I
- \
I ;
50.0 - I l
25.0 -
I
~
l I
~ !
0.0 I '
-25.0 ' '
O.0 7.0 14.0 0.0 7.0 14.0 Time (sec) Time (sec)
Level (inch-REF-SEP-SKRT) Wd Reactnmy
- - - - Vessel Sesam Flow . - - - - - Doppier Rosetmty 150.0 - --- Turbine Steam Flow 1.0 -
--- Scram Reactivny
Feedwater 4w - --- TotalReacevny
- \
\
~--
2c .
e 100.0 -
f.
g _ _ _ _ ---
- T. : '. ,, *
[0.o w- %- --
C j.
e *e e
O .- . ~ ~ ' ' <9 1 . ' e re -
- O o' ls.,e ', 3 s
i 50.0 I.
__8 -1.0-y' ll : ', l
. c: i
!!!:! \
l..'.
0.0 ' N ' l}
-2.0 '
O.0 7.0 14.0 0.0 7.0 14.0 Time (sec) Time (sec)
Figure 6 Plant Response to FW Controller Failure (BOC8 to EOC8 ICF & FFMTR aad Turbine Bypass Valves Out-of-Service) I i
l Page 19
f neauau s --
l l
,,/,
bu Flux Wasel Press Rae (psi)
! ---- Ave Su' tace Heat Flux
' - - - - Safety Valve Flow 150.0 - -- Cors in st Flow 300.0 - ---
I Relief Valve Fiow
--- Bypass Wlve Flow -
os.*
i *
/,\ee n '.
I~ \ '. v 200.0 -
h 100.0 % ~ ~'.s. 2,,
! C % C M r ~. ,, - #
50.0 -
100.0 -
o .,' .,
a 0.0 I
- I 0.0 0.0 4.0 8.0 0.0 4.0 8.0 Time (sec) . Time (sec)
Level (inch-REF-SEP-SKRT) '!oid Reace
- - - - - Vessel Steam Flow - - - - - Do r envity 200.0 - --- Turbine Steam Flow 1.0 - ---
m Reacevity
--- Feedwater Flw --- To l Reachvity G
.f.
en
- E c) . I g 100.0 m 0.0 '
,,,,, ., \ , ,.-
'. A s \ s l To c ,
. E . ,
.A g .
~. .,-:
3, ,sa .. ..-
,.m .
I 0.0 -
~M. .\ .N
, . .* , . , . - s 5
5en -1.0 -
\
'. a>
1 s c I
\
1 15
- 100.0 ' I \A
-2.0 '
O.0 4.0 8.0 0.0 4.0 8.0 Time (sec) Time (sec) l l Figure 7 Plant Response to MSIV Closure (Flux Scram)
Page 20 l
ac- ,
I Appendix A !
Analysis Conditions -
)
To reflect actual plant parameters accurately, the values shown in Table A-1 were used this cycle. -
Table A-1 Analysis Value i
Parameter ICF ICF&FFWTR l Thermal power, MWt 3323.0 3323.0 Core flow, Mlb/hr 113.9 I 113.9 Reactor pressure, psia 1017.0 1015.8 I Inlet enthalpy, BTU /lb Nore-fuel power fraction 526.1 516.6 0.039 0.039 Steam flow analysis, Mlb/hr 14.30 12.62 ,
I Dome pressure, psig 986.3 986.3 Turbine pressure, psig -933.1 945.2 No. of Dual Mode S/R Valves 17 17 Relief mode lowest serpoint, psig 1091.0 1091.0 Safety mode lowest serpoint, psig 1185.0 1185.0 Page 21
Appendix B Impact of Change to the APRM Flux Scram Setpoint l
'Ihe APRM High Neutron Flux Analytical Limit has been changed from 122.4% to 124.2%. This change has -
been incorporated into Overpressure Analysis described in Section 12. The review of non-pressurization events required by GESTAR is the responsibility of Comed. This change will not cause other non-limiting l AOO events to become limiting. This change has been incorporated in the the off-rated flow dependent limits defined in B33-0029p-03P, Updated Transient Analysis: AbnormalStart-up ofan Idle Recirculation Loop for LaSalle County Nuclear Station, Unlu 1 and 2, March 1998. -
Page 22
I A WUosJ e
(
t i
AdditionalInformation Regardmg l the Supplemental Reload Licensing Report Rev L l for l LaSalle 1 Reload 7 Cycle 8 Section 3. Reference Core Loading Pattern Cycle N-1 incremental exposure (nominal) 11.100 mwd /ST
)
Cycle N exposure increment
)
10,600 mwd /ST l Cycle N full power capability (if different from above) same Section 9 Core-wide AOO Analysis Results Increased Core Flow l
PV PV PV Limiting Power Flow Flux i Exposure Q/A (PSL) (Dome) (Bottom) !
"IYansient (%NBR) (%NBR) (%NBR) (%NBR) (psig) (psig) (psig)
LRNEP 100 105 491 119 1152 1162 1190 IEOC EOC MSIVF 102 105 467 130 1260 1264 1296 Increased Core Flow and Reduced Feedwater Temperature l
PV PV PV Limiting Power Flow Flux Exposure Q/A (PSL) (Dome) (Bottom)
Transient (% NBR) (%NBR) (%NBR) (%NBR) (psig) (psig) (psig)
EOC FWCF 100 105 345 121 1116 1118 1145 Increased Core Flow and RPT Out-of-Fervice PV PV PV Limiting Power Flow Flux Exposure Q/A (PSL) (Dome) (Bottom)
Transient (%NDR) (%NBR) (%NBR) (%NBR) (psig) (psig) (psig)
EOC LRNBP 100 105 595 124 1154 1163 1200*
Increased Core Flow and Turbine Bypass Valves Out-of-Service PV PV PV Limiting Power Flow Flux Exposure Q/A (PSL) (Dome) (Bottom)
"IYausient (%NUR) (%NBR) (%NUR) (%NBR) (psig) (psig) (psig)
EOC FWCF 100 105 529 125 1154 1162 1190 Page A-1
=--
ICF & FFWTR and Turbine Bypass Valves Out-of-Senice i
i PV I
PV PV Lhniting Power Flow Flux Q/A (PSL) (Dome) (Bottom)
Exposure Transient (%NBR) (%NBR) (%NBR) (%NBR) (psig) (psig) (psig)
EOC FWCF 100 105 467 126 1148 1157 1184 fi Were all resolved OPL-3 values used for safety and relief valve charaderistics? Yes f
1 Assumed MSIV closure characteristics:
' Time (bc) MSIV Area (ner tmiti
{
0.0 1.0 (fully opened) 0.6 1.0 1.7 0.01 3.0 0.0(fully closed)
Section 15. Stability Analysis Results1 \
Rodline Analyzed: Extrapolated rod block i Decay Ratio:
See Figure A-1 Reactor cott stability ratio, yfxo : 0.78 Channel hydrodynamic performance decay ratio, ygxo :
Hot Channel Decav Ratio GE 8x8NB 0.65
- 1. This stability analysis was performed using verified and NRCmviewed code::. and 'the current reloa procedure. This is not intended to be a bounding analysis however, and the calculated value for the core and channel decay ratios may not be indicative of plant stability under some conditions. Oscillations may occur in sp ratios which are calculated to be significantly less that 1.0. For this reason, Commonwealth Edison Compan not rely on this calculated decay ratio as a basis for avoiding or delaying implementation of the latest GE and NRC guidance in tids area.
The main usefulness of these analysis results is to provide a comparison with previous cycle analysis results re to core and fuel design differences.
Page A-2 I t
LASALLE1 Reload 7 A,. Namral Cuculadon B 105% Rod Line C Ultimate Performance Limit -
l 1.00 o A R 0.75 m
_o_
a:
>- j
< 0.50 s.>
w a
0.25 '
l
.-7 l 0.00 0.0 20.0 40.0 60.0 80.0 100.0 120.0 PERCENT POVER l
Figure A-1 Reactor Decay Ratio Page A-3
p Administrative Technical Rnquiramsnts - Appendix A L
L1C8 Reload Transient Analysis Results i
l l
l l
l i
l Attachment 3 l
l l
i ARTS Improvement Program Analysis, Supplement 1 (Excerpts) l 4
LISalle Unit 1 Cycle 8 May 1999
r Administrativa Technical Requircmants - Appsndix A L1C8 Reload Transient Analysis Results Attachment 3 ARTS Improvement Program Analysis, Supplement 1 (Excerpts)
Summary of Core Wide Transient Results AOO initial Peak Neutron Peak Heat Flux Equipment Out GE9B ACPR Power / Flow Flux (% NBR) (% initial) of Service ICF / Normal i Feedwater Temp l LRNBP 100/105 491 119 None 0.20 LRNBP 100/105 595 124 RPT 0.24 i FWCF 100/105 529 125 TBV 0.24 ICF / Reduced Feedwater Temp FWCF 100/105 345 121 None 0.18 FWCF 100/105 467 126 TBV 0.23 FWCF 25/105 59 154 None 0.59 FWCF 25/105 69 160 RPT 0.63 FWCF 25/105 80 164 TBV 0.70 OLMCPR and Kp Requirements Liraiting Power Equipment Out OLMCPR OLMCPR Calculated Generic Kp AOO of Service (Opt. A) (Opt. B) Kp l LRNBP 100 No EOOS 1.33 1.29 1.0 1.0 LRNBP 100 RPT OOS 1.37 1.33 1.0 1.0 FWCF 100 TBV OOS 1.35 1.33 1.0 1.0 FWCF 25 No EOOS 1.76 1.74 1.35 1.55 FWCF 25 RPT OOS 1.80 1.78 1.34 1.55 FWCF 25 TBV OOS 1.88 1.86 1.40 1.55 TOP / MOP and MAPFACp Requirements Limiting Power Equipment TOP MOP Calculated Generic AOO Out of MAPFACp MAPFACp Service LRNBP 100 No EOOS 24.9 25.2 1.0 1.0 LRNBP 100 RPT OOS 30.3 30.6 1.0 1.0 FWCF 100 TBV OOS 28.7 30.0 1.0 1.0 FWCF 25 No EOOS 50.1 52.0 0.83 0.61 FWCF 25 RPT OOS 57.1 59.0 0.83 0.61 FWCF 25 TBV OOS 62.7 64.5 0.79 0.61 L:salle Unit 1 Cycle 8 A3-2 May 1999
p Administrativa Technical R::quircmsnts - Appandix A L1C8 Reload Transient Analysis Results i
Attachment 4 l
l l ..
l TCV Slow Closure Analysis (Excerpts) l
! l i i l
t l
1 Ltsah Unit 1 cycle 8 - May 1999
'l
i Administrativa Technical Rcquiromants - Appandix A L1C8 Reload Transient Analysis Results Attachment 4
]
TCV Slow Closure Analysis (Excerpts) :
I Table 1 - Key Peak Values for the LRNBP Transient Events at Rated Power '
)
All TCV Fast Closure, 2 Seconds Slow Direct Scram and Closure, Flux Scram EOC-RPT and EOC-RPT OOS Neutron Flux (%) 491 421 i
Heat Flux (%) 119.5 124.4 Dome Pressure (Psig) 1162 1168 ,
i Vessel Pressure (Psig) 1190 1207 Table 2 - MCPR Operating Limits for LRNBP Event at Rated Power Uncorrected ACPR Option A Option B All TCV Fast Closure, Direct 0.20 1.33 1.29 Scram and EOC RPT 2 Seconds Slow Closure, Flux 0.25 1.37 1.33 Scram and EOC-RPT OOS Generic EOC-RPT OOS --
1.37 1.33 (Reference 2) ,
i Lasalle Unit 1 Cycle 8 A4-2 May 1999 l
1 l
Administrativa Technical Requircmsnts - Appandix A L1C8 Reload Transient Analysis Results Attachment 4 TCV Slow Closure Analysis (Excerpts)
Table 3. - Key Peak Values for the Off-Rated Transient Events LRNBP, One TCV Slow LRNBP, AllTCV Slow Closure at 50%/s,3 TCV Closure at 19%/s Fast Closure Neutron Flux (%) 170 109 Heat Flux (%) 62.0 75.1 Dome Pressure (psig) 1093 1125 Vessel Pressure (psig) 1120 1158 l
l ;
i l
l Lasalle Unit 1 Cycle 8 A4-3 May 1999
r
)
)
Administrativa Technical R:quircm:nts - App 3ndix A L1C8 Reload Transient Analysis Results I
i Attachment 4 TGV Slow Closure Analysis (Excerpts)
Table 4. - ACPR, TOP and MOP Values for the Off-rated Transient Events ,
l l LRNBP, One TCV Slow LRNBP, AllTGV Slow l Closure at 50%/s,3 TCV Closure at 19*/ds l l Fast Closure l Calculated ACPR 0.2648 0.6329 l Calculated TOP 26.17 49.27 .
Calculated MOP 26.17 55.30 l
l Adjusted ACPR 0.7420 1
Required MCPR Non-limiting 1.8120 l
l Reference MCPR No Evaluation 1.33 (a)
Required K(p) Performed 1.36 Limiting K(p) 1.53 (b)
Adjusted MOP 60.83
! i Required MOP 38.0 !
I i Required MAPFAC 0.62 L
Limiting MACFAC 0.60 (c)
Note : (a) Based on the bounding event (LRNBP with one TCV closing at 2 seconds) Option B MCPR value.
(b) Based on Figure 5.
(c) Based on Figure 6.
Lt_salle Unit 1 Cycle 8 A4 4 May 1999 L
Administrativa Tcchnical Rcquircments - Appendix A L1C8 Reload Transient Analysis Results '
Attachment 4 TCV Slow Closure Analysis (Excerpts) i
! NEUTRDilFLUX 1 VESSEL PRESS RIBE (PSI) list'Vlui'rtW " ,,,., !!.[fil ?it?!Itu
!...m m, . m -
a
, ... 4,\ ...
Is E
! ( %
\,.
~
\ ... s
'4.. ... ... ...
r..
TIE (ECOND81 TIME (E COND.)
I 1 LEVEL (: NCH-REF-SEP-SKRT) i VOID RI ACTIVITY 2 VESSEL STEAMrLOW 2 00PPLEl: REACTIVITY
., !!#E"! ,!4,^E' " ,., . !!!"i" !!!!!!!!!!
1 J
m.. ;
,n ,.
l ....
jwT :
x / C h
... ; .I ; ; ;
E ,..
f; i;
\
B
\\
-= 4.. ... ... ... -' i . . ... ... ...
Tg g ggEC00108) TIME (SECONOS)
Figure 1. LRNBP from Rated Power, All TCV Fast Closure, Direct ', cram, EOC-RPT Ltsalle Unit 1 Cycle 8 A4-5 May 1999 L
r-I Administriativa Technical Rsquirements - Appandix A !
L1C8 Reload Transient Analysis Results !
Attachment 4 TCV Slow Closure Analysis (Excerpts) l
! l l
l / l lEUTR0llFLUX 1 VESSEL PRESS RISE (PSI) t ggg,g IenE' Vll[ti'r T
- S RE[18vit?E rtu 4 Itf P ARE Vit VF Fi nu ggg ,,
=1 ..- d ..
N
.. ... : x ;
'4.. ... ... ... ' s.. - - - ...
f1E (IEC0Hotl IIE (SECONOS) 1 LEVEL (l NCH-REF-SEP-SKRT) VOID RI ACTIVITY 2 VESSEL STEAMFLOW 00PPLEl' REACTIVITY
,,, ! r i
,,, _ !! f!f!k!
E L -
2
=.. ; A ....
E
.. y ,
N- \ $ ,
- E l T
N -
W -i ..
l 4'
\
-*"4.. ... ... ... -... ... ... ... I TIME (SEC9N081 f!ME (SECONDS)
Figure 2. LRNBP from Rated Power, One TCV Slow Closure (50%/second)/Three TCV Fast Closure, Flux Scram, EOC-RPT OOS Lasalle Unit 1 Cycle 8 A4-6 May 1999
Administrative Technical Rcquirements - Appendix A L1C8 Reload Transient Analysis Results Attachment 4 TCV Slow Closure Analysis (Excerpts)
I E Ev!E!r0!*HEATFLUX ""
' 5* a'" a"
!!!m f,'n naJ E
il 4 r
~
... )/ N 4 ..
N .-
..., .._... .....=. i..
f! fit (MCON06) f!IIE (IEC0eIDS) 2 I k 2 L ITY
!m2!!,11,Ta" ,.,
!!!!tr!!fgim 5
, . . I /
Io ...: v c
i g
... g; [A .. %=:- -- 2 E ..
~...-- ... .... -**... ... ,...
fille ISEC0mptI iIllE (OEC000DS3
[ Figure 3. LRNBP from 50% Power, One TCV Slow Closure (50%/second)/Three TCV Fast Closure, Flux Scram Lasalle Unit 1 Cycle 8 A4 7 May 1999
Administrativa Technical Rsquirernants - Appendix A L1C8 Reload Transient Analysis Reruits l Attachment 4 i
l TCV Slow Closure Analysis (Excerpts) l I
1 NEurRON FLUX 1 VFS IEL PRESS RISE (PSI) 2 AVE SURFACE HEAT FLUX 2 St.F LTY VALVE FLOW
'c"'2"'""
,,,, ,,,,, " !M! !!f!! DJ l
t 1
l l . ..-T; <t
- V~ '
i i*
\- ..
t TIE (SEtte10tl TIE INCONOS) 1 LEV 'L(!NCH-REF-SEP-SKRT)
. i vol ) REACTIVITY 2 VES iEL STEANFLOW 2 00P5LER REACTIVITY l ,,,,, !M !!N!'lfE" ,,, !SEYb" !!!!H5!H s g : ;
' to N c
b E iw ';
s
- (
... : ~~i ;::: . ..
3
... ... u.. -*'... ... a..
Tint (KCON08) TIME IN CON 06)
Figure 4. LRNBP from 50% Power, All TCV Closure at 19%/second, Pressure Scram LaSalle Unit 1 Cycle 8 A4-8 May 1999 1
r Administrativa Technical R:quirements - Appsndix A '
L1C8 Reload Transient Analysis Results Attachment 5 Additional Reload Transient Analysis information (Excerpts) l l
I a
LisaUe Unit 1 Cycle 8 May 1999 L
Administrative Technical Requirements - Appendix A L1C8 Reload Transient Analysis Results Attachment 5 Additional Reload Transient Analysis Information (Excerpts)
Rod Withdrawal Error Aaalysis Rod Pattern Excerpted from Comed NFS Calculation BNDL:95-025 "A CPR RWE for LaSalle 1 Cycle 8", dated 10-13-95 0 0 0 0 24 0 0 0 0 0 0 ,
20 28 28 20 l 0 14 10 10 14 0 0 20 0 0 20 0 0 14 10 10 14 0 20 28 28 20 i 0 0 0 0 0 0 l 24 I O O O O l
Lasalle Unit 1 Cycle 8 A5-2 May 1999 L___