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Rev 2 to 960103, Neutronics Licensing Rept for LaSalle Unit 2,Cycle 8
ML20205L839
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 03/22/1999
From: Jennifer Fisher, Hsiao M, Pallatta A
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20205L820 List:
References
960103, 960103-R02, 960103-R2, NUDOCS 9904140314
Download: ML20205L839 (16)
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NUCLEAR FUELMANAGEMENTDEPARTMENT NUCLEAR DESIGN INFORMATION TRANSMITTAL, ~

E SAFETYRELATED Odginating Organlaation NDITNo.

960103 O NON.SAFETYRELATED ED NuclearFuelM- . Rev. No. 2 O REGULATORYRELATED D Other(specify) _ Page 1 of 15 Station 14Salle Unit 2 Cycle . _ 8 Generic To: D. A.Worthington E. A.McVey(12Salle) *

Subject:

g4Salle Unit 2 Cycle 8 Neutronics Ucensing Report Revision 2 Mint-Yuan Hsiao Pseparer

_ hh b Preparer's Sigraure 3 ~ 2-4 -9 7

. Date

  • lill T. Fisher Reviewer b'/ $ Y '

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% ewer's Sig Date Adelmo S. Pallosta /

NFM Supervisor 8 # f.

NFM Supervisor's Signature . Date Status ofInformation: @ Verified D Unverified O EngineeringJudgernent Method an' d Scheduse of Verification for Unverified NDITs: N/A Description of Information: LaSalic Unit 2 Cycle 8 Neutronics Ucensing Report, Revision 2 Purpose ofInforma' ion:

  • Rev. 0: Provide the station and BSS,proup LaSalle Unit 2 Cycle 8 Neutronics Ucensing Report (NLR).

Rev.1:Wis revision updatesSection IV." Control Rod Drop Accident," due to newinformation on Doppler coefficients.

Rev. 2: This revislor updates the NLR based or. NDfT NFM960082 Rev. 2. "LaSalle Unit 2 Cycle 8 Design Basis Loadin which is12/2/1998.

Energies," consistent with NDIT NFM9800166 Seq. I, "LaSalle 2 Cycle 8 Redesigned Final Ucensing Leadi Source ofInformat;on: As referenced in the NLR Supple' mental Distribution:

A. F. Ooss (13) LaSalle Central File Downer GroveCentralFile J. J. Reimer (LS) R. H. Jacobs A. S. Pallotta M. Y.Hsino (2 copies)

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NUCLEAR FUEL MANAGEMENT DEPARTMENT NDIT No. 960103 NUCLEAR DESGN WFORMATON TRANSMITTAL Rev. No. 2 Page 2 of 15 COMMONWEALTH EDISON COMPANY NUCLEAR FUEL SERVICES NEUTRONICS LICENSING REPORT for LaSalle Unit 2 Cycle 8 Revision 2 Prepared by: Nbb Date: 3~ /~ 7 Ming- uan Hsiao Reviewed by: /C I

, Date: s/z2/99 JillT. Fisher Approved by:- Date: 8 f.

Adelmo S. Pallotta e

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't NUCLEAR FUEL MANAGEMENT DEPARTMENT NDIT No. 960103 NUCLEAR DESGN INFORMATION TRANSMITTAL Rev. No. 2 Poge 3 of 15 Licensing Basis This document, in' conjunction with the references 1,2 and 4 in Section VM provide the licensing basis for LaSalle Unit 2 R'eload 7 Cycle 8. .

Table of Contents I. Nuclear Design Analysis I.1 Fuel Bundle Nuclear Design Analysis I.2 Core Nuclear Design Analysis 1.2.1 Core Configuration and Licensing Exposure Limits I.2.2 Core Reactivity Characteristics H. Control Rod Withdrawal Error E. Fuel Loading Error m.1 Fuel Mislocation Error III.2 Fuel Mistotation Error IV. Control Rod Drop Accident V. Loss of Feedwater Heating VI. Maximum Expo'sure Limit Compliance VH. Spent Fuel Pool and Fresh Fuel Vault Criticality Compliance 1

VH.1 Fresh Fuel Vault Criticality Compliance VH.2 L1 Spent Fuel Pool Criticality Compliance VB.3 L2 Spent Fuel Pool Criticality Compliance -

VM. References

- I preparer: ?WM, 4- #-9 9 i

NUCLEAR FUEL MANAGEMENT DEPARTMENT NDriNo. 960103 NUCLEAR DESIGN INFORMATION TRANSMITTAL Rev. No. 2 Pope 4 of 15

~

I. Nuclear Desien Analysis I.1 Fuel Bundle Nuclear Desien Analysis Assembly Average Enrichment (ATRIUM-9B), w/o U-235 .

SPCA9-381B-13GZ7-80M (High Gd) 3.81 SPCA9-384B-11GZ6-80M (Low Gd) 3.84 Axial Enrichment and Burnable Poison Distribution SPCA9-381B-13GZ7-80M (High Gd) Figure 1 SPCA9-384B-110.Z6-80M (Low Gd) Figure 1 Radial Enrichment and Burnable Poison Distribution SPCA9.-403L-13G7 Figure 2 SPCA9-430L-11G7 Figure 3 SPCA9-406L-11G6 Figure 4 SPCA9-434L-10G6 Figure 5 I.2 Core Nuclear Desien Analysis I.2.1 Core Conficuration and Licensine Exposure Limits Cycle Number.

Bundle Tyne Loaded in Core GE9B-P8CWB302-9GZ-100M-150-T 5 93 GE9B-P8CWB300-9GZ-100M-150-T 5 8 GE9B-P8CWB313-902100M-154 CECO 6 80 GE9B-P8CWB316-9GZ 100M 150-CECO 6 151 GE9B-P8CWB322-11GZ 100M-150-CECO 7 96 GE9B-P8CWB320-9GZ-100M-150-CECO 7 80 SPCA9-381B-13GZ7-80M 8 128 SPCA9-384B-11GZ6-80M 8 128-Cycle N-1 core average exposure at end of cycle (MWD /MTU) 26401 Cycle N-1 core average extiosure at end of cycle for shutdown consideration (MWD /MTU) 26401 preoarer: a.vd > e oc

NUCLEAR FUEL MANAGEMENT DEPARTMENT NC4T No. 960103 NUCLEAR DESGN INFORMATION TRANSM.TtAL Rev. No. 2 Page 5 of 15 Cycle N-1 core incremental exposure at end of cycle 9734 (MWD /MTU)

Cycle N-1 core incremental exposure at end of cycle for shutdown considerations (MWD /MTU) 9734 Cycle N core average exposure at beginning of cycle (MWD /MTU) 14142 Cycle N core incremental exposure at end of cycle (MWD /MTU) 13250 Cycle 8 neutronics analyses are analyzed for the actual EOC N-1 -

exposure given above. The exposure window that validates the pressurization transic.nts can be found in Reference 1.

I.2.2 Core Reactivity Characteristics All values reponed below are with zero xenon and are for 68'F moderator temperature. The MICROBURN-B cold BOC best estimate K-effective bias is 1.005. .

BOC Cold K-Effective, All Rods Out 1.11186 BOC Cold K-Effective All Rods In 0.95433 BOC Cold K-Effective, Strongest Rod Out 0.99187 BOC Shutdown Margin, % AK 1.313 Minimum Shutdown Margin, % AK 1.306 Reactivity Defect (R-value), % AK 0.007 Cycle Incremental Exposure Corresponding to Minimum Shutdown Margin R-Value (MWD /MTU) 250 Standby Liquid Control System Shutdown Margin, Cold Condition, (% AK) 18.742 LaSalle station has upgraded its Standby Liquid Control System so that the B-10 enrichment has been increased from 18.9% to 45%. The above SBLC analysis assumes 660 ppm with the boc enriched to 45% B-10.

__ _ _ _ neevver w !le *-**+OO

4 NUCLEAR FUEL MANAGEMENTDEPARTMENT NDIT No. 960103 NUCLEAR DESIGN NFORMATON TRANSMffTAL Rev. No. 2 Page 6 of 15 II. Control Rod Withdrawal Error The control rod withdrawal error event is analyzed at 100% of rated power,100% of rated flow and unblocked conditions only.

Distance Withdrawn (ft) A_QPR 12 CJnblocked) 0.28 l The design complies with the SPC 1% plastic strain and centerline melt criteria via conformance to the PAPT (Protection Against Power Transient) LHGR limits. The design complies with the GE centerline melt criteria via conformance to the GE thermal overpower protection (TOP) criteria. The design does not meet the GE mechanical overpower protection (MOP) criteria during a control rod withdrawal error event. However, a further analysis shows that the design complies with the GE 1% plastic strain criteria.

III. Fuel leading Error III.1 Fuel Mislocation Error

<< These data are to be fumished by SPC. >>

III.2 Fuel Misrotation Error

<< These data are to be furnished by SPC. >>

IV. Control Rod Dron Accident LaSalle is a banked position withdrawal sequence plant. In order to allow the site the optio inserting control rods using the simplified control rod sequence shown in Table 1, a control rod drop accident analysis was performed for the simplified sequence. The results de'monstrate that th 280 cal /gm Technical Specification Limit is not exceeded. The simplified sequence is thus va for.LaSalle 2 Cycle 8.

<< These data are to be furnished by SPC. >>

prenprer*_. MW E 2-1_l M ~

NUCLEAR FUEL MANAGEMENT DEPARTMENT NDIT No. 960103 flVCLEAR DESIGN INFORMATION TRANSMITTAL Rev. No. 2 Pope 7 of 15 V. Loss of Feedwater Heatine The loss of feedwater heating event is analyzed at 100% of rated power for 87%,100% and 105%

of rated flow and an assumed inlet temperature decrease of 145 F. The event was analyzed from BOC to EOC. The ACPR value reported below is bounding for both the SPC and the  ::sident GE fuel types and all the analyzed flows.

Event ACPR Loss of Feedwater Heating 0.190 The design complies with the SPC 1% plastic strain and centerline melt criteria via conformance to the PAPT (Protection Against Power Transient) LHGR limits. The design complies with the GE 1% plastic strain criteria via conformance to the mechanical overpower protection (MOP) limit.

The design does not meet the GE thermal overpower protection (TOP) criteria during a loss of feedwater heating event; hence, the MAPLHGR values in the COLR for the affected lattice are adjusted accordingly.

Maximum Value GE Criteria (Calculated) Limit Mechanical Overpower (MOP), % 36.1 45 Thermal Overpower (TOP), % 29.5 25 ,

VL Maximum Exposure Limit Compliance Note that the following exposures are based on the actual Cycle 7 EOC exposure of 9,734 MWD /MT and a nominal Cycle 8 exposure of 13,250 MWD /MT. If Cycle 8 reaches it's long window (approximately 500 MWD /MTU beyond the nominal Cycle 8 energy), the exposure limits will still be met.

Exposure GE9B GE9B ATRIUM-9B ATRIUM 9B (MWD /MT) Projected Limit Projected Limit (MWD /MT) (MWD /MT) (MWD /MT) (MWD /MT)

Peak Assembly 40,100 42,000* 19,176 48,000 Peak Pellet 54,464

~

60,000 30,432 66,000

  • Batch averaged value ff hh Yf __ _

NUCLEAR Ftr1. MANAGEMENT DEPARTMENT NDIT No. _960103 NUCLEAR DESGN INFORMATION TRANSMITTAL Rev. No. ,2 Pope 8 of M VII. Spent Fuel Pool and Fresh Fuel Vault Criticality Compliance For the L2C8 reload, there are two new SPC ATRIUM-9B assembly types consisting of 4 uniqu lattices, as identified in I.1 Fuel Bundle Nuclear Design Analysis.

VII.1 Fresh Fuel Vault Criticality Compliance The fuel storage vault criticality analysis that is detailed in Reference 5 remains valid for the above lattices. All the new (ATRIUM-9B) assemblies comply with the fresh fuel vault criticality limits, i.e., all lattices have an enrichment of less than 5.00 wt %

U-235 and a gadolinia content that is greater than 6 rods at 3.0 wt% Gd2 03 .

VII.2_I_.,1_ Spent Fuel Pool Criticality Compliance The LaSalle Unit I spent fuel pool criticality analysis that is detailed 10 Referenc'e 6 remains valid for the above lattices. All the new (ATRIUM-9B) assemolies comply with the spent fuel pool criticality limits, i.e., all lattices have an enrichment of less than 4.60 wt % U-235 and a gadolinia content that is greater than 8 rods at 3.0 wt%

Gd2 0. 3 VII.3 L2 Spent Fuel Pool Criticality Compliance The LaSalle Unit 2 spent fuel pool criticality analysis that is detailed in Reference 7 remains valid for the above lattices. As shown below, all.the new (ATRIUM-9B) assemblies comply with the LaSalle Unit 2 spent fuel pool criticality limit of k-eff < 0.95.

Lattice Type Maximem Maximum Spent Fuel k-inf* in-Rack Pool k-eff" k-eff Limit SPCA95  ;!

1.18428 < 0.85 0.95 SPCA9-4:- -11G7 1.20269 < 0.85 0.95 SPCA9-4GL -IlG6 1.21344 < 0.85 0.95 SPCA9-434L-10G6 1.22734 < 0.86 0.95

  • From 68 *F, uncontrolled CASMO-3G results. '

" From Figure 6.1 of Reference 7.

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NUCLEAR FUEL MANAGEMENT DEPARTMENT NDIT No. 960103 NUCLEAR DESIGN INFORMATION TRANSMlUAL Rev. No. 2 Poge 9 of 15 -

VHL References

1. "LaSalle Unit 2 Cycl'e 8 Reload Analysis", Siemens Power Corporation, EMF-96-125 Revision 2, dated March 1999.
2. "LaS'alle Unit 2 Cycle 8 Plant Transient, Analysis", Siemens Power Corporation, EMF-96-124(P), Revision 3, dated March 1999.
3. "LaSalle 2 Cycle 8 Redesigned Fmal Licensing Loading Plan a.4d Licensing Energy,"

NDIT NFM9800166 Seq.1, December 2,1998.

4. Commonwealth Edison, Nuclear Fuel Services, NFSR-0091, " Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods", as supplemented and approved.
5. " Criticality Safety Analysis for ATRIUM-9B Fuel, LaSalle Units'1 and 2 New Fuel Storage Vault," Siemens Power Corporation, EMF-95-134(P), December 1995.

6.

" Criticality Safety Analysis for ATRIUM-9B Fuel, LaSalle Unit 1 S3.st Fuel Storage Pool (BORAL Rack)," Siemens Power Corporation, EMF-96-117(P), April 1996.

7.

" Criticality Safety Analysis for ATRIUM 9B Fuel, LaSalle Unit 2 Spent Fuel Storage Pool (Boraflex Rack)," Siemens Power Corporation, EMF-95-088(P), February 1996.

8.

"LaSalle 2 Cycle 8 Re-design Standby Liquid Control System (SBLC) Worth Calculations,"

BNDL:98-018, Rev. O, December 9,1998.

9. "LaSalle 2 Cycle 8 256 Bundle Redesign LFWH Analysis," BNDL:98-016, Rev. 0, January 5,1999.

10.

"L2C8 RWE delta CPR Analysis," BNDL:99-003, Rev. O, February 18,1999.

11.

"L2C8 RWE MOP / TOP Analysis," BNDL:98-019, Rev. O, January 11,1999.

12.

"L2C8 SDM Calculations with As-Built Bundle Weights for the Redesigned L2C8 Core,"

BNDL:99-010, Rev. O, March 10,1999.

13.

" Revised M'APLHGR Values'for LaSalle 2- Rev.1," GE Letter WHC:99-008, fro n W. H. Hetzel to R. J. Chin, March 16,1999.

14.

"LaSalle 2 Cycle 8 RWE Clad Strain Compliance," GE Letter WHO:99-010, from W. H. H:tzel to R. J. Chin, March 19,1999.

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' asm-me NUCLEAR FUEL MANAGEMENT DEPARTMENT NOIT No. 960103 NUCLEAR DESIGN INFORMA110N TRANSMTTTAL Rev. No. 2 Pope 10 of 15 Table 1 L2C8 Shutdown Sequence Insenion BPWS Rod Grouo* (Bank) Comments 10 or 9 48 - 00 Either Group 10 or 9 may be insened 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 insened prior to insenion of any group 7 rods.

7 12 - 00 All group 7 rods must be banked at 12 before continuing insertionto 00.

S or 6 48- 00 . Groups 5 and 6 may be insened without banking anytime after Groups 9 and 10 have been insened and before Group 4.

4 48 - 00 All group 5-10 rods must be fully insened prior to insenion of any group 4 rods.

3B 48 - 00 All group 4 rods must be fully inserted prior to insenion of any group 3B rods.

34 48 - 00 All group 3B rods must be fully insened prior to insenion of any group 3A rods.

2 48- 00 Analyzed by Standard BPWS 1 48- 00 Analyzed by Standard BPWS

  • Group definitions are from LAP-100-13 Revision 20.

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NUCLEAR FUEL MANAGEMENT DEPARTMENT NDR No. _960103 NUCLEAR DESGN INFORMATION TRANSMITTAL Rev. No. 2 Pope 11 of 15 128 Bundles 128 Bundles Natural U 11" Natural U 11"

+-- See Figure 2 4.03 w/o 4.06 w/o 13G7.0 48* 11G6.0 42" See Figure 4 ---*

<---- See Figure 3 -

See Figure 5 --*

4.30 w/o 4.34 w/o 11G7.0 84" 10G6.0 90" Natural U 6" Natural U 6"

SPCA9-381B-13GZ7-80M SPCA9-384B-11GZ6-80M Figure 1. L2C8 FLLP Bundle Design 4

orenarer: WM. A-18

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NUCLEAR FUEL MANAGEMENT DEPARTMENT NDIT No. 960103 NUCLEAR DESIGN INFORMATION TRANSMITTAL Rev. No. 2 Pope 12 of 15 W .% ;c; cr =~ = ~ ? ,

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5 Rods ( 2) 4.19 w/o U-235 G Rods (13) 3.69 w/o U-235+7.0 w/o Gd203 FiguVe 2. SPCA9-403L-13G7 Lattice Enrichment Distribution preparer. 9nvn. .:t 2.#-oo - - -

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NUCLEAR FUEL MANAGEMENT DEPARTMENT NDriNo. 960103 NUCLEAR DESIGN INFORMATION TRANSM.'TTAL Rev. No. 2 ,

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l NUCLEAR FUEL MANAGEMENTDEPARTMENT NDIT No. 960103 NUCLEAR DESIGN INFORMATION TRANSMITTAL Rev. No. 2 Pope 14 of 15 g .. ; .. . .w .;p w: .. ;. r,:n:. ws s ,. . , cc. . .. . .: .~; e.':::.M 1 2 3 3 4 3 3 2 1 2.72 3.53 3.94 3.94 4. 53 3.94 3.94 3.53 2.72 I

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NUCLEAR FUEL MANAGEMENT DEPARTMENT NOIT No. 960103

, NUCLEAR DESGN INFORMATION TRAt4MITTAL Rev. No. 2 Pope 15 of 15 1

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Administrative Technical Requirements - Appendix B L2C8 Reload Transient Analysis Results Attachment 2 LaSalle Unit 2 Cycle 8 Reload Analysis LCSalle Unit 2 Cycle 8 March 1999

I NUCLEAR FUEL MANAGEMENT DEPARTMENT NUCLEAR DESIGN INFORMATION TRANSMITTAL .

@ SAFETY RELATED Originating Organization ~ NDIT No. NFM960146 O NON-SAFETY RELATED ' @ NuclearFuelManagement Seq. No. 1 O REGULATORYRELATED D Other(specify) - PageIof1 Station LaSalle Unit - 2 Cycle '8 Generic To: Mr. E. A.McVey (LaSalle)

Subject:

LaSalle 2 Cycle g Reload Analysis, Rpgision 2, EMF-%1t$ y Deborah A.Worthington j ppj Preparer PrepWir's SignaturV \ # / Dat(

Ming Y. Hsiao - M M /

f Reviewer ReviewVr's % nature Dat'e Thomas J. Rausch / WM 27[9p NFM Supervisor NFM Supervisor's Sgasture Date Status ofinformation: @ Verifed O Unverified O EngineeringJudgement Method and Schedule of Verification for Unverified NDITs:' N/A Description of information: This document summarizes the core design, thermal limits, and generic analyses needed to support L2C8 operation. This document supersedes NDIT %0146. Rev. O and L2C8 Reload Analysis EMF-%I25, Revision 1.

Purpose ofinformation:

Support L2C8 startup and operation.

Source ofinformation: Siemens Power Corporation. .

Supplemental Distribution: J. J. Reimer (LS) LaSalle Central File D. A.Worthington R. H. Jacobs A. S. Pallotta M. Y. Hsiao Downers Grove Central File b

. - m . __

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