ML013600520

LaSalle County Station Unit 1, Cycle 9 Core Operating Limits Report

ML013600520
Person / Time
Site:	LaSalle
Issue date:	11/29/2001
From:	Pardee C G Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
-RFPFR
Download: ML013600520 (217)

Text

Exelon,M Exelon Generation Company, LLC www.exeloncorp .co Nuclear LaSalle County Station 2601 North 21'tRoad Marseilles, IL 61341-9757 November 29, 2001 10 CFR 50.4 United States Nuclear Regulatory Commission Attention:

Document Control Desk Washington, D.C. 20555 LaSalle County Station, Unit 1 Facility Operating License No. NPF-1 1 NRC Docket No. 50-373

Subject:

Unit 1 Cycle 9 Core Operating Limits Report Exelon Generation Company (EGC), LLC, in a letter dated, September 21, 2001, notified the NRC that the refueling outage for Unit 1 had been changed to January 10, 2002. The change in the refueling outage has resulted in a need to revise the Core Operating Limits Report (COLR). The COLR revision incorporates new operating limits for operation beyond the current analyzed exposure and an update to a name change in the fuel manufacturer.

Other administrative changes have also been incorporated.

Refer to Section 1, page i, for a summary of changes.

In accordance with Technical Specification Section 5.6.5, "Core Operating Limits Report," and 10 CFR 50.4, "Written Communications," LaSalle County Station is submitting this revision to the COLR to the NRC. Should you have any questions concerning this letter, please contact Mr. William Riffer, Regulatory Assurance Manager, at (815) 415-2800.

Respectfully, harles G. Pardee Site Vice President LaSalle County Station Attachment cc: Regional Administrator

-NRC Region III NRC Senior Resident Inspector

-LaSalle County Station L)

Technical Requirements Manual -Appendix I Section 1 LaSalle Unit 1 Cycle 9 Core Operating Limits Report November 2001 Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report Issuance of Changes Summary LaSalle Unit 1 Cycle 9 Affected Affected Summary of Changes Date Section Pages All All Original Issue (Cycle 9) 10/99 All All Incorporated administrative changes (including updating the 11/99 date to be November 1999) All All Incorporated changes to thermal limits due to uprate and 5/00 MELLLA operation, revised LHGR and MAPLHGR limits, CBH penalties, and necessary administrative changes.

All All Incorporated ITS changes, RBM trip setpoint and allowable 5/01 value equation change for DLO and SLO, TIP symmetry Chi Squared testing, added information on the use of SUBTIP that allows operation with reduced number of TIPs, incorporated the results of revised thermal limits with correct thermal conductivity and ITS scram times, and other necessary administrative changes.

All All Incorporates coastdown thermal limits on tables 2-1, 2-2, and 11/01 3-1. Define the coastdown core average exposure limit. Add applicable references for coastdown analysis and evaluations.

Changed SPC/Siemens to Framatome-ANP or FANP where applicable.

Updated the Neutronic Licensing Report.November 2001 i Technical Requirements Manual -Appendix I LI C9 Core Operating Limits Report Table of Contents References

.........................................................................................................................

1. Average Planar Linear Heat Generation Rate (3.2.1) ..........................................

1-1 1.1 Tech Spec Reference

................................................................................

1-1 1.2 Description

.................................................................................................

1-1 2. M inim um Critical Power Ratio (3.2.2) ..................................................................

2-1 2.1 Tech Spec Reference

................................................................................

2-1 2.2 Description

.................................................................................................

2-1 3. Linear Heat Generation Rate (3.2.3) ....................................................................

3-1 3.1 Tech Spec Reference

................................................................................

3-1 3.2 Description

.................................................................................................

3-1 4. Control Rod W ithdrawal Block Instrum entation (3.3.2.1)

..................................

4-1 4.1 Tech Spec Reference

................................................................................

4-1 4.2 Description

.................................................................................................

4-1 5. Allowed Modes of Operation (B 3.2.2, B 3.2.3) ....................................................

5-1 6. Traversing In-Core Probe System (3.2.1, 3.2.2, 3.2.3) .........................................

6-1 6.1 Tech Spec Reference

................................................................................

6-1 6.2 Description

.................................................................................................

6-1 6.3 Bases .........................................................................................................

6-1 LaSalle Unit 1 Cycle 9 November 2001 ii Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report References

1. Commonwealth Edison Company Docket No. 50-373, LaSalle County Station, Unit 1 Facility Operating License, License No. NPF-1 1. 2. Letter from D. M. Crutchfield to All Power Reactor Licensees and Applicants, Generic Letter 88-16; Concerning the Removal of Cycle Specific Parameter Limits from Tech Specs, dated October 4, 1988. 3. LaSalle Unit 1 Cycle 9 Neutronics Licensing Report (NLR), TODI NFM9900149, Sequence No. 01, November 2001. 4. LaSalle Unit 1 Cycle 9 Reload Analysis, EMF-2276, Revision 1, October, 1999. 5. LaSalle Unit 1 Cycle 9 Plant Transient Analysis, EMF-2277, Revision 1, October, 1999. 6. LOCA Break Spectrum Analysis for LaSalle Units 1 and 2, EMF-2174(P), March 1999. 7. LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM-9B fuel, EMF-2175(P), March 1999. 8. LaSalle Extended Operating Domain (EOD) and Equipment Out of Service (EOOS) Safety Analysis for ATRIUM-9B Fuel, EMF-95-205(P), Rev. 2, June 1996. 9. ARTS Improvement Program analysis for LaSalle County Station Units 1 and 2, NEDC-31531P, December 1993 and Supplement 1, June 1998 (Removal of Direct Scram Bypassed Limit). 10. Lattice-Dependent MAPLHGR Report for LaSalle County Station Unit 1 Reload 7 Cycle 8, 24A5180AA Revision 0, December 1995. 11. Lattice-Dependent MAPLHGR Report for LaSalle County Station Unit I Reload 6 Cycle 7, 23A7231AA, Rev.0, December 1993. 12. LaSalle Unit 1 Cycle 9 Principal Transient Analysis Parameters, EMF-96-189, May 1999. 13. General Electric Standard Application for Reactor Fuel (GESTAR), NEDE-24011-P-A-14, June 2000. 14. "Project Task Report, LaSalle County Station, Power Uprate Evaluation, Task 407: ECCS Performance," GE report number GE-NE A1300384-39-01, Revision 0, Class 3, dated September 1999. 15. Evaluation of a Postulated Slow Turbine Control Valve Closure Event for LaSalle County Station, Units 1 and 2. GE-NE-187-13-0792, Revision 2, July 1998. 16. Transient Analysis Evaluation for LaSalle 3 TCV Operation at Power Uprate and MELLLA Conditions, NFM:BSA:00-025, R.W. Tsai to D. Bost, April 13, 2000. 17. Updated Transient Analysis:

Abnormal Start-up of an Idle Recirculation Loop for LaSalle County Nuclear Station, Units 1 and 2, B33-00296 03P, March 1998. 18. "TIP Symmetry Testing", JHR:97:021, J.H. Riddle to R. Chin, January 20, 1997 and "TIP Symmetry Testing", DEG:99:085, D.Garber to R. Chin, March 23, 1999 19. "Use of SUBTIP Methodology with TIP Symmetry Testing Above 50 Percent Power", DEG:99:087, D. Garber to R.Chin, March 24, 1999 20. "On-Site and Off-Site Reviews of the GE Turbine Control Valve Slow Closure Analysis", T.Rieck to G.Spedl, NFS:BSS:93-117, May 19, 1993. 21. "LaSalle Units 1 and 2 Operating Limits with Multiple Equipment Out of Service (EOOS)", NFS:BSA:95-024, April 6, 1995. 22. NFM Calculation No. BSA-L-99-07, MAPFACf Thermal Limit Multiplier for 105% Maximum Core Flow 23. "Evaluation of CBH Effects on Fresh Fuel for LaSalle Unit 1 Cycle 9", DEG:00:025, D. Garber to R. Chin, February 25, 2000. 24. "ComEd GE9/GE10 LHGR Improvement Program" J11-03692-LHGR, Revision 1, February 2000. 25. "LaSalle County Station Power Uprate Project", Task 201: Reactor Power/Flow Map, GE-NE-Al 300384-07-01, Revision 1, September 1999 26. "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications", NEDO-32465-P-A, August 1996.LaSalle Unit I Cycle 9 November 2001 iii Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report 27. "ANFB Critical Power Correlation", ANF-1 125(P)(A) and Supplements 1 and 2, Advanced Nuclear Fuels Corporation, April 1990. 28. Letter, Ashok C. Thadani (NRC) to R. A. Copeland (SPC), "Acceptance for Referencing of ULTRAFLOWTM Spacer on 9X9-IXIX BWR Fuel Design," July 28, 1993. 29. Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors/Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors:

Methodology for Analysis of Assembly Channel Bowing Effects/NRC Correspondence, XN NF-524(P)(A)

Revision 2 and Supplement 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation November 1990. 30. COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis, ANF-913(P)(A), Volume 1, Revision 1 and Volume I Supplements 2, 3, and 4, Advanced Nuclear Fuels Corporation, August 1990. 31. HUXY: A Generalized Multirod Heatup Code with 10CFR50, Appendix K Heatup Option, ANF-CC-33(PXA), Supplement 1 Revision 1; and Supplement 2, Advanced Nuclear Fuels Corporation, August 1986 and January 1991, respectively.

32. Advanced Nuclear Fuels Methodology for Boiling Water Reactors, XN-NF-80-19(P)(A), Volume 1, Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990. 33. Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads, XN-NF-80-19(P)(A), Volume 4, Revision 1, Exxon Nuclear Company, June 1986. 34. Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description, XN-NF-80 19(P)(A), Volume 3, Revision 2, Exxon Nuclear Company, January 1987. 35. .Generic-Mechanical -Desigrr-for-Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-85-67(P)(A)

Revision 1, Exxon Nuclear Company, September 1986. 36. Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9X9-IX and 9X9-9X BWR Reload Fuel, ANF-89-014(P)(A), Revision I and Supplements 1 and 2, October 1991. 37. Volume 1 -STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain, Volume 2 -STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain, Code Qualification Report, EMF-CC-074(P)(A), Siemens Power Corporation, July 1994. 38. RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, XN-NF-81-58(P)(A), Revision 2 Supplements 1 and 2, Exxon Nuclear Company, March 1984. 39. XCOBRA-T:

A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, XN-NF-84-105(P)(A), Volume 1 and Volume 1 Supplements 1 and 2; Volume 1 Supplement 4, Advanced Nuclear Fuels Corporation, February 1987 and June 1988, respectively.

40. Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation, January 1993. 41. Exxon Nuclear Methodology for Boiling Water Reactors -Neutronic Methods for Design and Analysis, XN-NF-80-19(P)(A)

Volume 1 and Supplements 1 and 2, Exxon Nuclear Company, Richland, WA 99352, March 1983. 42. Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF-79-71(P)(A), Revision 2 Supplements 1, 2, and 3, Exxon Nuclear Company, March 1986. 43. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A), Revision 1 and Revision 1 Supplement 1, Advanced Nuclear Fuels Corporation, May 1995. 44. Reference Deleted.November 2001 LaSalle Unit I Cycle 9 iv Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report 45. Commonwealth Edison Topical Report NFSR-0085, "Benchmark of BWR Nuclear Design Methods," November 1990, Revision 0. 46. Commonwealth Edison Topical Report NFSR-0085, Supplement 1, "Benchmark of BWR Nuclear Design Methods -Quad Cities Gamma Scan Comparisons," April 1991, Revision 0. 47. Commonwealth Edison Topical Report NFSR-0085, Supplement 2, "Benchmark of BWR Nuclear Design Methods -Neutronic Licensing Analyses," April 1991, Revision 0. 48. Commonwealth Edison Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods," Revision 0, Supplements 1 and 2, December 1991, March 1992, and May 1992, respectively; SER letter dated March 22, 1993. 49. BWR Jet Pump Model Revision for RELAX, ANF-91-048(P)(A), Supplement 1 and Supplement 2, Siemens Power Corporation, October 1997. 50. ANFB Critical Power Correlation Application for Coresident Fuel, EMF-1 125(P)(A), Supplement 1, Appendix C, Siemens Power Corporation, August 1997. 51. ANFB Critical Power Correlation Determination of ATRIUM-9B Additive Constant Uncertainties, ANF-1125(P)(A), Supplement 1, Appendix E, Siemens Power Corporation, September 1998. 52. "POWERPLEX-II CMSS Startup Testing", DEG:00:254, D. Garber to R. Chin, December 5, 2000. 53. "POWERPLEX-lI CMSS Startup Testing", DEG:00:256, D. Garber to R. Chin, December 6, 2000. 54. "LaSalle Unit 1 Cycle 9 Operating Limits for Proposed ITS Scram Times and corrected Fuel Thermal Conductivity", DEG: 01:045, D. Garber to R. Chin, March 22, 2001. 55. "LaSalle Unit 1 and Unit 2 Rod Block Monitor COLR Setpoint Change", NFM:MW:01-0106, A. Giancatarino to J. Nugent, April 3, 2001. 56. "LaSalle Unit 1 Cycle 9 Operating Limits for Proposed Cycle Extension", DEG:01:148, D. Garber to F. Trikur, September 21, 2001. 57. "LaSalle Unit 1 Cycle 9 GE9 Mechanical Limits for Proposed Cycle Extension", DEG:01:143, D. Garber to F. Trikur, September 18, 2001. 58. "Evaluation of L1C9 Cycle Extension Transient Analysis Results for Compliance with GE Fuel Mechanical Limits", NFM-MW:01-0335, C. de la Hoz to J. Nugent, October 25, 2001. 59. "LaSalle Unit 1 Cycle 9 Coastdown Evaluation", NFM-MW:01-0349 Revision 1, F. W. Trikur to J. Nugent, November 7, 2001.LaSalle Unit 1 Cycle 9 November 2001 V Technical Requirements Manual -Appendix I LI C9 Core Operating Limits Report 1. Average Planar Linear Heat Generation Rate (3.2.1) 1.1 Tech Spec

Reference:

Tech Spec 3.2.1 1.2

Description:

1.2.1 GE Fuel The MAPLHGR Limit is determined using the applicable Lattice-Type MAPLHGR limits from Tables 1.2-1, 1.2-2, 1.2-3, and 1.2-4. For Single Reactor Recirculation Loop Operation, the MAPLHGR limits in Tables 1.2-1, 1.2-2, 1.2-3, and 1.2-4 are multiplied by the MAPFAC multipliers provided in Figures 1.2-1 and 1.2-2.Table for Fuel-Type MAPLHGR Limits 1.2-1 1.2-2 1.2-3 1.2-4 Fuel Type (Reference

3) GE9B-P8CWB322-11GZ-100M-150-T GE9B-P8CWB320-9GZ-1 OOM-1 50-T GE9B-P8CWB343-12GZ-80M-150-T GE9B-P8CWB342-1 OGZ-80M-1 50-T 1.2.2 Framatome-ANP (FANP is formerly known as SPC) Fuel The MAPLHGR Limit is the- Lattice-Type MAPLHGR Limit. The Lattice-Type Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits are determined from the table given below: Fuel Type- Cycle First Inserted (References 3 and 4) SPCA9-393B-16GZ-100M 9 SPCA9-396B-12GZB-100M 9 SPCA9-384B-1 I GZ6-80M 9 SPCA9-396B-12GZC-10 OM 9 Planar Average Exposure MAPLHGR (kWlft) (GWdIMTU) (all FANP fuel types) (References 4 and 7)0.0 20.0 61.1 13.5 13.5 9.39 For single loop operation (or Abnormal Idle Loop Startup, UFSAR 15.4.4), the MAPLHGR multiplier for Framatome-ANP (FANP is formerly known as SPC) fuel is 0.90. (References 4, 6 and 7)LaSalle Unit 1 Cycle 9 Cycle First Inserted 7 7 8 8 1-1 November 2001 Technical Requirements Manual -Appendix I LIC9 Core Operating Limits Report Table 1.2-1 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) vs. Average Planar Exposure for Fuel Type GE9B-P8CWB322-1 1 GZ-1 OOM-1 50-T (References 11 and 24) Exposure (MWD/ST) Lattice Specific MAPLHGR limit (KW/ft) 0 12.74 12.09 11.65 11.25 12.11 12.74 200 12.67 12.13 11.70 11.32 12.15 12.67 1000 12.48 12.22 11.83 11.46 12.25 12.48 2000 12.42 12.35 12.00 11.61 12.39 12.42 3000 12.41 12.48 12.14 11.77 12.54 12.41 4000 12.44 12.62 12.28 11.94 12.70 12.44 5000 12.46 12.77 12.43 12.11 12.86 12.46 6000 12.49 12.90 12.58 12.29 13.02 12.49 7000 12.51 13.03 12.73 12.46 13.19 12.51 8000 12.54 11-3,16 12.88 12.64 13.33 12.54 9000 12.55 13.30 13.01 12.82 13.43 12.55 10000 12.57 13.42 13.12 12.98 13.44 12.57 12500 12.41 13.41 13.08 13.04 13.40 12.41 15000 12.04 13.05 12.78 12.77 13.06 12.04 20000 11.27 12.38 12.16 12.16 12.40 11.27 25000 10.49 11.74 11.51 11.51 11.76 10.49 27215.6 12.314 12.314 12.314 12.314 12.314 12.314 48080.8 10.800 10.800 10.800 10.800 10.800 10.800 58967.1 6.000 6.000 6.000 6.000 6.000 6.000 Lattice No. 733 1817 1818 1819 1820 1821 LaSalle Unit 1 Cycle 9 November 2001 1-2 Technical Requirements Manual -Appendix I LI C9 Core Operating Limits Report Table 1.2-2 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) vs. Average Planar Exposure for Fuel Type GE9B-P8CWB320-9GZ-1 OOM-1 50-T (References 11 and 24) Exposure (MWD/ST) Lattice Specific MAPLHGR limit (kwfft) 0 12.74 12.05 11.62 11.10 12.09 12.74 200 12.67 12.09 11.64 11.15 12.14 12.67 1000 12.48 12.19 11.73 11.27 12.25 12.48 2000 12.42 12.32 11.86 11.44 12.39 12.42 3000 12.41 12.44 11.99 11.62 12.53 12.41 4000 12.44 12.57 12.13 11.80 12.67 12.44 5000 12.46 12.70 12.27 11.96 12.81 12.46 6000 12.49 12.83 12.42 12.09 12.89 12.49 7000 12.51 12.97 12.54 12.23 12.98 12.51 8000 12.54 13.07_. 12.62 12.37 13.07 12.54 9000 12.55 13.15 12.70 12.51 13.15 12.55 10000 12.57 13.20 12.77 12.66 13.22 12.57 12500 12.41 13.19 12.70 12.67 13.20 12.41 15000 12.04 12.89 12.40 12.40 12.90 12.04 20000 11.27 12.29 11.82 11.82 12.30 11.27 25000 10.49 11.69 11.25 11.25 11.70 10.49 27215.6 12.314 12.314 12.314 12.314 12.314 12.314 48080.8 10.800 10.800 10.800 10.800 10.800 10.800 58967.1 6.000 6.000 6.000 6.000 6.000 6.000 Lattice No. 733 1812 1813 1814 1815 1816 LaSalle Unit 1 Cycle 9 1-3 November 2001 Technical Requirements Manual -Appendix I Li C9 Core Operating Limits Report Table 1.2-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) vs. Average Planar Exposure for Fuel Type GE9B-P8CWB343-12GZ-80M-l 50-T (References 10 and 24) Exposure (MWD/ST) Lattice Specific MAPLHGR limit (kw/ft) 0 12.66 11.69 11.37 10.92 12.66 200 12.59 11.71 11.43 10.99 12.59 1000 12.40 11.78 11.55 11.13 12.40 2000 12.34 11.95 11.72 11.33 12.34 3000 12.34 12.16 11.91 11.54 12.34 4000 12.37 12.40 12.11 11.76 12.37 5000 12.40 12.67 12.32 12.00 12.40 6000 12.43 12.90 12.53 12.24 12.43 7000 12.46 13.05 12.76 12.49 12.46 8000 12.48 13.21 12.98 12.75 12.48 9000 12.50 13.37 13.13 13.01 12.50 10000 12.51 13.54 13.30 13.22 12.51 12500 12.35 13.75 13.60 13.57 12.35 15000 11.98 13.48 13.23 13.21 11.98 20000 11.20 12.71 12.40 12.37 11.20 25000 10.42 11.92 11.60 11.57 10.42 27215.6 12.314 12.314 12.314 12.314 12.314 48080.8 10.800 10.800 10.800 10.800 10.800 58967.1 6.000 6.000 6.000 6.000 6.000 Lattice No. 732 2083 2084 2085 2086 LaSalle Unit I Cycle 9 November 2001 1-4 Technical Requirements Manual -Appendix I LI C9 Core Operating Limits Report Table 1.2-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) vs. Average Planar Exposure for Fuel Type GE9B-P8CWB342-1 OGZ-80M-1 50-T (References 10 and 24)Exposure (MWD/ST) Lattice Specific MAPLHGR limit (kw/ft) 0 12.66 12.04 12.25 11.72 12.09 12.66 200 12.59 12.08 12.28 11.77 12.12 12.59 1000 12.40 12.16 12.35 11.87 12.22 12.40 2000 12.34 12.28 12.45 12.00 12.37 12.34 3000 12.34 12.42 12.55 12.13 12.53 12.34 4000 12.37 12.57 12.65 12.27 12.70 12.37 5000 12.40 12.73 12.76 12.41 12.88 12.40 6000 12.43 12.89 12.87 12.56 13.07 12.43 7000 12.46 13.06 12.98 12.72 13.27 12.46 8000 12.48 13.24 13.10 12.88 13.47 12.48 9000 12.50 13.42 13.21 13.05 13.65 12.50 10000 12.51 13.61 13.31 13.21 13.76 12.51 12500 12.35 13.79 13.35 13.31 13.82 12.35 15000 11.98 13.50 13.06 13.05 13.51 11.98 20000 11.20 12.79 12.47 12.45 12.79 11.20 25000 10.42 11.95 11.67 11.63 11.95 10.42 27215.6 12.314 12.314 12.314 12.314 12.314 12.314 48080.8 10.800 10.800 10.800 10.800 10.800 10.800 58967.1 6.000 6.000 6.000 6.000 6.000 6.000 Lattice No. 732 2087 2088 2089 2090 2091 LaSalle Unit 1 Cycle 9 November 2001 1-5 Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report Figure 1.2-1 Power-Dependent SLO and Abnormal Idle Loop Startup MAPLHGR Multipliers for GE Fuel, MAPFACP (Reference 9 and 24)1 0.95 0.9 C., < 0.85 LL .0.8 = 0.75 0.7 75 0.65 0.6 0 0.55 .-J .0.5 S0.45 0 -J 0.4 U/) -0.35 "G) , 0.3 d) o 0.25 0.2 o 0.15 0. 0.1 0.05 0 25 30 35 40 45 50 55 60 65 70 Core Thermal Power (% Rated) 1-6 75 80 85 90 95 100 November 2001 LaSalle Unit 1 Cycle 9 For 25>P 1 MAPA0 =0 .1000240P-0 For 100>P: AF~p=10 No Thermal Limits Monitoring Required; If Official Monitoring is Desired, the Equations for _> 25% Power May Be Extrapolated for 25 > P, provided the Official monitoring is only performed with the TCV/TSV closure scrams and RPT enabled.

For 25:< P < 100 MAPFACp = 1.0+0.005224 (P-100) For 100 < P, MAPFACp = 1.00 P = % Rated Core Thermal Power Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report Figure 1.2-2 Flow-Dependent SLO and Abnormal Idle Loop Startup MAPLHGR Multiplier for GE Fuel, MAPFACF (References 9, 17, 22, and 24) 000 0-For 105% Maximum Attainable Core Flow 000000 00MAPFACF

= The Minimum of EITHER 1.0 0 C ,,"OR (0.6807 x (WT/100)+0.4672}

WT = % Rated Core Flow LaSalle Unit 1 Cycle 9 1 L. 0~ L 0 .6j U) CL 0. 0 MJ 0.9 0.8 0.7 0.6 0.5 0.4 0.3 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 Core Flow (% Rated)1-7 November 2001 Technical Requirements Manual -Appendix I Li C9 Core Operating Limits Report 2. Minimum Critical Power Ratio (3.2.2) 2.1 Tech Spec

Reference:

Tech Spec 3.2.2. 2.2

Description:

Prior to initial scram time testing for an operating cycle, the MCPR operating limit is to be based on the Technical Specification Scram Times. For Technical Specification requirements refer to Technical Specification table 3.1.4-1.

TIP symmetry Chi-squared testing shall be performed prior to reaching 500 MWd/MTU to validate the MCPR calculation.

MCPR limits from BOC to Coastdown are applicable up to a core average exposure of 29,439 MWd/MTU (which is the licensing basis exposure used by Framatome-ANP). (Reference

4) MCPR limits from Coastdown to EOL are applicable from a core average exposure of 29,439 MWd/MTU to a core average exposure of 31,062 MWd/MTU. (Reference

56) 2.2.1 Manual Flow Control MCPR Limits The Governing MCPR Operating Limit while in Manual Flow Control is either determined from 2.2.1.1 or 2.2.1.2, whichever is greater at any given power, flow condition.

2.2.1.1 Power-Dependent MCPR (MCPRp)* (Reference 3, 4, 23, and 56) 2.2.1.1.1 GE Fuel Table 2-1 gives the MCPRp limit as a function of core thermal power for Tech Spec Scram Speeds. 2.2.1.1.2 Framatome-ANP (formerly known as Siemens or SPC) Fuel Table 2-2 gives the MCPRp limit as a function of core thermal power for Tech Spec Scram Speeds. Note that the 1 OB rods are defined by the control cell locations 14 39, 22-15, 46-23, 38-47, 14-23, 38-15, 46-39, and 22-47. 2.2.1.2 Flow-Dependent MCPR (MCPRF) (Reference

4) Table 2-3 gives the MCPRF limit as a function of flow. 2.2.2 Automatic Flow Control MCPR Limits Automatic Flow Control MCPR Limits are not provided for Ll C9.

• For thermal limit monitoring cases at greater than 100%P, the 100% power MCPRp limits should be applied.November 2001 LaSalle Unit 1 Cycle 9 2-1 Technical Requirements Manual -Appendix I LIC9 Core Operating Limits Report Table 2-1 MCPRp for GE Fuel (References 3, 4, 5 54, and 56)Operation from BOC to Coastdown Percent Core Thermal Power* EOOS Combination 0 25 25(25.1) 60 80 80 (80.1) 100 NoEOOS 2.70 2.20 2.10 1.57 1.53 1.50 Single RR Loop Only 2.71 2.21 2.11 1.58 1.54 1.51 EOOS'* 2.85 2.35 2.35 1.71 1.69 1.58 EOOS/Single RR Loop** 2.86 2.36 2.36 1.72 1.70 1.59 Coastdown Operation Percent Core Thermal Power* EOOS Combination 0 25 25(25.1) 60 80 80(80.1) 100 No EOOSt 2.85 2.35 2.35 1.62 1.50 Single RR Loop Only 2.86 2.36 2.36 1.63 1.51 EOOS** 2.85 2.35 2.35 1.71 1.69 1.61 EOOS/Single RR Loop** 2.86 2.36 2.36 1.72 1.70 1.62

• Values are interpolated between relevant power levels. For operation at exactly 25% or 80% CTP, the more limiting value is used. 3489 MWt is rated power. ** Allowable EOOS conditions are listed in Section 5. Other EOOS conditions are not covered.

1 For coastdown operation the NO EOOS option includes final feedwater temperature reduction (FFTR) and /or feedwater heaters out of service up to 100 OF.LaSalle Unit 1 Cycle 9 November 2001 2-2 Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report Table 2-2 MCPRp for Framatome-ANP Fuel (References 3, 4, 5, 23, 54 and 56) For Operation at exposures from 11000 MWD/MTU to Coastdown All Framatome-ANP (formerly known as SPC) fuel except fuel type 36 in 10B cell locations and fuel type 46 and 47 in Al cell locations Percent Core Thermal Power* EOOS Combination 0 25 25 (25.1) 60 80 80 (80.1) 100 NoEOOS 2.70 2.20 2.05 1.56 1.51 1.46 Single RR Loop only 2.71 2.21 2.06 1.57 1.52 1.47 EOOS** 2.85 2.35 2.35 1.67 1.64 1.54 EOOS**/Single RR Loop 2.86 2.36 2.36 1.68 1.65 1.55 Framatome-ANP fuel that is fuel type 36 in 10B cell locations and fuel type 46 and 47 in Al cell locations Percent Core Thermal Power* EOOS Combination 0 25 25(25.1) 60 80 80 (80.1) 100 No EOOS 2.74 2.24 2.09 1.60 1.55 1.48 Single RR Loop only 2.75 2.25 2.10 1.61 1.56 1.49 EOOS* 2.89 2.39 2.39 1.71 1.66 1.56 EOOS**/Single RR Loop 2.90 2.40 2.40 1.72 1.67 1.57 Coastdown Operation All Framatome-ANP (formerly known as SPC) fuel except fuel type 36 in 10B cell locations Percent Core Thermal Power* EOOS Combination 0 25 25(25.1) 60 80 80 (80.1) 100 No EOOSt 2.85 2.35 2.35 1.62 1.46 Single RR Loop Only 2.86 2.36 2.36 1.63 1.47 EOOS** 2.85 2.35 2.35 1.69 1.64 1.57 EOOS/Single RR Loop** 2.86 2.36 2.36 1.70 1.65 1.58 Framatome-ANP fuel that is fuel type 36 in 10B cell locations Percent Core Thermal Power* EOOS Combination 0 25 25 (25.1) 60 80 80 (80.1) 100 No EOOSt 2.87 2.37 2.37 1.64 1.48 Single RR Loop only 2.88 2.38 2.38 1.65 1.49 EOOS* 2.87 2.37 2.37 1.71 1.66 1.59 EOOS*/Single RR Loop 2.88 2.38 2.38 1.72 1.67 1.60

• Values are interpolated between relevant power levels. For operation at exactly 25% and 80% CTP, the more limiting value is used. 3489 MWt is rated power. ** Allowable EOOS conditions are listed in Section 5. Other EOOS conditions are not covered.

t For coastdown operation the NO EOOS option includes final feedwater temperature reduction (FFTR) and /or feedwater heaters out of service up to 100 OF.November 2001 LaSalle Unit I Cycle 9 2-3 Technical Requirements Manual -Appendix I LIC9 Core Operating Limits Report Table 2-3 MCPRFfor GE and Framatome-ANP Fuel (References 4 & 5) MCPRf limits for 105% Maximum Attainable Core Flow ed) MCPRf ATRIUM-9BI 1.93 1.93 1.14 1.11 The MCPRf limits are applicable from BOC through coastdown and in all EOOS scenarios.

November 2001 LaSalle Unit 1 Cycle 9 2-4 Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report 3. Linear Heat Generation Rate (3.2.3) 3.1 Tech Spec

Reference:

Tech Spec 3.2.3. 3.2

Description:

3.2.1 GE Fuel a. The LHGR Limit is the product of the LHGR Limit in the following tables and the minimum of either the power dependent LHGR Factor*, LHGRFACp or the flow dependent LHGR Factor, LHGRFACF.

The LHGR Factors (LHGRFACp and LHGRFACF) for the GE fuel is determined from Figures 3.2-1 through 3.2-3. The following LHGR limits apply for the entire cycle exposure range: (References 9, 14 and 24) 1. GE9B-P8CWB322-11GZ-100M-150-T (bundle 3861 in Reference 24)Nodal Exposure (GWd/MT) LHGR Limit (KW/ft) 0.00 14.40 12.34 14.40 26.80 12.31 33.07 11.82 38.58 11.35 44.09 10.94 49.11 10.80 60.89 6.00 2. GE9B-P8CWB320-9GZ-100M-150-T (bundle 3860 in Reference

24) Nodal Exposure (GWd/MT) LHGR Limit (KW/ft) 0.00 14.40 12.14 14.40 26.19 12.31 48.16 10.80 59.93 6.00 3. GE9B-P8CWB343-12GZ-80M-150-T (bundle 3866 in Reference

24) Nodal Exposure (GWd/MT) LHGR Limit (KWIft) 0.00 14.40 12.33 14.40 27.86 12.31 49.76 10.80 61.18 6.00 4. GE9B-P8CWB342-10GZ-80M-150-T (bundle 3867 in Reference 24)Nodal Exposure (GWd/MT) LHGR Limit (KW/ft) 0.00 14.40 12.71 14.40 27.52 12.31 49.54 10.80 60.95 6.00 LaSalle Unit I Cycle 9 November 2001 3-1 Technical Requirements Manual -Appendix I LI C9 Core Operating Limits Report 3.2.2 Framatome-ANP (formerly known as SPC or Siemens) Fuel The LHGR Limit is the product of the Steady-State LHGR Limit and the minimum of either the power dependent LHGR Factor*, LHGRFACp or the flow dependent LHGR Factor, LHGRFACF.

The Steady-State LHGR limits are given below (Reference 4). LHGRFACp is determined from Table 3-1. LHGRFACF is determined from Table 3-2. FANP LHGRFACP multipliers in this COLR for BOC to coastdown are applicable up to a core average exposure of 29,439 MWd/MTU (Reference 4). FANP LHGRFACp multipliers in this COLR for coastdown operation are applicable up to a core average exposure of 31,062 MWd/MTU (Reference 56). Framatome-ANP Fuel Steady-State LHGR Limits for the following fuel types: 1. SPCA9-393B-16GZ-100M

2. SPCA9-396B-12GZB-100M

3. SPCA9-384B-1 GZ6-80M 4. SPCA9-396 B- 12GZC- 100M LHGR limits for all Framatome-ANP fuel from BOC through Coastdown (excluding fuel type 36 in 1 OB locations from rod pattern targeted for approximately 9000 MWD/MTU to rod pattern targeted approximately for 12,000MWD/IMTU)

Planar Average Exposure (GWd/MTU)

LHGR limit (kW/ft) (Reference

4) 0.0 14.4 15.0 14.4 61.1 8.32 LHGR limits for Framatome-ANP fuel type 36 in 10B locations (from rod pattern targeted at approximately 9000 MWD/MTU to rod pattern targeted at approximately 12,000 MWD/MT) Planar Average Exposure (GWd/MTU)

LHGR limit (kW/ft) (References 4 and 23) 0.0 14.05 15.0 14.05 61.1 7.97 Note that the 1OB rods are defined by the control cell locations 14-39, 22-15, 46-23, 38-47, 14-23, 38-15, 46-39, and 22-47.

• For thermal limit monitoring cases at greater than 100%P, the 100% power LHGRFACp limits should be applied.LaSalle Unit 1 Cycle 9 November 2001 3-2 Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report Figure 3.2-1 Power-Dependent LHGR Multipliers for GE fuel (formerly MAPFACp) (Reference 9 and 24)a. ".W -J C, ,1 C, -J "(D 4, CL 0 a) 0.I 0.95 0.9 0.85 0.8 0.75 0.7 0.65 0.6 0.55 0.5 0.45 0.4 0.35 0.3 0.25 0.2 0.15 0.1 0.05 (1 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Core Thermal Power (% Rated)3-3 November 2001 LaSalle Unit I Cycle 9 Technical Requirements Manual -Appendix I LIC9 Core Operating Limits Report Figure 3.2-2 Power-Dependent LHGR Multiplier for GE Fuel (TCV(s) Slow Closure) (formerly MAPFACp) (Reference 15 and 24)0. U CL _C, C a) 0 0. -r 0., a) 0 o o.10.9 0.95 0.98 0.85 0.8 0.75 0.7 0.65 0.6 0.55 0.5 0.45 0.4 0.35 0.3 0.25 0.2 0.15 0.1 0.05 0 3-4 November 2001 0 10 20 30 40 50 60 70 80 90 Core Thermal Power (% Rated)100 LaSalle Unit 1 Cycle 9 Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report Figure 3.2-3 Flow-Dependent LHGR Multiplier for GE Fuel (formerly MAPFACF) (Reference 9 and 17, 22, and 24)1 0.9 0 0.8 .2. *. 0.7 ., CD 05 0.6 CL 0.5 0 u. 0.4 40 45 50 55 60 65 70 Core Flow (% Rated)75 80 85 90 95 100 3-5 November 2001 For 105% Maximum Attainable Core Flow LHGRFACF = The Minimum of EITHER 1.0 0 OR {0.6807 x (WT/100)+0.4672}

WT = % Rated Core Flow For Abnormal Idle Loop Startup, LHGRFACF 0.40 SI I V I I I I I It LL 0.3 30 35 105 LaSalle Unit I Cycle 9 Technical Requirements Manual -Appendix I LIC9 Core Operating Limits Report Table 3-1 LHGRFACpfor Framatome-ANP Fuel (References 4, 5, 54 and 56) Operation from BOC to Coastdown[EOOS Combination No EOOS Single RR Loop only EOOS** EOOS/Single RR Loop-*Percent Core Thermal Power* 0 25 25 60 0,67 0.67 0.67 0.94 0.67 0.67 0.67 0.94 0.64 0.64 0.64 0.64 0.64 0.64 80 0.98 0.98 0.86 0.86 80 0.86 0.86 100 1.00 1.00 0.86 0.86 Coastdown Operation Percent Core Thermal Power* EOOS Combination 0 25 25 60 80 80 100 No EOOSI 0.64 0.64 0.64 0.91 0.93 Single RR Loop Only 0.64 0.64 0.64 0.91 0.93 EOOS** 0.64 0.64 0.64 0.83 0.83 0.83 EOOS/Single RR Loop** 0.64 0.64 0.64 0.83 0.83 0.83* Values are interpolated between relevant power levels. For operation at exactly 25% or 80% CTP, the more limiting value is used. ** Allowable EOOS conditions are listed in Section 5. t For coastdown operation the NO EOOS option includes final feedwater temperature reduction (FFTR) and bor feedwater heaters out of service up to 100 OF.LaSalle Unit 1 Cycle 9 I November 2001 3-6 I Technical Requirements Manual -Appendix I LI C9 Core Operating Limits Report Table 3-2 LHGRFACF for Framatome-ANP Fuel (References 4 & 5) Values Applicable for up to 105% Maximum Attainable Core Flow Flow (% rated) LHGRFACf ATRIUM-9B 0 0.69 30 0.69 76 1.00 105 1.00 These LHGRFACf multipliers apply from BOC through coastdown and in all EOOS scenarios.

LaSalle Unit 1 Cycle 9 3-7 November 2001 Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report 4. Control Rod Withdrawal Block Instrumentation (3.3.2.1) 4.1 Tech Spec

Reference:

Tech Spec Table 3.3.2.1-1.

4.2

Description:

The Rod Block Monitor Upscale Instrumentation Setpoints are determined from the relationships shown below: ROD BLOCK MONITOR UPSCALE TRIP FUNCTION Two Recirculation Loop Operation*

Single Recirculation Loop Operation*

TRIP SETPOINT 0.66 W + 51%** 0.66 W + 45.7%**ALLOWABLE VALUE 0.66 W + 54%** 0.66 W + 48.7%**This setpoint may be lower/higher and will still comply with the RWE Analysis, because RWE is analyzed unblocked.

Clamped, with an allowable value not to exceed the allowable value for recirculation loop flow (W) of 100%.LaSalle Unit 1 Cycle 9 4-1 November 2001 Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report 5. Allowed Modes of Operation (B 3.2.2, B 3.2.3) The Allowed Modes of Operation with combinations of Equipment Out-of-Service are as described below: -------.OPERATING REGION -....--Equipment Out of Service Options' Standard MELLLA ICF 7 Coastdown 9 None Yes Yes Yes Yes Feedwater Heaters 2 (Reference 9 and 56) Yes No 3 Yes Yes Single RR Loop 1° (Reference 9 and 56) Yes No 8 N/A Yes Turbine Bypass Valves (Reference

9) Yes Yes Yes No EOC Recirculation Pump Trip (Reference 9 and 56) Yes Yes Yes Yes TCV Slow Closure/EOC Recirculation Pump Trip (Reference 15 Yes Yes Yes Yes and 56) TCV Slow Closure/EOC Recirculation Pump Trip / Yes No 3 Yes Yes Feedwater Heaters 2 (Reference 15, 20,21, and 56) Turbine Bypass Valves / Feedwater Heaters 2 (Reference

9) No No No 5 No EOC Recirculation Pump Trip / Yes 4 No 3 Yes 4 Yes Feedwater Heaters 2 (Reference 9 and 56) TCV Stuck Closed 6 (Reference

16) Yes Yes Yes No 1. Each EOOS condition may be combined with one SRV OOS, up to two TIP Machines OOS or the equivalent number of TIP channels (100% available at startup from a refuel outage), a 20°F reduction in feedwater temperature (without Feedwater Heaters considered OOS), cycle startup with uncalibrated LPRMs (BOC to 500 MWd/MTU), and/or up to 50% of the LPRMs out of service.

2. Up to 1 00°F Reduction in Feedwater Temperature Allowed with Feedwater Heaters Out-of-Service or in combination with FFTR during coastdown.

Feedwater Heaters OOS may be an actual OOS condition, or an intentionally entered mode of operation to extend the cycle energy. As long as this condition is met, this is not an EOOS for coastdown.

3. If operating with Feedwater Heaters Out-of-Service, operation in MELLLA is supported by current transient analyses, but administratively prohibited due to core stability concerns.

4. EOC Recirculation Pump Trip OOS/Feedwater Heaters OOS is allowed during coastdown/non coastdown operation using the TCV Slow Closure/EOC Recirculation Pump Trip OOS/Feedwater Heaters OOS operating limits. 5. Only when operating in coastdown, otherwise this combination is not allowed. This is not applicable.

6. Operation is only allowed when less than 10.5 million Ibm/hr steam flow and when average position of 3 open TCVs is less than 50% open, with FCL <103%, and the MCFL setpoint _> 120%. TCV Stuck Closed may be in combination with any EOOS except TBVOOS or TCV Slow Closure. If in combination with other EOOS(s), thermal limits may require adjustment for the other EOOS(s) as designated in Sections 1, 2, and 3. 7. Increased Core Flow (ICF) is analyzed for up to 105% core flow. 8. The SLO boundary was not moved up with the incorporation of MELLLA. The flow boundary for SLO at uprated conditions remains the ELLLA boundary for pre-uprate conditions. (Reference

25) 9. Coastdown is defined to begin at a core average exposure of 29,439 MWd/MTU (which is the licensing basis exposure used by Framatome-ANP) (Reference

4) and applicable to a core average exposure of 31,062 MWd/MTU (Reference 56). 10. Single Loop Operation is allowed with any of the EOOS options listed in this table.LaSalle Unit 1 Cycle 9 November 2001 5-1 Technical Requirements Manual -Appendix I L1C9 Core Operating Limits Report 6. Traversing In-Core Probe System (3.2.1, 3.2.2, 3.2.3) 6.1 Tech Spec

Reference:

Tech Spec Sections 3.2.1, 3.2.2, 3.2.3 for thermal limits require the TIP system for recalibration of the LPRM detectors and monitoring thermal limits. 6.2

Description:

When the traversing in-core probe (TIP) system (for the required measurement locations) is used for recalibration of the LPRM detectors and monitoring thermal limits, the TIP system shall be operable with the following:

1. movable detectors, drives and readout equipment to map the core in the required measurement locations, and 2. indexing equipment to allow all required detectors to be calibrated in a common location.

For BOC to BOC + 500 MWD/MT, cycle analyses support thermal limit monitoring without the use of the TIPs. Following the first TIP set (required prior to BOC + 500 MWD/IMT), the following applies for use of the SUBTIP methodology:

With one or more TIP measurement locations inoperable, the TIP data for an inoperable measurement location may be replaced by data obtained from a 3-dimensional BWR core monitoring software system adjusted using the previously calculated uncertainties, provided the following conditions are met: 1. All TIP traces have previously been obtained at least once in the current operating cycle when the reactor core was operating above 20% power, (References 18, 52 and 53) and 2. The total number of simulated channels (measurement locations) does not exceed 42% (18 channels).

Otherwise, with the TIP system inoperable, suspend use of the system for the above applicable monitoring or calibration functions.

6.3 Bases

The operability of the TIP system with the above specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core. The normalization of the required detectors is performed internal to the core monitoring software system. Substitute TIP data, if needed, is 3-dimensional BWR core monitoring software calculated data which is adjusted based on axial and radial factors calculated from previous TIP sets. Since uncertainty could be introduced by the simulation and adjustment process, a maximum of 18 channels may be simulated to ensure that the uncertainties assumed in the substitution process methodology remain valid.LaSalle Unit 1 Cycle 9 November 2001 6-1 Technical Requirements Manual -Appendix I Section 2 LaSalle Unit 1 Cycle 9 Reload Transient Analysis Results November 2001 Technical Requirements Manual -Appendix I L1C9 Reload Transient Analysis Results Table of Contents Attachment 1 2 3 4 5 6 7 Preparer Exelon Siemens Power Corporation Siemens Power Corporation General Electric General Electric Framatome ANP Framatome ANP Document Neutronics Licensing Report Reload Analysis Plant Transient Analysis (Excerpts)

ARTS Improvement Program Analysis, Supplement I (Excerpts)

TCV Slow Closure Analysis (Excerpts)

LaSalle Unit 1 Cycle 9 Operating Limits for Proposed Scram Times and Corrected Fuel Thermal Conductivity LaSalle Unit I Cycle 9 Operating Limits for Proposed Cycle Extension LaSalle Unit 1 Cycle 9 November 2001 Technical Requirements Manual -Appendix I LI C9 Reload Transient Analysis Results Attachment 1 LaSalle Unit 1 Cycle 9 Neutronics Licensing Report LaSalle Unit I Cycle 9 November 2001 NUCLEAR FUEL MANAGEMENT TRANSMITTAL OF DESIGN INFORMATION 0 SAFETY RELATED Originating Organization NFM ID# NFM9900149 C NON-SAFETY RELATED 0 Nuclear Fuel Management Seq. No. 01 LI REGULATORY RELATED El Other (specify)

_ Page I of 27 Station: LaSalle Unit: I Cycle: 9 Generic: To: Kirk Peterman (LaSalle)

Subject:

/ LaSalle 1 Cycle 9 Neutronics Licensing ReorALR) Frank W. Trikur// o Preparer Prepa er's Signature Date Ming Y. Hsiao__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ Reviewer Reviewer'

ýiature Date Anthony D. Giancatarino

/1/0 0/ NFM Department Head Approver's Sipnature C\_ Date Status of Information:

0 Verfied D Unverified

[ Engineering Judgement Action Tracking # for Method and Schedule of Verification for Unverified DESIGN INFORMATION:

Description of Information:

LaSalle Unit I Cycle 9 Neutronics Licensing Report. Results and bases of neutronics licensings calculations for LaSalle I Cycle 9. These calculations cover operation with a rated core power up to 3489 MWt. Purpose of Information:

LaSalle Unit I Cycle 9 Neutronics Licensing Report Seq. 0: Original issue Seq. 1: Revised LI C9 peak fuel pellet burnup limit for the GE9B fuel design in the Maximum Exposure Limit Compliance Table.Source of Information:

NFM Calculation Note BNDL:99-050, Revision 0. Maximum peak pellet bumup limit for GE9B fuel is taken from the GE9/GEIO LHGR Improvement Program, J.l-03692-LHGR, Revision I, Class 3, February 2000 report transmitted by NDIT No. NFMO000067, Seq. No. 0. Supplemental Distribution:

Jeff Nugent (LaSalle), Norha Plumey (LaSalle), Adelmo S. Pallotta, Robert W. Tsai, Pedro L. Kong, LaSalle Central File, Cantara Central File Licensing Basis This document, in conjunction with References 1, 3 and 4 in Section VIII, provides the licensing basis for LaSalle County Station Unit I Reload 8, Cycle 9. Table of Contents I. Nuclear Design 1.1 New Reload Fuel Assembly Nuclear Design I.L1 Assembly Average Enrichment

1.1.2 Axial

Enrichment and Burnable Poison Distribution

. .3 -_RadialEnrichment and Burnable Poison-Distribution 1.2 Core Nuclear Design 1.2.1 Core Configuration and Licensing Exposure Limits 1.2.2 Core Reactivity Characteristics II. Control Rod Withdrawal Error I11. Fuel Loading Error IV. Control Rod Drop Accident V. Loss of Feedwater Heating VI. Maximum Exposure Limit Compliance VII. Spent Fuel Pool and Fresh Fuel Vault Criticality Compliance VII.1 Fresh Fuel Vault Criticality Compliance VII.2 L1 Spent Fuel Pool Criticality Compliance VII.3 L2 Spent Fuel Pool Criticality Compliance VIII. References 3 3 3 3 4 4 5 5 6 6 7 7 8 8 8 9 9'fy-) 10hill ;7ý NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. Page 3 of 27 I. Nuclear Design 1.1 New Reload Fuel Assembly Nuclear Desi2n 1.1.1 Assembly Average Enrichment Assembly Name Batch Identifier Enrichment (w/o U-235)SPCA9-393B-16GZ- IOOM SPCA9-396B-12GZB-1 OOM SPCA9-396B-12GZC-100M SPCA9-384B-I IGZ6-80M 1.1.2 Axial Enrichment and Burnable Poison Distribution Assembly Name Batch Identifier SPCA9-393B-16GZ-I OOM SPCA9-396B-12GZB-100M SPCA9-396B-12GZC-100M SPCA9-384B-1, GZ6-80M 1.1.3 Radial Enrichment and Burnable Poison Distribution Lattice Name SPCA9-4.56L-12G8.0/4G3.0-100M SPCA9-4.56L-12G8.0-l00M SPCA9-3.9 IL- 12G8.0-100M SPCA9-3.90L-8G5.0-I00M SPCA9-4.59L-12G8.0-1 OOM SPCA9-4.59L-12G7.0-1 OM SPCA9-3.96L-8G7.0/4G8.0-0I OM SPCA9-3.96L-8G5.0-100M SPCA9-4.58L-8G6.0/4G3.0-100M SPCA9-4.58L-8G6.0-1OOM SPCA9-4.06L-11G6.0-80M SPCA9-4.34L-I 0G6.0-80M Batch Found In 19A 19A 19A 19A 19B 19B 19B 19B and 19C 19C 19C 28B 28B 11&J 19A 19B 19C 28B 3.93 3.96 3.96 3.84 I9A 19B 19C 28B Fi2ure 2 Fieure 3 4 5 6 7 8 9 10 I1 12 13 14 1.2 Core Nuclear Design 1.2.1 Core Configuration and Licensing Exposure Limits Assembly Name GE9B-P8CWB322-1 IGZ-100M-I50-CECO GE9B-P8CWB320-9GZ-1 O0M-I 50-CECO GE9B-PSCWB343-12GZ-80M-150-CECO GE9B-P8CWB342-lOGZ-80M-150-CECO SPCA9-384B-1 IGZ6-80M SPCA9-393B-16GZ-10OM SPCA9-396B-12GZB- IOOM SPCA9-396B-12GZC-IOOM Exposure at EOC 8 (Cycle N-i) Nominal EOC 8 (MWDIMT) Short EOC 8 (MWD/MT) [for shutdown consideration)

Cycle Loaded 7 7 8 8 9 9 9 9 Core Average Exposure 27966.9 27455.9 Cycle 9 (Cycle N) neutronics analyses are valid for EOC 8 (Cycle N-i) exposures greater than 12000 MWD/MT. The exposure window that validates the pressurization transients can be found in the LIC9 reload analysis document (Reference 3).Exposure at BOC 9 (Cycle N) With Nominal EOC 8 (MWD/MT) With Short EOC 8 (MWD/MT)Core Average Exposure 10961.0 10634.9 The Cycle 9 incremental exposure to LFPC is 18000.0 MWD/MT (incremental energy to LFPC of 2418.0 GMD) based on a nominal EOC 8.Number in Core 56 89 104 .143 36 208 88 40 Core Incremental Exposure 12511.0 12000.0-4 101-7/99'1VO toll jqj 1.2.2 Core Reactivity Characteristics All values reported below are with zero xenon and are for 68°F moderator temperature.

The MICROBURN-B cold BOC K-effective bias is 1.0050 (Reference 11). The shutdown margin calculations are based on the short cycle 8 exposure given in Section 1.2.1. BOCCold K-Effective, All Rods Out 1.11710 BOC Cold K-Effective, All Rods In 0.96354 BOC Cold K-Effective, Strongest Rod Out 0.99407 BOC Shutdown Margin, % AK 1.09 Minimum Shutdown Margin, % AK 1.01 Cycle Exposure(s) of Minimum Shutdown Margin, MWD/MT 250.0 & 15000.0 Reactivity Defect (R-value)

Total, % AK 0.08 Standby Liquid Control System (SLCS) Shutdown Margin, Cold Condition, 660 ppm enriched Boron, % AK 17.81 Note that the SLCS analysis results credit a B-10 enrichment of 45% at LaSalle.

II. Control Rod Withdrawal Error Analysis was performed at a core power of 3489 MWt, 100% core flow (108.5 Mlbm/hr), unblocked (R.BM not credited) conditions only. Figure 15 is the initial rod pattern for the case that set the limit for the ATRIUM-9B fuel in the core. Figure 16 is initial rod pattern for the case that set the limit for the GE9B fuel in the core. These results bound operation with 3323 MWt as the rated power for the core. Distance ATRIUM-9B GE9B Withdrawn (ft) ACPR ACPR 12 0.29 0.31 The design complies with the SPC 1% plastic strain 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 complies with the GE 1% plastic strain criteria via conformance to updated GE mechanical overpower protection (MOP) criteria during a control rod withdrawal error event.

III. Fuel Loading Error The fuel loading error, including fuel mislocation and misorientation, is classified as an accident.

By demonsntating that the fuel loading error meets the more stringent Anticipated Operational Occurrence (AOO) requirements, the offsite dose requirement is assured to be met. Because the events listed below result in a ACPR value that is less than that of the limiting transient, the AOO requirements and hence the off-site dose requirements are met for the fuel loading error. The values reported below bound all fuel types found in the core. Event ACPR Mislocated Bundle 0.31 Misoriented Bundle 0.17 The design complies with the SPC 1% plastic strain and centerline melt criteria via conformance to the PAPT (Protection Against Power Transient)

LHGR limits. IN. Control Rb--Drop Accident LaSalle is a Banked Position Withdrawal Sequence (BPWS) plant. In order to allow the site the option of shutting down the reactor by inserting control rods using the simplified control rod sequences-shown in Table 1, the control rod drop accident analysis was performed for the simplified sequence.

The results from this simplified seqiuence anaiysis bound those where BPWS guidelines are followed.

The results demonstrate that the 280 cal/g Technical Specification limit for a control rod drop accident is not exceeded.

Note that the 0.32%Ak adder mentioned below is included in this analysis to account for possible rod mispositioning errors as well as clumping effects.

Dropped Control Rod Worth without 0.32 %Ak adder, %Ak 0.722 Dropped Control Rod Worth with 0.32 %Ak adder, %Ak 1.042 Doppler Coefficient used, (Akk/k)/°F

-9.50E-06 Effective Delayed Neutron Fraction used 0.0052 Four-Bundle Local Peaking Factor 1.358 Maximum Deposited Fuel Rod Enthalpy with 0.32 %Ak- -dder, (cal/gm) 184.1 Number of Rods Greater than 170 cal/gmn with 0.32%Ak adder 134 Note that thc limit on maximum deposited fuel rod enthalpy is 280 cal/gm and the (conservative) limit on the number of rods greater that 170 cal/gm (failed rods) is 770.

NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 7 of 27 V. Loss of Feedwater Heating The loss of feedwater heating event is analyzed at a core power of 3489 MWt for 81%, 100% and 105% rated flow with an assumed inlet temperature decrease of 1457F. These results bound operation with 3323 MWt as the rated power for the core.ATRIUM-9B ACPR Loss of Feedwater Heating 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 complies with the GE centerline melt criteria via conformance to the thermal overpower protection (TOP) limits. The analyses did not take credit for the thermal power scram function at the site. VI. Maximum Exposure Limit Compliance Note that the exposures listed below are based on the nominal Cycle 8 (Cycle N-I) exposure, 12511 MWD/MT, and the licensing basis (Reference

3) Cycle 9 (Cycle N) core average exposure of 29439 MWD/MT. GE9B GE9B ATRIUM-9B ATRIUM-9B Exposure Projected Exposure Exposure Limit Proiected Exposure Exposure Limit* Criteria (GWD/MT) (GWD/MT) (GWD/MT) (GWD/MT) Peak Fuel Assembly 44.9 48.0** 23.8 48.0 Peak Fuel Batch 40.4 42.0 N/A N/A Peak Fuel Rod N/A N/A 26.4 55.0 Peak Fuel Pellet 58.4 65.0 35.6 66.0

• The ATRIUM-9B exposure limits identified are not applicable until document EMF-85-74 is added to the Technical Specifications (Tech Specs). Until this document is added to the Tech Specs, the ATRIUM-9B exposure limits are 48.0 GWD/MT for Peak Fuel Assembly (no change), 50.0 GWD/MT for Peak Fuel Rod and 60.0 GWD/MT for Peak Fuel Pellet. ** There is no peak fuel assembly exposure limit for GE9B fuel. The limit reported above is based on the maximum channel exposure assumption used in developing the safety limit MCPR for LaSalle I Cycle 9./. f/.0 / hiyj ,/t-h-o Event 0.19 GE9B ACPR 0.18 VII. Spent Fuel Pool and Fresh Fuel Vault Criticality Compliance For the LIC9 reload, there are three new SPC ATRIUM-9B assembly types consisting of 10 unique enriched lattices as well as one SPC ATRJUM-9B assembly type with 2 unique enriched lattices which was initially manufactured for use in L2C8. These four (total) assembly and twelve (total) enriched lattice types are identified in 1.1 New Reload Fuel Assembly Nuclear Design. For the purpose of the following sections all four assembly types will be referred to as "new (ATRIUM-9B) assemblies".

VII.1 Fresh Fuel Vault Criticality Compliance The fuel storage vault criticality analysis that is detailed in Reference 6 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% GdO 3. VII.2 LI Spent Fuel Pool Criticality Compliance The LaSalle Unit I spent fuel pool criticality analysis that is detailed in Reference 7 remains valid for the above lattices.

All the new (ATRIUM-9B) assemblies comply -%Vith-sh-penrif(el 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% Gd,0 3. 1V.3.L2 Spent Fuel Pool Criticality Compliance The LaSalle Unit 2 spent fuel pool criticality analysis that is detailed in Reference 8 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 Maximum Maximum Spent Fuel k-inf* in-Rack Pool k-eff** k-eff Limit SPCA9-4.56L-12GS.014G3.0-IOOM 1.182 < 0.85 0.95 SPCA9-4.56L-12GS.0- IOOM 1.187 < 0.85 0.95 SPCA9-3.91 L- 12GS.0- I OOM 1.168 < 0.85 0.95 SPCA9-3.90L-SG5.0-IOOM 1.233 < 0.86 0.95 SPCA9-4.59L-12G8.0- lOOM 1.191 < 0.85 0.95 SPCA9-4.59J.1-12G7.0- I OOM 1.210 < 0.85 0.95 SPCA9-3.96L-8G7.0/4G8.0-I OOM 1. 186 < 0.85 0.95 SPCA9-3.96L-SGS.0-IOOM 1.231 < 0.86 0.95 SPCA9-4.58L-8G6.0/4G3.0-I OOM 1.233 < 0.86 0.95 SPCA9-4.5SL-8G6.0-OOM 1.236 < 0.86 0.95 SPCA9-4.06L-I I G6.0-80M 1.213 < 0.85 0.95 SPCA9-4.34L-I 0G6.0-80M 1.227 < 0.86 0.95

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

From Fiigure 6.1 or"Reference 8.I i NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 9 of 27 VIII. References

1. Commonwealth Edison Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods", Revision 0, Supplements 1 and 2, December 1991, March 1992, and May 1992, respectively; SER letter dated March 22, 1993.1 2. "LaSalle I Cycle 9 Core Design", NDITNFM9900038, Sequence 02, September 23, 1999. 3: "LaSalle Unit I Cycle 9 Reload Analysis", Siemens Power Corporation, EMF-2276.

4. "LaSalle Unit I Cycle 9 Plant Transient Analysis", Siemens Power Corporation, EMF-2277.

5. "Fuel Design Report for LaSalle Unit 1 Cycle 9 ATRIUM T M-9B Fuel Assemblies", Siemens Power Corporation, EMF-2249(P), Revision 1, September 1999. 6. "Criticality Safety Analysis for ATRJUM-9B Fuel, LaSalle Units I and 2 New Fuel Storage Vault", Siemens Power Corporation, EMF-95-134(P), December 1995. [NDIT 960089, Revision 0 (sic).] -7. "Criticality Safety Analysis for ATRJUM-9B Fuel, LaSalle Unit 1 Spent Fuel Storage Pool (BORAL Rick)",si~me-P, o ra-ion EMF-96-117(P), April 1996. [NDIT 960087, Revision 0 (sic).] 8. -"Criticality-Safety Analysis for ATRIUM-9B Fuel,- LaSalle Unit2 Spent Fuel Storage Pool (Boaflx Rck", iemns owrtCorporation, EMF-95-088(P), February 1996. [NDIT 960088, Revision 0 (sic).] 9. "LSI C9 Preliminary Bundle Design", NFM Calculation Note, BNDL:99-009, Revision 0, March 3, 1999. 10. "LSI Projection to EOC N-I for Cycle 9 FLLP", NFM Calculation Note, BNDL:99-027, Revision 0, April 22, 1999. 11. "LaSalle I Cycle 9 Design Basis for FLLP", NFM Calculation Note, BNDL:99-028, Revision 0, June 2, 1999. 12. "LaSalle I Cycle 9 Final Licensing Loading Pattern (FLLP)", NFM Calculation Note, BNDL:99-029, Revision 0, April 29, 1999. 13. "LaSalle I Cycle 9 LFWH", NFM Calculation Note, BNDL:99-035, Revision 0, June 25, 1999. 14. "LaSalle I Cycle 9 Standard RWE", NFM Calculation Note, BNDL:99-036, Revision 0, July 2, 1999. 15. "LaSalle I Cycle 9 MOPs/TOPs RWE", NFM Calculation Note, BNDL:BNDL:99-037, Revision 0, July 7, 1999.

NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 10 of 27 VIII. References

16. "LaSalle I Cycle 9 SLCS, CARO and CARl", NFM Calculation Note, BNDL:99-039, Revision 0, June 23, 1999. 17. "LaSalle I Cycle 9 Control Rod Drop Accident Analysis", NFM Calculation Note, BNDL:99-040, Revision 0, July 27, 1999. 18. "LaSalle Unit I Cycle 9 Bundle Misorientation Analysis", NFM Calculation Note, BNDL:99-054, Revision 0, August 20, 1999. 19. "LaSalle I Cycle 9 Fuel Assembly Mislocation Calculations", NFM Calculation Note, BNDL:99-055, Revision 0, July 29, 1999. 20. "LaSalle I Cycle 9 -L I C8 Data for GE Plastic Strain Analysis", NFM Calculation Note, BNDL:99-057, Revision 0, August 20, 1999. 21. "LaSalle I Cycle 9 -LI C9 Data for GE Plastic Strain Analysis", NFM Calculation Note, BNDL:99-058, Revision 0, August 18, 1999. 22. "LaSalle Unit I Cycle 9 New Fuel Storage", NDIT NFM9900119, Sequence 00, May 26,1999.

23. "LaSalle Unit 2 Cycle 8 Neutronics Licensing Report, Revision 2", NDIT NFM960103, Revision (sic) 2, March 22, 1999. 24. "LaSalle I Cycle 9 RWE Clad Strain Compliance", GE Proprietary Letter WHC:99-031 from William H. Hetzel to Dr. R.J. Chin, dated September 27, 1999. 25. "LaSalle I Cycle 9 NLR", NFM Calculation Note, BNDL:99-050, Revision 0, October 1999.

Table 1 LaSalle 1 Cycle 9 Simplified Shutdown Sequences Shutdown From an Al Sequence Insertion Rod Group (Bank) Comments*

7 or S 48-00 Either Group 7 or 8 may be inserted first. 10 48-10 Groups 7 and 8 must be fully inserted prior to inserting any Group 10 rod. 10 10-00 Group 10 must be at 10 prior to inserting any Group 10 rod to 00. 9 48-10 Group 10 must be fully inserted prior to inserting any Group 9 rod. 9 10-00 Group 9 must be at 10 prior to inserting any Group 9 rod to 00. 5 or 6 48-00 Groups 5 and 6 may be inserted without banking anytime after .. .Groups 7 and 8 have been inserted and before Group 4 is inserted.

4 48-00 Groups 5 through 10 must be fully inserted prior to inserting any Group 4 rod. 3 48-10 Group 4 must be fully inserted prior to inserting anyGroup 3 rod. -3 10-00 Group 3 must be at 10 prior to inserting any Group 3 rod to 00. 2 48-00 Group 3 must be fully inserted prior to inserting any Group 2 rod. 1 4 8-00 Group 2 must be fully inserted prior to inserting any Group I rod. Shutdown from an A2 Sequence Insertion Rod Group (Bank) Comments*

9 or 10 48-00 Either Group 9 or 10 may be inserted first. S 48-00 Groups 9 and 10 must be fully inserted prior to inserting any Group 8 rod. 7 48-10 Group 8 must be fully inserted prior to inserting any Group 7 rod. 7 10-00 Group 7 must be at 10 prior to inserting any Group 7 rod 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 is inserted.

4 48-00 Groups 5 through 10 must be fully inserted prior to inserting any _ _Group 4 rod. 3 48-10 Group 4 must be fully inserted prior to inserting any Group 3 rod. 3 10-00 Group 3 must be at 10 prior to inserting any Group 3 rod to 00.. 2 48-00 Group 3 must be fully inserted prior to inserting any Group 2 rod. 1 48-00 Group 2 must be fully inserted prior to inserting any Group I rod.

• The standard BPWS rules concerning out-of-service rods apply to the shutdown sequences.=_.

BATCH 19A Natural Uranium SPCA9-3.90L 8G5.0-100M SPCA9 3.91L-12G8.0 -looM SPCA9 4.56L-12G8.0 -looM SPCA9-4.56L 12GS.0/4G3.0 -looM Natural Uranium SPCA9-393B-16GZ-1OOM I 1 " 12" 30" 72" 18" BATCH 19B Natural Uranium SPCA9-3.96L 8G5.0-1 0OM SPCA9 3.96L-8G7.0/4G8.0 -loom SPCA9 4.59L- 12G7.0 -looM SPCA9 4.59L-12GS.0 -looM II," 12" 30" 24" 66" Natural Uranium SPCA9-396B-12GZB- IOOM BATCH 19C Natural Uranium SPCA9 3.96L-SG5.0 -looM SPCA9 4.58L-8G6.0 -looM SPCA9 4.58L-SG6.0/4G3.0 -looM Natural Uranium SPCA9-396B-12GZC-IOOM Figure 1 LlC9 ATRIUM-9B Assembly Axial Designs (1OOM Channels)III" 42" 24r 66" 6" NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 13 of 27 BATCH 28B Natural Uranium SPCA9 4.06L-1 1 G6.0 -80M SPCA9 4.34L-10G6.0

-80M Natural Uranium 411" " 42" 90" 6"1 SPCA9-384B-11GZ6-80M Figure 2 LIC9 ATRIUM-9B Assembly Axial Designs (80M Channels);TR- C/,I11 I Rods (4) 2 Rods (8) 3 Rods (8) 4 Rods (36) GI Rods (4) G2 Rods (8) G3 Rods (4)3.00 w/o U-235 4.00 w/o U-235 4.70 wlo U-235 4.95 w/o U-235 4.70 w/o U-235+8.0 wh/o Gd203 4.20 w/o U-235+8.0 w/o Gd203 4.00 w/o U-235+3.0 w/o Gd203 Figure 3 SPCA9-4.56L-12G8.0/4G3.0-1OOM (19A) Enrichment Distribution a

I Rods (4) 2 Rods (12) 3 Rods (8) 4 Rods (36) GI Rods (4) G2 Rods (8)3.00 w/o U-235 4.00 wv/o U-235 4.70 w/o U-235 4.95 w/o U-235 4.70 w/o U-235+8.0 w/o Gd203 4.20 w/o U-235+8.0 w/o Gd203 Figure 4 SPCAg-4.56L-12G8.0-100M (19A) Enrichment Distribution NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. _ Page 16 of 27 I Rods (4) 2 Rods (12) 3 Rods (8) 4 Rods (36) GI Rods (12)2.60 w/o U-235 3.40 w/o U-235 3.80 w/o U-235 4.40 w/o U-235 3.40 w/o U-235+8.0 w/o Gd203 Figure 5 SPCA9-3.91L-12G8.0-100M (19A) Enrichment Distribution 1 2 3 4 4 4 3 2 1 S2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 2 2 G 4 2 4 -" 2 2 3.40 3.40 AO 4.40 3.40 4.40 3.40 3.40 3 4 4 4 4 4I 3 3.80 4.40 4.40 4.40 4.40 4.40 3.80 4 4 4 * .4 4 4 -"l rn l .4 0440.4 ? 4.40 4.40 4.40 4.4 4.044 .4 2 4 ate -4 2 4 4.40 3.40-- 4.40 4.40 3.40 4.40 4 4 4 *4 4 4 4.40 4.40 4.40 --? 4.40 4.40 4.40 3 4 4 4 4 4 G1÷ 5&i 3 3.80 4.40 4.40 4.40 4.40 4.40 4 3.80 2 2 2 G~4 2 4 2 2 3.40 3.40 ' 4.40 3.40 4.40 iý3.4O' 3.40 3.40 1 2 3 4 4 4 3 2 1 2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 I Rods (4) 2 Rods (16) 3 Rods (8) 4 Rods (36) GI Rods (8)2.60 w/o U-235 3.40 w/o U-235 3.80 wlo U-235 4.40 w/o U-235 3.40 w/o U-235+5.0 wo Gd203 Figure 6 SPCA9-3.90L-8G5.0-IOOM (19A) Enrichment Distribution I Rods (4) 2 Rods (8) 3 Rods (8) 4 Rods (40) GI Rods (12)3.00 4.00 4.70 4.95 4.20 w/o U-235 w/o U-235 w/o U-235 w/o U-235 w/o U-235+8.0 w/o Gd203 Figure 7 SPCA9-4.59L-12G8.0-100M (19B) Enrichment Distribution I Rods (4) 2 Rods (8) 3 Rods (8) 4 Rods (40) Gi Rods (12)3.00 w/o U-235 4.00 w/o U-235 4,70 w/o U-235 4.95 w/o U-235 4.20 w/o U-235+7.0 w/o Gd203 Figure 8 SPCAg-4.59L-12G7.0-100M (19B) Enrichment Distribution4ILa)

NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 20 of 27 1 Rods (4) 2 Rods (8) 3 Rods (8) 4 Rods (40) G I Rods (8) G2 Rods (4)2.60 w/o U-235 3.40 w/o U-235 3.80 w/o U-235 4.40 w/o U-235 3.40 wlo U-235+7.0 w/o Gd203 3.40 w/o U-235+8.0 wlo Gd203 Figure 9 SPCA9-3.96L-8G7.0/4G8.O-O00M (19B) Enrichment Distribution 4D 6 W/,S alzqgi I Rods (4) 2 Rods (12) 3 Rods (8) 4 Rods (40) GI Rods (8)2.60 sy/o 3.40 w/o 3.80 wv/o 4.40 w/o 3.40 w/o U-235 U-235 U-235 U-235 U-235+5.0 w/o Gd203 Figure 10 SPCA9-3.96L-8G5.0-IOOM (19B and 19C) Enrichment Distribution . .v,3,,i' NUCLEAR FUEL MANAGEMENT NFM ID4 NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Pape 22 of 27 1 Rods (4) 2 Rods (8) 3 Rods (8) 4 Rods (40) GI Rods (8) G2 Rods (4)3.00 w/o U-235 4.00 wlo U-235 4.70 wlo U-235 4.95 w/o U-235 4.20 w/o U-235+6.0 w/o Gd203 4.00 w/o U-235+3.0 w/o Gd203 Figure 11 SPCA9-4.58L-8G6.0/4G3.0-100M (19C) Enrichment Distribution f"Y, EILi4 NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 23 of 27 I Rods (4) 2 Rods (12) 3 Rods (8) 4 Rods (40) GI Rods (8)3.00 w/o U-235 4.00 wlo U-235 4.70 w/o U-235 4.95 w/o U-235 4.20 w/o U-235+6.0 w/o Gd203 Figure 12 SPCA9-4.58L-8G6.0-10OM(19C)

Enrichment Distribution NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 24 of 27 Rods (4) Rods ( 8) Rods (16) Rods (33) Rods (11)2.72 w/o U-235 3.53 w/o U-235 3.94 wlo U-235 4.53 w/o U-235 3.69 wlo U-235+6.0 w/o Gd203 Figure 13 SPCA9-4.06L-11G6.0-80M (28B) Enrichment Distribution ell- Io/q/1 2 3 4 G NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 25 of 27 I Rods (4) 2 Rods (8) 3 Rods (16) 4 Rods (34) G Rods (10)2.72 w/o U-235 3.78 w/o U-235 4.19 w/o U-235 4.78 wlo U-235 4.19 wlo U-235+6.0 w/o Gd203 Figure 14 SPCA9-4.34L-10G6.0-80M (28B) Enrichment Distribution NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 26 of 27 Cycle 9 Exposure Core Average Exposure 13000.0 MWd/MTU 1746.3 GWd 23961.4 MWd/MTU Delta E: MWd/MTU, (GWd) Power: MWt Core Pressure:

psia Inlet Subcooling:

Btu/ibm Flow: Mlb/hr.0 ( .00 1 3489.0 (100.00 %) 1020.1 -18.28 108.50 (100.00 %J 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 ----16 ......-16----0 -.16 16 . .. .0 -- 8-0-- 0-0 -- 0--- 16 1- .. .--0 -- 8 ---16-..--11 -7 3 0-- 0-- 0 ----------.---

16 ----N Top 25 24 23 22 21 20 19 18 17 16 15 59 55 51 47 0-- 0-- 8-- 0---- 43 -.16-- 39 16 --16-- 0-16-- 16 -- 0 - 0-- 8--Axial Profile Power Exposure .153 3.961 .284 6.830 .663 16.284 .806 19.848 .874 21.821 .935 23.364 .978 24.366 1.006 25.256 1.010 26.344 1.011 2.7.653 1. 63-- 28.409 14 1.030 28.934 0 0 16 16 - 0-- 0 -- 0 --- 16 --- --- 16 ---- 31 -- 27 -- 23 -- 19 15 11 7 3 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 Control Rod Density: % k-effective:

Void Fraction:

Core Delta-P: psia Core Plate Delta-P: psia Coolant Temp: Deg-F In Channel Flow: Mlb/hr Source Convergence 20.54 1.00388 .448 21.675 17.213 545.8 93.26 .00008 13 1.056 12 1.087 --- --. 11 1.128 10 1.177 9 1.231 8 1.279 7 1.347 6 1.421 5 1.485* 4 1.477 3 1.308 2 .962 Bottom 1 .279 29.323 29.530 29.908 30.365 30.778* 30.556 30.740 30.648 29.817 27.607 23.608 17.267 4.864% AXIAL TILT -18.447 -10.750 AVG BOT 8ft/12ft 1.1042 1.0860 Active Channel Flow: Nlb/hr 93.26 Figure 15 Initial RWE Rod Pattern for Limiting ATRIUM-9B Case Error Rod is 34-43 ~,io 59 55 51 47 43 39 35 31 27 23 19 15 NUCLEAR FUEL MANAGEMENT TRANSMITTAL OF DESIGN INFORMATION NFM ID# NFM9900149 Seq. No. I Pape 27 of 27 Cycle 9 Exposure .0 MWd/MTU .0 GWd Core Average Exposure 10961.0 MWd/MTU Delta E: MWd/MTU, (GWd) Power: MWt Core Pressure:

psia Inlet Subcooling:

Btu/ibm Flow: Mlb/hr 2 6 10 14 18 -.. .. 12 ---20.-- 0-0 --24---20-- 0-12 -2 6 10 14 18 22 20 0 24 0 24 0 20 22 Control Rod Density: k-effective:

Void Fraction:

Core Delta-P: psia.0 1 .00 ) 3489.0 (100.00 %) 1020.1 -18.28 108.50 (100.00 %)26 30 34 38 42 46 50 54 58 59 -- 0 --20 -- 55 51 -- 24 -- 0 -- 12 47 --43 -- 00--24_--

0-- 20 --,39 -.. ... ....-35--2A.-- 0 --24-- 0 --31 --27 -- 0 --24-- 0 --20 --23 --. -24-- 0 --12... 15 -- 0 --20 .-- 7 3 26 30 34 38 42 46 50 54 58 15.23 1.00250 .430 21.559 Axial Profile N Power Exposure Top 25 .134 1.558 24 .242 2.748 23 .614 6.567 22 .753 8.305 21 .813 9.766 20 .885 10.769 19 .947 11.473 18 1.008 12.074 17 1.084 12.760 16 1.209 13.155 15 1.271. 13.418 14 1.290 13.606 13 1.288 13.747 12 1.256 13.859 11 1.254 13.965 10 1.273 14.034* 9 1.304 14.008 8 1.326 13.527 7 1.357 13.424 6 1.365* 13.358 5 1.316 13.185 4 1.164 12.622 3 .968 11.120 2 .693 8.223 Bottom 1 .187 2.308 AXIAL TILT -13.547 -10.296 Core Plate Delta-P: psia Coolant Temp: Deg-F In Channel Flow: Mlb/hr Source Convergence 17.096 545.5 93.36 .00007 AVG BOT 8ft/12ft 1.1172 Active Channel Flow: mlb/hr*.Figure 16 Initial RWE Rod Pattern for Limiting GE9B Case Error Rod is 30-39 59 55 51 47 43 39 35 31 27 23 19 15 11 7 3 1.0898 93.36 Technical Requirements Manual -Appendix I LI C9 Reload Transient Analysis Results Attachment 2 LaSalle Unit 1 Cycle 9 Reload Analysis LaSalle Unit 1 Cycle 9 November 2001 IjiJT- U dm U i c\ oo'i SIEMENS EMF-2276 Revision 1 LaSalle Unit I Cycle 9 Reload Analysis October 1999 Siemens Power Corporation Nuclear Division Siemens Power Corporation ISSUED IN SPC ON-LINE DOCUMENT DATE: / LZq LaSalle Unit I Cycle 9 Reload Analysis Prepared: Haun. Engineer Neutronics Prepared:

Concurred:

D. G. Carr, Team Leader BWR Safety Analysis H. D. Curet, Manager Product Licensing Approved: Approved: Approved:

Approved: 0. C. Brow , anager BWR Neutronics M. E. Garrett, Manager BWR Safety Analysis .M. Howe, Manager Product Mechanical Engineering Denver, Manager Commercial Operations

/&/- 7 /fr f_. Date i0- 6 Date Date Date Date 1o/099 Date Date Siemens Power Corporation EMF-2276 Revision 1 gdh EMF-2276 Revision 1 Page ii LaSalle Unit 1 Cycle 9 Ke\~auU "d y Nature of Changes Description and Justification

1. 5-1 2. 5-10 3. 9-1 Discussion added to indicate MCPRI limits and LHGRFACI multipliers are provided for maximum core flows of 102.5% and 105% of rated. Revised Figure 5.1 to include MCPRI limits for both 102.5% and 105% maximum core flows. Updated references to revised plant transient analysis and fuel design reports.

The changed items are further identified by a vertical line (I ) in the right-hand margin.If PM P EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page iii Contents 1.0 Introduction

..................................................................................................................

1-1 2.0 Fuel Mechanical Design Analysis ................................................................................

2-1 3.0 Thermal-Hydraulic Design Analysis .............................................................................

3-1 3.2 Hydraulic Characterization

...............................................................................

3-1 3.2.1 Hydraulic Compatibility

......................................................................

3-1 3.2.3 Fuel Centerline Temperature

.............................................................

3-1 3.2.5 Bypass Flow ......................................................................................

3-1 3.3 MCPR Fuel Cladding Integrity Safety Limit (SLMCPR) .....................................

3-1 3.3.1 Coolant Thermodynamic Condition

....................................................

3-1 3.3.2 Design Basis Radial Power Distribution

.............................................

3-2 3.3.3 Design Basis Local Power Distribution

...............................................

3-2 4.0 Nuclear Design Analysis ...................................................-...........................................

4-1 4.1 Fuel Bundle Nuclear Design Analysis ...............................................................

4-1 4.2 Core Nuclear Design Analysis ..........................................................................

4-2 4.2.1 Core Configuration

.............................................................................

4-2 4.2.2 __ Core Reactivity Characteristics

...........................................................

4-2 4.2.4 Core Hydrodynamic Stability

..............................................................

4-2 5.0 Anticipated Operational Occurrences

...........................................................................

5-1 5.1 Analysis of Plant Transients at Rated Conditions

.............................................

5-1 5.2 Analysis for Reduced Flow Operation

..............................................................

5-1 5.3 Analysis for Reduced Power Operation

............................................................

5-2 5.4 ASME Overpressurization Analysis ...................................

5-2 5.5 Control Rod W ithdrawal Error ...........................................................................

5-2 5.6 Fuel Loading Error ...........................................................................................

5-2 5.7 Determination of Thermal Margins ...................................................................

5-2 6.0 Postulated Accidents

...................................................................................................

6-1 6.1 Loss-of-Coolant Accident .................................................................................

6-1 6.1.1 Break Location Spectrum ...................................................................

6-1 6.1.2 Break Size Spectrum ................................... " ......................................

6-1 6.1.3 MAPLHGR Analyses .........................................................................

6-1 6.2 Control Rod Drop Accident .....................................

6-1 6.3 Spent Fuel Cask Drop Accident .......................................................................

6-1 7.0 Technical Specifications

..............................................................................................

7-1 7.1 Limiting Safety System Settings .......................................................................

7-1 7.1.1 MCPR Fuel Cladding Integrity Safety Limit .........................................

7-1 7.1.2 Steam Dome Pressure Safety Limit ....................................................

7-1 7.2 Limiting Conditions-for Operation

.....................................................................

7-1 7.2.1 Average Planar Linear Heat Generation Rate ....................................

7-1 7.2.2 Minimum Critical Power Ratio .............................................................

7-1 7.2.3 Linear Heat Generation Rate ..............................................................

7-2 LaSalle Unit 1 Cycle 9 EMF-2276 Reload Analysis Revision 1 Page iv 5.0 M ethodology References

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8-1 9.0 Additional References

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9-1 Tables 1.1 EOD and EOOS Operating Conditions

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1-2 4.1 Neutronic Design Values ..............................................

4-4 5.1 EOC Base Case and EOOS MCPRp Limits and LHGRFACP Multipliers for TSSS Insertion Times for Prepower Uprate Conditions (3323 MWt Rated Pow er) .........................................................................................................................

5-4 5.2 Base Case MCPRp Limits and LHGRFACp Multipliers for NSS Insertion Times for Prepower Uprate (3323 MWt Rated Power) .................................................

5-6 5.3 EOC Base Case and EOOS MCPRp Limits and LHGRFACp for Multipliers for TSSS Insertion Times for Power Uprate Conditions (3489 MWt Rated Pow er) .........................................................................................................................

5-7 5.4 EOC MCPR, Limits and LHGRFACP Multipliers for NSS Insertion Times for Power Uprate Conditions (3489 MWt Rated Power) ..............................................

5-9 Figures 3.1 Radial Power Distribution for SLMCPR Determination

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3-3 3.2 LaSalle Unit 1 Cycle 9 Safety Limit Local Peaking Factors SPCA9-393B-1 6GZ-1 0DM With Channel Bow .............................................................

3-4 3.3 LaSalle Unit 1 Cycle 9 Safety Limit Local Peaking Factors SPCA9-396B-1 2GZB-1 0DM and SPCA9-396B-1 2GZC-1 0DM With Channel Bow ................

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35 3.4 LaSalle Unit 1 Cycle 9 Safety Limit Local Peaking Factors SPCA9-384B-1i1GZ-80M With Channel Bow ........................

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3-6 4.1 SPCA9-4.56L-12G8.014G3.0-10DM Enrichment Distribution

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4-5 4.2 SPCA9-4.56L-12G8.0-100M Enrichment Distribution

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4-6 4.3 SPCA9-3.91L-12G8.0-100M Enrichment Distribution

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4-7 4.4 SPCA9-3.90L-8G5.0-100M Enrichment Distribution

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4-8 4.5 SPCA9-4z59L-12G8.O-100M Enrichment Distribution

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4-9 4.6 SPCA9-4.59L-12G7.0-100M Enrichment Distribution

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4-10 4.7 SPCA9-3.96L-8G7.0/4G8.0-100M Enrichment Distribution

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4-11 4.8 SPCA9-3.96L-SG5.0-100M Enrichment Distribution

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4-12 4.9 SPCA9-4.58L-8G6.0/4G3.0-10DM Enrichment Distribution

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4-13 4.10 SPCA9-4.58L-8G6.0-OOM-Enrichment Distribution

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4-14 4.11 SPCA9-4.06L-11G6.0-80M Enrichment Distribution

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4-15 4.12 SPCA9-4.34L-10G6.0-80M Enrichment Distribution

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4-16 4.13 ATRIUM-9B LSA-1 19A Assembly Design .................................................................

4-17 EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page v 4.14 ATRIUM-9B LSA-1 19B Assembly Design .................................................................

4-19 4.15 ATRIUM-9B LSA-1 19C Assembly Design .................................................................

4-21 4.16 ATRIUM-9B SPCA9-384B-11GZ-80M Assembly Design ...........................................

4-23 4.17 LaSalle Unit 1 Cycle 9 Reference Loading Map .........................................................

4-25 5.1 Flow Dependent MCPR Limits for Manual Flow Control Mode ...................................

5-10 5.2 Flow Dependent LHGR Multipliers for ATRIUM-9B Fuel ............................................

5-11 5.3 Base Case Power Dependent MCPR Limits for ATRIUM-9B Fuel -TSSS Insertion Times. ......................................

...........................

...... 5-12 5.4 Base Case Power Dependent MCPR Limits for GE9 Fuel -TSSS Insertion Tim es ..........................................................................................................

5-13 5.5 Base Case Power Dependent MCPR Limits for ATRIUM-9B Fuel -NSS Insertion Tim es .........................................................................................................

5-14 5.6 Base Case Power Dependent MCPR Limits for GE9 Fuel -NSS Insertion Times .........................................................

5-15 5.7 Starting Control Rod Pattern for Control Rod Withdrawal Analysis ............................

5-1.6 7.1 Protection Against Power Transient LHGR Limit for ATRIUM-9B Fuel .........................

7-3 Sintions Powor Corporation LaSalle Unit I Cycle 9 Reload Analysis Nomenclature abnormal operational occurrence beginning of cycle critical power ratio effective full power hour end of cycle extended operating domain equipment out of service feedwater heater out of service feedwater controller failure increased core flow LFWH LHGR LHGRFAC LPRM LRNB MAPLHGR MCPR MELLLA MSIV NSS PAPT PCT RPT SLMCPR SLO SPC SRVOOS TBVOOS TCV TIP TIPOOS TSSS UFSAR ACPR loss of feedwater heater linear heat generation rate LHGR multiplier local power range monitor load rejection no bypass maximum average planar linear heat generation rate minimum critical power ratio maximum extended load line limit area main steam isolation valve nominal scram speed protection against power transient peak clad temperature recirculation pump trip safety limit minimum critical power ratio single-loop operation Siemens Power Corporation safety/relief valve out of service turbine bypass valves out of service turbine control valve traversing in-core probe traversing in-core probe out of service technical specification scram speed updated final safety analysis report change in critical power ratio Siemens Power Corporation EMF-2276 Revision 1 Page vi AOO BOC CPR EFPH EOC EOD EOOS FHOOS FWCF ICF EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 1-1 1.0 Introduction This report provides the results of the analysis performed by Siemens Power Corporation (SPC) as part of the reload analysis in support of the Cycle 9 reload for LaSalle Unit 1. This report is intended to be used in conjunction with the SPC topical Report XN-NF-8O-19(P)(A), Volume 4, Revision 1, Application of the ENC Methodology to BWR Reloads, which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list. Section numbers in this report are the same as corresponding section numbers in XN-NF-80-19(P)(A), Volume 4, Revision 1. Methodology used in this report-which supersedes XN-NF-80-19(P)(A), Volume 4, Revision 1, is referenced in Section 8.0. The NRC Technical Limitations presented in the methodology documents, including the documents referenced in Section 8.0, have been satisfied by these analyses.

Analyses performed by Commonwealth Edison Company (CornEd) are described elsewhere.

This document alone does not necessarily identify the limiting events or the appropriate operating limits for Cycle 9. The limiting events and operating limits must be determined in conjunction with results from ComEd analyses.

The Cycle 9 core consists of a total of 764 fuel assemblies, including 372 unirradiated ATRIUM'T m-9B assemblies and 392 irradiated GE9 assemblies.

The reference core configuration is described in Section 4.2. The design and safety analyses reported in this document were based on the design and operational assumptions in effect for LaSalle Unit 1 during the previous operating cycle. The effects of channel bow are explicitly accounted for in the safety limit analysis.

The extended operating domain (EOD) and equipment out of service (EOOS) conditions presented in Table 1.1 are supported.

Analyses were performed to support end-of-cycle (EOC) operating limits. This report provides limits for both pre-power uprate (3323 MWt) and power uprate (3489 MWt) conditions.

The analyses upon which the operating limits are based were performed such that both the pre power uprate and power uprate limits are applicable for all of Cycle 9. ATRIUM is a trademark of Siemens.Siemens Power Corporation EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 1-2 Table 1.1 EOD and EOOS Operating Conditions Extended Operating Domain (EOD) Conditions Increased Core Flow Maximum Extended Load Line Limit Analysis (MELLLA) Equipment Out of Service (EOOS) Conditions*

Feedwater Heaters Out .of Service (FHOOS) Single-Loop Operation (SLO) -Recirculation Loop Out of Service Turbine Bypass Valves Out of Service (TBVOOS) Recirculation Pump Trip Out of Service (No RPT) Turbine Control Valve (TCV) Slow Closure and/or No RPT -Safety-Relief Valve Out-of Service (SRVOOS) Up to 2 TIP Machine(s)

Out of Service (or the equivalent number of TIP channels)

Up to 50% of the LPRMs Out of Service TCV Slow Closure, FHOOS and/or No RPT EOOS conditions are supported for EOD conditions as well as the standard operating domain. Each EOOS condition combined with I SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels) and/or up to 50% of the LPRMs out of service is supported.

Sioniens Power Corporation EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 2-1 2.0 Fuel Mechanical Design Analysis Applicable SPC Fuel Design Reports References 9.1 & 9.2 To assure that the power history for the ATRIUM-9B fuel to be irradiated during Cycle 9 of LaSalle Unit 1 is bounded by the assumed power history in the fuel mechanical design analysis, LHGR operating limits have been specified in Section 7.2.3. In addition, LHGR limits for . Anticipated Operational Occurrences have been specified in Reference 9.1 and are presented in Section 7.2.3 as Figure 7.1.Siemens Power Corporation LaSalle Unit 1 Cycle 9 Reload Analysis EMF-2276 Revision 1 Page 3-1 3.0 Thermal-Hydraulic Design Analysis 3.2 Hydraulic Characterization

3.2.1 Hydraulic

Compatibility Component hydraulic resistances for the fuel types in the LaSalle Unit 1 Cycle 9 core have been determined in single-phase flow tests of full-scale assemblies.

The hydraulic demand curves for SPC ATRIUM-9B and GE9 fuel in the LaSalle Unit 1 core are provided in Reference

9.1. Figure

4.2. 3.2.3 Fuel Centerline Temperature Applicable Report ATRIUM-9e Reference 9.1, Figure 3.3 3.2.5 Bypass Flow Calculated Bypass Flow at 100%P/1 00%F (includes water channel flow)14.7 Mlb/hr Reference 9.3 3.3 MCPR Fuel Cladding Integrity Safety Limit (SLMCPR)Two-Loop Operation*

Single-Loop Operation*

1.11 1.12 3.3.1 Coolant Thermodynamic Condition Thermal Power (at SLMCPR) Feedwater Flow Rate (at SLMCPR) Core Exit Pressure (at Rated Conditions)

Feedwater T-emperature Reference 9.3 5232.35 MWt 22.7 Mlbmlhr 1031.35 psia 426.5°F Includes the effects of channel bow, up to 2 TIPOOS (or the equivalent number of TIP channels), a 2000 EFPH LPRM calibration interval, and up to 50% of the LPRMs out of service.Siemens Power Corporation EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 3-2 3.3.2 Desion Basis Radial Power Distribution Figure 3.1 shows the radial power distribution used in the MCPR Fuel Cladding Integrity Safety Limit analysis.

3.3.3 Desion

Basis Local Power Distribution Figures 3.2. 3.3 and 3.4 show the local power peaking factors used in the MCPR Fuel Cladding Integrity Safety Limit analysis.

SPCA9-393B-16GZ-100M Figure 3.2 SPCA9-396B-12GZB-100M and Figure 3.3 SPCA9-396B-12GZC-100M SPCA9-384B-11GZ-50M Figure 3.4 Siemens Power Corporation LaSalle Unit I Cycle 9 Reload Analysis 200 175 150 cn "D 125 M 5 100 Q) -o E--75 I z 50 25 0 EMF-2276 Revision 1 Page 3-3.0 .1 .2 .3 .4 .5 .6 .7 .8 .9 1.0 1.1 1.2 1.3 1.4 1.5 1.6 Radiol Power Peaking Figure 3.1 Radial Power Distribution for SLMCPR Determination Siemens Power Corporation LaSalle Unit I Cycle 9 A: Il "~neC " ont r o I Rod Co rn e r Figure 3.2 LaSalle Unit I Cycle 9 Safety Limit Local Peaking Factors SPCA9-393B-16GZ-1OOM With Channel Bow EMF-2276 Revision 1 Page 3-4 C 0 n t r 0 R 0 d C 0 r n e r 1.023 1.055 1.068 1.112 1.099 1.102 1.049 1.023 .977 1.055 .958 .894 1.016 .894 1.007 .877 .927 1.002 1.068 .894 1.031 1.065 1.084 1.056 1.010 .863 1.011 1.112 1.016 1.065 Internal 1.044 .980 1.051 1.099 .894 1.084 Water 1.063 .863 1.038 1.102 1.007 1.056' Channel 1.035 .971 1.041 1.049 .877 1.010 1.044 1.063 1.035 .990 .846 .992 1.023 .927 .863 .980 .863 .971 .846 .895 .970 .977 1.002 1.011 1.051 1.038 1.041 .992 .970 .931 Reloa ,h-t 'jl0"a y ,.

LaSalle Unit 1 Cycle 9 1: l r 14 A ,J lc (C 0 n t r 0 R 0 d C 0 r n e r EMF-2276 Revision 1 Page 3-5 ontrol Rod Corner 1.013 1.042 1.056 1.110 1.098 1.100 1.037 1.010 .967 1.042 .944 1.025 .879 1.014 .871 1.005 .912 .989 1.056 1.025 1.018 1.064 1.081 1.055 .997 .989 .999 1.110 .879 1.064 Internal 1.043 .848 1.047 1.098 1.014 1.081 Water 1.059 .978 1.035 1.100 .871 1.055 Channel 1.034 .840 1.037 1.037 1.005 .997 1.043 1.059 1.034 .977 .968 .979 1.010 .912 .989 .848 .978 .840 .968 .881 .956 .967 .989 .999 1.047 1.035 1.037 .979 .956 .921 Figure 3.3 LaSalle Unit I Cycle 9 Safety Limit Local Peaking Factors SPCA9-396B-1 2GZB-1 0DM and SPCAg-396B-1 2GZC-1 0DM With Channel Bow Siemens Power Corporation Re oa LaSalle Unit 1 Cycle 9 C IUL) 3 iI=J EMF-2276 Revision 1 Page 3-6 C 0 n t r 0 R 0 d C 0 r n e r Figure 3.4 LaSalle Unit I Cycle 9 Safety Limit Local Peaking Factors SPCA9-384B-1 1 GZ-80M With Channel Bow Siemens Power Corporation ontrol Rod Corner 1.022 1.056 1.061 1.035 1.102 1.028 1.045 1.029 .982 1.056 .947 1.018 1.003 '879 .997 1.004 .919 1.011 1.061 1.018 1.001 1.050 1.081 1.048 .996 .992 1.012 1.035 1.003 1.050 Internal .926 .983 .987 1.102 .879 1.081 Water 1.077 .853 1.049 1.028 .997 1.048 Channel 1.040 .970 .979 1.045 1.004 .996 .926 1.077 1.040 .859 .980 .996 1.029 .919 .992 .983 .853 .970 .980 .891 .983 .982 1.011 1.012 .987 1.049 .979 .996 .983 .941 LaSalle Unit 1 Cycle 9 Reload Analysis EMF-2276 Revision 1 Page 4-1 4.0 Nuclear Design Analysis 4.1 Fuel Bundle Nuclear Design Analysis Assembly Average Enrichment (ATRIUM-9B fuel) SPCA9-393B-16GZ-100M SPCAg-396B-12GZB-100M SPCA9-384B-11GZ-80M SPCA9-396B-12GZC-100M Radial Enrichment Distribution SPCA9-4.56L-12G8.D/4G3.0-1 0DM SPCA9-4.56L-12G8.0-1OOM SPCA9-3.91 L-1 2G8.0-1 0DM SPCA9-3.90L-8G5.0-100M SPCA9-4.59L-12G8.0-100M SPCA9-4.59L_-12G7.0-100M SPCA9-3.96L-8G7.0/4G8.0-1 0DM SPCA9-3.96L-8G5.0-10DM SPCA9-4.58L-8G6.0/4G3.0-100M SPCA9-4.5BL-8G6.0-1 0DM SPCA9-4.06L-1 1G6.0-80M SPCA9-4.34L-1 0G6.0-80M Figure 4.1 Figure 4.2 Figure 4.3 Figure 4.4 Figure 4.5 Figure 4.6 Figure 4.7 Figure 4.8 Figure 4.9 Figure 4.10 Figures 4.11 Figures 4.12 Axial Enrichment Distribution Burnable Absorber Distribution Non-Fueled Rods Neutronic Design Parameters Figures 4.13-4.16 Figures 4.13-4.16 Figures 4.1-4.12 Table 4.1-Fuel Storage LaSalle New Fuel Storage Vault Reference 9.4 The LSA-1 Reload Batch fuel designs meet the fuel design limitations defined in Table 2.1 of Reference 9.4 and therefore can be safely stored in the vault.LaSalle Unit 1 Spent Fuel Storage Pool (BORAL Racks)Reference 9.5 The LSA-1 Reload Batch fuel designs meet the fuel design limitations defined in Table 2.1 of Reference 9.5 and therefore can be safely stored in the pool.Siemens Power Corporation 3.93 wt% 3.96 wt% 3.84 wt% 3.96 wt%

LaSalle Unit I Cycle 9 An LaSalle Unit 2 Spent Fuel Storage Pool Reference 9.6 The LSA-1 Reload Batch fuel designs can be safely stored as long as the fuel assembly reactivity limitations defined in Reference 9.6 are met. <ComEd has responsibility to confirm that fuel meets reactivity limitations.

> 4.2 Core Nuclear Design Analysis 4.2.1 Core Confiouration Core Exposure at EOC8, MWd/MTU (nominal value) Core Exposure at BOC9, MWd/MTU (from nominal EOCS). Core Exposure at EOC9, MWd/MTU (licensing basis)Figure 4.17 27.957 10,962 29,439 NOTE: Analyses in this report are applicable to a core exposure of 29,439 MWdIMTU.

< Cycle 9 short window exposure to be determined by CoinEd. > 4.2.2 Core Reactivity Characteristics

< This data is to be furnished by CornEd. >Core Hydrodynamic Stability Reference

8.7 LaSalle

Unit 1 utilizes the BWROG Interim Corrective Actions (ICAs) to address thermal hydraulic instability issues. This is in response to Generic Letter 94-02. When the long term solution OPRM is fully implemented, the ICAs will remain as a backup to the OPRM system. In order to support the ICAs and remain cognizant of the relative stability of one cycle compared with previous cycles, decay ratios are calculated at various points on the power to flow map and at various points in the cycle. This satisfies the following functions.

1. Provides trending information to qualitatively compare the stability from cycle to cycle. 2. Provides decay ratio sensitivities to rod line and flow changes near the ICA regions.EMF-2276 Revision 1 Paqe 4-2 4.2.4 Reload Ana lvsis EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 4-3 3. Allows CornEd to review this information to determine if any administrative conservatisms are appropriate beyond the existing requirements.

The NRC approved STAIF computer code was used in the core hydrodynamic stability analysis performed in support of LaSalle Unit 1 Cycle 9. The power/flow state points used for this analysis were chosen to assist CornEd in performing the three functions described above. The Cycle 9 licensing basis control rod step-through projection was used to establish expected core depletion conditions.

For each power/flow point, decay ratios were calculated at multiple cycle exposures to determine the highest expected decay ratio throughout the cycle. the results from this analysis are shown below. Power/Flow Maximum Maximum M" Globalt Regional t 29.6/26.6 0.73 0.61 30.3/29.2 0.48 0.53 51.9/26.6

>1.1 >1.1 54.4/29.2

>1.1 >1.1 61.9/50.0 0.46 0.69 73.6/50.0 0.67 1.04 78.1/55.0 0.57 0.90 82.4/60.0 0.49 0.79 70.0/55.0 0.44 0.69 For reactor operation under conditions of power coastdown, single-loop operation, final feedwater temperature reduction (FFTR) and/or operation with feedwater heaters out of service, it is possible that higher decay ratios could be achieved than are shown for normal operation.

NOTE: % power is based on 3489 MWt as rated. % flow is based on 108.5 Mlb/hr as rated. NOTE: Decay ratios greater than 1.1 are outside the range of the STAIF methodology applicability.

These points should be considered unstable without quantitative comparison.

LaSalle Unit I Cycle 9 Reload Arialysis Table 4.1 Neutronic Design Values Number of Fuel Assemblies Rated Thermal Power, MWt Rated Core Flow, Mlbm/hr Core Inlet Subcooling, Btu/lbm Moderator Temperature, 'F Channel Thickness, inch Fuel Assembly Pitch, inch Wide Water Gap Thickness, inch* Narrow Water Gap Thickness, inch Control Rod Datat Absorber Material Total Blade Support Span, inch Blade Thickness, inch Blade Face-to-Face Internal Dimension, inch Absorber Rod OD, inch Absorber Rod ID, inch Percentage B 4 C, %TD 764 3489 108.5 18.1 548.8 0.080 & 0.100 6.0 0.28110.261 0.281/02.61 B 4 C 1.580 0.260 0.200 0.188 0.138 70 The water gap thicknesses presented are based on 80/1 00-mil channels for ATRIUM-9B fuel. The control rod data represents original equipment control blades at LaSalle and were used in the neutronic calculations.

Siemens Power Corporation EMF-2276 Revision 1 Pagje 4-4 I LaSalle Unit I Cycle 9 Reload Analysis IJL -F EMF-2276 Revision 1 Page 4-5 Axial Location In Assembly 1- 2 5 7 7 7 5 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 2 3 4 6 4 3 2 4.00 4.20 7 4.70 7 4.20 4.00 2 4.00 3.00 8.00 8.00 8.00 3.00 400 5 4.20 7 7 7 7 4.20 5 4.70 8.00 4.95 4.95 4.95 4.95 8.004 " 4.70 7 7 7. 7 7 7 4.95 4.95 4.95 4.95 4.95 4.95 6 " 6_ 7 4.70 7 Water Channel 7 4.70 7 4.95 8.00 4.95 4.95 8.00 4-95 7 7 7 7 7 7 4.95 4.95 4.95 4.95 4.95 4.95 4 74 5 420 4.20 5 4.70 4.95 4.95 4.95 4.95 4.95 8.00 4.70 4.00 4.0 7000~> 2 3 4 7 6 7 4: 3 2 4.00 4.00 4.20 4.708 7 4.20' :4.00" 2 3.00 8.00 8.00 .00 3.00 4.00 1 2 5 7 7 7 5 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 Pellet Type 1 2 3 4 5 6 :7_Quantity 4 8 4 8 8 .4 36 U 2 3 5+ Gd 2 0 3 Concentration (wt%)3.00 4.00 4.00+ 3.00 4.20 + 8.00 4.70 4.70+ 8.00 4.95 Figure 4.1 SPCA9-4.56L-12G8.0/4G3.0-1OOM .Eurichment Distribution Siemens Power Corporation I

LaSalle Unit 1 Cycle 9 Reload Analysis_JL "3 F" Pellet Type Quantity 1 4 2 12 3 8 4 8 5 6 4 36 U 2 3 5 + Gd 2 0 3 Concentration (wt%) 3.00 4.00 4.20+ 8.00 4.70 4.70 + 8.00 4.95 Figure 4.2 SPCA9-4.56L-12G8.0-100M " -Enrichment Distribution Axial Location In Assembly EMF-227 Revision Page 4-1 2 4 6 6 6 4 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 23 5 6 3 2 2 2 2 4.20 6 4.70 4.20 4.00 4.00 8.00 8.00 8.00 4.00 3 3 4 3 6 6 6 6 6 4.20 4 4.70 8.00 4.995 4.95 46 .5 49 8.00 47 6 6 6 6 6 6 4.95 4.95 4.95 4.95 4.95 4.95 5 5 6 4.70 6 Water Channel 6 -4.70 6 8.00 4.95 4.95 8.00 4.95 6 6 6 6 6 6 4.95 4.95 4.95 4.95 4.95 4.95 3 3 4 20 6 6 6 6 6 4.20 4.70 8.20 4.95 4.95 4.95 4.95 4.95 4.20 4.70 .3 5 3 2 2 4.20 6 4.70 6 4.20 2 2 4.00 4.00 8.00 1 .00 8.00 4.00 4.00 1 2 4 6 6 6 4 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 Reload Anafvsis LaSalle Unit 1 Cycle 9 Reload Analysis I L --1 F Pellet Type Quantity 1 4 2 12 3 12 4 8 5 36 U 2 3 5+ Gd 2 0 3 Concentration (wt%)2.60 3.40 3.40+ 8.0Q 3.80 4.40 Figure 4.3 SPCA9-3.91L-12G8.0-100M Enrichment Distribution Siemens Power Corporation Axial Location In Assembly EMF-2276 Revision 1 Page 4-7 1 2 4 5 5 5 4 2 1 2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 2 2 3.40 5 340 5 340 2 2 3.40 3.40 8.00 8.00 440 8.00 3.40 3.40 3 3 3.40 5 5] 5 5 3.40 4 3.80 8.00 4.40 4.40 4.40 4.40 4.40 8.00 3.80 5 5 5 5 5 5 4A0 4.40 4.40 4.40 4.40 4.40 3 -3 4. 3.40 4 Water Channel s 34O. 4 44 8:0 4.40 4.40 4.40 4.0 8.00 .8.00 . 5 5 5 5 5 5 4.40 4.40 4.40 4.40 4.40 4.40 3 3 4 340 5 5 5 5 5 3.40 3.80 8.00 4.40 4.40 4.40 4.40 4.40 8.00 3.80 3* 3 2 32

• 5 340 -'3-.40--

---3.40 2 2 3.40 3.40 8.00 4.40 8.00 8.00 3.40 3.40 1 2 4 5 5 5 4 2 1 2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 I l LaSalle Unit 1 Cycle 9 P.Relr4 Anni SiS Axial Location In Assembly n Pellet Type Quantity 1 4 2 16 3 8 4 8 5 36 U235+ Gd 2 0 3 Concentration (wt%) 2.60 3.40 3.40 -+ 5.00 3.80 4.40 Figure 4.4 SPCA9-3.90L-8G5.0-I0OM Enrichment Distribution Siemens Power Corporation IL I-EMF-2276 Revision 1 Page 4-8 I 1 2 4 5 5 5 4 2 1 2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 3 3 2 2 3.40 5 2 5 3.40 2 2 3.40 3.40 5.00 4.40 3.40 4.40 500 3 3 3 3.40 40 5 3.40 4 3.80 5.00 4.40 4.40 4.40 4A0 4.40 .5.00 3.80 5 5 5 5 5 5 4Ao 4.40 4.40 4.40 4A0 4A0 5 2 5 WaterChannel 5 2 5 4.40 3.40 C4.40 4.40 3.40 4.40 5 5 5 5 5 5 4A0 4.40 4.40 4A0 4A0 4.40 4 3 5 5 5 5 3 4 3.80 3.40 4.40 4.40 4.40 40 4.40 340 3.80 5.00 5.00 3 3 2 2 340 5 2 5 3.40 2 2 ! 3.40 3.40 5:00 4.40 3.40 4.40 3.40 3.40 3.40 1 2 4 5 5 5 4 2 1 2.60 3.40 3.80 4.40 4.40 4.40 3.80 3A0 2.60 LaSalle Unit 1 Cycle 9 Rp!ond Analysis IL E MF-2276 Revision 1 Page 4-9 Axial Location in Assembly Pellet Type Quantity 1 4 2 8 3 12 4 8 5 40 U 2 3 5 + Gdt0 3 Concentration (wt%) 3.00 4.00 4.20 + 8.00 4.70 4.95 Figure 4.5 SPCA9-4.59L-12G8.0-100M Enrichment Distribution Siemens Power Corporation 1 2 4 5 5 5 4 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 3 3 3 5 30 2 2 4.20 5 4.20 5 4.20 4 .2 4.00 8.00 8.00 8.00 -8.00 4 5 5 5 5 5 5 5 4 4.70 4.95 4.95 4.95 4.95 4.95 4.95 4.95 4.70 3 3 5 4.20 5 5 4.20 5 8.00 4.95 8.00 5 WaterChannel 5 4.95 4.95 4.95 4.95 4.95 4.95 3 3. 5 4.20 5.20 4.95 8.00 4.95 4.95 8.00' 4.95 4 5 5 5 5 5 5 5 4 4.70 4.95 4.95 4.95 4.95 4.95 4.95 4.95 4.70 2 3 3 3 5 3 2 2 4.20 5 4.20 5 4.20 4.20 2 4.00 8.00 8.00 8.00 8.00 4.00 1 2 4 5 5 5 4 2 1 3.00 4.00 4.70 4.95 4,95 4.95 4.70 4.00 3.00 Reload Analvsis LaSalle Unit I Cycle 9 Reload Analysis Axial Location In Assembly Pellet Type Quantity 1 4 2 8 3 12 4 8 5 40 U 2 3 5+ Gd 2 O 3 Concentration (wt%)3.00 4.00 4.20 + 7.00. 4.70 4.95 Figure 4.6 SPCA9-4.59L-12G7.0-1O0M Enrichment Distribution Siol)ian Powor Corporation IL EMF-2276 Revision 1 Page 4-10 7423U 1 1 2 4 5 5 5 4 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 3 3 3 3 2 4.20 5 4.20 5 4.20 5 4.20 2 4.00 700 7.00 7.00 7.00 4.00 4 5 5 5 5 5 5 5 4 4.70 4.95 4.95 4.95 4.95 4.95 " 4.95 4.95 4.70 3 .3 5 4.20 5 5 420 5 7.00 7.00 5 5 5 Water Channel 5 5 5 4.95 4.95 4.95 4.95 4.95 4.95 3 33. 5 420 4.20 4.95 7.00 4.95 4.95 7.00 4.95 4 5 5 5 5 5 5 5 4 4.70 4.95 4.95 4.95 4.95 4.95 4.95 4.95 4.70 3 3 3 3 2 420 4-20 4.20 4.20 2 4.00 7.00 7.00 7.00 7.00 4.00 1 2 4 5 5 5 4 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 I LaSalle Unit 1 Cycle 9 PIr==Iv4 I L i-EMF-2276 Revision 1 Page 4-11 Axial Location In Assembly Pellet Type 1 2 3 4 5 6 Quantity 4 8 8 4 8 40 U 2 3 5 + Gd 2 0 3 Concentration (wt%) 2.60 3.40 3.40 + 7.00 3.40 + 8.00 3.80 4.40 Figure 4.7 SPCA9-3.96L-8G7.014G8.0-100M Enrichment Distribution Siemens Power Corporation 1 2 5 6 6 6 5 2 1 2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 4 3 3 4 2 340 6 3AO 6 3.40 6 340 2 3.40 8.00 4.40 7.00 4.40 7.00 4.40 8.00 3.40 5 6 6 6 6 6 6 6 5 3.80 4.40 4.40 4.40 4.40 4.40 4.40 4.40 3.80 3 3 6 3.40 6 6 340 6 4.40 7.00 4.40 4.40 7.00 4.40 6 6 6 Water Channel 6 6 6 4.40 4.40 4.40 4.40 4.40 4.40 3 .3 : 6 3.40 6 6 7.30" 6 4.40 7.00. 4.40 4.40 :7.0 4.40 5 6 6 6 6 6 6 *6 5 3.80 4.40 4.40 4.40 4.40 4.40 4.40 4.40 3.80 2 4 3 3 4" 3.40 340 6 3.40 6 -340 2 3.40 8.00 4.40 700 4.40 7.00 4.40 8:00 3.40 1 2 5 6 6 6 5 2 1 2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 PoInnrl Analysis LaSalle Unit 1 Cycle 9 Reload Analysis Axial Location In Assembly Pellet Type Quantity 1 4 2 12 3 8 4 8 5 40 U 2 3 5 + Gd 2 0 3 Concentration (wt%) 2.60 3.40 3.40 + 5.00 3.80 4.40 Figure 4.8 SPCA9-3.96L-8G5.0-1OOM

_ Enrichment Distribution Siemens Power Corporation EMF-2276 Revision 1 Page 4-12 IL 3-. 7 7, -r -z I 1 2 4 5 5 5 4 2 1 2.60 3.40 3.80 4.40 4.40 4.4 3.80 3.40 2.60 3 3 23.40 5 340 5 2 2 3.40 3.40 4.40 500 4.40 4.40 3.40 3.40 4 5 5 5 5 5 5 5 4 3.80 4.40 4.40 4.40 4.40 4.40 4.40 4.40 3.80 3 53. s 3.40 s 3.40 4.40 5.00 4.40 4.40 5.00 4.40 440 4 4.40 Water Channel 440 4.40 440 3.0 44 .40 4.40 4.40 4.40 4.40 5.00 5.00 4 5 5 5 5 5 5 5 4 3.80 4.40 4.40 4.40 4.40 4.40 4.40 4.40 3.80 2 2 5 3 5 3 5 2 2 3.40 3.40 4.40 5.00 4.40 4.40 3.40 3.40 1 2 4 5 5 5 4 2 1 2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 I !/

LaSalle Unit 1 Cycle 9 Reload Analysis I L I r" Pellet Type 1 2 3 4 5 6 Quantity 4 8 4 8 8 40 U 2 3 5 + Gd 2 0 3 Concentration (wt%) 3.00 4.00 4.00 + 3.00 4.20 + 6.00 4.70 4.95 Figure 4.9 SPCAS-4.58L-8G6.0/4G3.0-1 OM Enrichment Distribution Siemens Power Corporation Axial Location In Assembly EMF-2276 Revision 1 Page 4-13 1 2 5 6 6 6 5 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 3 4 4 3 2 2 4.00 6 4.20 6 4.20 6 4.00 2 4.00 3.00 4 6.00 6.00 3.00 . 5 6 6 6 6 6 6 6 5 4.70 4.95 4.95 4.95 4.95 4.95 4.95 4.95 4.70 6 420 6 6 420 6 6.00 4 4.95 6.00 6 6 6 6 6 6 4.95 4.95 9 Wate Channel 495 4.95 4.95 44 6 .20 6 .2 6 4.95 6.00 495 6.00 5 5 6 6 6 6 6 6 6 5 4.70 4.95 4.95 4.95 4.95 4.95 4.95 4.95 4.70 2 3 4 4 6 6 2 4.00 -4.20 4 4 .20 400 400 .00 6.00 6.00 ' 3.00 4.00 1 ---- 2 5 6 6 .. 6 5 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 LaSalle Unit 1 Cycle 9 Reload Analysis Axial Location In Assembly I L Figure 4.10 SPCA9-4.58L-8G6.0-100M SEnrichment Distribution cSipmpn; Pnwor rnrrnmrtfll 1 2 4 5 5 5 4 2 1 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 -3 5 2 2 2 2 5 420 4.20 4.005 4.002.9 4.95 4.95 4.00 4.00 4.00 4.00 4.95 6.09.00 4 4 5 5 5 5 5 5 5 4 4.70 4.95 4.95 4.95 4.95 4.95 4.95 4.95 4.70 5 3 5.3 4.20.4.20 4.9 4.*20 4.95 4.95 ..0 4.95 6.00 6.00o" 495 5 Water Channel 5 4.95 495 4.9 4.95 4.95 4s .5 49 5 3."3. 4.2 ~4.20 4.5 4.20 4.95 4.95 40 4.95 6.00 6.00 4 5 5 5 .5 5 5 5 4 4.70 4.95 4.95 4.95 4.95 4.95 4.95 4.95 4.70 3 3 2 2 5 3 5 3 5 2 2 2 2 5 4.20 .4.202 2 4.204-2 4.95 4.00 4.00 4.00 4.00 4.95 6.00 6.00 1 2 4 5 5 5 4 2 1 3.00 4.00 4.70 4.95 -4.95 4.95 4.70 4.00 3.00 Pellet Type Quantity -U 2 3 5 + Gd 2 Q 3 Concentration (wt%) 1 4 3.00 2 12 4.00 3 8 4.20 + 6.00 4 8 4.70 5 40 4.95 EMF-2276 Revision 1 Page 4-14 I d LaSalle Unit 1 Cycle 9 PDIJrt nrjannv rk I L "1 Fl Pellet Type 1 2 3 4 G Quantity 4 8 16 33 11 U 2 3 5 + Gd 2 O 3 Concentration (wt*lo) 2.72 3.53 3.94 4.53 3.69 + 6.00 Figure 4.11 SPCA9-4.06L-11G6.0-80M Enrichment Distribution Siemens Power Corporation Axial Location In Assembly EMF-2276 Revision I Page 4-15 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 G G G 23,69 39 3.69 2 3.5 3, 69 4.53 4.53 3.9 4.53 4.53 3.6 ,353 3,3 6.00 6 ,00 6.00 3 4 4 4 4 4 4 4 3 3.94 4.53 4.53 4.53 4.53 4.53 4.53 4.53 3.94 3 4 A 4 3 3. 4.53 4.53 6900 4.53 3.94 G "G 4 3.69 4 Water Channel 4 -369 4 6.00 4,53 4.53 6.00- 4.53 3 4 4 4 4 3 3.94 4.53 4.53 4.53 4.53 3.94 3G G 3369 4 369 4 3 3.94 4.53 4.53 '3.69 4.5 4.53 3690 4.53 3.94 G G *G. 2 3.69 4.53 453 69 4.53 4.53 2353 3,53 6.00 6.00 6,00 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 LaSalle Unit 1 Cycle 9 V A ncIy 0lC IL -I I-EMF-2276 Revision 1 PaQe 4-16 Axial Location In Assembly Pellet Type 1 2 3 4 G Quantity 4 8 16 34 10 U 2 3 5 + Gd 2 0 3 Concentration (wt%) 2.72 3.78 4.19 4.78 4.19+6.00 Figure 4.12 SPCA9-4.34L-1OG6.0-BOM Enrichment Distribution Siemens Power Corporation 1 2 3 3 4 3 3 2 1 2.72 3.78 4.19 4.19 4.78 4.19 4.19 3.78 2.72 G G 4 4 G 2 .94.19 4:19 2 0 4.19 4.78 4.78 6.00 4.78 4.78 6.00 3.78 3.8 6.00 6.00 6.00 3 4 4 4 4 4 4 4 3 4.19 4.78 4.78 4.78 4.78 4.78 4.78 4.78 4.19 3 4 4 4.19 3 4.19 4.78 4.78 6.00 4.78 4.19 4 G 44 G 4.78 Water Channel 4 4.19 4.78 4.78 el 4.78 60,4.78 4.78 e6.00 6.00 3 4 4 4 4 .3 4.19 4.78 4.78 4.78 4.78 4.19 4 G 4 4 4 4 3 4.19 4.78 4.78 4.19 4.78 4.78 4.78 4.78 4.19 419 .8 47 6.00 2 G G 2 4.19 4.19 4.19 3.8 4.00 4 4 4.78 4.78 .19 3.78 3.78 6.00 4.78 4.78 6.00 6.00 1 2 3 3 4 3 3 2 1 2.72 3.78 4.19 4.19 4.78 4.19 4.19 3.78 2.72 Y t EMF-2276 Revision 1 Page 4-17 LaSalle Unit 1 Cycle 9 Annlvsis S PCAD-393B-16QZ-100M Natural Uranium 3.90-8G5.0 3.91-12G8.0 4-56-12G8.0 4.56-12G8.014G3.0 Natural Uranium Lattice SPCA9-0.72L-0.GO.O-1OM SPCA9-3.90L-8G5.O-1 0OM SPCA9-3.91 L-12GS.0-100M 2 3 11 12 12 12 11 3 2 3 4 6 12 10 12 6 4 3 11 6 12 12 12 12 12 6 11 12 12 12 W W W. 12 12 12 12 10 12 W W W 12 10 12 12 12 12 W W W 12 12 12 11 6 12 12 12 12 12 6 11 3 4 6 12 10 12 6 4 3 2 3 11 12 12 12 11 3 2 SPCA9-4.56L-12G8.0-1OOM SPCA9-4.56L 12G8.O/4G3.0-100M SPCA9-0.72L-O-GO.0-1 DOM Fuel Rod Type 2 3 4 6 10 11 12 No. Rods 4 8 4 8 4 8 36 Figure 4.13 ATRIUM-9B LSA-1 19A Assembly Design Siemens Power CorporaLion Reload A LaSalle Unit 1 Cycle 9 Reload Analysis SPCA9-393B-16GZ-100M ROD _2 E2 E3 E4 E2 ROD 3 E2 E5 El0 E2 Lattice Index Enrichment

+ Gd "E2 0.72 wt% U-235 E3 2.60 wt% U-235 E4 3.00 wt% U-235 E5 3.40 wt% U-235 E6 3.40 wt% U-235 + 5.0 wt% Gd 2 0 3 E7 3.40 wt% U-235 + 7.0 wt% Gd 2 03 E8 3.40 wt% U.2235 + 8.0 wt% Gd 2 0I E9 3.80 wt% U-235 E10 4.00 wt% U-235 Lattice Index Enrichment

+ Gd Eli 4.00 wt% U-235 + 3.0 wt% Gd 2 0 3 E12 4.20 wt% U-235 + 6.0 wt% Gd 2 0 3 E13 4.20 wt 0/' U-235 + 7.0 wt% Gd 2 0 3 E14 4.20 wt% U-235 -t 8.0 wt% Gd 2 0 3 E15 4.40 wt% U-235 E16 4.70 wt% U-235 E17 4.70 wt% U-235.+ 8.0 wt% Gd 2 0 3 E18 4.95 wt% U-235 Figure 4.13 ATRIUM-9B LSA-1 19A Assembly Design (continued)

Siemens Power Corporation EMF-2276 Revision 1 Page 4-18 ROD 10 E2 E5 ROD _6 F E2 E6 E8 E14 ROD E2 E5 El0 Ell E2 ROD E2 E9 E16 E2 ROD 12 E2 E15 E18 E2 E17 LaSalle Unit 1 Cycle 9 SPCAS-396B-1 2GZB-1 COM Natural Uranium 3.96-8G5.0 3.96-BG7.0/4G8.0 I 4.59-12G7.0 4.59-12G8.0 Natural Uranium Lattice SPCA9-o0.72L-O.GO.O-100M SPCA9-3.96L-8G5.0-100M SPCA9-3.96L-8G7.O/4G8.O looM 2 3 11 12 12 12 11 3 2 3 7 12 8 12 8 12 7 3 11 12 12 12 12 12 12 12 11 12 8 12 W W W. 12 8 12 12 12 12 W W W 12 12 12 12 .8 12 W W W 12 8 12 11 12 12 12 12 12 12 12 11' 3 7 12 8 12 8 12 7 3 2 3 11 12 12 12 11 3 2 SPCA9-4.59L-12G7.0-100M SPCA9-4.59L-12GS.O-100M SPCA9-0.72L-O.GO.O-IOOM Fuel Rod Type 2 3 7 8 11 12 No. Rods 4 8 4 8 8 40 Figure 4.14 ATRIUM-9B LSA-1 19B Assembly Design Siemens Power Corporation ReIo EMF-2276 Revision 1 Page 4-19 ad Ana ys -

LaSalle Unit 1 Cycle 9 Reload Analysis SPCA9-396B-12GZB-1 00M ROD E2 !E2 E3 E4 E2 ROD 3 E2 E5 El0 E2 Lattice Index Enrichment

-+ Gd E2 0.72 wt% U-235 E3 2.60 wt% U-235 E4 3.00 wt% U-235 E5 3.40 wt% U-235 E6 3.40 wt% U-235 + 5.0 wt% Gd 2 0 3 E7 3.40 wt% U-235 + 7.0 wt% Gd 2 0 3 ES 3.40 wtrU-235 + 8.0 wt% Gd 2 0 3 E9 3.80 wt% U-235 E10 4.00wt% U-235 Lattice Index Enrichment

+ Gd Eli 4.00 wt% U-235 + 3.0 wt% Gd 2 O 3 E12 4.20 wt% U-235 + 6.0 wt% Gd 2 0 3 E13 4.20 wt% U-235 +.7.0 wt% Gd 2 0 3 E14 4.20 wt% U-235 + 8.0 wt% Gd 2 0 3 E15 4.40 wt% U-235 E16 4.70 wt% U-235 E17 4.70 wt% U-235 + 8.0 wt% Gd 2 0 3 E18 4.95 wt% U-235 Figure 4.14 ATRIUM-9B LSA-1 19B Assembly Design (continued)

Siemens Power Corporation EMF-2276 Revision 1 Page 4-20 ROD 7 E2 E5 E8 E13 E14 E2 ROD _8 E2 E6 E7 E13 E14 E2 ROD 11 E9 E16 E2'ROD 12 E2 E15 E18 E2 EMF-2276 Revision 1 Page 4-21 LaSalle Unit 1 Cycle 9 Reload Analysis SPCA9-396B-12GZC-1 OOM Natural Uranium 3.96-8G5.0 4.58-8G6.0 4.58-8G6.014G3.0 Natural Uranium Lattice SPCA9-0.72L-0.GO.0 looM SPCA9-3.96L-8G5.0-1 DOM SPCA9-4.58L-8G6.0-100M SPCA9-4.58L 8G6.014G3.0-100M SPCA9-0.72L-0.GO.0 1 0DM 2 3 11 12 12 12 11 3 2 3 5 12 9 12 9 12 5 3 11 12 12 12 12 12 12 12 11 12 9 12 W W W. 12 9 12 12. 12 12 W W W 12 12 12 12 9 12 W W W 12 9 12 11 12 12 12 12 12 12 12 11 3 5 12 9 12 9 12 5 3 2 3 11 12 12 12 11 3 2 Fuel Rod Type 2 3 5 9 11 12 No. Rods 4 8 4 8 8 40 Figure 4.15 ATRIUM-9B LSA-1 19C Assembly Design Siemens Power Corporation LaSalle Unit 1 Cycle 9 d Analysis ROD 2 E2 E3 E4 E2 SPCA9-396B-1 2GZC-1 0DM ROD 3 F E2 E5 El0 E2 Lattice Index Enrichment

+ Gd -E2 0.72 wt% U-235 E3 2.60 wt% U-235 E4 3.00 wt% U-235 E5 3.40 wt% U-235 E6 3.40 wt% U-235 + 5.0 wt% Gd 2 0 3 E7 3.40 wt% U-235 + 7.0 wt% Gd 2 0 3 ES 3.40 wt% U-2-S5 + 8.0 wt% Gd 2 0 3 E9 3.80 wt% U-235 E1O 4.00 wt% U-235 Lattice Index Enrichment

+ Gd Ell 4.00 wt% U-235 + 3.0 wt% Gd 2 0 3 E12 4.20 wt% U-235 + 6.0 wt% Gd 2 0 3 E13 4.20 wt% U-235 + 7.0 wt% Gd 2 03 E14 4.20 wt% U-235 + 8.0 wt% Gd 2 0 3 E15 4.40 wt% U-235 E16 4.70 wt% U-235 E17 4.70 wt% U-235 + 8.0 wt% Gd 2 0 3 E18 4.95 wt% U-235 Figure 4.15 ATRIUM-gB LSA-1 19C Assembly Design (continued)

Sionions Power Corporation EMF-2276 Revision 1 Page 4-22 ROD 5 E2 E5 Ell E2 ROD 9 E2 E6 E12 E2 ROD 11 E2 E9 E16 E2 ROD 12 E18 E2 Reload Analvsis LaSalle Unit 1 Cycle 9 Reload Analysis SPCA9-384B-11 GZ-80M SPCA9-0.72L-OGO.0-80M SPCA9-4.06L-1 1 G6.0-80M SPCA9-4.34L-10G6.0-OM 1 2 3 3 4 3 3 2 1 2 GI 4 4 G1 4 4 G1 2 3 4 4 4 4 4 4 4 3 Fuel Rod Type 1 2 3 4 G1 G2 3 4 3 3 2 1 4 G1 4 4 GI 2 4 4 4 4 4 3 W" W W G1 4 3 W W W 4 G1 4 W W W 4 4 3 GI 4 4 G2 4 3 4 GI 4 4 GI 2 3 4 3 3 2 1 No. Rods 4 8 16 33 10 1 SPCA9-0.72L-OGO.0-80M Figure 4.16 ATRIUM-9B SPCA9-384B-11GZ-80M Assembly Design Sioniens Power Corporation EMF-2276 Revision 1 Page 4-23 EMF-2276 Revision 1 Paqe 4-24 LaSalle Unit 1 Cycle 9 Re lnd Analvsis SPCA9-384B-11GZ-80M 1 A B A B C D E F G H J 2 A hC D A 0.72 wt% 2.72 wt% 3.53 wt% 3.78 wt% 3.94 wt% 4.19 wt% 4.53 wt% 4.78 wt% 3.69 wt% 4.19 wt%3 A E F A 4 A G H A G1 A A G2 A i H A U-235 U-235 U-235 U-235 U-235 U-235 U-235 U-235 U-235 + 6.00 wt% Gd 2 03 U-235 + 6.00 wt% Gd 2 0 Figure 4.16 ATRIUM-9B SPCA9-384B-11GZ-BOM Assembly Design

• _ (continued)

Siemens Power Corporation Reload Analvsis LaSalle Unit 1 Cycle 9 EMF-2276 Revision 1 Page 4-25 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 60 58 56 54 52 50 48 46 44 42 40 38 36 34 32 30 28 26 24 22 20 1I 16 14 12 10 8 6 4 2 Bundle Name GE9B-.PBCWB322-11GZ-100M-150 GE9B.PSCW8320-SGZ-1004M-150 GE9B-PF`CWB343-12GZ-80M-150 GE9B-P8CWB342-IOGZ-80M-15O SPCA9-393B-16GZ-IO0M SPCAS-396B-12GZB-1 OM SPCAB-384B-11GZ-80M SPCA9-396B-12GZC-1 DOM Number Load' of Bundles cle 56 89 104 143 208 68 36 40 Figure 4.17 LaSalle Unit I Cycle 9 Reference Loading Map Siemens Power Corporation Fuel 1 2 4 5 6 7 a 9 7 7 8 8 9 9 9 EMF-2276 Revision 1 Paqe 5-1 LaSalle Unit 1 Cycle 9 Reload Analysis 5.0 Anticipated Operational Occurrences Applicable Disposition of Events 5.1 Analysis of Plant Transients at Rated Conditions Limiting Transients:

Scram Transient Speed LRNB FWCFt LRNB' FWCF t LFWH*TSSS TSSS NSS NSS Reference

9.7 Reference

9.3 Load Rejection No Bypass (LRNB) Feedwater Controller Failure (FWCF) Loss of Feedwater Heating (LFWH)" Peak Neutron Flux (% Rated)460.2 371.3 401.1 342.9 zt Peak Heat Flux (% Rated)126.5 122.6 121.3 120.5 Peak Lower Plenum Pressure (psig)1206 1167 1203 1164 5.2 Analysis for Reduced Flow Operation Limiting Transient:

Slow Flow Excursion MCPR, Manual Flow Control -ATRIUM-9B and GE9 Fuel LHGRFAC 1 -ATRIUM-9B Fuel MAPFACf -GE9 Fuel ACPR ATRIUM-9B/GE9 0.341 0.38 0.30 / 0.33 0.31 /0.34 0.28 / 0.31 Reference

9.3 Figure

5.1 Figure 5.2 :T MCPRI and LHGRFAC, results are applicable at all Cycle 9 exposures and in all EOD and EOOS scenarios presented in Table 1.1. MCPR, limits are provided for maximum core flows of 102.5% and 105% of rated. The LHGRFACf multipliers provided in Figure 5.2 are applicable for maximum core flows of 102.5% and 105% of rated. Based on 100%P/81%F conditions.

Based on 100%P/105%F conditions.

This data to be furnished by CornEd.Siemens Power Corporation LaSalle Unit 1 Cycle 9 Reload Analysis 5.3 Analysis for Reduced Power Operation Limiting Transient:

Reference 9.3 Load Rejection No Bypass (LRNB) Feedwater Controller Failure (FWCF)MCPRp Base Case Operation LHGRFACp Base Case Operation' MCPRp, EOOS Conditions LHGRFACp, EOOS Conditions*

MAPFACP -All Operating Conditions' 5.4 ASME Overpressurization Analysis Limiting Event Worst Single Failure Maximum Vessel Pressure (Lower Plenum) Maximum Steam Dome Pressure 5.5 Control Rod Withdrawal Error Starting Control Pattern for Analysis Tables 5.1-5.4 Figures 5.3-5.6 Tables 5.1-5.4 Tables 5.1-5.4 Tables 5.1-5.4 <To be furnished by ComEd.> Reference 9.3 MSIV. Closure Valve Position Scram 1320 psig 1291 psig Figure 5.7< This data is to be furnished by CornEd. > 5.6 Fuel Loading Error < This data is to be furnished by ComEd. > 5.7 Determith-tion of Thermal Margins The results of the analyses presented in Sections 5.1-5.3 are used for the determination of the operating limit. Section 5.1 provides the results of analyses at rated conditions.

Section 5.2 provides for the determination of thie MCPR and LHGR limits at reduced flow (MCPR(, Figure LHGRFAC, values presented are applicable to SPC fuel. GE MAPFAC, limits will continue to be applied to GE9 fuel at off-rated power.Siemens Power Corporation EMF-2276 Revision 1 Page 5-2 EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 5-3 5.1; LHGRFACf, Figure 5.2). Section 5.3 provides for the determination of the MCPR and LHGR limits at conditions of reduced power (Figures 5.3-5.6, Tables 5.1-5.4).

Limits are presented for base case operation and the EOD and EOOS scenarios presented in Table 1.1. The results presented are based on the analyses performed by SPC. As indicated above, the final Cycle 9 MCPR operating limits need to be established in conjunction with the results from CoinEd analyses.Siemens Power Corporation EMF-2276 Revision 1 Page 5-4 LaSalle Unit 1 Cycle 9 Neioad Ana Idys Table 5.1 EOC Base Case and EOOS MCPRP Limits and LHGRFACP Multipliers for TSSS Insertion Times for Prepower Uprate Conditions (3323 MWt Rated Power)ATRIUM-9B Fuel EOOS Condition Base Case Operation Feedwater Heaters Out of Service (FHOOS) Single-Loop Operation Turbine Bypass Valves Out of Service (TBVOOS)Power (% Rated)0 25 25 63 84 100 0 25 25 63 100 0 25 25 63 84 100 0 25 25 63 100 MCPRP 2.70 2.22 2.07 1.56 1.51 1.46 2.85 2.38 2.38 1.62 1.47 2.71 2.23 2.08 1.57 1.52 1.47 2.70 2.22 2.17 1.63 1.49 LHGRFACP 0.66 0.66 0.66 0.94 0.98 0.99 0.64 0.64 0.64 0.90 0.99 0.66 0.66 0.66 0.94 0.98 0.99 0.66 0.66 0:66 0.90 0.94 GE9 Fuel MCPRP"2.70 2.22 2.12 1.57 1.53 1.50 2.85 2.38 2.38 1.62 1.51 2.71 2.23 2.13 1.58 1.54 1.51 2.70 2.22 2.17 1.65 1.53 Limits support operation with any combination of one SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), up to a 20°F reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF and MELLLA regions of the power/flow map.Siemens Power Corporation EMF-2276 Revision 1 Page 5-5 LaSalle Unit 1 Cycle 9 Analysis Table 5.1 EOC Base Case and EOOS MCPRP Limits and LHGRFACp Multipliers for TSSS Insertion Times for Prepower Uprate Conditions (3323 MWt Rated Power) (continued)

ATRIUM-9B Fuel EOOS/EOD Condition Recirculation Pump Trip Out of Service (No RPT) Turbine Control Valve (TCV) Slow Closure and/or No RPT TCV Slow Closure/ FHOOS and/or No RPT Idle Loop Startup Power (% Rated)0 25 25 63 100 0 25. 25 84 84 100 0 25 25 84 84 100 0 25 25 63 100 MCPRP 2.70 2.22 2.07 1.60 1.51 2.70 2.22 2.16 1.65 1.63 1.56 2.85 2.38 2.38 1.65 1.63 1.56 2.54 2.54 2.54 2.54 2.54 LHGRFACP 0.66 0.66 0.66 0.86 0.86 0.66 0.66 0.66 0.86 0.86 0.86 0.63 0.63 0.63 0.86 0.86 0.86 0.40 0.40 0.40 0.40 0.40 GE9 Fuel MCPR, 2.70 2.22 2.12 1.63 1.56 2.70 2.22 2.16 1.69 1.67 1.60 2.85 2.38 2.38 1.69 1.67 1.60 2.54 2.54 2.54 2.54 2.54 Limits support operation with any combination of one SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), up to a 20°F reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF and MELLLA regions of the powerlflow map.

EMF-2276 Revision I Page 5-6 LaSalle Unit 1 Cycle 9 Reload Analysis Table 5.2 Base Case MCPRP Limits and LHGRFACP Multipliers for NSS Insertion Times for Prepower Uprate (3323 MWt Rated Power)f EOOS Condition Base Case Operation Power (% Rated)0 25 25 63 84 100 ATRIUM-9B Fuel MCPRP LHGRFAC, 2.70 2.22 2.07 1.54 1.48 1.43 0.74 0.74 0.74 0.95 1.00 1.00 Limits support operation with any combination of one SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), up to a 20°F reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF and MELLLA regions for the power/flow map.Siemens Power Corporation GE9 MCPRP 2.70 2.22 2.07 1.56 1.51 .1.46 LaSalle Unit 1 Cycle 9 EMF-2276 Revision I Page 5-7 Table 5.3 EOC Base Case and EOOS MCPRp Limits and LHGRFACp for Multipliers for TSSS Insertion Times for Power Uprate Conditions (3489 MWt Rated Power)*ATRIUM-9B Fuel EOQS/EOD Condition Base Case Operation Feedwater Heaters Out of Service (FHOOS) Single-Loop Operation Turbine Bypass Valves Out of Service (TB VOOS)Power (% Rated)0 25 25 60 80 100 0 25 25 60 100 0 25 25 60 80 100 0 25 25 60 100 MCPRP 2.70 2.20 2.05 1.56 1.51 1.45 2.85 2.35 2.35 1.62 1.45 2.71 2.21 2.06 1.57 1.52 1.46 2.70 2.20 2.15 1.63 1.47 LHGRFAC, 0.67 0.67 0.67 0.94 0.98 1.00 0.65 0.65 0.65 0.90 1.00 0.67 0.67 0.67 0.94 0.98 1.00 0.67 .0.67 0.67 0.90. 0.94 GE9 Fuel MCPRP"2.70 2.20 2.10 1.57 1.53 1.49 2.85 2.35 2.35 1.62 1.49 2.71 2.21 2.11 1.58 1.54 1.50 2.70 2.20 2.15 1.65 1.51 Limits support operation with any combination of one SRVOOS. up to 2 TIPOOS (or the equivalent number of TIP channels), up to a 20°F reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF and MELLLA regions of the power/flow map.Siemens Power Corporation

ý W. j LaSalle Unit 1 Cycle 9 R4=In~r An~ivsis Table 5.3 EOC Base Case and EOOS MCPRP Limits and LHGRFACP Multipliers for TSSS Insertion Times for Power Uprate Conditions (3489 MWt Rated Power) (continued)

ATRIUM-9B Fuel EOOS/EOD Condition Recirculation Pump Trip Out of Service (No RPT) Turbine Control Valve (TCV) Slow Closure and/or No RPT TCV Slow Closure/ FHOOS and/or No RPT Idle Loop Restart Power (% Rated)0 25 25 60 100 0 25 25 80 80 100 0 25 25 80 80 100 0 25 25 60 100 MCPRP 2.70 2.20 2.05 1.60 1.50 2.70 2.20 2.15 1.65 1.63 1.54 2.85 2.35 2.35 1.65 1.63 1.54 2.54 2.54 2.54 2.54 2.54 LHGRFACP 0.67 0.67 0.67 0.86 0.86 0.67 0.67 0.67 0.86 0.86 0.86 0.64 0.64 0.64 0.86 0.86 0.86 0.40 0.40 0.40 0.40 0.40 EMF-2276 Revision 1 Page 5-8 GE9 Fuel MCPRP"2.70 2.20 2.10 1.63 1.55 -2.70 2.20 2.15 1.69 1.67 1.58 2.85 2.35 2.35 1.69 1.67 1.58 2.54 2.54 2.54 2.54 2.54 Limits support operation with any combination of one SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), up to a 201F reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of service in the-standard, ICF and MELLLA regions of the power/flow map.Siemens Power Corporation Reload Analvsis EMF-2276 Revision 1 Pace 5-9 LaSalle Unit 1 Cycle 9 tA~4 Andlv~i Table 5.4 EOC MCPR, Limits and LHGRFAC, Multipliers for NSS Insertion Times for Power Uprate Conditions (3489 MWt Rated Power)EOOS Condition Base Case Operation Power (% Rated)0 25 25 60 80 100 ATRIUM-9B Fuel MCPRP LHGRFACP 2.70 2.20 2.05 1.54 1.48 1.42 0.75 0.75 0.75 0.95 1.00 1.00 Limits support operation with any combination of one SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), up to a 20°F reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF and MELLLA regions of the power/flow map.Siemens Power Corporation GE9 MCPR, 2.70 2.20 2.05 1.56 1.51 1.45 PalmnH Analysis LaSalle Unit 1 Cycle 9 nI"loadU ta-Illys z 0. 0 1.00 0 20 40 60 80 100 Flow ('/% of Ratd)102.5% Maximum Core Flow 105% Maximum Core Flow MCPR, ATRIUM-9B 1.85 1.85 1.11 1.11 MCPi 1 GE9 1.85 1.85 1.11 1.11 MCPR 1 ATRIUM-gB 1.93 1.93 1.14 1.11 Figure 5.1. Flow Dependent MCPR Limits for Manual Flow Control Mode Siemens Power Corporation EMF-2276 Revision 1 Page 5-10 120 Flow (% rated) 0 30 102.5 105 MCPR 1 GE9 1.93 1.93 1.14 1.11 LaSalle Unit I Cycle 9 Reload Analysis 1.1 1 0.80.8-40 50 60 70 Percent of Rated Flow Flow (% rated) 0 30 76 105 LHGRFACf 0.69 0.69 1.00 1.00 Figure 5.2 Flow Dependent LHGR Multipliers for ATRIUM-9B Fuel Siemens Power Corporation EMF-2276 Revision 1 Page 5-11 LaSalle Unit I Cycle 9 D 1 14 18 1l .ei..l 2.75 2.65 2.55 2A5 2.35 2-25 2.15 2.05 1.95 1.85 1.75 1.65 1.55 1.45 1.35 1.25 I *I5 0 500 1000 1500 2000 2500 3000 3500 4000 Povmr (MWth)3323 MWt Rated Power Power (%) MCPRp Limits 100 1.46 84 1.51 63 1.56 25 2.07 25 2.22 0 2.70 3489 MWt Rated Power Power (%) MCPRp Limits 100 1.45 80 1.51 60 .1.56 25 2.05 25 2.20 0 2.70 Figure 5.3 Base Case Power Dependent MCPR Limits for ATRIUM-9B Fuel -TSSS Insertion Times Siemens Power Corporation EMF-2276 Revision 1 Paqe 5-12ý W- "a Yý115 LaSalle Unit 1 Cycle 9 I.2.75 2A5 2.55 2-45 2.35 225 2.15 2.05 1.95S 1,85 1.75 1.65 1.55 1.45 1MS5 1.25 1.15 0 1000 .1500. ---- 2000 -- 2500 3000 -3500 4000 Poer (MIWth)3323 MVVt Rated Power Power (%) MCPRp Limits 100 1.50 84 1.53 63 1.57 25 2.12 25 2.22 0 2.70 3489 MWt Rated Power Power (%) MCPRp Limits 100 1.49 80 1.53 60 1.57 25 2.10 25 2.20 0 2.70 Figure 5.4 Base Case Power Dependent MCPR Limits for GES Fuel -TSSS Insertion Times SioMnons Powor Corporation EMF-2276 Revision 1 Page 5-13 D 1 4 AýýI I't EMF-2276 Revision I Page 5-14 LaSalle Unit 1 Cycle 9 Reload Analysis 2.75, 2.65 2.55 2.45 235 2.25 2.15 2.05

• 1.95 1.75 1.65 1.55 1.45 1.35 1.25 1.15 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWih)3323 MWt Rated Power Power (%) MCPR, Limits 100 1.43 84 1.48 63 1.54 25 2.07 25 2.22 0 2.70 3489 MWt Rated Power Power (%) MCPRp Limits 100 1.42 80 1.48 60 1.54 25 2.05 25 2.20 0 2.70 Figure 5.5 Base Case Power Dependent MCPR Limits for ATRIUM-9B Fuel -NSS Insertion Times Siemens Power Corporation LaSalle Unit 1 Cycle 9 Reload Analysis 2.75 2.65 2.55 2A5 2.35 225 2.15 2.05 S1 95 1.85 1.75 1.65 1.55 1A5 1.35 1.25 1.15 0 -o500 1000 1500 -2000 2500 3000 3500 4000 Power (MWfh)3323 MWt Rated Power Power (%) MCPRp Limits 100 84 63 25 25 0 1.46 1.51 1.56 2.07 2.22 2.70 3489 MWt Rated Power Power (%) MCPRp Limits 100 80 60 25 25 0 1.45 1.51 1.56 2.05 2.20 . 2.70 Figure 5.6 Base Case Power Dependent MCPR Limits for GE9 Fuel -NSS Insertion Times Siemens Power Corporation EMF-2276 Revision 1 Page 5-15 LaSalle Unit 1 Cycle 9< This data is to be furnished by ComEd.> Figure 5.7 Starting Control Rod Pattern for Control Rod Withdrawal Analysis 0.Zipmanc Pnuiar Cre'mnm#*^.n EMF-2276 Revision 1 Page 5-16 LaSalle Unit 1 Cycle 9 ED 1 4 A-=i cis 6.0 Postulated Accidents 6.1 Loss-of-Coolant Accident 6.1.1 Break Location Spectrum 6.1.2 Break Size Spectrum EMF-2276 Revision I Page 6-1 Reference

9.8 Reference

9.8 6.1.3 MAPLHGR Analyses The MAPLHGR limits presented in Reference 9.9 are valid for LaSalle Unit 1 ATRIUM-9B (LSA-1) fuel for Cycle 9 operation.

Limiting Break: 1.1 ft 2 Break Recirculation Pump Discharge Line High Pressure Core Spray Diesel Generator Single Failure Peak clad temperature and peak local metal water reaction results for the Cycle 9 ATRIUM-9B reload fuel are 1795°F and 0.72% respectively.

These results are bounded by the results presented in Reference 9.11, which support the Reference

9.9 MAPLHGR

limits. The maximum core-wide metal-water reaction for Cycle 9 remains less than 0.16%. LOCA/heatup analysis results for LaSalle ATRIUM-9B are presented below (from Reference 9.11): Maximum PCT (°F)Peak Local Metal-Water Reaction (%)ATRIUM-9B Fuel 1825 0.79 The maximum core wide metal-water reaction is < 0.16%. 6.2 Control Rod Drop Accident < This data is to be furnished by CornEd. > 6.3 Spent Fuel Cask Drop Accident The radiological consequences of a spent fuel cask drop accident have been evaluated for SPC ATRIUM fuel designs in conformance with the analysis described in the LSCS UFSAR Section The peak local metal water reaction result is consistent with the limiting PCT analysis results reported in Reference 9.11.Siemens Power Corporation

-ý W_ I EMF-2276 LaSalle Unit 1 Cycle 9 Revision I Reload Analysis Page 6-2 15.7.5. The analysis is assumed to occur 360 days following shutdown of the reactor, and it is assumed that all 32 fuel assemblies in the cask completely fail as a result of the accident.

Because the accident is assumed not to occur sooner than 360 days following shutdown of the reactor, the source term for the accident will be very low due to fission product decay. Hence, the commensurate radiological whole-body and thyroid doses will be very low. The results of this analysis demonstrate that spent fuel cask drop accidents involving SPC"ATRIUM fuel will not exceed the established radiological whole-body and thyroid dose limits which-are a small fraction of the 10 CFR 100 limits for radiological exposures.

EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 7-1 7.0 Technical Specifications

7.1 Limiting

Safety System Settings 7.1.1 MCPR Fuel Cladding Integrity Safety Limit MCPR Safety Limit (all fuel) -two-loop operation MCPR Safety Limit (all fuel) -single-loop operation

7.1.2 Steam

Dome Pressure Safety Limit Pressure Safety Limit 7.2 Limiting Conditions for Operation 7 2 1 Averaoe Planar Linear Heat Generation Rate ATRIUM-9E MAPLHGR Average Planar Exposure (GWd/MTU)0.0 20.0 61.1 3 Fuel Limits MAPLHGR (kWtft) 13.5 13.5 9.39 Single Loop Operation MAPLHGR Multiplier for SPC Fuel is 0.90 7.2.2 Minimum Critical Power Ratio Rated Conditions MCPR Limit 1 1.11' 1.12 1325 psig Reference 9.9 GE9-Fuel MAPLHGR Limits < To be furnished by CornEd. > Reference 9.9 Flow Dependent MCPR Limits: Mfanual Flow Control Figure 5.1 Includes the effects of channel bow, up to 2 TIPOOS (or the equivalent number of TIP channels), a 2000 EFPH LPRM calibration interval and up to 50% of the LPRMs out of service.

This data is to be furnished by ComEd.Siemens Power Corporation J m EMF-2276 Revision 1 Paqe 7-2 LaSalle Unit I Cycle 9 Reload Analysis Power Dependent MCPR Limits: Base Case Operation

-TSSS Insertion Times Base Case Operation

-NSS Insertion Times EOD and EOOS Operation Figures 5.3 & 5.4 Figures 5.5 & 5.6 Tables 5.1-5.4 7.2.3 Linear Heat Generation Rateý ATRIUM-9B -Fuel- Steady-State LHGR Limits Average Planar Exposure LHGR (GWd/MTU) (kWVft) 0.0 14.4 15.0 14.4 61.1 8.32 Reference 9.1 GE9 Fuel Steady-State LHGR Limits < To be furnished by CornEd. >The protection against power transient (PAPT) linear heat generation rate curve for ATRIUM-9B fuel is identified in Reference 9.1 and is presented here as Figure 7.1 for convenience.

LHGRFAC 1 and LHGRFACp multipliers are applied directly to the steady-state LHGR limits at reduced power, reduced flow and/or EOD/EOOS conditions to ensure the PAPT LHGR limits are not violated during an AOO. Comparison of the Cycle 9 nodal power histories for the rated power pressurization transients with the approved bounding curves to show compliance with the 1% strain criteria for GE9 fuel is discussed in Reference 9.10. LHGRFAC Multipliers for Off-Rated Conditions

-ATRIUM-9B Fuel: LHGRFACt LHGR.FACp Figure 5.2 Tables 5.1-5.4 MAPFAC Multipliers for Off-Rated Conditions

-GE9 Fuel: MAPFACt MAPFACP< To be furnished by CornEd. > < To be furnished by CornEd. >R l a ... .. ... ..,r ..

EMF-2276 Revision 1 Page 7-3 LaSalle Unit 1 Cycle 9C, 3 20 18 16 14 12 10-8 6 4 2-i I I -10 15 20 25 30 35 40 45 50 Average Planar Exposure, GWd/MTU!55 s0 65 770 Figure 7.1 Protection Against Power-Transient LHGR Limit for ATRIUM-9B Fuel Siemens Power Corporation (0.19.4) (15.19.4)

(61.1.11)0 0 5 Reload Ana ys -V EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 8-1 8.0 Methodology References See XN-NF-80-19(P)(A)

Volume 4 Revision 1 for a complete bibliography.

8.1 ANF-913(P)(A)

Volume I Revision 1 and Volume 1 Supplements 2, 3 and 4, COTRANSA2.:

A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990. 8.2 ANF-524(P)(A)

Revision 2 and Supplement 1 Revision 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990. 8.3 ANF-1 125(P)(A) and-ANF-1 125(P)(A), Supplement 1, ANFB Critical Power Correlation, Advanced Nuclear Fuels Corporation, April 1990. 8.4 EMF-1 125(P)(A), Supplement 1 Appendix C, ANFB Critical Power Correlation Application for Co-Resident Fuel. Siemens Power Corporation, August 1997. 8.5 ANF-1 125(P)(A), Supplement 1 Appendix E Revision 0, ANFB Critical Power Correlation Determination of A TRIUM-9B Additive Constant Uncertainties, Siemens Power Corporation, September 1998. 8.6 XN-NF-80-19(P)(A)

Volume 1 Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Methodology for Boiling Water Reactors:

Benchmark Results for CASMO-3G/MICROBURN-B Calculation Methodology, Advanced Nuclear Fuels Corporation, November 1990. 8.7 EMF-CC-074(P)(A)

Volume 1, STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain, and Volume 2, STAIF -A Computer Program for BWR Stability Analysis in the Frequency Domain -Code Qualification Report, Siemens Power Corporation.

July 1994.Siemens Power Comnralinn EMF-2276 Revision1 LaSalle Unit 1 Cycle 9 Paoe 9-1 Reload Analysis 9.0 Additional References 9.1 EMF-2249(P)

Revision 1, Fuel Design Report for LaSalle Unit I Cycle 9 ATRIU MFuel Assemblies, Siemens Power Corporation, September 1999. 9.2 ANF-89-014(P)(A)

Revision I and Supplements 1 and 2, Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels 9x9-IX and 9x9-9X BWR Reload Fuel, Advanced Nuclear Fuels Corporation, October 1991. 9.3 EMF-2277 Revision 1, LaSalle Unit I Cycle 9 Plant Transient Analysis, Siemens Power Corporation, October 1999. 9.4 EMF-95-134(P), Criticality Safety Analysis for ATRIUM'm-9B Fuel, LaSalle Units I and 2 New Fuel Storage Vault, Siemens Power Corporation, December 1995. 9.5 EMF-96-117(P), Criticality Safety Analysis for ATRIUM'm-9B Fuel, LaSalle Unit 1 Spent Fuel Storage Pool (BORAL Rack), Siemens Power Corporation, April 1996. 9.6 EMF-95-088(P), Criticality Safety Analysis for A Fuel, LaSalle Unit 2 Spent Fuel Storage Pool (Boraflex Rack), Siemens Power Corporation, February 1996. 9.7 EMF-95-205(P)

Revision 2. LaSalle Extended Operating Domain (EOD) and Equipment Out of Service (EOOS) Safety Analysis for Fuel, Siemens Power -G-orpor-atierk,-June

-1996, 9.8 EMF-2174(P), LOCA Break Spectrum Analysis for LaSalle Units I and 2, Siemens Power Corporation, March 1999. 9.9 EMF-2175(P), LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUMTM-9B Fuel, Siemens Power Corporation, March 1999. 9.10 Letter, D. E. Garber (SPC) to R. J. Chin (CoinEd), "LaSalle Unit 1 Cycle 9 Mechanical Limits for GE9 Fuel." DEG:99:213, August 4, 1999. -9.11 Letter, D. E. Garber (SPC) to R. J. Chin (CoinEd), "10 CFR 50.46 Reporting for the LaSalle Units," DEG:99:129, May 6, 1999.

Technical Requirements Manual -Appendix I LI C9 Reload Transient Analysis Results Attachment 3 LaSalle Unit 1 Cycle 9 Plant Transient Analysis (Excerpts)

LaSalle Unit 1 Cycle 9 November 2001 EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 1-4.110 100 90 80 a) "0 "06 "w cr 70 60 50 40 30 20 10 0 0 10 20 30 40 50 60 70 Percent of Rated Flow 80 90 100 110 120 Figure 1.1 LaSalle County Nuclear Station Power / Flow Map Siemens Power Corporation LaSalle Unit 1 Cycle 9 1.80 1.70' 1.60 IM 1.40 1.30 1.20 1.10 -EMF-2277 Revision 1 Page 2-9-MCPf Unrrit- 102.5% Max Fow 0 GE -105% MaxFlow "A ATPJUM-gB

-105% Max Flow .........MC U it- 105% Max F.o i ,I S QI 60 Flow (% of Rated)102.5% Maximum Core Flow MCPRr GE9 MCPR, (penalty ATRIUM-9B included)1.85 1.85 1.11 1.11 _1.85 1.85 1.11 1.11 105% Maximum Core Flow MCPR 1 GE9 MCPR l(penalty ATRIUM-9B included)1.93 1.93 1.14 1.11 1.93 1.93 1.14 1.11 Figure 2.1 Flow-Dependent MCPR Limits for Manual Flow Control Mode Siemens Power rnrn^,n,;.-

a ATPJUM-gB-10275% Max Flow 4 GE9- 1025% Max Fko 0 20 40 80 100 120 Flow (% rated) 0 30 102.5 105 DI _r ý;.-t Anal sis EMF-2277 Revision 1 Page 2-10 LaSalle Unit 1 Cycle 9 Plant Transient Analysis 40 50 60 70 Percent of Rated Flow Flow (% rated)0 30 .76 105 LHGRFACt 0.69 0.69 1.00 1.00 Figure 2.2 Flow-Dependent LHGRFAC Multipliers for ATRIUM-9B Fuel Siemens Power Corporation 1,, 0:: 0.8' (5 -J LaSalle Unit 1 Cycle 9 Plant Transient Analysis Table 3.1 LaSalle Unit 1 Plant Conditions at Rated Power and Flow Reactor Thermal Power Total Core Flow Core Active Flow Core Bypass Flow Core Inlet Enthalpy Vessel Pressures Steam Dome Core Exit (upper plenum) Lower Plenum TCV Inlet Pressure Feedwater/Steam Flow Feedwater Enthalpy Recirculating Pump Flow (per pump) Core Average Gap Coefficient (EOC)3489 MWth 108.5 Mlbm/hr 93.8 Mlbm/hr 14.7 Mlbm/hr 523.9 Btu/Ibm 1001 psia 1013 psia 1038 psia 948 psia 15.145 Mlbm/hr 406.6 Btu/lbm 15.83 Mlbm/hr 1173 Btu/hr-ft 2-°FPower Carporntion EMF-2277 Revision 1 Page 3-9"Includes water channel flow.

EMF-2277 Revision 1 Page 3-10 LaSalle Unit 1 Cycle 9 ndII -P Ana Si Table 3.2 Scram Speed Insertion Times Control Rod Position (Notch)48 (full-out) 48* 45 39 25 5 0 (full-in)TSSS Time (sec)0.000 0.200'* 0.430 0.860 1.930 3.490 3.880 As indicated in Reference 8, the delay between scram signal and control rod motion is conservatively modeled. Sensitivity analyses indicate that using no delay provides conservative results.Siemens Power Corporation NSS Time (sec)0.OQO 0.200* 0.380 0.680 1.680 2.680 2.804 jant rans en Y LaSalle Unit 1 Cycle 9 Plant Transient Analysis EMF-2277 Revision 1 Page 3-1 1 Table 3.3 EOC Base Case LRNB Transient Results Power'/ Flow Pea k ATRIUM-9B ATRIUM-9B GE9 Neutron Flux ACPR LHGRFACp ACPR M% rated)Peak Heat Flux (% rated)TSSS Insertion Times 100/105 100/81 80/1 05 80/57.2 60/105 60/35.1 40/105 25/105 23.81/105 20/105 NDS 0.325 0.337 0.332 0.377 0.319 0.301 0.260" 0.191*° 0.186" 1.008 1.000 1.000 1.000 1.000 1.033 1.098 1.113 1.202 1.211 0.706 0.362 0.377 0.368 0.410 0.349 0.289 0.271 0.177" 0.171 0.980 438.8 460.2 367.6 323.7 253.1 135.0 106.0"" 44.9*

• 41.6"* 44.6 123.7 126.5 98.2 99.3 72.3 67.1 45.6" 27.0" 25.6"* 38.1 NSS Insertion Times 100/105 100/81 801105 80/57.2 60/105 0.304 0.288 0.317 0.278 0.309 1.000 1.000 1.009 1.014 1.038 0.338 0.323 0.351 0.306 0.338 401.1 409.5 347.5 256.1 245.9 121.3 122.0 96.9 94.2 71.6 Power presented relative to uprated power (3489 MWth). The analysis results presented are from an earlier cycle exposure.

The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.Siemens Power Corporation LaSalle Unit 1 Cycle 9 Table 3.4 EOC Base Case FWCF Transient Results Power*/ Flow ATRIUM-9B ATRIUM-9B ACPR LHGRFACP GE9 ACPR Peak Neutron .Flux (% rated)TSSS Insertion Times 100/105 100/81 80/105 80/57.2 60/105 60/35.1 40/105 25/105 23.811105 20/105 NDS 0.299 0.280 0.355 0.294 0.431" 0.251 0.582"" 0.884"" 0.936** 1.119 1.019 1.032 0.986 1.063 0.955 1.143 0.8914" 0.767"" 0.750"m 0.688 0.322 0.301 0.376 0.310 0.440 0.252 0.546"* 0.913"* 0.964*

• 1.029 371.3 303.1 327.2 203.7 218.3"* 104.7 128.0** 62.5"* 61.1. 70.2 NSS Insertion Times 100/105 80/105 60/105 40/105 25/105 23.81/105 0.280 0.341 0.417 , 0.570* 0.861

• 0.901"*Power presented relative to uprated power (3489 MWth). "4 The analysis results presented are from an earlier cycle exposure.

The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.Siemens Power Corooralion EMF-2277 Revision 1 Page 3-12 Peak Heat Flux (% rated)122.6 121.8 101.9 96.0 79.3** 67.0 58.8** 43.2** 42.2* 43.9 1.033 1.000 0.959 0.895'* 0.777"" 0.760"" 0.301 0.360 0.430 0.535' 0.871" 0.923'*342.9 312.3 229.7. 124.6"' 76.7"* 73.8'*120.5 100.8 79.9 58.5* 44.0' 42.9"" V "II Q " I iI I IJ.f'J u '

LaSalle Unit 1 Cycle 9 Plant Transient Analysis Table 3.5 Input for MCPR Safety Limit Analysis Fuel Related Uncertainties Parameter ANFB Correlation*

ATRIUM-9B GE9 Radial Power Local Peaking Factor Assembly Flow Rate (mixed core) Channel Bow Local Peaking Source Document Reference 17 Reference 12 Reference 16 Reference 5 Reference 5 Function of nominal and bowed local peaking and standard deviation of bow data (see Reference 18).Nominal Values and Plant Measurement Uncertainties Parameter Feedwater Flow Rate`' (Mlbm/hr)

Feedwater Temperature (fF) Core Pressure (psia) Total Core Flow (Mlbm/hr)

Core Power` (MWth)Value 22.7 426.5 1031.35 113.9 5232.35 Uncertainty

(%) (Reference

8) 1.76 0.76 0.50 2.50 Statistical Treatment Convoluted Convoluted Convoluted Convoluted Convoluted Convoluted Statistical Treatment Convoluted Convoluted Convoluted Convoluted

° Additive constant uncertainties values are used.

• Feedwater flow rate and core power were increased above design values to attain desired core MCPR for safety limit evaluation consistent with Reference 5 methodology.

Siemens Power Corporation EMF-2277 Revision 1 Page 3-13 LaSalle Unit 1 Cycle 9 Plant Transient Analysis Table 3.6 Flow-Dependent MCPR Results 102.5% Maximum Core Flow GE9 ATRIUM-9B 1.775 1.821 1.645 1.693 1.552 1.597 1.464 1.501 1.379 1.406 1.295 1.308 1.209 1.214 1.129 1.129 1.110 1.110 1.110 1.110 105% Maximum Core Flow GE9 ATRIUM-9B 1.866 1.914 1.711 1.761 1.604 1.649 1.505 1.543 1.412 1.439 1.322 1.336 1.232 1.237 1.149 1.149 1.110 7 1.110 EMF-2277 Revision 1 Page 3-14 Core Flow M% rated) 30 40 50 60 70 80 90 100 102.5 105-- fl...... V' LaSalle Unit 1 Cycle 9 Plant Transient Analysis a w b. 0 I-. z U Id a.TIME. SECONDS Figure 3.1 EOC Load Rejection No Bypass at 100/81 -TSSS Key Parameters Siemens Power Corporation.

EMF-2277 Revision 1 Page 3-15 LaSalle Unit 1 Cycle 9 -. ~A I TIME, SECONDS Figure 3.2 EOC Load Rejection No Bypass at 100/81 -TSSS Vessel Water Level qipmernv Pnwer Corormtion EMF-2277 Revision 1 Page 3-16 0 w N I-I,) z 0 z_ -J LiP ILd -J V) V) w 5.0 Mant I ransient "0 Y. ý LaSalle Unit 1 Cycle 9 Dis,,t Tr~nci~nt W 0 U) (L LiJ -:2 0~TIME, SECONDS Figure 3.3 EOC Load Rejection No Bypass at 100/81 -TSSS Dome Pressure Siemens Power Corporation EMF-2277 Revision 1 Page 3-17 DI i Trýneiont nalvsis LaSalle Unit 1 Cycle 9 Pl=.

A.n~lvsis Arflo 3OO.O 3W. -- 200.0 LL 0 z () 10.0 .0-100.0 1 I CORE POWER HEAT FLUX CORE FLOW STEAM FLOW FEED FLOW L --r ----------i ----I --I ---.0 5.0 10.0 15.0 TIME, SECONDS 20.0 Figure 3.4 EOC Feedwater Controller Failure at 100/105 -TSSS Key Parameters Siemens Power Corporation EMF-2277 Revision 1 Page 3-18 25.0 01 -r-s-ent nalysis I--1 I 11 LaSalle Unit 1 Cycle 9 Plant Transient Analysis 0 Ir LI N I.-cn z W 0 im -J W hin Li I--J W, IAJ TIME. SECONDS Figure 3.5 EOC Feedwater Controller Failure at 100/105 -TSSS Vessel Water Level Siomens Power Corporation EMF-2277 Revision 1 Page 3-19 LaSalle Unit 1 Cycle 9 (n a. w (n En w 0 0 EMF-2277 Revision 1 Page 3-20 TIME. SECONDS Figure 3.6 EOC Feedwater Controller Failure at 100/105 -TSSS Dome Pressure Siemeins Pnwpr rnrnnrtinn Pl- 'r -i-ent Analysis LaSalle Unit 1 Cycle 9 Pi~ni Tr~n~tpn?

Anilysis 200 175 150 0) "0 125 C 0 100 L -o E 75 z 50 25 0.0 .1 .2 .3 .4 .5 .6 .7 .8 .S 1.0 1.1 Rodiol Power Peaking 1.2 1.3 1.4 1.5 1.6 Figure 3.7 Radial Power Distribution for SLMCPR Determination Sionions Power Corooralion EMF-2277 Revision 1 Page 3-21 Plant Transient Analysis EMF-2277 Revision 1 Page 3-22 LaSalle Unit 1 Cycle 9 D1 'rU ; -- AnCISIv. *C 0 n t r 0 R 0 d C 0 r n e r-I- Figure 3.8 LaSalle Unit 1 Cycle 9 Safety Limit Local Peaking Factors SPCA9-393B-16GZ-1 0DM With Channel Bow (Assembly Exposure of 22,500 MWd/MTU)D.....-.ontrol Rod Corner 1.023 1.055 1.068 1.112 1.099 1.102 1.049 1.023 0.977 1.055 0.958 0.894 1.016 0.894 1.007 0.877 " 0.927 1.002 1.068 0.894 1.031 1.065 1.084 1.056 1.010 0.863 1.011 1.112 1.016 1.065 Internal 1.044 0.980 1.051 1.099 0.894 1.084 Water 1.063 0.863 1.038 1.102 1.007 1.056 Channel 1.035 0.971 1.041 1.049 0.877 1.010 1.044 1.063 1.035 0.990 0.846 0.992 1.023 0.927 0.863 0.980 0.863 0.971 0.846 0.895 0.970 0.977 1.002 1.011 1.051 1.038 1.041 0.992 0.970 0.931 EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 3-23 ontrol Rod Corner 1.013 1.042 1.056 1.110 1.098 1.100 1.037 1.010 0.967 1.042 0.944 1.025 0.879 1.014 0.871 1.005 "0.912 0.989 1.056 1.025 1.018 1.064 1.081 1.055 0.997 0.989 0.999 1.110 0.879 1.064 Internal 1.043 0.848 1.047 1.098 1.014 1.081 Water 1.059 0.978 1.035 1.100 0.871 1.055 Channel 1.034 0.840 1.037 1.037 1.005 0.997 1.043 1.059 1.034 0.977 0.968 0.979 1.010 0.912 0.989 0.848 0.978 0.840 0.968 0.881 0.956 0.967 0.989 0.999 1.047 1.035 1.037 0.979 0.956 0.921-Figure 3.9 LaSalle Unit 1 Cycle 9 Safety Limit Local Peaking Factors SPCA9-396B-12GZB-1OOM and SPCA9-396B-1 2GZC-1 OOM With Channel Bow (Assembly Exposure of 25,000 MWd/MTU)Pnwer itrntoration R 0 d C 0 r n e r EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 3-24 Control Rod Corner Figure 3.10 LaSalle Unit 1 Cycle 9 Safety Limit Local Peaking Factors SPCA9-384B-1 1 GZ6-80M With Channel Bow (Assembly Exposure of 20,000 MWdMTU)0 n t r 0 R 0 d C 0 r n e r 1.022 1.056 1.061 1.035 1.102 1.028 1.045 1.029 0.982 1.056 0.947 1.018 1.003 0.879 0.997 1.004 0.919 1.011 1.061 1.018 1.001 1.050 1.081 1.048 0.996 0.992 1.012 1.035 1.003 1.050 Internal 0.926 0.983 0.987 1.102 0.879 1.081 Water 1.077 0.853 1.049 1.028 0.997 1.048 Channel 1.040 0.970 0.979 1.045. 1.004 0.996 0.926 1.077 1.040 0.859 0.980 0.996 1.029 0.919 0.992 0.983 0.853 0.970 0.980 0.891 0.983 0.982 1.011 1.012 0.987 1.049 0.979 0.996 0.983 0.941 LaSalle Unit 1 Cycle 9 Plant Transient Analysis 0. U 0 500 1000 1500 2000 2500 3000 3500 Power (MWth)3323 MWth Rated Power Power (%) MCPRP Limit 100 84 63 25 25 0 1.46 1.51 1.56 2.07 2.22 2.70 3489 MWth Rated Power Power -(%) MCPRP Limit 100 80 60 25 25 0 1.45 1.51 1.56 2.05 2.20 2.70 *Figure 3.11 EOC Base Case Power-Dependent MCPR Limits for ATRIUM-98 Fuel -TSSS Insertion Times Siemens Power Corporation EMF-2277 Revision 1 Page 3-25 4000 LaSalle Unit 1 Cycle 9 Plant Transient Analysis 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWth) -3323 MWth Rated Power Power (%) MCPRP Limit 100 84 63 25 25 0 1.50 1.53 1.57 2.12 2.22 2.70 3489 MWth Rated Power Power (%) MCPRP Limit 100 80 60 25 25 0 1.49 1.53 1.57 2.10 2.20 2.70 Figure 3.12 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel -TSSS Insertion Times EMF-2277 Revision 1 Page 3-26 LaSalle Unit 1 Cycle 9 nt Transient Analysis 0 500 1000 1500 2000 2500 3000 3500 4000 Power IMWth)3323 MWth Rated Power Power (%) MCPRP Limit 100 84 63 25 25 0 1.43 1.48 1.54 2.07 2.22 2.70 3489 MWth Rated Power Power (%) MCPRp Limit 100 80 60 25 25 0 1.42 1.48 1.54 2.05 2.20 2.70 Figure 3.13 EOC Base Case Power-Dependent MCPR Limits for ATRIUM-9B Fuel -NSS Insertion Times Sienienn Pnwpr .nrnnrsrinn EMF-2277 Revision 1 Page 3-27 Pl2nt Transient Analysis EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 3-28 2.75 2.65 2.55 2.45 2.35 2.2S 2.15 2.05 1.95 1.55 1.75 1.65 1.55 1.45 1.35 1.25 1.15 0 500 1000 1500 2000 2500 3000 3500 4000 Power IMWth)3323 MWth Rated Power Power (%) MCPR, Limit 100 84 63 25 25 0 1.46 1.51 1.56 2.07 2.22 2.70 3489 MWth Rated Power Power (%) MCPRp Limit 100 80 60 25 25 0 1.45 1.51 1.56 2.05 2.20 2.70 Figure 3.14 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel -NSS Insertion Times LaSalle Unit 1 Cycle 9 1.30 1.20 1.10 " 1.00 -J0.90 0.10 0.70 0.60 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWth)3323 MWth Rated Power Power (%) LHGRFACP.100 84 63 25 25 0 0.99 0.98 0.94 0.66 0.66 0.66 3489 MWth Rated Power Power (%) LHGRFACP 100 80 60 25 25 0 1.00 0.98 0.94 0.67 0.67 0.67 Figure 3.15 EOC Base Case Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel -TSSS Insertion Times Siemens Power Corooration EMF-2277 Revision 1 Page 3-29 01-* Trnnsient Analysis LaSalle Unit 1 Cycle 9 Dih Trn;ant Ann[lVsis C. U II. 0 -J 0.60 0 500 1000 1500 2000 2500 3000 3500 Power (MWth)3323 MWth Rated Power Power (%M LHGRFACP 100 84 63 25 25 0 1.00 1.00 0.95 0.74 0.74 0.74 3489 MWth Rated Power Power (%) LHGRFACP 100 80 60 25 25 0 1.00 1.00 0.95 0.75 0.75 0.75 Figure 3.16 EDC Base Case Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel -NSS Insertion Times EMF-2277 Revision 1 Page 3-30 4000 O"L I a". F LaSalle Unit 1 Cycle 9 Plant Transient Analysis EMF-2277 Revision 1 Page 5-5 Table 5.1 EOC Feedwater Heater Out-of-Service Analysis Results ATRIUM-9B GE9 Power*/Flow

(%ratedl%rated) tCPR LHGRFACP "ACPR 100/105 100/81 80/105 80/57.2 60/105 60/35.1 401105 25/105 23.811105 0.311 0.286 0.376 0.306 0.482** 0.216 0.698** 1.142'* 1.209*'1.000 1.034 0.968 1.055 0.927' 1.123 0.831"* 0.6784* 0.662'0.329 0.302 0.390 0.320 0.472 0.228 0.677*' 1.156"' 1.216 `w* Power presented relative to uprated power (3489 MWth). The analysis results presented are from an earlier cycle exposure.

The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.Siemens Power Corporation Event FWCF FWCF FWCF FWCF FWCF FWCF FWCF FWCF FWCF LaSalle Unit 1 Cycle 9 Plant Transient Analysis Table 5.2 Abnormal Recirculation Loop Startup Analysis Results ATRIUM-99B Power*/Flow FCV (%rated/%rated)

Position ACPR* LHGRFACp 33.33/47 27% open 1.40 0.425* Power presented relative to uprated power (3489 MWth). 6CPR results for ATRIUM-9B fuel are conservatively applicable for GE9 fuel.EMF-2277 Revision 1 Page 5-6 LaSalle Unit 1 Cycle 9 Plant Tr~n~ip~nt Analysis EMF-2277 Revision 1 Page 5-7 Table 5.3 EOC Turbine Bypass Valves Out-of-Service Analysis Results ATRIUM-9B GE9 Power*/Flow

(%rated/%rated)

ACPR LHGRFACP ACPR 100/105 0.359 0.968 0.392 100/81 0.359 0.947 0.394 80/105 0.418 0.942 0.449 80/57.2 0.417 0.957 0.447 60/105 0.499,0 0.917 0.514 60/35.1 0.327 1.032 0.332 40/105 0-658' 0.859" 0.619 25/105 0.962°

• 0.750""' 0.952" 23.81/105 1.004.* 0.736* 1.002** Power presented relative to uprated power (3489 MWth). The analysis results presented are from an earlier cycle exposure.

The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.Siemens Power Corporation Event FWCF FWCF FWCF FWCF FWCF FWCF FWCF FWCF FWCF Plant Transient Analvsis LaSalle Unit 1 Cycle 9 Plant Transient Analysis EMF-2277 Revision 1 Page 5-8 Table 5.4 EOC Recirculation Pump Trip Out-of-Service Analysis Results.ATRIUM-9B GE9 Power"/Flow

(%ratedl%rated)

ACPR LHGRFA.Cp 6CPR 100/105 100/81 80/105 80157.2 100/105 100/81 80/105 80/57.2 60/105 60/35.1 40/105 25/105 23.81/105 0.382 0.371 0.389 0.391 0.353 0.308 0.403 0.295 0.466 0.190 0.596' 0.858'" 0.896**0.909 0.868 0.923 0.899 0.942 0.948 0.920 1.003 0.901 1.120 0.857'" 0.757"" 0.743"" 0.433 0.430 0.438 0.439 0.391 0.349 0.438 0.327 0.492 0.193 0.581 0.861" 0.910**Power presentod rolativo to uprated power (3489 MWth) The analysis results presontod are from an earlier cycle exposure.

The ACPR and LHGRFACp results nra consorvitivoly used to establish the thermal limits.Event LRNB LRN8 LRNB LRNB FWCF FWCF FWCF FWCF FWCF FWCF FWCF FWCF FWCF LaSalle Unit 1 Cycle 9 Table 5.5 EOC Turbine Control Valve Slow Closure Analysis Results ATRIUM-9B Slow Valve Characteristics 1 TCV closing at 2.0 sec 1 TCV closing at 2.7 sec 2 TCVs closing at 7.75 sec 1 TCV closing at 2.0 sec 2 TCVs closing at 7.75 sec 1 TCV closing at 2.0 sec 2 TCVs closing at 2.0 sec 2 TCVs closing at 7.75 sec 1 TCV closing at 2.0 sec 2 TCVs closing at 2.0 sec 2 TCVs closing at 7.75 sec, 2 TCVs closing at 2.7 sec 1 TCV closing at 2.0 sec 2 TCVs closing at 7.75 sec 1 TCV closing at 2.0 sec 1 TCV closing at 2.0 sec 1 TCV closing at 2.0 sec 1 TCV closing at 2.0 sec 2 TCVs closing at 7.75 sec 1 TCV closing at 2.0 sec Power*/Flow

(%rated/%rated) 100/105*

• 100/105°*

100/105

• 1 00/81 0

• 1 00/8140 80/105*4 80/105t 80/105** 80/57.2o 80157.2f 80/57.2

• 60/105** 601105°o 60/35.1'" 4011 O5t 25/105i 23.81/105 t 23.81/1051 Event LRNB LRNB LRNB LRNB LRNB LRNB LRNB LRNB LRNB LRNB LRNB LRNB LRNB LRNB LRNB LRNB "-LRNB LRNB LRNB w/ FHOOS ACPR 0.420 0.419 0.219 0.369 0.199 0.432 0.527 0.293 0.504 0.520 0.277 0.432 0.346 0.591 0.8281 0.9921 1.011 0.977 0.363 LHGRFACP 0.902 0.8"99 1.057 0.928 1.107 0.911 0.882 1.014 0.911 0.928 1.115 0.932 1.001 0.991 0.759$ 0.707 0.6991 0.721"0.944 Power presented relative to uprated power (3489 MWth). Scram initiated by high neutron flux. I Scram initiated by high dome pressure.

I The analysis results presented are from an earlier cycle exposure.

The "CPR and LHGRFACp results are conservatively used to establish the thermal limits.Rip.mens Power Corooration EMF-2277 Revision 1 Page 5-9 GE9 100/105*°ACPR 0.461 0.461 0.233 0.421 0.223 0.466 0.568 0.314 0.548 0.564 0.305 0.461 0.370 0.5B5 0.824 0.9841 1.0081 0.954 0.396:

• A-=I cis Plant I rans en LaSalle Unit 1 Cyclit 9 Plant Transient Analysis EMF-2277 Revision 1 Page 5-10 Table 5.6 EOC Recirculation Pump Trip and Feedwater Heater Out-of-Service Analysis Results ATRIUM-9B GE9 Power 4 /Flow (%rated/%rated)

ACPR LHGRFACG " ACPR 100/1 05 0.332 0.954 0.375 100/81 0.305 0.948 0.354 100/105 0.358 0.933 0.391 100/81 0.307 0.968 0.342 80/105 0.419 0.911 0.448 80/57.2 0.300 1.007 0.329 60/105 0.508` 0.882 0.523 60/35.1 0.212 1.104 0.226 40/105 0.705' 0.804" 0.664 25/105 1.073'* 0.673"* 1.092" 23.81/105 1.125 *' 0.658** 1.146`* Power presented relative to uprated power (3489 MWth). The analysis results presented are from an earlier cycle exposure.

The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.Event LRNB LRNB FWCF FWCF FWCF FWCF FWCF FWCF FWCF FWCF FWCF LaSalle Unit 1 Cycle 9 Plant Transient Analysis CL a. UE 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWIh)3323 MWth Rated Power Power (%) MCPR. Limit 100 63 25 25 0 1.47 1.62 2.38 2.38 2.85 3489 MWth Rated Power Power (%)100 60 25 25 0 MCPRP Limit 1.45 1.62 2.35 2.35 2.85 Figure 5.1 EOC Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel EMF-2277 Revision 1 Page 5-11 LaSalle Unit 1 Cycle 9 am 0 0"a I " I C. C.) II 0 -J-- 0 500 1000 1500 2000 2500 3000 3500 Power IMWth)3323 MWth Rated Power 3489 MWth Rated Power Power (%)100 63 25 25 0 LHGRFACP 0.99 0.90 0.64 0.64 0.64 Power (%)100 60 25 25 0 LHGRFACP 1.00 0.90 0.65 0.65 0.65 Figure 5.2 EOC Feedwater Heaters Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel EMF-2277 Revision 1 Page 5-12 4000 LaSalle Unit 1 Cycle 9 Plant Transient Analysis C.)0 500 1000 1500 2000 2500 3000 3500 Power iMWth)3323 MWth Rated Power MCPRP Limit 1.51 1.62 2.38 2.38 2.85 3489 MWth Rated Power Power (%) MCPRP Limit 100 60 25 25 0 1.49 1.62 2.35 2.35 2.85 Figure 5.3 EOC Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel gint onnn Pn%AwsF rn-~.-,.-EMF-2277 Revision 1 "Page 5-13 4000 Power (%)-4 100 63 25 25 0 LaSalle Unit 1 Cycle 9 Tanci~fnt Analysis V ICEL I C. C.2.65 2.55 2.4S 2.35 2.25 2.15 2.05 1.95 1.8s 1.75 1.65 1.55 1.45 1.35 1.25 1.15 0[ Idle Loop Restart SOLMCPR I 500 3323 MWth Rated Power Power (%) MCPRp Limit 100 63 25 25 0 2.54 2.54 2.54 2.54 2.54 3489 MWth Rated Power Power (%) MCPRp Limit 100 60 25 25 0 2.54 2.54 2.54 2.54 2.54 Figure 5.4 Abnormal Idle Recirculation Loop Startup Power-Dependent MCPR Limits for ATRIUM-9B Fuel EMF-2277 Revision 1 Page 5-14 1000 1500 2000 Power (MWth)2500 3000 3500 4000 3500 2500 3000 LaSalle Unit 1 Cycle 9 Plant Transient Analysis 1.30 1.25 1.20 1.15 1.10 1.05s 1.00 2 0.95 C. 0.90 LL. 0.85 w 3 0.80 "- 0.75 0.70 0.65 0.60 0.55 0.50. 0.45 0.40 0.35 0 500 3323 MWth Rated Power Power (%) LHGRFACp 100 63 25 25 0 0.40 0.40 0.40 0.40 0.40 3489 MWth Rated Power Power (%)100 60 25 25 .0 LHGRFACP 0.40 0.40 0.40 0.40 .0.40 Figure 5.5 Abnormal Idle Recirculation Loop Startup Power-Dependent LHGR Multipliers for ATRIUM-SB Fuel Siemens Power Corporation EMF-2277 Revision 1 Page 5-15 IdeLoop Restart ALGFAP 1000 1500 2000 Power (MWlh)2500 3000 3500 4000 2 LaSalle Unit 1 Cycle 9 Plnnr Tr~nn.ient Analysis 2.65 2.55 2.45' 2.35 2.25 2.15 2.05 1.95, 1.85, 1.75 1.65 1.55 1.45 1.35 1.25 1.15 0 500 3323 MWth Rated Power. Power (%) MCPRp Limit 100 63 25 25 0 2.54 2.54 2.54 2.54 2.54 3489 MWth Rated Power Power (%) MCPRp Limit 100 60 25 25 0 2.54 2.54 2.54 2.54 2.54 Figure 5.6 Abnormal Idle Recirculation Loop Startup Power-Dependent MCPR Limits for GE9 Fuel EMF-2277.

Revision 1 Page 5-16 4000 1000 is00 IdlLLop estart] _OMPRJ 2000 Power (MWth)2500 3000 3500 Plant Transient Analysis LaSalle Unit 1 Cycle 9 Plant Transient Analysis C. a.2.75 2.65 2.55 2.45 2.35 2.25' 2.15' 2.05' 1.95 1.85 1.75' 1.65 1.55' 1.45' 1.35' 1.25' 1.15 0* FWCF Power Uprate OLMCPR ....Pre-Power Uprate OLMCPR 500 3323 MWth Rated Power Power M%) MCPRP Limit 100 63 25 25 0 1.49 1.63 2.17 2.22 2.70 3489 MWth Rated Power Power (%) MCPRP Limit 100 60 25 25 0 1.47 1.63 2.15 2.20 2.70 Figure 5.7 EOC Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits for ATRIUM-SB Fuel EMF-2277 Revision 1 Page 5-17 1000 1500 2000 Power (MWth)2500 3000 3500 4000 3000 3500 2500 EMF-2277 LaSalle Unit 1 Cycle 9 PeM 5-2271 Plant Transient Analysis Page 5-18 0.) ,-I .J 1.30 1.25 1.20 1.15 1.10 1.05 1.00 0.95 0.90 0.85 0.0 0.75 0.70 0.65 0.60 0 4000 500 1000 1500 2000 2500 3000 3500 Power (MWIh)3323 MWth Rated Power Power (%)100 63 25 25 0 LHGRFACo 0.94 0.90 0.66 0.66 0.66 3489 MWth Rated Power Power (%) LHGRFACp 100 60 25 25 0 0.94 0.90 0.67 0.67 0.67 Figure 5.8 EOC Turbine Bypass Valves Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel Power (%)

EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 5-19 I.2.75 2.65 2.55 2.45 2.35 2.25 2.15 2.05 1.95 1.55 1.75 1.65 1.55 1.45 1.35 1.25 1.15 0 4000 500 1000 1500 2000 2500 3000 3500 Pow.r (MWIh)3323 MWth Rated Power Power (%) MCPRP Limit 100 63 25 25 0 1.53 1.65 2.17 2.22 2.70 3489 MWth Rated Power Power (%)100 60 25 25 0 MCPRP Limit 1.51 1.65 2.15 2.20 2.70 Figure 5.9 EOC Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel.. ... t ri' leNr l tnn LaSalle Unit 1 Cycle 9 Plant Transient Analysis a.2.75 2.65 2.55 2.45 2.35 2.25 2.15 2.05 1.95 1.,5 1.75 1.65 1.55 1.45 1.35 1.25 1.15 0 500 1000 1500 2000 2500 3000 3500 Power IMWIh)3323 MWth Rated Power Power (%) MCPRP Limit 100 63 25 25 0 1.51 1.60 2.07 2.22 2.70 3489 MWth Rated Power Power 1%) MCPRP Limit 100 60 25 25 0 1.50 1.60 2.05 2.20 2.70 Figure 5.10 EOC Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel EMF-2277 Revision I Page 5-20 4000 EMF-2277 LaSalle Unit 1 Cycle 9 Revision I Plant Transient Analysis Page 5-21 0 500 1000 1500 2000 2500 3000 3500 Power (MWth)4000 3323 MWth Rated Power Power (%) LHGRFACP 100 63 25 25 0 0.86 0.86 0.66 0.66 0.66 3489 MWth Rated Power (%) L 100 60 25 25 0 Power HGRFACP 0.86 0.86 0.67 0.67 0.67 Figure 5.11 EOC Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel Siemens Power Corporation LaSalle Unit 1 Cycle 9 Pl~ni T r'nc:;pnt Analysis 0, U 0 500 1000 1500 2000 2500 3000 -3500 Power (MWth)3323 MWth Rated Power Power (%) MCPRV Limit 100 63 25 25 0 1.56 1.63 2.12 2.22 2.70 3489 MWth Rated Power Power (%M 100 60 25 25 0 MCPRP Limit 1.55 1.63 2.10 2.20 2.70 Figure 5.12 EOC Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel EMF-2277 Revision 1 Page 5-22 4000 Pinnt Trmnsient Analysis t LaSalle Unit 1 Cycle 9 Plant Transipnt Analysis 2.75 2.65 2.55 2.45 2.35 2.25 2.15 2.05 1.95 1.85 1.75 1.65 1.55 145 1.35 1.25 1.15 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWth)3323 MWth Rated Power MCPRP Limit 1.56 1.63 1.65 2.16 2.22 2.70 3489 MWth Rated Power Power (%) MCPRP Limit 100 80 80 25 25 0 1.54 1.63 1.65 2.15 2.20 2.70 Figure 5.13 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel Siemens Power Cornoratinn EMF-2277 Revision 1 Page 5-23 Power (%)100 84 84 25 25 0 LaSalle Unit 1 Cycle 9 Plant Transient Analysis 1.30 1.25 1.20 1.15 1.10 1.05 0. 1.00 u C: 0.95 -J 0.90 0.85 0.50 0.75 0.70 0.65 0.60 0 3323 MWth Rated Power Power (%)100 84 84 25 25 0LHGRFACP 0.86 0.86 0.86 0.66 0.66 0.66 3489 MWth Rated Power wer (%) LHGRFACP 100 80 80 25 25 0 0.86 0.86 0.86 0.67 0.67 0.67 Figure 5.14 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel a__.nrOn.a flnrnnrint;n

.Slow TCV Closure

• LRNB No RPT a FWCF No RPT -Power Uprale LHGRFACp .....Pre-Power Uprale LHGRFACo

• C Ue EMF-2277 Revision 1 Page 5-24 4000 500 1000 1500 2000 Power IMWth)2500 3000 3500 Power (%) LHGRFACp LaSalle Unit 1 Cycle 9 Plant Transient Analysis Z. C.2.75 2.65 2.55 2.45 2.35 2.25 2.15 2.05 1.95 1.85 1.75, 1.65 1.55, 1.45 1.35 1.25 1.1s, 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWlh)3323 MWth Rated Power 3489 MWth Rated Power MCPRP Limit 1.60 1.67 1.69 2.16 2.22 2.70 Power (%)100 80 80 25 25 0 MCPRP Limit 1.58 1.67 1.69 2.15 2.20 2.70 Figure 5.15 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel-. ..... )... .UA n. .i'% ,tin EMF-2277 Revision 1 Page 5-25 Power 1%)100 84 84 25 25 0 S.EMF-2277 Revision 1 Page 5-25 LaSalle Unit 1 Cycle 9 Plant Transient Analysis C cc CL 0 500 1000 1500 2000 2500 3000 3500 Power (MWth)3323 MWth Rated Power .Power (%) MCPRP Limit 100 84 84 25 25 0 1.56 1.63 1.65 2.38 2.38 2.85 3489 MWth Rated Power Power (%) MCPRP Limit 100 80 80 25 25 0 1.54 1.63 1.65 2.35 2.35 2.85 Figure 5.16 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip and Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9g Fuel EMF-2277 Revision 1 Paqe 5-26 4000 LaSalle Unit 1 Cycle 9 Plant Transient Analysis U) a'. =, 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWth)3323 MWth Rated Power Power (%) LHGRFACP 100 84 84 25 25 0 0.B6 0.86 0.86 0.63 0.63 0.63 3489 MWth Rated Power Power (%)100 80 80 25 25 0 LHGRFACP 0.86 0.86 0.86 0.64 0.64 0.64 Figure 5.17 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip and Feedwater Heaters Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel Siemens Power Corporation EMF-2277 Revision 1 Page 5-27 LaSalle Unit 1 Cycle 9 Plant Transient Analysis CL 0 S00 1000 1500 2000 2500 3000 3500 Power (MWth)3323 MWth Rated Power 3489 MWth Rated Power MCPRP Limit 1.60 1.67 1.69 2.38 2.38 2.85 Power (%)100 80 80 25 25 0 MCPRP Limit 1.58 1.67 1.69 2.35 2.35 2.85 Figure 5.18 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip and Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel Siemens Power Corporation EMF-2277 Revision 1 Page 5-28 A000 Power (%)100 84 84 25 25 0 LaSalle Unit 1 Cycle 9 Plant Transient Analysis Table 7.1 ASME Overpressurization Analysis Results 102%P/105%F Peak Peak Neutron Heat Event Flux Flux (%rated) (%rated)MS IV TCV TSV 425.43 70S.96 710.29 135.28 142.84 142.85 Maximum Vessel Pressure Lower Plenum (psig) 1320.26 1318.41 1318.41 s'ininiS Pnwtir Coriorrntinn EMF-2277 Revision 1 Page 7-2 Maximum Dome Pressure S(psig) 1291.12 1287.56 1287.55 LaSalle Unit 1 Cycle 9 I~lnt~ T;r~ncipnt Analvsis 0 Li I.-W. 0 I z 0d 0..0 .0 2.0 3.0 4.0 TIME, SECONDS"Figure 7.1 Overpressurization Event at 102/105 MSIV Closure Key Parameters Siomons Power Corporation EMF-2277 Revision 1 Page 7-3 5.0 I1n- Trn~sian Analysis .... ..

LaSalle Unit 1 Cycle 9 Plant Analysis 0 N id 0 z -J w I ,J W 2-0 TOAE, SECONDS Figure 7.2 Overpressurization Event at 102/105 MSIV Closure Vessel Water Level Siemens Power Coronration EMF-2277 Revision 1 Page 7-4.Plant Transient Analysis LaSalle Unit 1 Cycle 9 Plant Transient Analysis a. D: V) z (/L 0. Id .0 nj.0 1.0 2.0 3.0 TIME, SECONDS 4.0=Figure 7.3 Overpressurization Event at 102/105 MSIV Closure Lower Plenum Pressure Siemens Power Corporation EMF-2277 Revision 1 Page 7-5 5.0 LaSalle Unit 1 Cycle 9 Plant Transient Analysis Lj (n w 0.. Lii M 0 0 TIME, SECONDS Figure 7.4 Overpressurization Event at 102/105 MSIV Closure Dome Pressure c;.m-ae .. ,O ,ar r .nnrst*;'n EMF-2277 Revision 1 Page 7-6 LaSalle Unit 1 Cycle 9 SRV BANK 1 SRV BANK 2 SRV BANK 3 SRV BANK 4 L,0.0 :L: I :; it I:w .0I !i.S0 TIME, SECONDS Fr~ure 7.5 Overpressurization Event at 102/105 MSIV Closure Safety/Relief Valve Flow Rates Sicinoiir Powor Corporation EMF-2277 Revision 1 Page 7-7 qvnfV n..150c m -J ~.1000 0 -J LL CL LA 500.0 ,.0o 3.0 4.0 5.0 rialL ..- io ýa u o Technical Requirements Manual -Appendix I LI C9 Reload Transient Analysis Results Attachment 4 ARTS Improvement Program Analysis, Supplement 1 (Excerpts)

LaSalle Unit 1 Cycle 9 November 2001 Technical Requirements Manual -Appendix I LIC9 Reload Transient Analysis Results 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 (a) Based on the GE9/10 LHGR Improvement Report, the MAPFACs are applied to LHGR (Reference 24)LaSalle Unit 1 Cycle 9 November 2001 Technical Requirements Manual -Appendix I LI C9 Reload Transient Analysis Results Attachment 5 TCV Slow Closure Analysis (Excerpts)

LaSalle Unit 1 Cycle 9 November 2001 Technical Requirements Manual -Appendix I L1C9 Reload Transient Analysis Results Table 4.- TOP and MOP Values for the Off-rated Transient Events LRNBP, One TCV Slow LRNBP, All TCV Slow Closure at 50%/s, 3 TCV Fast Closure at 19%/s Closure Calculated TOP 26.17 49.27 Calculated MOP 26.17 55.30 Adjusted MOP 60.83 Required MOP 38.0 Required MAPFAC 0.62 Limiting MACFAC 0.60 (a)Note: (a) Based on Figure 3.2-2 in COLR. (b) Based on the GE9/10 LHGR Improvement Report, the MAPFACs are applied to LHGR (Reference 24)LaSalle Unit I Cycle 9 November 2001 Administrative Technical Requirements

-Appendix A LI C9 Reload Transient Analysis Results PRESS RISEP![ VALVE FLOW VALVE FLOW VAIE nu 19.10-. 0.4 a Figure 1. LRNBP from Rated Power, All TCV Fast Closure, Direct Scram, EOC-RPT LaSalle Unit I Cycle 9"a MaI .5 I VESSEL 2 SAFElY 3 RELIEF a DYDACC TIME ISECOOS!I a T7NE 2CECONOi2 4.m I C MI = a. U a a.' a.. a I 6.' C MI C TinE tSECONDS)TInE (SECONDS)I I)2.0 I November 1999 Administrative Technical Requirements

-Appendix A LIC9 Reload Transient Analysis Results TINE ISECONOSI TZNE ISECONIS)a 9 Mi a a. = C ii b. U I.a Mi a 1.. .l Ja. 0.O 2.6 TInE (SECONDS)8.0 1.I TIME (SECONDS)LRNBP from Rated Power, One TCV Slow Closure(50%/second)/Three TCV Fast Closure, Flu> Scram, EOC-RPT OOS LaSalle Unit I Cycle 9 C Mi a U 9 a Mi VOID RI ACTIVITY IACT IVITY 3 EACTIVITT

'EAC7VITYV I.-Figure 2.November 1999 VO1O RI OOPPLEI .3SCRAMIq

'n'Ta!

Administrative Technical Requirements

-Appendix A LIC9 Reload Transient Analysis Results I NEUTRON FLUX 2 AVE SURFACE HEAT FLUX 5 COR INLET FLOW *10.0 0.0§TIME ISECONOI)U, a a. = C LI a S a LI d a I YES 2 SAF 3 REL 4 AYp EL PRESS RISECPSI)

TY VALVE FLOW EF VALVE FLOW i VA! VE ringU I ¶O.O .. .... ., , TIME ISCONOS) I VOU REACTIVITY 2 OOP ýLER REACTIVITY 3 SCR REACTIVITY i oe"-ae..g TIME ISECONOS)$.0 TIME ISECONOS)LRNBP from 50% Power, One TCV Slow Closure(50%/second)/Three TCV Fast Closure, Fit Scram LaSalle Unit I Cycle 9=. a" La Figure 3.i3.L November 1999.0q ,O I Administrative Technical Requirements

-Appendix A LIC9 Reload Transient Analysis Results.NEU RON FLUX 2 AVE SURFACE HEAT FLUX 5 COR INLET FLOW T.0 .E S.6O is. TINE ISECONDS)TInE ISECONOS)Figure 4. LRNBP from 50% Power, All TCV Closure at 19%/second, Pressure Scram LaSalle Unit I Cycle 9 a SM 5 U I I.' U SM a.SEL PRESS RISEIPSI)

YEl VALVE FLOW [EF VALVE FLOW lee vii uC rl Iu-SE.ON TIME ISECOIIOS)

'em.. 10.Za 0a IM'-0 TIME ISECONOS)I VES 2 SAF 3 REL A QVD November 1999 Technical Requirements Manual -Appendix I Li C9 Reload Transient Analysis Results Attachment 6 LaSalle Unit 1 Cycle 9 Operating Limits For Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity LaSalle Unit I Cycle 9 November 2001 Framatome ANP Richland, Inc. Proprietary fFRAMATOME AN P March 22, 2001 DEG:01:045 Dr. R. J. Chin Nuclear Fuel Services (Suite 400) Exelon Corporation 1400 Opus Place Downers Grove, IL 60515-5701

Dear Dr. Chin:

LaSalle Unit I Cycle 9 Operating Limits for Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity Ref: 1: LaSalle County Nuclear Station Unit I Technical Specifications, as amended.

Ref: 2: EMF-2277 Revision 1, LaSalle Unit I Cycle 9 Plant Transient Analysis, Siemens Power Corporation, October 1999. Ref: 3: EMF-2276 Revision 1, LaSalle Unit 1 Cycle 9 Reload Analysis, Siemens Power Corporation, October 1999. Ref: 4: Letter,-DE--Garber-(FRA-ANP) to R. J. Chin (Exelon), "LaSallefnit-l-Cycle 9 Base Case Operating Limits for Proposed ITS Scram Times," DEG:01:013, January 18,2001.

Ref: 5: Letter, D. E. Garber (FRA-ANP) to R. J. Chin (Exelon), "Transmittal of Condition Report 9191," DEG:01:038, February 27, 2001. Exelon has proposed replacing the current Technical Specifications (Reference

1) with Improved Technical Specifications (ITS) during LaSalle Unit 1 Cycle 9 (LI C9) operation.

The operating limits for L1C9 (References 2 and 3) were established consistent with the scram times presented in Reference I and are not consistent with the proposed ITS surveillance times. Exelon has requested that FRA-ANP perform analyses to address a mid-cycle transition to the ITS for base case operation and one equipment out-of-service (EOOS) scenario.

Reference 4 describes the determination of analytical scram times consistent with the ITS and provided base case operating limits. Reference 5 identifies an error in the fuel thermal conductivity used in the transient analyses for LaSalle, including the analyses provided-in Reference

4. Framatome ANP Richland, Inc.2101 Horn Rapids Road Richland, WA 99352 Tel: (509) 375-8100 Fax: (509) 375-8402 Dr. R. J. Chin DEG:O1:045 March 22, 2001 Page 2 The attachment provides the LIC9 base case and slow TCV closure/FHOOS and or no RPT transient analysis results and operating limits using the analytical scram times and the corrected fuel thermal conductivity.

The base case operation limits provided in the attachment supercede those transmitted in Reference

4. Very truly yours, David Garber Project Manager slg Enclosure cc: P. Kong DEG:01:045 Attachment Page A-1 LaSaile Unit I Cycle 9 Operating Limits for Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity Limiting Condition for Operation (LCO) 3.1.3.3 of the current LaSalle Unit 1 Technical Specifications (Reference

1) specifies the average scram insertion times of all operable control rods. The average control rod insertion times must not exceed the scram times for the requirements of LCO 3.1.3.3 to be met. Exelon is planning to implement Improved Technical Specifications (ITS) for LaSalle Unit 1 during Cycle 9. The scram surveillance times in the proposed ITS are slightly more restrictive than those presented in Reference

1. Additionally, the surveillance requirement for the ITS is that each rod must meet the scram times. The LaSalle Unit I Cycle 9 (LI C9) operating limits (References 2 and 3) are based on the average scram times presented in Reference

1. Therefore, the limiting base case and equipment out-of-service transient analyses used to set the operating limits provided in References 2 and 3 must be reanalyzed with revised scram times in order to support the mid-cycle implementation of the ITS. FRA-ANP provided proposed ITS surveillance scram times to Exelon in Reference 4, Table 1. The Reference 4 analytical scram times are presented in Table 1 for completeness.

FRA-ANP informed Exelon of an error in the fuel thermal conductivity used in COTRANSA2 calculations (Reference 5). The analysis results presented in Tables 2 and 3 include the effect of the corrected fuel thermal conductivity.

Reference 9 provided a disposition of LOCA and UFSAR events for ITS scram times for LaSalle.

The Reference 9 disposition remains applicable.

Base Case Operation Reference 4 provided base case operating limits for the proposed ITS scram times. After Reference 4 was issued, FRA-ANP informed Exelon of an error in the fuel thermal conductivity used in COTRANSA2 calculations (Reference 5). The analyses provided in Reference 4 have been reanalyzed using the corrected fuel thermal conductivity.

The results of these analyses are presented in Table 2.

..............-...

a .I.~. a IWIJAMLIL y DEG:01:045 Attachment Page A-2 Figures 1 and 2 present the revised base case MCPRp limits for the and GE9 fuel, respectively.

The sum of the LlC9 safety limit MCPR (1.11 per Reference

2) and the ACPR results from Table 2 are also presented in Figures 1 and 2. The Reference 2 base case LHGRFACp multipliers and the LHGRFACp results from Table 2 are presented in Figure 3. Review of Figure 3 shows that all of the ATRIUM-9B LHGRFACp results are above the LHGRFACp multipliers, and therefore, the Reference 2 base case LHGRFACp multipliers remain applicable for the proposed ITS scram times. TCV Slow Closure/FHOOS and/or No RPT Exelon requested that FRA-ANP provide operating limits for the most limiting equipment out-of service (EOOS) scenario provided in Reference

2. Review of the Reference 2 limits shows that the most limiting two-loop operation EOOS scenario is TCV slow closure/FHOOS and/or no RPT. The TCV slow closure/FHOOS and/or no RPT limits consider transient analysis results from the following scenarios:

TCV slow closure (up to all four valves), EOC RPT OOS, FHOOS, and a combination of FHOOS and EOC RPT OOS. (Note: TCV slow closure analyses with FHOOS are bound by TCV slow closure analyses at nominal feedwater temperature, and therefore, no specific analyses are required for this scenario.)

In order to reduce the workscope required to establish new limits, only a subset of the analyses reported in Reference 2 have been reanalyzed.

Review of Figures 5.16, 5.17 and 5.18 in Reference 2 show that the TCV slow closure analyses are limiting for all power levels above 25% power (872.25 MWt); the FWCF no RPT with FHOOS id limiting at 25% power. Additionally, these figures show that there is considerable margin between the analysis results and the limits at power levels of 40% (1395.6 MWt) and 60% (2093.4 MWt). Table 5.5 of Reference 2 was reviewed to determine which specific TCV slow closure analyses required reanalysis to establish the limits. Tables 5.1 (FHOOS) and 5.4 (EOC RPT OOS) of Reference 2 were also reviewed since the limits are applicable for EOC RPT OOS or FHOOS only. Table 3 presents the analysis results required to adequately establish the slow TCV closure/FHOOS and/or no RPT limits. Figures 4 and 5 present the revised slow TCV closure/FHOOS and/or no RPT MCPRp limits for the ATRIUM-98 and GE9 fuel, respectively.

The sum of the LIC9 safety limit MCPR (1.11 per Reference

2) and the ACPR results from Table 3 are also presented in Figures 4 and 5.* ATRIUM is a trademark of Framatome ANP.

i-ramatome ANI- Kicnland, Inc. Proprietary DEG:01:045 Attachment Page A-3 The Reference 2 slow TCV closure/FHOOS and/or no RPT LHGRFACp multipliers and the LHGRFACp results from Table 3 are presented in Figure 6. Review of Figure 6 shows that all of the ATRIUM-9B LHGRFAC, results are above the LHGRFACP multipliers, and therefore, the Reference 2 slow TCV closure/FHOOS and/or no RPT LHGRFACp multipliers remain applicable.

The MCPRp limits and LHGRFACp multipliers provided in Figures 4-6 protect operation with up to four TCVs closing slowly, EOC RPT OOS, FHOOS and any combination of up to four TCVs closing slowly, EOC RPT OOS and FHOOS. The only equipment out-of-service scenarios provided in Reference 2 not explicitly protected by the slow TCV closure/FHOOS and/or no RPT limits are single-loop operation (discussed below), turbine bypass valves OOS, and abnormal startup of an idle loop. Comparison of turbine bypass valves OOS and the TCV slow closure/FHOOS and/or no RPT limits in Table 2.2 of Reference 3 shows the TCV slow closure/FHOOS and/or no RPT limits clearly bound the turbine bypass valves OOS limits. Consequently, applying the TCV slow closure/FHOOS and/or no RPT limits will protect operation with the turbine bypass OOS. No analyses were. performed to address the abnormal startup of an idle loop limits with ITS scram times and the corrected fuel thermal conductivity.

Single-Loop Operation Figures 1-3 provide the two-loop operation (TO) MCPRp limits and LHGRFACp multipliers for base case operation.

Reference 7 indicates that the consequences of base case pressurization transients in single-loop operation (SLO) are bound by the consequences of the same transient initiated from the same power/flow conditions in TLO and that the TLO base case ACPRs and the LHGRFACp multipliers remain applicable for SLO. Reference 2 indicates the LIC9 TLO safety limit MCPR is 1.11 and the SLO safety limit MCPR is 1.12. Since the TLO ACPR results are applicable to SLO, the SLO ATRIUM-9B and GE9 MCPRp limits can be determined by adding 0.01 to the base case operation MCPRp limits provided in Figures 1 and 2 to account for the increase in safety limit MCPR. The base case operation LHGRFACP multipliers presented in Figure 3 remain applicable for SLO. The conclusion that TLO ACPR results generally bound SLO results has been demonstrated for both base case operation and some equipment out-of-service scenarios for other BWRs. Although specific LIC9 analyses for a combination of TCV slow closure/FHOOS and/or no RPT in SLO have not been performed, FRA-ANP expects the TLO operation ACPR results would remain applicable in DEG:01:045 Attachment Page A-4 SLO for this scenario.

Therefore, SLO MCPRp limits for TCV slow closure/FHOOS andlor no RPT can be determined by adding 0.01 to the TCV slow closure/FHOOS and/or no RPT MCPRp limits reported in Figures 4 and 5 to account for the increase in safety limit MCPR. The Figure 6 TCV slow closure/FHOOS and/or no RPT LHGRFACp multipliers remain applicable for SLO. GE9 Mechanical Limits Reference 6 provides an evaluation of the GE mechanical limits for LI C9. An evaluation of the GE9 mechanical limits for the rated power analyses reported in Tables 2 and 3 was performed.

It was demonstrated that the maximum nodal power ratio history curve for the analyses are bound by either the LIC9 or L2C8 curves. It is FRA-ANP's position that the GE mechanical limits criteria have been met for the implementation of ITS provided no GE9 LHGR set down was required for either LIC9 or L2C8; if an LHGR set down was required for the GE9 fuel for LIC9 or L2C8, further evaluation may be required.

....References

1. LaSalle County Nuclear Station Unit I Technical Specifications, as amended.

2. EMF-2277 Revision 1, LaSalle Unit I Cycle 9 Plant Transient Analysis, Siemens Power Corporation, October 1999. 3. EMF-2276 Revision 1, LaSalle Unit I Cycle 9 Reload Analysis, Siemens Power Corporation, October 1999. 4. Letter, D. E. Garber (FRA-ANP) to R. J. Chin (Exelon), "LaSalle Unit I Cycle 9 Base Case Operating Umits for Proposed ITS Scram Times," DEG:01:013, January 18, 2001. 5. Letter, D. E. Garber (FRA-ANP) to R. J. Chin (Exelon), "Transmittal of Condition Report 9191," DEG:01:038, February 27, 2001. 6. Letter, D. E. Garber (SPC) to R. J. Chin (ComEd), "LaSalle Unit I Cycle 9 Mechanical Limits for GE9 Fuel," DEG:99:213, August 4, 1999. 7. EMF-95-205(P)

Revision 2, LaSalle Extended Operating Domain (EOD) and Equipment Out of Service (EOOS) Safety Analysis forATRlUMU T-9B Fuel, Siemens Power Corporation, ' l bj~~e196.

8. EMF-96-1 89 Revision 0, LaSalle Unit I Cycle 9 Principal Transient Analysis Parameters, Siemens Power Corporation, May 1999. 9. Letter D. E. Garber (SPC) to R. J. Chin (CornEd), "Evaluation of Improved Technical Specification Scram Times at Dresden, LaSalle and Quad Cities Station,'

DEG:99:195, July 26, 1999..---fv 1%-8 08" lit o p i ary

...........ý46 o iim.. Fa I 6 IatdlI DEG:01:045 Attachment Page A-5 Table I Proposed ITS Scram Insertion Times The 0.20-second delay is considered a nominal value that cannot be verified by the plant Therefore, the transient analysis calculations are performed to bound a range of no delay (linear insertion from start signal to notch 45) to a delay value just before notch 45. This is consistent with the information provided in Reference

8.

r l l r-ir Ililo, Inc. rropnetary DEG:01:045 Attachment Page A-6 Table 2 Base Case Transient Analysis Results With Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity Peak Peak Power ATRIUM-9B ATRIUM-9B GE9 Neutron Flux Heat Flux /Flow ACPR LHGRFACp ACPR (% rated) (% rated) LRNB FWCF The analysis results presented are from an exposure prior to EOC. The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.

I $s5areOLW811 PMIrE riULitaeaU, liet. rroprietary DEG:01:045 Attachment Page A-7 Table 3 EOOS Transient Analysis Results With Proposed-ITS Scram Times and Corrected Fuel Thermal Conductivity Slow TCV Closure 100 / 105" 1 TCV closing in 2.0 seconds 0.424 100 / 105* 1 TCV closing in 2.7 seconds 0.422 80 / 57.2* 1 TCV closing in 2.0 seconds 0.530 80 / 1 0 5 t 2 TCV closing in 2.0 seconds 0.540 80 / 57.2t 2 TCV closing in 2.0 seconds 0.560 25 / 105t 1 TCV closing in 2.0 seconds 1.007: LRNB No RPT FWCF With FHOOS 25/105 NA 1.9*0.6640 1.202* FWCF No RPT With FHOOS 25/105 INA 1.108t 0.6W0 1. 130*Scram initiated by high neutron flux. Scram initiated by high dome pressure.

The analysis results presented are from an exposure prior to EOC. The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.

rg-a011nLUmul/-tlr rllimrill inc. rroprietary DEG:01:045 2.7 26 2.5 2.4 2Z3' 2.2 21! 2.1O= z,, a- 1.95 Attachment Page A-8 0 500 1000 1500 2000 2500 3000 3500 4000 Powr MV%)Power MCPRp (%) Limit 100 1.46 80 1.51 60 1.56 25 2.05 25 2.20 0 2.70 Figure I EOC Base Case Power-Dependent MCPR Limits for ATRIUM-9B Fuel With Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity S...... r a up. ftasy DEG:01:045 2.7: 2.35 2Z25 2.15 2.05 9I.95 e, Attachment Page A-9 0 5M0 1000 1500 2000 2500 3000 3500 400M Powr (Mth)Power MCPRp (%) Limit 100 1.50 80 1.53 60 1.57 25 2.10 25 2.20 0 2.70 Figure 2 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel With Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity riamlldIUIUIle Rivr rlcnlIanlO, InC. -roprietary DEG:01:045 1.30 1.20 1.10 CL 1.00 U-' 0.90, Attachment Page A-10 0.60 0 500 1000 1500 2000 2500 3000 3500 .. Pow(" .ft)Power LHGRFACp (%) Multiplier 100 1.00 80 0.98 60 0.94 25 0.67 25 0.67 0 0.67 4000:.Figure 3 EOC Base Case Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel With Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity rramaxome ANI- Inc. Proprietary DEG:01:045 2.95 2.85 2.75 .65 2.55 2.45 2.35 2.25 2.15 2.05-Attachment Page A-11 0 500 1000 1500 2000 2500 3000 3500 4000 Power M )Power MCPRp (%) Limit 100 1.54 80 1.64 80 1.67 25 2.35 25 2.35 0 2.85 Figure 4 EOC Slow TCV Closure/FHOOS and/or No RPT Power-Dependent MCPR Limits for ATRIUM-9B Fuel With Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity asLIII; matIiIsdiU, InC. rroprletary DEG:01:045 2- W 2.7.= 2.65 255 2.45 2.35 2.25 06 2.15 S2.05 UE Attachment Page A-12 0 500 1000 1500 2000 2500 3000 3500 PMWM jO Power MCPRp (%) Limit 100 1.58 80 1.69 80 1.71 25 2.35 25 2.35 0 2.85 4000 Figure 5 EOC Slow TCV ClosurelFHOOS and/or No RPT Power-Dependent MCPR Limits for GE9 Fuel With Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity

---

, **DEG:01 :045 1 1. 1. 0 Attachment Page A-13 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWQ Power LHGRFAC, (%) Multiplier 100 0.86 80 0.86 80 0.86 25 0.64 25 0.64 0 0.64 Figure 6 EOC Slow TCV ClosureiFHOOS andlor No RPT Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel With Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity Technical Requirements Manual -Appendix I LI C9 Reload Transient Analysis Results Attachment 7 LaSalle Unit 1 Cycle 9 Operating Limits For Proposed Cycle Extension LaSalle Unit 1 Cycle 9 November 2001 Framatome ANP, Inc. Proprietary

/FRAMATOME AR"LP September 21, 2001 DEG:01:148 Mr. F. W. Trikur Exelon Nuclear Nuclear Fuel Management 4300 Winfield Road Warrenville, IL 60555

Dear Mr. Trikur:

LaSalle Unit I Cycle 9 Operating Limits for Proposed Cycle Extension Ref: 1: Contract for Fuel Fabrication and Related Components and Services dated as of October 24, 2000 between Siemens Power Corporation and Commonwealth Edison Company for LaSalle Nuclear Plant. Exelon has proposed operating LaSalle Unit 1 Cycle 9 beyond the currently licensed exposure of 18,477 MWd/MTU. The attachment provides the operating limits to support the planned cycle extension.

Very truly yours, David Garber Manager, Customer Projects Enclosures Framatome ANP, Inc.2101 Horn Rapids Road Richland, WA 99352 Tel: (509) 375-8100 Fax: (509) 375-8402 Framatome ANP, Inc. Proprietary DEG:01:148 Attachment Page A-1 LaSalle Unit 1 Cycle 9 Operating Limits for Proposed Cycle Extension Exelon has informed FRA-ANP of plans to extend LaSalle Unit 1 Cycle 9 (L1 C9) beyond the current licensing core exposure of 29,439 MWd/MTU (page 4-2 of Reference 1, corresponding to a cycle exposure of 18,477 MWd/MTU) by implementing a combined FFTR/coastdown.

Exelon has requested that FRA-ANP provide operating limits for base case Technical Specification scram speed (TSSS) and slow TCV closure and/or no RPT operation for the cycle extension.

This letter report summarizes the transient analysis results and operating limits required to support the LI C9 cycle extension.

Cycle Extension LIC9 was originally licensed to a cycle exposure of 18,477 MWd/MTU. The data provided-in Reference 2 indicates the LI C9 full-power capability is projected to continue to a cycle exposure of 18,800 MWd/MTU with a final coastdown exposure of 19,600 MWd/MTU using a coastdown rate of 14.9% power/1000 MWd/MTU. Per discussions with Exelon, the L1C9 coastdown will include a final feedwater temperature reduction (FFTR) of 1 00°F. The approach used to model the Ll C9 cycle extension is consistent with the L2C9 FFTR/coastdown extension described in Item II.A of Reference

3. FRA-ANP began with the latest projection-to the licensing EOC exposure of 18,477 MWd/MTU which includes core follow data to a cycle exposure of 11,564.3 MWd/MTU (References 4 and 5). The cycle was increased by 24 EFPD to account for the full-power capability extension due to the FFTR which corresponds to a cycle exposure of 19,100 MWd/MTU. Operation was then assumed to continue at a coastdown rate of 10% power/1,000 MWd/MTU. In order to protect a 10% power increase due to a Xenon transient, an additional 1,000 MWd/MTU of full power capability is included.

Based on this approach, LIC9 is conservatively modeled to operate at rated power to a cycle exposure of 20,100 MWd/MTU.

Operating Limits Reference 6 provided Li C9 EOC (18,477 MWd/MTU) operating limits for base case TSSS and slow TCV closure/FHOOS and/or no RPT scenarios to support the implementation of Improved Technical Specifications (ITS) and to correct an error in the fuel thermal conductivity.

Tables 2 and 3 of Reference 6 list the transient analyses required to support the LIC9 EOC limits. A similar set of analyses is required to establish the L1C9 combined FFTRPcoastdown limits. Analyses are only Framatome ANP, Inc. Proprietary DEG:01:148 Attachment Page A-2 required at 105% of core flow to support the combined FFTR/coastdown, consistent with the L2C9 analysis approach presented in Table 4 of Reference

3. FFTR/coastdown analyses at 105% flow protect operation for all flows within the power/flow map provided in Figure 1.1 of Reference

7. In general, performing analyses at higher exposures produces higher results. As a result, analyses performed at coastdown exposures tend to be more conservative than those performed at EOC. The LIC9 extension includes a 100°F temperature reduction to extend the full-power capability by 24 EFPD, and therefore, FFTRlcoastdown exposures are higher than standard coastdown exposures.

LRNB analyses tend to be more conservative for high feedwater temperatures (FWT) while low FWT produce higher results for FWCF analyses.

It is obvious that FWCF analyses performed at the FFTR FWT and an FFTR/coastdown exposure bound all operation during the FFTR/coastdown.

However, it is unclear if the combination of FFTR FWT and an FFTR/coastdown exposure would produce more conservative results than the upper bound FWT and a coastdown exposure-for LRNB analyses.

Therefore, in order to protect any operating scenario during FFTR/coastdown, the LRNB analyses were performed with the upper bound FWT at the FFTR/coastdown exposure.

All L1C9 FFTR/coastdown analyses were performed at a cycle exposure of 20,100 MWd/MTU. The transient analyses were performed with the ITS scram times shown in Table 1 of Reference 6 and include the correct fuel thermal conductivity.

Table 1 presents the base case TSSS analysis results for the combined FFTR/coastdown.

Figures 1 and 2 present the base case TSSS MCPRp limits for the ATRIUM-9B and GE9 fuel, respectively.

The sum of the Ll C9 SLMCPR of 1.11 and the ACPR results from Table 1 are also presented in Figures 1 and 2. The FFTR/coastdown base case TSSS LHGRFACp multipliers and the LHGRFACp results from Table 2 are presented in Figure 3. Table 2 presents the slow TCV closure and no RPT analysis results for the combined FFTR /coastdown.

Figures 4 and 5 present the slow TCV closure and/or no RPT MCPRp limits for the ATRIUM-9B and GE9 fuel, respectively.

The sum of the Ll C9 SLMCPR of 1.11 and the ACPR results from Table 2 are also presented in Figures 4 and 5. The FFTRlcoastdown slow TCV closure and/or no RPT LHGRFACp multipliers and the LHGRFACp results from Table 2 are presented in Figure 6.

Framatome ANP, Inc. Proprietary DEG:01:148 Attachment Page A-3 Licensing Applicability Reference 1 summarizes the LI C9 licensing analyses and limits for which FRA-ANP was responsible to a cycle exposure of 18,477 MWd/MTU. Licensing analyses performed by Exelon in support of LlC9 are presented elsewhere.

In addition to the analyses listed in Tables 1 and 2, FRA-ANP has performed evaluations to determine the applicability of the Reference 1 analysis results and limits to the LI C9 cycle extension.

The evaluations demonstrated that the Reference 1 licensing analysis results and limits remain applicable for the L1C9 cycle extension with the exception of the MCPRp _limits-and LHGlRFACpmultipliers provided-in.Figures_

1 through_6.

Reference 2 describes the planned LIC9 FFTRlcoastdown as 14.9% power/1,000 MWd/MTU beginning at a cycle exposure of 18,800 MWd/MTU. The L1C9 operating limits provided in References 6 and 8 remain applicable to a cycle exposure of 18,477 MWd/MTU (core exposure of 29,439 MWd/MTU).

The MCPRp limits and LHGRFACp multipliers presented in Figures 1 through 6 must be used for operation beyond a cycle exposure of 18,477 MWd/MTU. In the event that the actual operation deviates significantly from the planned FFTR/coastdown,.the following requirements must be met in order satisfy the coastdown analysis assumptions:

Coastdown operation must begin prior to a cycle exposure of 19,100 MWd/MTU.

Thermal power during FFTR/coastdown operation must be reduced at a rate faster than 10% power/1,000 MWd/MTU The limits and multipliers presented in Figures 1 through 6 are applicable to a cycle exposure of 20,100 MWd/MTU. The MCPRp limits and LHGRFACp multipliers are valid for any feedwater temperature within the bounds defined in Reference 7, Item 3.12. Comparison of the Cycle 9 FFTRlcoastdown nodal power histories for the rated power pressurization transients with the approved bounding curves to show compliance with the 1% clad strain and centerline melt criteria for GE9 fuel is discussed in Reference

9.

Framatome ANP, Inc. Proprietary DEG:01:148 Attachment Page A-4 References

1. EMF-2276 Revision 1, LaSalle Unit I Cycle 9 Reload Analysis, Siemens Power Corporation, October 1999. 2. Exelon TODI NFM0100051, "LaSalle Unit 1 Cycle 10 Final Licensing Loading Plan (FLLP)," September 11, 2001. 3. Letter, D. E. Garber (SPC) to R. J. Chin (ComEd), "LaSalle Unit 2. Cycle 9 Post Analysis Calculation Plan," DEG:00:231, October 20, 2000. 4. Letter, J. K. Wheeler (Exelon) to D. E. Garber (SPC), "LaSalle Unit 1 Cycle 9 Core Follow Data through October 7, 2000," NFMO100004, January 5, 2001. 5. Letter, J. T. Fisher (Exelon) to D. E. Garber (FRA-ANP), "LaSalle Unit 1 Cycle 9 Core Follow Data October 8, 2000 through February," NFMO100037, March 27, 2001. 6. Letter, D. E. Garber (FRA-ANP) to R. J. Chin (Exelon), "LaSalle Unit 1 Cycle 9 Operating Limits for Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity," DEG:01:045, March 22, 2001. 7. EMF-96-189 Revision 0, LaSalle Unit I Cycle 9 Principal Transient Analysis Parameters, Siemens Power Corporation, May 1999. 8. Letter, D. E.-Garber. (FRA-ANP-)

to R. J. Chin (Exelon), "LaSalle Unit 1 Cycle 9 NSS Base Case and TBVOOS or FHOOS Operating Limits for Proposed ITS Scram Times With Corrected Fuel Thermal Conductivity," DEG:01:074, May 15, 2001. 9. Letter, D. E. Garber (FRA-ANP) to F. W. Trikur (Exelon), "LaSalle Unit I Cycle 9 GE9 Mechanical Limits for Proposed Cycle Extension," DEG:01:143, September 18, 2001.

Framatome ANP, Inc. Proprietary DEG:01:148 Attachment Page A-5 Table 1 Base Case TSSS FFTR/Coastdown Transient Analysis Results Power ATRIUM-9B ATRIUM-9B GE9 (% rated) ACPR LHGRFACp ACPR LRNB 100 0.35 0.93 0.39* 80 0.39* 0.97 0.42* FWCF 100 0.31 1.01* 0.34*0.38 0:98 0.39*60 050* 0.91* 049* 40 0.64 0.88* 0.60 25 1.20* 0.66* 1.21** The analysis results presented are from an exposure prior to 20,100 MWd/MTU. The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.AA Framatome ANP, Inc. Proprietary Attachment Page A-6 Table 2 EOOS FFTR/Coastdown Transient Analysis Results Power Slow Valve ATRIUM-9B ATRIUM-9B GE9 (% rated) Characteristics ACPR LHGRFACp ACPR Slow TCV Closure 100* 1 TCV closing in 2.0 seconds 0.46 0.83 0.50 100* 1 TCV closing in 2.7 seconds 0.46 0.83 0.50 80* 1 TCV closing in 2.0 seconds 0.53t 0.87 0.58t 80* 1 TCV closing in 2.0 seconds 0.58 0.85 0.60t 25* 1 TCV closing in 2.0 seconds 1.01t 0.70' 1.01t LRNB No RPT 100 0.42 0 .0.46 FWCF No RPT With FHOOS 25 1.11t 0.66t 1.13t Scram initiated by high neutron flux. The analysis results presented are from an exposure prior to 20,100 MWd/MTU. The ACPR and LHGRFACp results are conservatively used to establish the thermal limits. Scram initiated by high dome pressure.t *DEG:01:148 Framatome ANP, Inc. Proprietary DEG:01:148 2.95 2.85 2.75 2.65 2.55 2.45 2.35 2.25 2.15 2.05 1.95 1.75 1.45 1.15 Attachment Page A-7* LRNB

• FWCF -OLMCPR a 0 10 20 30 40 50 60 70 80 90 100 110 Power (% rated)Power MCPRp (%) Limit 100 1.46 60 1.62 25 2.35 25 2.35 0 2.85 Figure 1 FFTRlCoastdown Base Case Power-Dependent MCPR Limits for ATRIUM-9B Fuel a. i. IL, Framatome ANP, Inc. Proprietary DEG:01:148 2.95 285 2.75 2.65* 2.55* 2.45 2.35 2.25 2,15 " 2.05 C.) 1.95 1.85 1.75 1.65 1.55 1.45 1.35 Attachment Page A-8 0 10 20 30 40 50 60 70 80 90 100 110 Power (0/ rated)Power MCPRp (%) Limit 100 1.50 60 1.62 25 2.35 25 2.35 0 2.85 Figure 2 FFTR/Coastdown Base Case Power-Dependent MCPR Limits for GE9 Fuel Framatome ANP, Inc. Proprietary DEG:01:148 1.30 1.25 1.20 OL 1.00 U. S0.95 -J 0.90 0.85 0.80 0.75 0.70 Attachment Page A-9* FW -LHI U CF GRFACp a a 10 20 30 40 50 60 Power (% rated)70 80 90 100 .110 Power LHGRFACp (%) Multiplier 100 0.93 60 0.91 25 0.64 25 0.64 0 0.64 Figure 3 FFTR/Coastdown Base Case Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel 0.65 0.60 Framatome ANP, Inc. Proprietary Attachment Page A-10 DEG:01:148

• Slow TCV Closure

• LRNB No RPT

• FVCF No RPT -OLMCPR U S 0 10 20 30 40 50 60 70 80 90 100 110 -Power (% rated)Power MCPRp (%) Limit 100 1.57 80 1.64 80 1.69 25 2.35 25 2.35 0 2.85 Figure 4 FFTR/Coastdown Slow TCV Closure andlor No RPT Power-Dependent MCPR Limits for ATRIUM-9B Fuel 2.85 2.75 2.65 2.55 2.45 2.05 1.95 1.85 1.75 1.65 1.55 1.45 1.35 1.25 Framatome ANP, Inc. Proprietary DEG:01:148

2. 2 2 2 2 ..' 0j Attachment Page A-1I.85.75.65 L55 .45 2.35 2-25 1.85 1.75 1.65 1.55 1.45 1.35 1.25* Saw TCV Closure

• LRNB No RPT

• FVCF No RPT -OLMCPR U 0 10 20 30 40 50 60 70 80 90 100 110 Power (* rated)Power (%) 100 80 80 25 25 0 MCPRp Limit 1.61 1.69 1.71 2.35 2.35 2.85 Figure 5 FFTR/Coastdown Slow TCV Closure and/or No RPT Power-Dependent MCPR Limits for GE9 Fuel 1. 1;)

Framatome ANP, Inc. Proprietary Attachment Page A-12 DEG:01:148

• Slow TCV Closure

• LRNB No RPT

• FWCF No RPT -LHGRFACp UbU --,--.-------.-----,--------

-'0 10 20 30 40 50 60 70 80 90 100 110 --Power(% rated)Power LHGRFACp (%) Multiplier 100 0.83 80 0.83 80 0.83 25 0.64 25 0.64 0 0.64 Figure 6 FFTR/Coastdown Slow TCV Closure andlor No RPT Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel 1.25 1.20 1.15 1.10 1.05 0. 1.00, S0.95 (9 -I-' 0.90--A-0.85 0.80. 0.75 0.70 0.65