ML013600520
| ML013600520 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 11/29/2001 |
| From: | Pardee C Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| -RFPFR | |
| Download: ML013600520 (217) | |
Text
Exelon,M Nuclear Exelon Generation Company, LLC www.exeloncorp .co 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 Affected Summary of Changes Date Affected Section Pages All Original Issue (Cycle 9) 10/99 All All All Incorporated administrative changes (including updating the 11/99 date to be November 1999)
All Incorporated changes to thermal limits due to uprate and 5/00 All 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 Incorporates coastdown thermal limits on tables 2-1, 2-2, and 11/01 All 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 LaSalle Unit 1 Cycle 9 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. Minim 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 ii November 2001
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 iii November 2001
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 XN Power Methodology for Boiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing Effects/NRC Correspondence, NF-524(P)(A) Revision 2 and Supplement 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation November 1990.
I
- 30. COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis, ANF-913(P)(A), Volume 1, Revision 1 and Volume Supplements 2, 3, and 4, Advanced Nuclear Fuels Corporation, August 1990.
1; and
- 31. HUXY: A Generalized Multirod Heatup Code with 10CFR50, Appendix K Heatup Option, ANF-CC-33(PXA), Supplement 1 Revision 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.
Company,
- 35. .Generic- Mechanical -Desigrr-for- Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-85-67(P)(A) Revision 1, Exxon Nuclear September 1986.
Reload
- 36. Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9X9-IX and 9X9-9X BWR 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 July for BWR Stability Analysis in the Frequency Domain, Code Qualification Report, EMF-CC-074(P)(A), Siemens Power Corporation, 1994.
Nuclear
- 38. RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, XN-NF-81-58(P)(A), Revision 2 Supplements 1 and 2, Exxon Company, March 1984.
1
- 39. XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis, XN-NF-84-105(P)(A), Volume 1 and Volume 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.
Volume 1 and
- 41. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, XN-NF-80-19(P)(A)
Supplements 1 and 2, Exxon Nuclear Company, Richland, WA 99352, March 1983.
3, Exxon
- 42. Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF-79-71(P)(A), Revision 2 Supplements 1, 2, and Nuclear Company, March 1986.
Nuclear
- 43. Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A), Revision 1 and Revision 1 Supplement 1, Advanced Fuels Corporation, May 1995.
- 44. Reference Deleted.
LaSalle Unit I Cycle 9 iv November 2001
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 V November 2001
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 Fuel Type Cycle First (Reference 3) Inserted MAPLHGR Limits 1.2-1 GE9B-P8CWB322-11GZ-100M-150-T 7 1.2-2 GE9B-P8CWB320-9GZ-1 OOM-1 50-T 7 1.2-3 GE9B-P8CWB343-12GZ-80M-150-T 8 1.2-4 GE9B-P8CWB342-1 OGZ-80M-1 50-T 8 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-10OM 9 Planar Average Exposure MAPLHGR (kWlft)
(GWdIMTU) (all FANP fuel types)
(References 4 and 7) 0.0 13.5 20.0 13.5 61.1 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 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 1GZ-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 1-2 November 2001
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 1-4 November 2001
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 1-5 November 2001
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 For 25>P 1 0.7 No Thermal Limits Monitoring Required; If Official Monitoring is Desired, the Equations 75 0.65 MAPA0 for _>25% Power .1000240P-0
=0May Be Extrapolated for 25 >
P, provided the Official monitoring is only 0.6 performed with the TCV/TSV closure scrams 0 0.55 and RPT enabled.AF~p=10
.-J For 100>P:
. 0.5 For 25:< P < 100 S0.45 MAPFACp = 1.0+0.005224 (P-100) 0
-J 0.4 U/)
- 0.35 For 100 < P, MAPFACp = 1.00 "G) P = % Rated Core Thermal Power
, 0.3 d) o 0.25 0.2 o 0.15 0.
0.1 0.05 0
45 50 55 60 65 70 75 80 85 90 95 100 25 30 35 40 Core Thermal Power (% Rated) 1-6 November 2001 LaSalle Unit 1 Cycle 9
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) 1 L.
0.9 105% Maximum Attainable Core Flow
- 000 0-For 0~
0.8 000000 00MAPFACF = The Minimum of EITHER 1.0 L
0 C,,"OR (0.6807 x (WT/100)+0.4672}
0 0.7
.6j U) WT = % Rated Core Flow 0.6 CL 0.
0 0.5 MJ 0.4 0.3 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 Core Flow (% Rated)
LaSalle Unit 1 Cycle 9 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:
2.2
Description:
is to be Prior to initial scram time testing for an operating cycle, the MCPR operating limit Specification Scram Times. For Technical Specification based on the Technical requirements refer to Technical Specification table 3.1.4-1.
MWd/MTU to TIP symmetry Chi-squared testing shall be performed prior to reaching 500 validate the MCPR calculation.
exposure of MCPR limits from BOC to Coastdown are applicable up to a core average 29,439 MWd/MTU (which is the licensing basis exposure used by Framatome-ANP).
(Reference 4) of 29,439 MCPR limits from Coastdown to EOL are applicable from a core average exposure 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 flow determined from 2.2.1.1 or 2.2.1.2, whichever is greater at any given power, 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 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.
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.
limits should be applied.
- For thermal limit monitoring cases at greater than 100%P, the 100% power MCPRp 2-1 November 2001 LaSalle Unit 1 Cycle 9
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.
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 2-2 November 2001
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*
0 25 25 (25.1) 60 80 80 (80.1) 100 EOOS Combination 2.70 2.20 2.05 1.56 1.51 1.46 NoEOOS 2.71 2.21 2.06 1.57 1.52 1.47 Single RR Loop only 2.85 2.35 2.35 1.67 1.64 1.54 EOOS**
2.86 2.36 2.36 1.68 1.65 1.55 EOOS**/Single RR Loop 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*
0 25 25(25.1) 60 80 80 (80.1) 100 EOOS Combination 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 2.89 2.39 2.39 1.71 1.66 1.56 EOOS*
2.90 2.40 2.40 1.72 1.67 1.57 EOOS**/Single RR Loop 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 2.85 2.35 2.35 1.62 1.46 No EOOSt 2.86 2.36 2.36 1.63 1.47 Single RR Loop Only 2.85 2.35 2.35 1.69 1.64 1.57 EOOS**
2.86 2.36 2.36 1.70 1.65 1.58 EOOS/Single RR Loop**
Framatome-ANP fuel that is fuel type 36 in 10B cell locations Percent Core Thermal Power*
0 25 25 (25.1) 60 80 80 (80.1) 100 EOOS Combination 2.87 2.37 2.37 1.64 1.48 No EOOSt 2.88 2.38 2.38 1.65 1.49 Single RR Loop only EOOS* 2.87 2.37 2.37 1.71 1.66 1.59 2.88 2.38 2.38 1.72 1.67 1.60 EOOS*/Single RR Loop
- 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.
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.
LaSalle Unit I Cycle 9 2-3 November 2001
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 all EOOS scenarios.
The MCPRf limits are applicable from BOC through coastdown and in 2-4 November 2001 LaSalle Unit 1 Cycle 9
Technical Requirements Manual - Appendix I L1C9 Core Operating Limits Report
- 3. Linear Heat Generation Rate (3.2.3) 3.1 Tech Spec
Reference:
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 3-1 November 2001
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 3-2 November 2001
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)
I 0.95 0.9 0.85 a.
0.8 0.75
-J
".W 0.7 C,
0.65 0.6
,1 0.55 0.5
-J 0.45 C,
4, 0.4
"(D 0 0.35 CL a) 0.3 0.25
- 0. 0.2 0.15 0.1 0.05 (1
65 70 75 80 85 90 95 100 25 30 35 40 45 50 55 60 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) 10.9 0.95 0.
0.98 U 0.85 0.8
_ 0.75 C,
CL 0.7 0 0.65 0.6 C
0.55 a) 0.5 0.
-r 0.45 0.4 0.35 0., 0.3 a) 0 0.25 o 0.2 o.
0.15 0.1 0.05 0
60 70 80 90 100 0 10 20 30 40 50 Core Thermal Power (% Rated) 3-4 November 2001 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 For 105% Maximum Attainable Core Flow 0 0.8 LHGRFACF = The Minimum of EITHER 1.0 0 OR {0.6807 x (WT/100)+0.4672}
.2.
- . 0.7 WT = % Rated Core Flow CD For Abnormal Idle Loop Startup, LHGRFACF 0.40 05 0.6 CL 0.5 SI I V 0
u.
I I 0.4 I I I It LL 0.3 85 90 95 100 105 30 35 40 45 50 55 60 65 70 75 80 Core Flow (% Rated)
LaSalle Unit I Cycle 9 3-5 November 2001
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 Percent Core Thermal Power*
[EOOS Combination 0 25 25 60 80 80 100 0.67 0.94 0.98 1.00 No EOOS 0,67 0.67 Single RR Loop only 0.67 0.67 0.67 0.94 0.98 1.00 EOOS** 0.64 0.64 0.86 0.86 0.86 0.64 EOOS/Single RR Loop-* 0.64 0.64 0.64 0.86 0.86 0.86 I I
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 3-6 November 2001
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 TRIP SETPOINT ALLOWABLE VALUE Two Recirculation Loop 0.66 W + 51%** 0.66 W + 54%**
Operation*
Single Recirculation Loop 0.66 W + 45.7%** 0.66 W + 48.7%**
Operation*
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 Coastdown9 None Yes Yes Yes Yes Feedwater Heaters 2 (Reference 9 and 56) Yes No 3 Yes Yes Single RR Loop1° (Reference 9 and 56) Yes No 8 N/A Yes Turbine Bypass Valves (Reference 9) Yes Yes Yes No EOC Recirculation Pump Trip (Reference 9and 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 No3 Yes Yes Feedwater Heaters 2 (Reference 15, 20,21, and 56)
Turbine Bypass Valves / Feedwater Heaters 2 (Reference 9) No No No5 No EOC Recirculation Pump Trip / Yes 4 No3 Yes 4 Yes Feedwater Heaters 2 (Reference 9 and 56)
TCV Stuck Closed6 (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 5-1 November 2001
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 6-1 November 2001
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 Preparer Document Exelon Neutronics Licensing Report 1
Siemens Power Corporation Reload Analysis 2
Siemens Power Corporation Plant Transient Analysis (Excerpts) 3 4 General Electric ARTS Improvement Program Analysis, Supplement I (Excerpts)
General Electric TCV Slow Closure Analysis 5
(Excerpts)
Framatome ANP LaSalle Unit 1 Cycle 9 Operating 6
Limits for Proposed Scram Times and Corrected Fuel Thermal Conductivity Framatome ANP LaSalle Unit I Cycle 9 Operating 7
Limits for Proposed Cycle Extension November 2001 LaSalle Unit 1 Cycle 9
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 3 1.1 New Reload Fuel Assembly Nuclear Design 3 3
I.L1 Assembly Average Enrichment 3
1.1.2 Axial Enrichment and Burnable Poison Distribution
. .3 -_RadialEnrichment and Burnable Poison-Distribution 1.2 Core Nuclear Design 4 1.2.1 Core Configuration and Licensing Exposure Limits 4 1.2.2 Core Reactivity Characteristics 5 II. Control Rod Withdrawal Error 5 I11. Fuel Loading Error 6 IV. Control Rod Drop Accident 6 V. Loss of Feedwater Heating 7 VI. Maximum Exposure Limit Compliance 7 VII. Spent Fuel Pool and Fresh Fuel Vault Criticality Compliance 8 VII.1 Fresh Fuel Vault Criticality Compliance 8 VII.2 L1 Spent Fuel Pool Criticality Compliance 8 VII.3 L2 Spent Fuel Pool Criticality Compliance 9 VIII. References 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 19A 3.93 SPCA9-396B-12GZB-1 OOM 19B 3.96 SPCA9-396B-12GZC-100M 19C 3.96 SPCA9-384B-I IGZ6-80M 28B 3.84 1.1.2 Axial Enrichment and Burnable Poison Distribution Assembly Name Batch Identifier Fi2ure SPCA9-393B-16GZ-I OOM I9A SPCA9-396B-12GZB-100M 19B SPCA9-396B-12GZC-100M 19C SPCA9-384B- 1, GZ6-80M 28B 2 1.1.3 Radial Enrichment and Burnable Poison Distribution Lattice Name Batch Found In Fieure SPCA9-4.56L- 12G8.0/4G3.0-100M 19A 3 SPCA9-4.56L-12G8.0-l00M 19A 4 SPCA9-3.9 IL- 12G8.0-100M 19A 5 SPCA9-3.90L-8G5.0-I00M 19A 6 SPCA9-4.59L- 12G8.0-1 OOM 19B 7 SPCA9-4.59L- 12G7.0-1 OM 19B 8 SPCA9-3.96L-8G7.0/4G8.0-0I OM 19B 9 SPCA9-3.96L-8G5.0- 100M 19B and 19C 10 SPCA9-4.58L-8G6.0/4G3.0-100M 19C I1 SPCA9-4.58L-8G6.0-1OOM 19C 12 SPCA9-4.06L- 11G6.0-80M 28B 13 SPCA9-4.34L- I0G6.0-80M 28B 14 11&J goI4/O
1.2 Core Nuclear Design 1.2.1 Core Configuration and Licensing Exposure Limits Cycle Number Loaded in Core Assembly Name 7 56 GE9B-P8CWB322-1 IGZ-100M- I50-CECO 7 89 GE9B-P8CWB320-9GZ-1 O0M-I 50-CECO 8 104 GE9B-PSCWB343-12GZ-80M-150-CECO 8 .143 GE9B-P8CWB342-lOGZ-80M-150-CECO 9 36 SPCA9-384B- 1IGZ6-80M 9 208 SPCA9-393B-16GZ- 10OM 9 88 SPCA9-396B- 12GZB- IOOM 9 40 SPCA9-396B-12GZC- IOOM Core Core Average Incremental Exposure Exposure Exposure at EOC 8 (Cycle N-i) 27966.9 12511.0 Nominal EOC 8 (MWDIMT) 27455.9 12000.0 Short EOC 8 (MWD/MT) [for shutdown consideration)
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).
Core Average Exposure Exposure at BOC 9 (Cycle N)
With Nominal EOC 8 (MWD/MT) 10961.0 With Short EOC 8 (MWD/MT) 10634.9
-4 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.
'1VO toll jqj 101-7/99
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 GE9B Event ACPR ACPR Loss of Feedwater Heating 0.19 0.18 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,/t-h-o hiyj
I 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- IOOM 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- IOOM 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 i
SPCA9-4.06L- I IG6.0-80M 1.213 < 0.85 0.95 SPCA9-4.34L- I0G6.0-80M 1.227 < 0.86 0.95
- From 68 *F, uncontrollcd CASMO-3G results.
From Fiigure 6.1 or"Reference 8.
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 ATRIUMTM -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 48-00 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.
- 7 bjZ
BATCH BATCH BATCH 19A 19B 19C Natural Uranium I1 " Natural Uranium II," Natural Uranium III" SPCA9-3.90L SPCA9-3.96L 8G5.0-100M 12" 8G5.0-1 0OM 12" SPCA9 SPCA9 SPCA9 3.96L-SG5.0 42" 3.91L-12G8.0 30" 3.96L-8G7.0/4G8.0 30" -looM
-looM -loom SPCA9 SPCA9 4.59L- 12G7.0 24" 4.58L-8G6.0 24r
-looM -looM SPCA9 4.56L-12G8.0 72"
-looM SPCA9 SPCA9 4.59L-12GS.0 66" 4.58L-SG6.0/4G3.0 66"
-looM -looM SPCA9-4.56L 12GS.0/4G3.0 18"
-looM Natural Uranium Natural Uranium Natural Uranium 6" SPCA9-393B- 16GZ-1OOM SPCA9-396B- 12GZB- IOOM SPCA9-396B-12GZC- IOOM Figure 1 LlC9 ATRIUM-9B Assembly Axial Designs (1OOM Channels)
NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 13 of 27 BATCH 28B Natural Uranium 411" SPCA9 4.06L-1 1G6.0 "42"
-80M SPCA9 4.34L-10G6.0
-80M 90" Natural Uranium 6"1 SPCA9-384B-11GZ6-80M Figure 2 LIC9 ATRIUM-9B Assembly Axial Designs (80M Channels)
- TR- C/,I11
I Rods (4) 3.00 w/o U-235 2 Rods (8) 4.00 w/o U-235 3 Rods (8) 4.70 wlo U-235 4 Rods (36) 4.95 w/o U-235 GI Rods (4) 4.70 w/o U-235+8.0 wh/o Gd203 G2 Rods (8) 4.20 w/o U-235+8.0 w/o Gd203 G3 Rods (4) 4.00 w/o U-235+3.0 w/o Gd203 a
Figure 3 SPCA9-4.56L-12G8.0/4G3.0-1OOM (19A)
Enrichment Distribution
I Rods (4) 3.00 w/o U-235 2 Rods (12) 4.00 wv/o U-235 3 Rods (8) 4.70 w/o U-235 4 Rods (36) 4.95 w/o U-235 GI Rods (4) 4.70 w/o U-235+8.0 w/o Gd203 G2 Rods (8) 4.20 w/o U-235+8.0 w/o Gd203 Figure 4 SPCAg-4.56L-12G8.0-100M (19A)
Enrichment Distribution k
NFM ID# NFM9900149 NUCLEAR FUEL MANAGEMENT TRANSMITTAL OF DESIGN INFORMATION Seq. No. _
Page 16 of 27 I Rods (4) 2.60 w/o U-235 2 Rods (12) 3.40 w/o U-235 3 Rods (8) 3.80 w/o U-235 4 Rods (36) 4.40 w/o U-235 GI Rods (12) 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
? 4.40 4.40 4.40 4.4 4.044
- *'*- rn"l l -*.... . 4 0440.4
. 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) 2.60 w/o U-235 2 Rods (16) 3.40 w/o U-235 3 Rods (8) 3.80 wlo U-235 4 Rods (36) 4.40 w/o U-235 GI Rods (8) 3.40 w/o U-235+5.0 wo Gd203 Figure 6 SPCA9-3.90L-8G5.0-IOOM (19A)
Enrichment Distribution
I Rods (4) 3.00 w/o U-235 2 Rods (8) 4.00 w/o U-235 3 Rods (8) 4.70 w/o U-235 4 Rods (40) 4.95 w/o U-235 GI Rods (12) 4.20 w/o U-235+8.0 w/o Gd203 Figure 7 SPCA9-4.59L-12G8.0-100M (19B)
Enrichment Distribution
I Rods (4) 3.00 w/o U-235 2 Rods (8) 4.00 w/o U-235 3 Rods (8) 4,70 w/o U-235 4 Rods (40) 4.95 w/o U-235 Gi Rods (12) 4.20 w/o U-235+7.0 w/o Gd203 Figure 8 SPCAg-4.59L-12G7.0-100M (19B)
Enrichment Distribution go(ieI 4ILa)
NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 20 of 27 1 Rods (4) 2.60 w/o U-235 2 Rods (8) 3.40 w/o U-235 3 Rods (8) 3.80 w/o U-235 4 Rods (40) 4.40 w/o U-235 GI Rods (8) 3.40 wlo U-235+7.0 w/o Gd203 G2 Rods (4) 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) 2.60 sy/o U-235 2 Rods (12) 3.40 w/o U-235 3 Rods (8) 3.80 wv/oU-235 4 Rods (40) 4.40 w/o U-235 GI Rods (8) 3.40 w/o 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) 3.00 w/o U-235 2 Rods (8) 4.00 wlo U-235 3 Rods (8) 4.70 wlo U-235 4 Rods (40) 4.95 w/o U-235 GI Rods (8) 4.20 w/o U-235+6.0 w/o Gd203 G2 Rods (4) 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) 3.00 w/o U-235 2 Rods (12) 4.00 wlo U-235 3 Rods (8) 4.70 w/o U-235 4 Rods (40) 4.95 w/o U-235 GI Rods (8) 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 1 Rods (4) 2.72 w/o U-235 2 Rods ( 8) 3.53 w/o U-235 3 Rods (16) 3.94 wlo U-235 4 Rods (33) 4.53 w/o U-235 G Rods (11) 3.69 wlo U-235+6.0 w/o Gd203 Figure 13 SPCA9-4.06L-11G6.0-80M (28B)
Enrichment Distribution Io/q/
ell-
NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION Seq. No. I Page 25 of 27 I Rods (4) 2.72 w/o U-235 2 Rods (8) 3.78 w/o U-235 3 Rods (16) 4.19 w/o U-235 4 Rods (34) 4.78 wlo U-235 G Rods (10) 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 13000.0 MWd/MTU 1746.3 GWd Core Average Exposure 23961.4 MWd/MTU Delta E: MWd/MTU, (GWd) .0 ( .00 1 Axial Profile Power: MWt 3489.0 (100.00 %) N Power Exposure Core Pressure: psia 1020.1 Top 25 .153 3.961 Inlet Subcooling: Btu/ibm -18.28 24 .284 6.830 Flow: Mlb/hr 108.50 (100.00 %J 23 .663 16.284 22 .806 19.848 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 .874 21.821 59 59 .935 20 23.364 55 ---- 16 ......- 16---- 55 19 .978 24.366 51 0- 0-- 0-- 0 --- 51 18 1.006 25.256 47 .16 16 . -------.--- 16 ---- 47 17 1.010 26.344 43 ... 0 -- 8-0-- 0-- 8-- 0---- 43 16 1.011 2.7.653 39 16-- 39 -. 15 1. 63-- 28.409 35 0-- 0- 16 -- 16-- 0- 14 1.030 28.934 31 31
-- 13 1.056 29.323 27 0 -- 0- 16-- 16 -- 0 - 0 -- 27 12 1.087 29.530 23 -- 16 .. 1-.-- 16 -- 23 --- --. 11 1.128 29.908 19 0 -- 8 -
0-- 0-- 8-- 0 -- 19 10 1.177 30.365 15 -- 16-..-- 15 16 - 9 1.231 30.778*
11 - 0 -- 0-- 0 -- 0 8 1.279 30.556 11 7 --- 16 --- --- 16 -- 7 7 1.347 30.740 3 3 6 1.421 30.648 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.485* 29.817 4 1.477 27.607 Control Rod Density: % 20.54 3 1.308 23.608 2 .962 17.267 k-effective: 1.00388 Bottom 1 .279 4.864 Void Fraction: .448 Core Delta-P: psia 21.675 % AXIAL TILT -18.447 -10.750 Core Plate Delta-P: psia 17.213 AVG BOT 8ft/12ft 1.1042 1.0860 Coolant Temp: Deg-F 545.8 In Channel Flow: Mlb/hr 93.26 Active Channel Flow: Nlb/hr 93.26 Source Convergence .00008 Figure 15 Initial RWE Rod Pattern for Limiting ATRIUM-9B Case Error Rod is 34-43
~,io
NUCLEAR FUEL MANAGEMENT NFM ID# NFM9900149 TRANSMITTAL OF DESIGN INFORMATION 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) .0 1 .00 ) Axial Profile Power: MWt 3489.0 (100.00 %) N Power Exposure Core Pressure: psia 1020.1 Top 25 .134 1.558 Inlet Subcooling: Btu/ibm -18.28 24 .242 2.748 Flow: Mlb/hr 108.50 (100.00 %) 23 .614 6.567 22 .753 8.305 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 21 .813 9.766 59 59 20 .885 10.769 55 20 -- 0 -- 20 -- 55 19 .947 11.473 51 51 18 1.008 12.074 47 -.. .. 12 - 0 -- 24 -- 0 -- 12 47 17 1.084 12.760 43 43 -- 16 1.209 13.155 39 -- 20.-- 0- 24 -- 0-- 20 -- ,39 00--24_-- 15 1.271. 13.418
... .* *. . . . .- 35 -..
35 14 1.290 13.606 31 0 -- 24- 0 -- 2A.-- 0 -- 24-- 0 -- 31 13 1.288 13.747 27 27 -- 12 1.256 13.859 23 -- 20-- 0- 24 -- 0 -- 24-- 0 -- 20 -- 23 11 1.254 13.965 19 - 19 -. 10 1.273 14.034*
15 12 - 0 -- 24-- 0 -- 12... 15 9 1.304 14.008 11 8 1.326 13.527 7 20 -- 0 -- 20 .-- 7 7 1.357 13.424 3 3 6 1.365* 13.358 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 5 1.316 13.185 4 1.164 12.622 Control Rod Density: 15.23 3 .968 11.120 2 .693 8.223 k-effective: 1.00250 Bottom 1 .187 2.308 Void Fraction: .430 Core Delta-P: psia 21.559 AXIAL TILT -13.547 -10.296 Core Plate Delta-P: psia 17.096 AVG BOT 8ft/12ft 1.1172 1.0898 Coolant Temp: Deg-F 545.5 In Channel Flow: Mlb/hr 93.36 Active Channel Flow: mlb/hr*. 93.36 Source Convergence .00007 Figure 16 Initial RWE Rod Pattern for Limiting GE9B Case Error Rod is 30-39
Technical Requirements Manual - Appendix I LI C9 Reload Transient Analysis Results Attachment 2 LaSalle Unit 1 Cycle 9 Reload Analysis November 2001 LaSalle Unit 1 Cycle 9
IjiJT- U dm i c\oo'i SIEMENS EMF-2276 Revision 1 LaSalle Unit I Cycle 9 Reload Analysis October 1999 Siemens Power Corporation Nuclear Division
ISSUED INSPC ON-LINE Siemens Power Corporation DOCUMENT SYSTE*L-DATE: / LZq EMF-22761 Revision LaSalle Unit I Cycle 9 Reload Analysis Prepared:
/&/- 7/fr f_.
Date Haun. Engineer Neutronics Prepared:
i0- 6 Date D. G. Carr, Team Leader BWR Safety Analysis Concurred:
Date H. D. Curet, Manager Product Licensing Approved: Date
- 0. C. Brow , anager BWR Neutronics Approved: Date M. E. Garrett, Manager BWR Safety Analysis 1o/099 Approved: Date
.M. Howe, Manager Product Mechanical Engineering Approved: Date Denver, Manager Commercial Operations gdh Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 Page ii Ke\~auU "d y Nature of Changes Pmn*P PM Description and Justification If
- 1. 5-1 Discussion added to indicate MCPRI limits and LHGRFACI multipliers are provided for maximum core flows of 102.5% and 105% of rated.
- 2. 5-10 Revised Figure 5.1 to include MCPRI limits for both 102.5% and 105%
maximum core flows.
- 3. 9-1 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.
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 Revision 1 Reload Analysis Page iv 8-1 5.0 Methodology References .............................................................................................
9-1 9.0 Additional References .............................................................................................
Tables 1-2 1.1 EOD and EOOS Operating Conditions ........................................................................ 4-4 4.1 Neutronic Design Values ..............................................
for 5.1 EOC Base Case and EOOS MCPRp Limits and LHGRFACP Multipliers 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 Power) ......................................................................................................................... 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 ................................................ 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 ................ .................................... 35 3.4 LaSalle Unit 1 Cycle 9 Safety Limit Local Peaking Factors SPCA9-384B-1i1GZ-80M With Channel Bow ........................ : ...................................... 3-6 4.1 SPCA9-4.56L-12G8.014G3.0-10DM Enrichment Distribution ........................................ 4-5 4.2 SPCA9-4.56L-12G8.0-100M Enrichment Distribution .................................................. 4-6 4.3 SPCA9-3.91L-12G8.0-100M Enrichment Distribution ....................... :.......................... 4-7 4.4 SPCA9-3.90L-8G5.0-100M Enrichment Distribution ............................................... 4-8 4.5 SPCA9-4z59L-12G8.O-100M Enrichment Distribution .................................................. 4-9 4.6 SPCA9-4.59L-12G7.0-100M Enrichment Distribution ................................................ 4-10 4.7 SPCA9-3.96L-8G7.0/4G8.0-100M Enrichment Distribution ........................................ 4-11 4.8 SPCA9-3.96L-SG5.0-100M Enrichment Distribution .................................................. 4-12 4.9 SPCA9-4.58L-8G6.0/4G3.0-10DM Enrichment Distribution ........................................ 4-13 4.10 SPCA9-4.58L-8G6.0-OOM-Enrichment Distribution .................................................. 4-14 4.11 SPCA9-4.06L-11G6.0-80M Enrichment Distribution .................................................. 4-15 4.12 SPCA9-4.34L-10G6.0-80M Enrichment Distribution .................................................. 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 Times .......................................................................................................... 5-13 5.5 Base Case Power Dependent MCPR Limits for ATRIUM-9B Fuel - NSS Insertion Times ......................................................................................................... 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
EMF-2276 LaSalle Unit I Cycle 9 Revision 1 Reload Analysis Page vi Nomenclature AOO abnormal operational occurrence BOC beginning of cycle CPR critical power ratio EFPH effective full power hour EOC end of cycle EOD extended operating domain EOOS equipment out of service feedwater heater out of service FHOOS FWCF feedwater controller failure ICF increased core flow LFWH loss of feedwater heater LHGR linear heat generation rate LHGRFAC LHGR multiplier LPRM local power range monitor LRNB load rejection no bypass MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MELLLA maximum extended load line limit area MSIV main steam isolation valve NSS nominal scram speed PAPT protection against power transient PCT peak clad temperature RPT recirculation pump trip SLMCPR safety limit minimum critical power ratio SLO single-loop operation SPC Siemens Power Corporation SRVOOS safety/relief valve out of service TBVOOS turbine bypass valves out of service TCV turbine control valve TIP traversing in-core probe TIPOOS traversing in-core probe out of service TSSS technical specification scram speed UFSAR updated final safety analysis report ACPR change in critical power ratio Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 Page 1-1 Reload Analysis 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
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis 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 14.7 Mlb/hr Reference 9.3 at 100%P/1 00%F (includes water channel flow) 3.3 MCPR Fuel Cladding Integrity Safety Limit (SLMCPR)
Two-Loop Operation* 1.11 Reference 9.3 Single-Loop Operation* 1.12 3.3.1 Coolant Thermodynamic Condition Thermal Power (at SLMCPR) 5232.35 MWt Feedwater Flow Rate (at SLMCPR) 22.7 Mlbmlhr Core Exit Pressure (at Rated Conditions) 1031.35 psia Feedwater T-emperature 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
EMF-2276 Revision 1 LaSalle Unit I Cycle 9 Reload Analysis Page 3-3 200 175 150 cn "D 125 M
100 5
Q)
-o E--75 I
z 50 25 0
.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
EMF-2276 Revision 1 LaSalle A:UnitIl I "~neC Cycle 9 Page 3-4 Reloa
- ,h-t 'jl0"a y ,.
C ont r o I Rod Co rn e r 0
n t 1.023 1.055 1.068 1.112 1.099 1.102 1.049 1.023 .977 r
0 1.055 .958 .894 1.016 .894 1.007 .877 .927 1.002 R
0 d 1.068 .894 1.031 1.065 1.084 1.056 1.010 .863 1.011 C
0 Internal 1.044 .980 1.051 1.112 1.016 1.065 r
n e
1.099 .894 1.084 Water 1.063 .863 1.038 r
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 Figure 3.2 LaSalle Unit I Cycle 9 Safety Limit Local Peaking Factors SPCA9-393B-16GZ-1OOM With Channel Bow
EMF-2276 Revision 1 LaSalle 1:
l Unit 1 Cycle 9 r14 A ,J lc ( Page 3-5 Re oa C ontrol Rod Corner 0
n t 1.013 1.042 1.056 1.110 1.098 1.100 1.037 1.010 .967 r
0 1.042 .944 1.025 .879 1.014 .871 1.005 .912 .989 R
0 1.056 1.025 1.018 1.064 1.081 1.055 .997 .989 .999 d
C 0 1.110 .879 1.064 Internal 1.043 .848 1.047 r
n e 1.059 .978 1.035 1.098 1.014 1.081 Water r
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
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-6 CIUL) 3 iI=J C ontrol Rod Corner 0
n t 1.022 1.056 1.061 1.035 1.102 1.028 1.045 1.029 .982 r
0 1.056 .947 1.018 1.003 '879 .997 1.004 .919 1.011 R
0 d 1.061 1.018 1.001 1.050 1.081 1.048 .996 .992 1.012 C
0 Internal .926 .983 .987 1.035 1.003 1.050 r
n e
1.102 .879 1.081 Water 1.077 .853 1.049 r
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 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
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Page 4-1 Reload Analysis 4.0 Nuclear Design Analysis 4.1 Fuel Bundle NuclearDesign Analysis Assembly Average Enrichment (ATRIUM-9B fuel)
SPCA9-393B-16GZ-100M 3.93 wt%
SPCAg-396B-12GZB-100M 3.96 wt%
SPCA9-384B-11GZ-80M 3.84 wt%
SPCA9-396B-12GZC-100M 3.96 wt%
Radial Enrichment Distribution SPCA9-4.56L-12G8.D/4G3.0-1 0DM Figure 4.1 SPCA9-4.56L-12G8.0-1OOM Figure 4.2 SPCA9-3.91 L-1 2G8.0-1 0DM Figure 4.3 SPCA9-3.90L-8G5.0-100M Figure 4.4 SPCA9-4.59L-12G8.0-100M Figure 4.5 SPCA9-4.59L_-12G7.0-100M Figure 4.6 SPCA9-3.96L-8G7.0/4G8.0-1 0DM Figure 4.7 SPCA9-3.96L-8G5.0-10DM Figure 4.8 SPCA9-4.58L-8G6.0/4G3.0-100M Figure 4.9 SPCA9-4.5BL-8G6.0-1 0DM Figure 4.10 SPCA9-4.06L-1 1G6.0-80M Figures 4.11 SPCA9-4.34L-1 0G6.0-80M Figures 4.12 Axial Enrichment Distribution Figures 4.13-4.16 Burnable Absorber Distribution Figures 4.13-4.16 Non-Fueled Rods Figures 4.1-4.12 Neutronic Design Parameters 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
EMF-2276 LaSalle Unit I Cycle 9 Revision 1 An Iv_*_
Rplt=r,*. Paqe 4-2 Reload Ana lvsis 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 Figure 4.17 Core Exposure at EOC8, MWd/MTU 27.957 (nominal value)
Core Exposure at BOC9, MWd/MTU 10,962 (from nominal EOCS).
Core Exposure at EOC9, MWd/MTU 29,439 (licensing basis)
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. >
4.2.4 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 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 Regionalt 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.
EMF-2276 LaSalle Unit I Cycle 9 Revision 1 Reload Arialysis Pagje 4-4 Table 4.1 Neutronic Design Values Number of Fuel Assemblies 764 Rated Thermal Power, MWt 3489 Rated Core Flow, Mlbm/hr 108.5 Core Inlet Subcooling, Btu/lbm 18.1 Moderator Temperature, 'F 548.8 Channel Thickness, inch 0.080 & 0.100 Fuel Assembly Pitch, inch 6.0 Wide Water Gap Thickness, inch* 0.28110.261 Narrow Water Gap Thickness, inch 0.281/02.61 Control Rod Datat Absorber Material B4 C Total Blade Support Span, inch 1.580 Blade Thickness, inch 0.260 Blade Face-to-Face Internal Dimension, inch 0.200 Absorber Rod OD, inch 0.188 Absorber Rod ID, inch 0.138 Percentage B4 C, %TD 70 The water gap thicknesses presented are based on 80/1 00-mil channels for ATRIUM-9B fuel.
I The control rod data represents original equipment control blades at LaSalle and were used in the neutronic calculations.
Siemens Power Corporation
EMF-2276 LaSalle Unit I Cycle 9 Revision 1 Reload Analysis Page 4-5 IJL Axial Location
- F 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 4.00 3.00 4.20 8.00 7 4.70 7 4.20 4.00 2 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 2 4.00 3 4.0 4 7 7000~>
6 7 4: 3 2 4.00 4.00 4.20 4.708 7 4.20' 3.00 8.00 8.00.00 :4.00" 3.00 2 4.00 1 2 5 7 7 7 5 2 1 I
3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 lul LJ Pellet Type Quantity U2 35 + Gd 2 0 3 Concentration (wt%)
1 4 3.00 2 8 4.00 3 4 4.00+ 3.00 4 8 4.20 + 8.00 5 8 4.70 6 .4 4.70+ 8.00
- 7_ 36 4.95 Figure 4.1 SPCA9-4.56L-12G8.0/4G3.0-1OOM
.Eurichment Distribution Siemens Power Corporation
EMF-227 LaSalle Unit 1 Cycle 9 Revision Reload Analysis Page 4-Reload Anafvsis
_JL Axial Location "3 F" In Assembly 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 4.00 8.00 8.00 8.00 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 4 20 6 6 6 6 6 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 Iul L2J Pellet Type Quantity U235 + Gd 2 0 3 Concentration (wt%)
1 4 3.00 2 12 4.00 3 8 4.20+ 8.00 4 8 4.70 5 4 4.70 + 8.00 6 36 4.95 Figure
" - 4.2 SPCA9-4.56L-12G8.0-100M Enrichment Distribution
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 4-7 I L Axial Location In Assembly
-- 1 F 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. 44 8:0 3.40 4.0 8.00 4.40 4 Water Channel 4.40 s 34O. .
.8.00 4.40 4
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 32 340 5* -'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 Pellet Type Quantity U23 5 + Gd 2 0 3 Concentration (wt%)I 1 4 2.60 2 12 3.40 3 12 3.40+ 8.0Q 4 8 3.80 5 36 4.40 Figure 4.3 SPCA9-3.91L-12G8.0-100M Enrichment Distribution Siemens Power Corporation
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 P.Relr4 Anni SiS Page 4-8 IL Axial Location In Assembly I-1 2.60 2
3.40 4
3.80 5
4.40 5
4.40 5
4.40 4
3.80 2
3.40 1
2.60 n 2
3.40 2
3.40 3
3.40 5.00 5
4.40 2
3.40 5
4.40 3
3.40 500 2 2 I 3 3 3 3.40 40 5 3.40 4 5.00 4.40 4.40 4.40 4A0 4.40 .5.00 3.80 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 lul Pellet Type Quantity U235+ Gd 2 0 3 Concentration (wt%)
1 4 2.60 2 16 3.40 3 8 3.40 -+5.00 4 8 3.80 5 36 4.40 Figure 4.4 SPCA9-3.90L-8G5.0-I0OM Enrichment Distribution Siemens Power Corporation
E MF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Rp!ond Analysis Page 4-9 Reload Analvsis IL Axial Location in Assembly 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 4.95 8.00' 4.95 4.95 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 2
2 3 4.20 5 3 4.20 5 3 4.20 5 3 4.20 2 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 I7I Iul Li Pellet Type Quantity U2 35 + Gdt0 3 Concentration (wt%)
1 4 3.00 2 8 4.00 3 12 4.20 + 8.00 4 8 4.70 5 40 4.95 Figure 4.5 SPCA9-4.59L-12G8.0-100M Enrichment Distribution Siemens Power Corporation
EMF-2276 LaSalle Unit I Cycle 9 Revision 1 Reload Analysis Page 4-10 IL Axial Location In Assembly 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 7423U 2 4.00 3
4.20 700 5
3 4.20 7.00 5
3 4.20 7.00 5
3 4.20 7.00 2
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 I
7.00 7.00 5 5 5 Water Channel 5 5 5 4.95 4.95 4.95 4.95 4.95 4.95 333.
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 Iu'I Pellet Type Quantity U 23 5 + Gd 2 O 3 Concentration (wt%)
1 4 3.00 2 8 4.00 3 12 4.20 + 7.00.
4 8 4.70 5 40 4.95 Figure 4.6 SPCA9-4.59L-12G7.0-1O0M Enrichment Distribution Siol)ian Powor Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 PIr==Iv4 An~Iv~i* Page 4-11 PoInnrl Analysis I L Axial Location In Assembly i-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
4.40 6
4.40 6
4.40 Water Channel 6 4.40 6
4.40 6
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" 2 3.40 340 6 3.40 6 -340 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 luI L2J Pellet Type Quantity U235 + Gd 2 0 3 Concentration (wt%)
1 4 2.60 2 8 3.40 3 8 3.40 + 7.00 4 4 3.40 + 8.00 5 8 3.80 6 40 4.40 Figure 4.7 SPCA9-3.96L-8G7.014G8.0-100M Enrichment Distribution Siemens Power Corporation
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 4-12 IL / Axial Location In Assembly 3-.
7, 7
1 2.60 2
3.40 4
3.80 5
4.40 5
4.40 5
4.4 4
3.80 2
3.40 1
2.60 I 3 3 23.40 5 340 5 2 2 3.40 3.40 4.40 500 4.40 4.40 3.40 3.40
-r 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 s 3 3.40 53.
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
-z 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 Iul Li Pellet Type Quantity U2 35 + Gd 2 0 3 Concentration (wt%)
1 4 2.60 2 12 3.40 3 8 3.40 + 5.00 4 8 3.80 5 40 4.40 Figure 4.8 SPCA9-3.96L-8G5.0-1OOM
_ Enrichment Distribution I
Siemens Power Corporation
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 4-13 I L Axial Location In Assembly I r" 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 4.00 6 4
-4.20 4 44.20 6 400 2 400 .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 lul L1J Pellet Type Quantity U 2 35 + Gd 2 0 3 Concentration (wt%)
1 4 3.00 2 8 4.00 3 4 4.00 + 3.00 4 8 4.20 + 6.00 5 8 4.70 6 40 4.95 Figure 4.9 SPCAS-4.58L-8G6.0/4G3.0-1 OM Enrichment Distribution Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 Page 4-14 Reload Analysis I L Axial Location In Assembly
-11 -
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 4.005 2 4.002.9 5 420 4.95 4.20 4.95 4.00 4.00 2
4.95 6.09.00 4 4.00 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 5 5.34.20.4.20 3
. .0 4.95 4.9 4.*20 4.95 4.95 I
6.00 6.00o" 495 5 Water Channel 5 4.95 495 4.9 4.95 4.95 4s .5 49 5 4.2 3."3. ~4.20 4.95 40 4.95 4.5 4.20 4.95 6.00 6.00 5 5 5 .5 5 5 5 4 4
4.95 4.95 4.95 4.95 4.95 4.95 4.95 4.70 4.70 3 3 2 2 2 5 3 4.20 5 3
.4.202 5 2 2 4.204-2 4.95 4.00 4.00 d
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 23 5 + Gd 2 Q3 Concentration (wt%)
1 4 3.00 2 12 4.00 3 8 4.20 + 6.00 4 8 4.70 5 40 4.95 Figure 4.10 SPCA9-4.58L-8G6.0-100M SEnrichment Distribution cSipmpn; Pnwor rnrrnmrtfll
EMF-2276 Revision I LaSalle Unit 1 Cycle 9 PDIJrt nrjannvrk Page 4-15 I L Axial Location In Assembly "1 Fl 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 23,69 G 39 G 3.69 G 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 4.53
-369 6.00- 4.53 4 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 luI Pellet Type Quantity U2 3 5 + Gd 2O 3 Concentration (wt*lo) 1 4 2.72 2 8 3.53 3 16 3.94 4 33 4.53 G 11 3.69 + 6.00 Figure 4.11 SPCA9-4.06L-11G6.0-80M Enrichment Distribution Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 PaQe 4-16 V ncIy A
Y0lC IL Axial Location In Assembly
-I I- t 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 .94.19 2 G 4 4 4:19 G 2 3.8 6.00 4.190 4.78 4.78 6.00 6.00 4.78 4.78 6.00 6.00 3.78 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 4.78 4.78 Water Channel el 4.78 4 60,4.784.19 4.78 e6.00 6.00 3 4 4 4 4 . 3 4.19 4.78 4.78 4.78 4.78 4.19 4G 4 4 4 4 3 4.19419 .8 4.78 47 4.78 6.00 4.19 4.78 4.78 4.78 4.78 4.19 2 3.8 4.00 G 4 G 44.19 4.78 4.78 .19 4.19 2 3.78 4.19 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 IuI Li Pellet Type Quantity U2 35 + Gd 2 0 3 Concentration (wt%)
1 4 2.72 2 8 3.78 3 16 4.19 4 34 4.78 G 10 4.19+6.00 Figure 4.12 SPCA9-4.34L-1OG6.0-BOM Enrichment Distribution Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 P*I*rI Annlvsis Page 4-17 Reload A S PCAD-393B-16QZ-100M Lattice 2 3 11 12 12 12 11 3 2 3 4 6 12 10 12 6 4 3 Natural Uranium SPCA9-0.72L-0.GO.O-1OM 11 6 12 12 12 12 12 6 11 3.90-8G5.0 SPCA9-3.90L-8G5.O-1 0OM 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 3.91-12G8.0 SPCA9-3.91 L-12GS.0-100M 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 4-56-12G8.0 SPCA9-4.56L-12G8.0-1OOM Fuel Rod No.
Type Rods 2 4 3 8 4 4 6 8 SPCA9-4.56L 4.56-12G8.014G3.0 12G8.O/4G3.0-100M 10 4 11 8 Natural Uranium SPCA9-0.72L-O-GO.0-1 DOM 12 36 Figure 4.13 ATRIUM-9B LSA-1 19A Assembly Design Siemens Power CorporaLion
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 4-18 SPCA9-393B-16GZ-100M ROD ROD ROD ROD ROD ROD ROD
_2 3 _6 10 12 E2 E2 E2 F E2 E2 E2 E2 E3 E5 E5 E6 E5 E9 E15 E8 E4 El0 El0 E14 E16 E18 E17 Ell E2 E2 E2 E2 E2 Lattice Lattice Index Enrichment + Gd Index Enrichment + Gd "E2 0.72 wt% U-235 Eli 4.00 wt% U-235 + 3.0 wt% Gd 20 3 E3 2.60 wt% U-235 E12 4.20 wt% U-235 + 6.0 wt% Gd 20 3 E4 3.00 wt% U-235 E13 4.20 wt0/' U-235 + 7.0 wt% Gd 2 0 3 E5 3.40 wt% U-235 E14 4.20 wt% U-235 -t 8.0 wt% Gd 203 E6 3.40 wt% U-235 + 5.0 wt% Gd 20 3 E15 4.40 wt% U-235 E7 3.40 wt% U-235 + 7.0 wt% Gd 203 E16 4.70 wt% U-235 E8 3.40 wt% U.2235 + 8.0 wt% Gd 2 0I E17 4.70 wt% U-235.+ 8.0 wt% Gd 203 E9 3.80 wt% U-235 E18 4.95 wt% U-235 E10 4.00 wt% U-235 Figure 4.13 ATRIUM-9B LSA-1 19A Assembly Design (continued)
Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 Page 4-19 ReIo ad Ana ys -
SPCAS-396B-1 2GZB-1 COM Lattice 2 3 11 12 12 12 11 3 2 3 7 12 8 12 8 12 7 3 Natural Uranium SPCA9-o0.72L-O.GO.O-100M 11 12 12 12 12 12 12 12 11 SPCA9-3.96L-8G5.0-100M 12 8 12 W W W. 12 8 12 3.96-8G5.0 12 12 12 W W W 12 12 12 12 .8 12 W W W 12 8 12 SPCA9-3.96L-8G7.O/4G8.O 3.96-BG7.0/4G8.0 I looM 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 4.59-12G7.0 SPCA9-4.59L-12G7.0-100M Fuel Rod No.
Type Rods SPCA9-4.59L-12GS.O-100M 2 4 4.59-12G8.0 3 8 7 4 8 8 11 8 12 40 Natural Uranium SPCA9-0.72L-O.GO.O-IOOM Figure 4.14 ATRIUM-9B LSA-1 19B Assembly Design Siemens Power Corporation
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 4-20 SPCA9-396B-12GZB-1 00M ROD ROD ROD ROD ROD ROD E2 3 7 _8 11 12
!E2 E2 E2 E2 E2 E3 E5 E5 E6 E9 E15 E8 E7 E4 El0 E13 E13 E16 E18 E14 E14 E2 E2 E2 E2 E2' E2 Lattice Lattice Index Enrichment -+Gd Index Enrichment + Gd E2 0.72 wt% U-235 Eli 4.00 wt% U-235 + 3.0 wt% Gd2 O3 E3 2.60 wt% U-235 E12 4.20 wt% U-235 + 6.0 wt% Gd 2 0 3 E4 3.00 wt% U-235 E13 4.20 wt% U-235 +.7.0 wt% Gd 2 03 E5 3.40 wt% U-235 E14 4.20 wt% U-235 + 8.0 wt% Gd 2 0 3 E6 3.40 wt% U-235 + 5.0 wt% Gd 20 3 E15 4.40 wt% U-235 E7 3.40 wt% U-235 + 7.0 wt% Gd 2 0 3 E16 4.70 wt% U-235 ES 3.40 wtrU-235 + 8.0 wt% Gd 20 3 E17 4.70 wt% U-235 + 8.0 wt% Gd 2 03 E9 3.80 wt% U-235 E18 4.95 wt% U-235 E10 4.00wt% U-235 Figure 4.14 ATRIUM-9B LSA-1 19B Assembly Design (continued)
Siemens Power Corporation
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 4-21 SPCA9-396B-12GZC-1 OOM Lattice 2 3 11 12 12 12 11 3 2 SPCA9-0.72L-0.GO.0 3 5 12 9 12 9 12 5 3 Natural Uranium looM 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 3.96-8G5.0 SPCA9-3.96L-8G5.0-1 DOM 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 4.58-8G6.0 SPCA9-4.58L-8G6.0-100M Fuel Rod No.
SPCA9-4.58L Type Rods 4.58-8G6.014G3.0 8G6.014G3.0-100M 2 4 3 8 5 4 9 8 11 8 12 40 SPCA9-0.72L-0.GO.0 Natural Uranium 10DM Figure 4.15 ATRIUM-9B LSA-1 19C Assembly Design Siemens Power Corporation
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 R*_Io* d Analysis Page 4-22 Reload Analvsis SPCA9-396B-1 2GZC-1 0DM ROD ROD ROD ROD ROD ROD 2 3 5 9 11 12 E2 F E2 E2 E2 E2 E3 E5 E5 E6 E9 E4 El0 E12 E16 E18 Ell E2 E2 E2 E2 E2 E2 Lattice Lattice Index Enrichment + Gd Index Enrichment + Gd
-E2 0.72 wt% U-235 Ell 4.00 wt% U-235 + 3.0 wt% Gd 2 0 3 E3 2.60 wt% U-235 E12 4.20 wt% U-235 + 6.0 wt% Gd 20 3 E4 3.00 wt% U-235 E13 4.20 wt% U-235 + 7.0 wt% Gd 203 E5 3.40 wt% U-235 E14 4.20 wt% U-235 + 8.0 wt% Gd 20 3 E6 3.40 wt% U-235 + 5.0 wt% Gd2 03 E15 4.40 wt% U-235 E7 3.40 wt% U-235 + 7.0 wt% Gd 2 0 3 E16 4.70 wt% U-235 ES 3.40 wt% U-2-S5 + 8.0 wt% Gd 20 3 E17 4.70 wt% U-235 + 8.0 wt% Gd 20 3 E9 3.80 wt% U-235 E18 4.95 wt% U-235 E1O 4.00 wt% U-235 Figure 4.15 ATRIUM-gB LSA-1 19C Assembly Design (continued)
Sionions Power Corporation
EMF-22761 Revision LaSalle Unit 1 Cycle 9 Page 4-23 Reload Analysis SPCA9-384B-11 GZ-80M 1 2 3 3 4 3 3 2 1 2 GI 4 4 G1 4 4 GI 2 SPCA9-0.72L-OGO.0-80M 3 4 4 4 4 4 4 4 3 3 4 4 W" W W G1 4 3 4 G1 4 W W W 4 G1 4 SPCA9-4.06L-1 1 G6.0-80M 3 4 4 W W W 4 4 3 3 4 4 GI 4 4 G2 4 3 2 G1 4 4 GI 4 4 GI 2 1 2 3 3 4 3 3 2 1 Fuel Rod No.
Type Rods 4
SPCA9-4.34L-10G6.0-OM 1 2 8 3 16 4 33 G1 10 G2 1 SPCA9-0.72L-OGO.0-80M Figure 4.16 ATRIUM-9B SPCA9-384B-11GZ-80M Assembly Design Sioniens Power Corporation
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Re lnd Analvsis Paqe 4-24 Reload Analvsis SPCA9-384B-11GZ-80M 1 2 3 4 G1 G2 A A A A A A i
B hC E G D F H H A A A A A A 0.72 wt% U-235 B 2.72 wt% U-235 C 3.53 wt% U-235 D 3.78 wt% U-235 E 3.94 wt% U-235 F 4.19 wt% U-235 G 4.53 wt% U-235 H 4.78 wt% U-235 3.69 wt% U-235 + 6.00 wt% Gd 203 J 4.19 wt% U-235 + 6.00 wt% Gd 2 0 Figure 4.16 ATRIUM-9B SPCA9-384B-11GZ-BOM Assembly Design
- _ (continued)
Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 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 Number Load' Fuel Bundle Name of Bundles cle 56 7 1 GE9B-.PBCWB322-11GZ-100M-150 89 7 2 GE9B.PSCW8320-SGZ-1004M-150 104 8 4 GE9B-PF`CWB343-12GZ-80M-150 143 8 5 GE9B-P8CWB342-IOGZ-80M-15O 208 9 6 SPCA9-393B-16GZ-IO0M 68 9 7 SPCAS-396B-12GZB-1 OM 36 9 a SPCAB-384B-11GZ-80M 9 40 SPCA9-396B-12GZC-1 DOM Figure 4.17 LaSalle Unit I Cycle 9 Reference Loading Map Siemens Power Corporation
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Paqe 5-1 5.0 Anticipated Operational Occurrences Applicable Disposition of Events Reference 9.7 5.1 Analysis of Plant Transients at Rated Conditions Reference 9.3 Limiting Transients: Load Rejection No Bypass (LRNB)
Feedwater Controller Failure (FWCF)
Loss of Feedwater Heating (LFWH)"
Peak Peak Peak Lower Neutron Heat Plenum Scram Flux Flux Pressure ACPR Transient Speed (% Rated) (% Rated) (psig) ATRIUM-9B/GE9 LRNB TSSS 460.2 126.5 1206 0.341 0.38 FWCFt TSSS 371.3 122.6 1167 0.30 / 0.33 LRNB' NSS 401.1 121.3 1203 0.31 /0.34 FWCF t NSS 342.9 120.5 1164 0.28 / 0.31 zt LFWH*
5.2 Analysis for Reduced Flow Operation Reference 9.3 Limiting Transient: Slow Flow Excursion MCPR, Manual Flow Control - ATRIUM-9B and GE9 Fuel Figure 5.1 LHGRFAC -1 ATRIUM-9B Fuel Figure 5.2
- T MAPFACf - GE9 Fuel 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
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 5-2 5.3 Analysis for Reduced Power Operation Reference 9.3 Limiting Transient: Load Rejection No Bypass (LRNB)
Feedwater Controller Failure (FWCF)
MCPRp Base Case Operation Tables 5.1-5.4 Figures 5.3-5.6 LHGRFACp Base Case Operation' Tables 5.1-5.4 MCPRp, EOOS Conditions Tables 5.1-5.4 LHGRFACp, EOOS Conditions* Tables 5.1-5.4 MAPFACP - All Operating Conditions' <To be furnished by ComEd.>
5.4 ASME OverpressurizationAnalysis Reference 9.3 Limiting Event MSIV. Closure Worst Single Failure Valve Position Scram Maximum Vessel Pressure (Lower Plenum) 1320 psig Maximum Steam Dome Pressure 1291 psig 5.5 Control Rod Withdrawal Error Starting Control Pattern for Analysis 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 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 LaSalle Unit 1 Cycle 9 Page 5-4 Neioad AnaIdys 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 GE9 Fuel Power LHGRFACP MCPRP EOOS Condition (% Rated) MCPRP Base Case Operation 0 2.70 0.66 "2.70 25 2.22 0.66 2.22 25 2.07 0.66 2.12 63 1.56 0.94 1.57 84 1.51 0.98 1.53 100 1.46 0.99 1.50 0 2.85 0.64 2.85 Feedwater Heaters 25 2.38 0.64 2.38 Out of Service 25 2.38 0.64 2.38 (FHOOS) 1.62 63 1.62 0.90 100 1.47 0.99 1.51 0 2.71 0.66 2.71 Single-Loop 2.23 25 2.23 0.66 Operation 2.13 25 2.08 0.66 63 1.57 0.94 1.58 84 1.52 0.98 1.54 100 1.47 0.99 1.51 0 2.70 0.66 2.70 Turbine Bypass 25 2.22 0.66 2.22 Valves Out of 25 2.17 0:66 2.17 Service (TBVOOS) 1.65 63 1.63 0.90 100 1.49 0.94 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 LaSalle Unit 1 Cycle 9 Page 5-5 P*Il~d 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 GE9 Fuel EOOS/EOD Power MCPR, Condition (% Rated) MCPRP LHGRFACP 0.66 2.70 Recirculation Pump 0 2.70 25 2.22 0.66 2.22 Trip Out of Service 2.12 25 2.07 0.66 (No RPT) 1.63 63 1.60 0.86 100 1.51 0.86 1.56 0 2.70 0.66 2.70 Turbine Control 2.22
- 25. 2.22 0.66 Valve (TCV) Slow 2.16 25 2.16 0.66 Closure and/or 1.69 84 1.65 0.86 No RPT 1.67 84 1.63 0.86 100 1.56 0.86 1.60 0 2.85 0.63 2.85 TCV Slow Closure/ 2.38 25 2.38 0.63 FHOOS and/or 2.38 25 2.38 0.63 No RPT 1.69 84 1.65 0.86 84 1.63 0.86 1.67 100 1.56 0.86 1.60 0 2.54 0.40 2.54 Idle Loop 2.54 25 2.54 0.40 Startup 2.54 25 2.54 0.40 63 2.54 0.40 2.54 100 2.54 0.40 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 LaSalle Unit 1 Cycle 9 Page 5-6 Reload Analysis Table 5.2 Base Case MCPRP Limits and LHGRFACP Multipliers for NSS Insertion Times for Prepower Uprate (3323 MWt Rated Power)f ATRIUM-9B Fuel GE9 Power MCPRP EOOS Condition (% Rated) MCPRP LHGRFAC, 0 2.70 0.74 2.70 Base Case Operation 2.22 25 2.22 0.74 2.07 0.74 2.07 25 1.54 0.95 1.56 63 1.48 1.00 1.51 84 1.43 1.00 .1.46 100 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 for FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF and MELLLA regions the power/flow map.
Siemens Power Corporation
EMF-2276 Revision I LaSalle Unit 1 Cycle 9 Page 5-7
ý W. j 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 GE9 Fuel Power
(% Rated) MCPRP LHGRFAC, MCPRP EOQS/EOD Condition Base Case Operation 0 2.70 0.67 "2.70 25 2.20 0.67 2.20 25 2.05 0.67 2.10 60 1.56 0.94 1.57 80 1.51 0.98 1.53 100 1.45 1.00 1.49 0 2.85 0.65 2.85 Feedwater Heaters 25 2.35 0.65 2.35 Out of Service 25 2.35 0.65 2.35 (FHOOS) 60 1.62 0.90 1.62 100 1.45 1.00 1.49 0 2.71 0.67 2.71 Single-Loop 25 2.21 0.67 2.21 Operation 25 2.06 0.67 2.11 60 1.57 0.94 1.58 80 1.52 0.98 1.54 100 1.46 1.00 1.50 0 2.70 0.67 2.70 Turbine Bypass 25 2.20 .0.67 2.20 Valves Out of 25 2.15 0.67 2.15 Service (TBVOOS) 60 1.63 0.90. 1.65 100 1.47 0.94 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
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 R4=In~r An~ivsis Page 5-8 Reload Analvsis 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 GE9 Fuel EOOS/EOD Power Condition (% Rated) MCPRP LHGRFACP MCPRP Recirculation Pump 0 2.70 0.67 "2.70 Trip Out of Service 25 2.20 0.67 2.20 (No RPT) 25 2.05 0.67 2.10 60 1.60 0.86 1.63 100 1.50 0.86 1.55 Turbine Control 0 2.70 0.67 - 2.70 Valve (TCV) Slow 25 2.20 0.67 2.20 Closure and/or 25 2.15 0.67 2.15 No RPT 80 1.65 0.86 1.69 80 1.63 0.86 1.67 100 1.54 0.86 1.58 TCV Slow Closure/ 0 2.85 0.64 2.85 FHOOS and/or 25 2.35 0.64 2.35 No RPT 25 2.35 0.64 2.35 80 1.65 0.86 1.69 80 1.63 0.86 1.67 100 1.54 0.86 1.58 Idle Loop 0 2.54 0.40 2.54 Restart 25 2.54 0.40 2.54 25 2.54 0.40 2.54 60 2.54 0.40 2.54 2.54 100 2.54 0.40 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
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 P* tA~4 Andlv~i Pace 5-9 PalmnH Analysis Table 5.4 EOC MCPR, Limits and LHGRFAC, Multipliers for NSS Insertion Times for Power Uprate Conditions (3489 MWt Rated Power)
ATRIUM-9B Fuel GE9 Power EOOS Condition (% Rated) MCPRP LHGRFACP MCPR, Base Case Operation 0 2.70 0.75 2.70 25 2.20 0.75 2.20 25 2.05 0.75 2.05 60 1.54 0.95 1.56 80 1.48 1.00 1.51 100 1.42 1.00 1.45 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 LaSalle Unit 1 Cycle 9 Page 5-10 nI"loadU ta-Illysz 0.
0 1.00 40 60 80 100 120 0 20 of Ratd)
Flow ('/%
102.5% Maximum Core Flow 105% Maximum Core Flow Flow MCPR, MCPi 1 MCPR1 MCPR1 ATRIUM-9B GE9 ATRIUM-gB GE9
(% rated) 0 1.85 1.85 1.93 1.93 30 1.85 1.85 1.93 1.93 1.11 1.11 1.14 1.14 102.5 1.11 1.11 1.11 1.11 105 Figure 5.1. Flow Dependent MCPR Limits for Manual Flow Control Mode Siemens Power Corporation
EMF-2276 Revision 1 Page 5-11 LaSalle Unit I Cycle 9 Reload Analysis 1.1 1
0.8
- 0.8-40 50 60 70 Percent of Rated Flow Flow (% rated) LHGRFACf 0 0.69 0.69 30 1.00 76 1.00 105 Fuel Figure 5.2 Flow Dependent LHGR Multipliers for ATRIUM-9B Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit I Cycle 9 1l. ei..l Paqe 5-12 D ý 1W- 14 18 "a Yý 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 115 0 500 1000 1500 2000 2500 3000 3500 4000 Povmr (MWth) 3323 MWt Rated Power 3489 MWt Rated Power Power (%) MCPRp Limits Power (%) MCPRp Limits 100 1.46 100 1.45 84 1.51 80 1.51 63 1.56 60 .1.56 25 2.07 25 2.05 25 2.22 25 2.20 0 2.70 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 LaSalle Unit 1 Cycle 9 Page 5-13 D 1 4 AýýI I't 2.75 2A5 2.55 2-45 2.35 225 2.15 2.05 I.
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 3489 MWt Rated Power Power (%) MCPRp Limits Power (%) MCPRp Limits 100 1.50 100 1.49 84 1.53 80 1.53 63 1.57 60 1.57 25 2.12 25 2.10 25 2.22 25 2.20 0 2.70 0 2.70 Figure 5.4 Base Case Power Dependent MCPR Limits for GES Fuel - TSSS Insertion Times SioMnons Powor Corporation
EMF-2276 Revision I LaSalle Unit 1 Cycle 9 Page 5-14 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 3489 MWt Rated Power MCPR, Limits Power (%) MCPRp Limits Power (%)
100 1.42 100 1.43 1.48 80 1.48 84 1.54 60 1.54 63 2.07 25 2.05 25 2.22 25 2.20 25 2.70 0 2.70 0
Figure 5.5 Base Case Power Dependent MCPR Limits for ATRIUM-9B Fuel - NSS Insertion Times Siemens Power Corporation
EMF-2276 LaSalle Unit 1 Cycle 9 Revision 1 Reload Analysis Page 5-15 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 3489 MWt Rated Power Power (%) MCPRp Limits Power (%) MCPRp Limits 100 1.46 100 1.45 84 1.51 80 1.51 63 1.56 60 1.56 25 2.07 25 2.05 25 2.22 25 2.20 .
0 2.70 0 2.70 Figure 5.6 Base Case Power Dependent MCPR Limits for GE9 Fuel - NSS Insertion Times Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 Page 5-16
< 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 I LaSalle Unit 1 Cycle 9 Page 6-1 ED 1 4 A-=i cis
- ý W_ I 6.0 Postulated Accidents 6.1 Loss-of-Coolant Accident 6.1.1 Break Location Spectrum Reference 9.8 Break Size Spectrum Reference 9.8 6.1.2 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 ft2 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 Peak Local Metal-Water Reaction
(°F) (%)
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
EMF-2276 Revision I LaSalle Unit 1 Cycle 9 Page 6-2 Reload Analysis shutdown of the reactor, and it is 15.7.5. The analysis is assumed to occur 360 days following a result of the accident.
assumed that all 32 fuel assemblies in the cask completely fail as following shutdown of the Because the accident is assumed not to occur sooner than 360 days to fission product decay. Hence, reactor, the source term for the accident will be very low due will be very low. The results of the commensurate radiological whole-body and thyroid doses involving SPC"ATRIUM fuel will this analysis demonstrate that spent fuel cask drop accidents dose limits which-are a small not exceed the established radiological whole-body and thyroid fraction of the 10 CFR 100 limits for radiological exposures.
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 Page 7-1 Reload Analysis 7.0 Technical Specifications 7.1 Limiting Safety System Settings 7.1.1 MCPR Fuel Cladding Integrity Safety Limit 1.11' MCPR Safety Limit (all fuel) - two-loop operation 1.12 MCPR Safety Limit (all fuel) - single-loop operation 7.1.2 Steam Dome Pressure Safety Limit 1325 psig Pressure Safety Limit 7.2 Limiting Conditions for Operation Reference 9.9 72 1 J m*l m Averaoe Planar Linear Heat Generation Rate GE9-Fuel ATRIUM-9E3 Fuel MAPLHGR Limits MAPLHGR Limits Average Planar Exposure MAPLHGR (GWd/MTU) (kWtft) < To be furnished by CornEd. >
0.0 13.5 20.0 13.5 61.1 9.39 Single Loop Operation MAPLHGR Multiplier Reference 9.9 for SPC Fuel is 0.90 7.2.2 Minimum Critical Power Ratio 1
Rated Conditions MCPR Limit 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
EMF-2276 LaSalle Unit I Cycle 9 Revision 1 Reload Analysis Paqe 7-2 R... l .. a ... .. ,r ..
Power Dependent MCPR Limits:
Base Case Operation - TSSS Insertion Times Figures 5.3 & 5.4 Base Case Operation - NSS Insertion Times Figures 5.5 & 5.6 EOD and EOOS Operation Tables 5.1-5.4 7.2.3 Linear Heat Generation Rate Reference 9.1
ý ATRIUM-9B -Fuel- GE9 Fuel Steady-State LHGR Limits Steady-State LHGR Limits Average Planar Exposure LHGR (GWd/MTU) (kWVft) < To be furnished by CornEd. >
0.0 14.4 15.0 14.4 61.1 8.32 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.
LHGRFAC1 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 Figure 5.2 LHGR.FACp Tables 5.1-5.4 MAPFAC Multipliers for Off-Rated Conditions - GE9 Fuel:
MAPFACt < To be furnished by CornEd. >
MAPFACP < To be furnished by CornEd. >
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 Page 7-3 Reload Ana ys -
"*'3 20 (0.19.4) (15.19.4) 18 16 14 12 (61.1.11) cz C, 10-3 8
6 4
2-I - I 0 i V
!55 s0 65 770 0 5 10 15 20 25 30 35 40 45 50 Average Planar Exposure, GWd/MTU Figure 7.1 Protection Against Power-Transient LHGR Limit for ATRIUM-9B Fuel Siemens Power Corporation
EMF-2276 Revision 1 LaSalle Unit 1 Cycle 9 Page 8-1 Reload Analysis 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 Programfor Boiling Water Reactor TransientAnalyses, Advanced Nuclear Fuels Corporation, August 1990.
8.2 ANF-524(P)(A) Revision 2 and Supplement 1 Revision 2, ANF CriticalPower 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 CriticalPowerCorrelation, Advanced Nuclear Fuels Corporation, April 1990.
8.4 EMF-1 125(P)(A), Supplement 1 Appendix C, ANFB CriticalPower 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 CriticalPower 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 NuclearFuels Methodology for Boiling Water Reactors:
Benchmark Results for CASMO-3G/MICROBURN-B CalculationMethodology, Advanced Nuclear Fuels Corporation, November 1990.
8.7 EMF-CC-074(P)(A) Volume 1, STAIF - A Computer Programfor BWR StabilityAnalysis in the FrequencyDomain, and Volume 2, STAIF - A Computer Programfor BWR Stability Analysis in the Frequency Domain - Code Qualification Report, Siemens Power Corporation. July 1994.
Siemens Power Comnralinn
EMF-2276 Revision1 Paoe 9-1 LaSalle Unit 1 Cycle 9 Analysis Reload 9.0 Additional References M
Unit I Cycle 9 ATRIU *'-9B 9.1 EMF-2249(P) Revision 1, Fuel Design Report for LaSalle Fuel Assemblies, Siemens Power Corporation, September 1999.
Advanced NuclearFuels 9.2 ANF-89-014(P)(A) Revision I and Supplements 1 and 2, Fuels 9x9-IX and 9x9-9X CorporationGeneric MechanicalDesign for Advanced Nuclear 1991.
BWR Reload Fuel, Advanced Nuclear Fuels Corporation, October Analysis, Siemens Power 9.3 EMF-2277 Revision 1, LaSalle Unit I Cycle 9 Plant Transient Corporation, October 1999.
Fuel, LaSalle Units I and 2 9.4 EMF-95-134(P), CriticalitySafety Analysis for ATRIUM'm-9B 1995.
New Fuel Storage Vault, Siemens Power Corporation, December Fuel,LaSalle Unit 1 Spent 9.5 EMF-96-117(P), CriticalitySafety Analysis for ATRIUM'm-9B Corporation, April 1996.
Fuel Storage Pool (BORAL Rack), Siemens Power Fuel, LaSalle Unit 2 Spent 9.6 EMF-95-088(P), CriticalitySafety Analysis for A TRIUM*-9B February 1996.
Fuel Storage Pool (Boraflex Rack), Siemens Power Corporation, Domain (EOD) and Equipment 9.7 EMF-95-205(P) Revision 2. LaSalle Extended Operating Siemens Power Out of Service (EOOS) Safety Analysis for ATRIUM-*'9B Fuel,
-G-orpor-atierk,-June -1996, Units I and 2, Siemens 9.8 EMF-2174(P), LOCA Break Spectrum Analysis for LaSalle Power Corporation, March 1999.
for ATRIUMTM-9B Fuel, 9.9 EMF-2175(P), LaSalle LOCA-ECCS Analysis MAPLHGR Limits Siemens Power Corporation, March 1999.
Unit 1 Cycle 9 Mechanical 9.10 Letter, D. E. Garber (SPC) to R. J. Chin (CoinEd), "LaSalle Limits for GE9 Fuel." DEG:99:213, August 4, 1999.
Reporting for the
- 9.11 Letter, D. E. Garber (SPC) to R. J. Chin (CoinEd), "10 CFR 50.46 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 Revision 1 LaSalle Unit 1 Cycle 9 Page 1-4 Plant Transient Analysis
.110 100 90 80 70 a)
"0 60 "06 "w
cr 50 40 30 20 10 0
0 10 20 30 40 50 60 70 80 90 100 110 120 Percent of Rated Flow Figure 1.1 LaSalle County Nuclear Station Power / Flow Map Siemens Power Corporation
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 2-9 DI _r ý;.-t Anal sis a ATPJUM-gB- 10275% Max Flow 4 GE9- 1025% MaxFko
- MCPfUnrrit- 102.5% Max Fow 0 GE -105% MaxFlow 1.80 "AATPJUM-gB - 105% Max Flow MC U it- 105% Max F.o .........
1.70' 1.60 i,I S IM 1.40 QI 1.30 1.20 1.10 -
0 20 40 60 80 100 120 Flow (% of Rated) 102.5% 105%
Maximum Core Flow Maximum Core Flow MCPRr MCPR1 GE9 GE9 Flow MCPR, (penalty MCPR l(penalty
(% rated) ATRIUM-9B included) ATRIUM-9B included) 0 1.85 1.85 1.93 1.93 1.85 1.85 1.93 1.93 30 1.11 1.11 1.14 1.14 102.5 1.11 _ 1.11 1.11 1.11 105 Figure 2.1 Flow-Dependent MCPR Limits for Manual Flow Control Mode Siemens Power rnrn^,n,;.-
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 2-10 Plant Transient Analysis 1,,
0:: 0.8' (5
-J 40 50 60 70 Percent of Rated Flow Flow
(% rated) LHGRFACt 0 0.69 30 0.69
.76 1.00 105 1.00 Figure 2.2 Flow-Dependent LHGRFAC Multipliers for ATRIUM-9B Fuel Siemens Power Corporation
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 3-9 Table 3.1 LaSalle Unit 1 Plant Conditions at Rated Power and Flow Reactor Thermal Power 3489 MWth Total Core Flow 108.5 Mlbm/hr Core Active Flow 93.8 Mlbm/hr Core Bypass Flow 14.7 Mlbm/hr Core Inlet Enthalpy 523.9 Btu/Ibm Vessel Pressures Steam Dome 1001 psia Core Exit (upper plenum) 1013 psia Lower Plenum 1038 psia TCV Inlet Pressure 948 psia Feedwater/Steam Flow 15.145 Mlbm/hr Feedwater Enthalpy 406.6 Btu/lbm Recirculating Pump Flow (per pump) 15.83 Mlbm/hr Core Average Gap Coefficient (EOC) 1173 Btu/hr-ft 2 -°F "Includeswater channel flow.
Sin*mnw Power Carporntion
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-10 jant ndII rans en -P Ana YSi Table 3.2 Scram Speed Insertion Times Control Rod Position TSSS Time NSS Time (sec) (sec)
(Notch) 48 (full-out) 0.000 0.OQO 0.200'* 0.200*
48*
0.430 0.380 45 0.860 0.680 39 1.930 1.680 25 3.490 2.680 5
3.880 2.804 0 (full-in)
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
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 3-1 1 Table 3.3 EOC Base Case LRNB Transient Results Pea k Peak Power'/ ATRIUM-9B ATRIUM-9B GE9 Neutron Flux Heat Flux Flow ACPR LHGRFACp ACPR M%rated) (% rated)
TSSS Insertion Times 100/105 0.325 1.000 0.362 438.8 123.7 100/81 0.337 1.000 0.377 460.2 126.5 80/1 05 0.332 1.000 0.368 367.6 98.2 80/57.2 0.377 1.000 0.410 323.7 99.3 60/105 0.319 1.033 0.349 253.1 72.3 60/35.1 0.301 1.098 0.289 135.0 67.1 40/105 0.260" 1.113 0.271 106.0"" 45.6" 25/105 0.191*° 1.202 0.177" 44.9*
- 27.0" 23.81/105 0.186" 1.211 0.171 41.6"* 25.6"*
20/105 NDS 1.008 0.706 0.980 44.6 38.1 NSS Insertion Times 100/105 0.304 1.000 0.338 401.1 121.3 100/81 0.288 1.000 0.323 409.5 122.0 801105 0.317 1.009 0.351 347.5 96.9 80/57.2 0.278 1.014 0.306 256.1 94.2 60/105 0.309 1.038 0.338 245.9 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
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-12 V "II I Q iI
" I IJ.f'J u '
Table 3.4 EOC Base Case FWCF Transient Results Peak Neutron Peak ATRIUM-9B GE9 . Flux Heat Flux Power*/ ATRIUM-9B ACPR (% rated) (% rated)
Flow ACPR LHGRFACP TSSS Insertion Times 100/105 0.299 1.019 0.322 371.3 122.6 1.032 0.301 303.1 121.8 100/81 0.280 0.355 0.986 0.376 327.2 101.9 80/105 1.063 0.310 203.7 96.0 80/57.2 0.294 0.431" 0.955 0.440 218.3"* 79.3**
60/105 0.251 1.143 0.252 104.7 67.0 60/35.1 0.582"" 0.8914" 0.546"* 128.0** 58.8**
40/105 0.767"" 0.913"* 62.5"* 43.2**
25/105 0.884""
0.750"m 0.964*
- 61.1. 42.2*
23.811105 0.936**
0.688 1.029 70.2 43.9 20/105 NDS 1.119 NSS Insertion Times 100/105 0.280 1.033 0.301 342.9 120.5 0.341 1.000 0.360 312.3 100.8 80/105 0.959 0.430 229.7. 79.9 60/105 0.417 0.895'* 0.535' 124.6"' 58.5*
40/105 , 0.570*
0.861
- 0.777"" 0.871" 76.7"* 44.0' 25/105 0.760"" 0.923'* 73.8'* 42.9""
23.81/105 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 LaSalle Unit 1 Cycle 9 Page 3-13 Plant Transient Analysis Table 3.5 Input for MCPR Safety Limit Analysis Fuel Related Uncertainties Source Statistical Parameter Document Treatment ANFB Correlation*
ATRIUM-9B Reference 17 Convoluted GE9 Reference 12 Convoluted Radial Power Reference 16 Convoluted Local Peaking Factor Reference 5 Convoluted Assembly Flow Rate Reference 5 Convoluted (mixed core)
Channel Bow Local Function of nominal and bowed Convoluted Peaking local peaking and standard deviation of bow data (see Reference 18).
Nominal Values and Plant Measurement Uncertainties Uncertainty (%) Statistical Parameter Value (Reference 8) Treatment Feedwater Flow Rate`' (Mlbm/hr) 22.7 1.76 Convoluted Feedwater Temperature (fF) 426.5 0.76 Convoluted Core Pressure (psia) 1031.35 0.50 Convoluted Total Core Flow (Mlbm/hr) 113.9 2.50 Convoluted Core Power` (MWth) 5232.35
° 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 LaSalle Unit 1 Cycle 9 Page 3-14 Plant Transient Analysis Table 3.6 Flow-Dependent MCPR Results 102.5% Maximum Core Flow 105% Maximum Core Flow Core Flow ATRIUM-9B GE9 ATRIUM-9B M%rated) GE9 1.775 1.821 1.866 1.914 30 1.693 1.711 1.761 40 1.645 1.552 1.597 1.604 1.649 50 1.501 1.505 1.543 60 1.464 1.406 1.412 1.439 70 1.379 1.308 1.322 1.336 80 1.295 1.214 1.232 1.237 90 1.209 1.129 1.149 1.149 100 1.129 102.5 1.110 1.110 7 1.110 1.110 105 1.110 1.110
-- fl...... V'
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-15 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 LaSalle Unit 1 Cycle 9 Page 3-16
-. ~A Mant I ransient "0 I Y. ý 0
w N
I-I,)
z 0
z_
-J LiP ILd
-J V)
V) w 5.0 TIME, SECONDS Figure 3.2 EOC Load Rejection No Bypass at 100/81 - TSSS Vessel Water Level qipmernv Pnwer Corormtion
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-17 Dis,,t Tr~nci~nt Andlv_*iS DI i Trýneiont nalvsis (L
LiJ
-:2 U)
W 0
0~
TIME, SECONDS Figure 3.3 EOC Load Rejection No Bypass at 100/81 - TSSS Dome Pressure Siemens Power Corporation
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-18 Pl=. Trpne;*nf A.n~lvsis 01 -r-s-ent nalysis Arflo CORE POWER HEAT FLUX CORE FLOW STEAM FLOW FEED FLOW I-3W.
3OO.O
-- 200.0 LL 0
z
() 10.0 L--r-------- --i ----I --I- --
I
.0 -1 11
-100.0 1 I
.0 5.0 10.0 15.0 20.0 25.0 TIME, SECONDS Figure 3.4 EOC Feedwater Controller Failure at 100/105 - TSSS Key Parameters Siemens Power Corporation
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-19 Plant Transient Analysis 0
Ir LI N
I.-
cn z
W 0
im
-J W
Li I-hin
-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 LaSalle Unit 1 Cycle 9 Page 3-20 Pl- 'r -i-ent Analysis a.
(n (n
w En w
0 0
TIME. SECONDS Figure 3.6 EOC Feedwater Controller Failure at 100/105 - TSSS Dome Pressure Siemeins Pnwpr rnrnnrtinn
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Pi~ni Tr~n~tpn? Anilysis Page 3-21 Plant Transient Analysis 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 1.2 1.3 1.4 1.5 1.6 Rodiol Power Peaking Figure 3.7 Radial Power Distribution for SLMCPR Determination Sionions Power Corooralion
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-22 D1 'rU ; --
- AnCISIv.
C ontrol Rod Corner 0
n 1.023 1.055 1.068 1.112 1.099 1.102 1.049 1.023 0.977 t
r 0
1.055 0.958 0.894 1.016 0.894 1.007 0.877 " 0.927 1.002 R
0 1.068 0.894 1.031 1.065 1.084 1.056 1.010 0.863 1.011 d
C Internal 1.044 0.980 1.051 1.112 1.016 1.065 0
r n 1.038 e 1.099 0.894 1.084 Water 1.063 0.863 r
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
-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.....-.
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-23 Plant Transient Analysis 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 R
0 1.056 1.025 1.018 1.064 1.081 1.055 0.997 0.989 0.999 d
C Internal 1.043 0.848 1.047 1.110 0.879 1.064 0
r n
e 1.098 1.014 1.081 Water 1.059 0.978 1.035 r
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) q;ame.n* Pnwer itrntoration
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-24 Plant Transient Analysis Control Rod Corner 0
n t 1.022 1.056 1.061 1.035 1.102 1.028 1.045 1.029 0.982 r
0 1.056 0.947 1.018 1.003 0.879 0.997 1.004 0.919 1.011 R
0 1.061 1.018 1.001 1.050 1.081 1.048 0.996 0.992 1.012 d
C Internal 0.926 0.983 0.987 1.035 1.003 1.050 0
r n
e 1.102 0.879 1.081 Water 1.077 0.853 1.049 r
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 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)
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 3-25 0.
U 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPRP Limit Power -(%) MCPRP Limit 100 1.46 100 1.45 84 1.51 80 1.51 63 1.56 60 1.56 25 2.07 25 2.05 25 2.22 25 2.20 0 2.70 0 2.70
- Figure 3.11 EOC Base Case Power-Dependent MCPR Limits for ATRIUM-98 Fuel - TSSS Insertion Times Siemens Power Corporation
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 3-26 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWth) -
3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPRP Limit Power (%) MCPRP Limit 100 1.50 100 1.49 84 1.53 80 1.53 63 1.57 60 1.57 25 2.12 25 2.10 25 2.22 25 2.20 0 2.70 0 2.70 Figure 3.12 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel - TSSS Insertion Times
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-27 PI* nt Transient Analysis Pl2nt Transient Analysis 0 500 1000 1500 2000 2500 3000 3500 4000 Power IMWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPRP Limit Power (%) MCPRp Limit 100 1.43 100 1.42 84 1.48 80 1.48 63 1.54 60 1.54 25 2.07 25 2.05 25 2.22 25 2.20 0 2.70 0 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 LaSalle Unit 1 Cycle 9 Page 3-28 Plant Transient Analysis 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 1000 1500 2000 2500 3000 3500 4000 0 500 Power IMWth) 3323 MWth Rated Power 3489 MWth Rated Power MCPR, Limit Power (%) MCPRp Limit Power (%)
100 1.46 100 1.45 1.51 80 1.51 84 1.56 60 1.56 63 25 2.05 25 2.07 2.22 25 2.20 25 2.70 0 2.70 0
Figure 3.14 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel - NSS Insertion Times
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 3-29 01-* Trnnsient Analysis 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 3489 MWth Rated Power Power (%) LHGRFACP. Power (%) LHGRFACP 100 0.99 100 1.00 0.98 80 0.98 84 0.94 60 0.94 63 25 0.66 25 0.67 0.66 25 0.67 25 0.66 0 0.67 0
Figure 3.15 EOC Base Case Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel - TSSS Insertion Times Siemens Power Corooration
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Dih Trn;ant Ann[lVsis Page 3-30 O"L I a". F C.
U II.
0
-J 0.60 1500 2000 2500 3000 3500 4000 0 500 1000 Power (MWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%M LHGRFACP Power (%) LHGRFACP 100 1.00 100 1.00 84 1.00 80 1.00 63 0.95 60 0.95 25 0.74 25 0.75 25 0.74 25 0.75 0 0.74 0 0.75 Figure 3.16 EDC Base Case Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel - NSS Insertion Times
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 5-5 Table 5.1 EOC Feedwater Heater Out-of-Service Analysis Results ATRIUM-9B GE9 Power*/Flow Event (%ratedl%rated) tCPR LHGRFACP "ACPR FWCF 100/105 0.311 1.000 0.329 FWCF 100/81 0.286 1.034 0.302 FWCF 80/105 0.376 0.968 0.390 FWCF 80/57.2 0.306 1.055 0.320 FWCF 60/105 0.482** 0.927' 0.472 FWCF 60/35.1 0.216 1.123 0.228 FWCF 401105 0.698** 0.831"* 0.677*'
FWCF 25/105 1.142'* 0.6784* 1.156"'
FWCF 23.811105 1.209*' 0.662' 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
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 5-6 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 LaSalle Unit 1 Cycle 9 Plant Tr~n~ip~nt Analysis Page 5-7 Plant Transient Analvsis Table 5.3 EOC Turbine Bypass Valves Out-of-Service Analysis Results ATRIUM-9B GE9 Power*/Flow Event (%rated/%rated) ACPR LHGRFACP ACPR FWCF 100/105 0.359 0.968 0.392 FWCF 100/81 0.359 0.947 0.394 FWCF 80/105 0.418 0.942 0.449 FWCF 80/57.2 0.417 0.957 0.447 FWCF 60/105 0.499,0 0.917 0.514 FWCF 60/35.1 0.327 1.032 0.332 FWCF 40/105 0-658' 0.859" 0.619 FWCF 25/105 0.962°
- 0.750""' 0.952" FWCF 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
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 5-8 Table 5.4 EOC Recirculation Pump Trip Out-of-Service Analysis Results
. ATRIUM-9B GE9 Power"/Flow Event (%ratedl%rated) ACPR LHGRFA.Cp 6CPR LRNB 100/105 0.382 0.909 0.433 LRN8 100/81 0.371 0.868 0.430 LRNB 80/105 0.389 0.923 0.438 LRNB 80157.2 0.391 0.899 0.439 FWCF 100/105 0.353 0.942 0.391 FWCF 100/81 0.308 0.948 0.349 FWCF 80/105 0.403 0.920 0.438 FWCF 80/57.2 0.295 1.003 0.327 FWCF 60/105 0.466 0.901 0.492 FWCF 60/35.1 0.190 1.120 0.193 FWCF 40/105 0.596' 0.857'" 0.581 FWCF 25/105 0.858'" 0.757"" 0.861" FWCF 23.81/105 0.896** 0.743"" 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.
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 5-9 Plant I rans : en
- A-=I cis Table 5.5 EOC Turbine Control Valve Slow Closure Analysis Results ATRIUM-9B GE9 Slow Valve Power*/Flow Characteristics (%rated/%rated) ACPR LHGRFACP ACPR Event 0.461 100/105*
- 0.420 0.902 LRNB 1 TCV closing at 2.0 sec 0.461 100/105°* 0.419 0.8"99 LRNB 1 TCV closing at 2.7 sec 0.233 100/105
- 0.219 1.057 LRNB 2 TCVs closing at 7.75 sec 0.421 1 00/81 0
- 0.369 0.928 LRNB 1 TCV closing at 2.0 sec 0.199 1.107 0.223 1 00/8140 LRNB 2 TCVs closing at 7.75 sec 0.466 80/105*4 0.432 0.911 LRNB 1 TCV closing at 2.0 sec 0.568 80/105t 0.527 0.882 LRNB 2 TCVs closing at 2.0 sec 0.314 80/105** 0.293 1.014 LRNB 2 TCVs closing at 7.75 sec 0.548 80/57.2o 0.504 0.911 LRNB 1 TCV closing at 2.0 sec 0.564 80157.2f 0.520 0.928 LRNB 2 TCVs closing at 2.0 sec 0.305 80/57.2
- 0.277 1.115 LRNB 2 TCVs closing at 7.75 sec, 2 TCVs closing at 2.7 sec 0.461 60/105** 0.432 0.932 LRNB 1 TCV closing at 2.0 sec 0.370 601105°o 0.346 1.001 LRNB 2 TCVs closing at 7.75 sec 0.591 0.991 0.5B5 60/35.1'"
LRNB 1 TCV closing at 2.0 sec 0.824 4011 O5t 0.8281 0.759$
LRNB 1 TCV closing at 2.0 sec 0.9921 0.707 0.9841 25/105i LRNB 1 TCV closing at 2.0 sec 1.011 0.6991 1.0081 23.81/105 t
"-LRNB 1 TCV closing at 2.0 sec 0.977 0.721 0.954 23.81/1051 LRNB 2 TCVs closing at 7.75 sec 100/105*° 0.363 "0.944 0.396 LRNB w/ 1 TCV closing at 2.0 sec FHOOS 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 LaSalle Unit 1 Cyclit 9 Revision 1 Plant Transient Analysis Page 5-10 Table 5.6 EOC Recirculation Pump Trip and Feedwater Heater Out-of-Service Analysis Results ATRIUM-9B GE9 Power4 /Flow Event (%rated/%rated) ACPR LHGRFACG " ACPR LRNB 100/1 05 0.332 0.954 0.375 LRNB 100/81 0.305 0.948 0.354 FWCF 100/105 0.358 0.933 0.391 FWCF 100/81 0.307 0.968 0.342 FWCF 80/105 0.419 0.911 0.448 FWCF 80/57.2 0.300 1.007 0.329 FWCF 60/105 0.508` 0.882 0.523 FWCF 60/35.1 0.212 1.104 0.226 FWCF 40/105 0.705' 0.804" 0.664 FWCF 25/105 1.073'* 0.673"* 1.092" FWCF 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.
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 5-11 CL a.
UE 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWIh) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPR. Limit Power (%) MCPRP Limit 100 1.47 100 1.45 63 1.62 60 1.62 25 2.38 25 2.35 25 2.38 25 2.35 0 2.85 0 2.85 Figure 5.1 EOC Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 5-12 0
am I 0"a " I C.
C.)
II 0
-J 500 1000 1500 2000 2500 3000 3500 4000
-- 0 Power IMWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) LHGRFACP Power (%) LHGRFACP 100 0.99 100 1.00 63 0.90 60 0.90 25 0.64 25 0.65 25 0.64 25 0.65 0 0.64 0 0.65 Figure 5.2 EOC Feedwater Heaters Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis "Page 5-13 C.)
0 500 1000 1500 2000 2500 3000 3500 4000 Power iMWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPRP Limit Power (%) MCPRP Limit
-4 100 1.51 100 1.49 63 1.62 60 1.62 25 2.38 25 2.35 25 2.38 25 2.35 0 2.85 0 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 LaSalle Unit 1 Cycle 9 Page 5-14 DI*.,.. Tanci~fnt Analysis V ICEL I 2.65 2.55 2.4S 2.35 2.25 2.15 2.05 C. 1.95 C.
1.8s 1.75 1.65 1.55
[ Idle Loop Restart 1.45 SOLMCPR I 1.35 1.25 1.15 3000 3500 4000 2500 3500 500 1000 1500 2000 2500 3000 0
Power (MWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPRp Limit Power (%) MCPRp Limit 100 2.54 100 2.54 2.54 2.54 60 63 2.54 2.54 25 25 2.54 2.54 25 25 2.54 2.54 0 0
Figure 5.4 Abnormal Idle Recirculation Loop Startup Power-Dependent MCPR Limits for ATRIUM-9B Fuel
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Plant Transient Analysis Page 5-15 1.30 1.25 1.20 1.15 IdeLoop Restart 1.10 1.05s 1.00 2 0.95 C. 0.90 LL. 0.85 w
3 0.80 ALGFAP
"- 0.75 0.70 0.65 0.60 0.55 0.50.
0.45 0.40 0.35 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWlh) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) LHGRFACp Power (%) LHGRFACP 100 0.40 100 0.40 63 0.40 60 0.40 25 0.40 25 2 0.40 25 0.40 25 0.40 0 0.40 .0 . 0.40 Figure 5.5 Abnormal Idle Recirculation Loop Startup Power-Dependent LHGR Multipliers for ATRIUM-SB Fuel Siemens Power Corporation
EMF-2277.
LaSalle Unit 1 Cycle 9 Revision 1 Plnnr Tr~nn.ient Analysis Page 5-16 Plant Transient 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 IdlLLopestart]
_OMPRJ 1.25 1.15 0 500 1000 is00 2000 2500 3000 3500 4000 Power (MWth) 3323 MWth Rated Power. 3489 MWth Rated Power Power (%) MCPRp Limit Power (%) MCPRp Limit 100 2.54 100 2.54 63 2.54 60 2.54 25 2.54 25 2.54 25 2.54 25 2.54 0 2.54 0 2.54 Figure 5.6 Abnormal Idle Recirculation Loop Startup Power-Dependent MCPR Limits for GE9 Fuel
EMF-22771 Revision LaSalle Unit 1 Cycle 9 Page 5-17 Plant Transient Analysis 2.75 2.65
- FWCF 2.55 Power Uprate OLMCPR
.... Pre-Power Uprate OLMCPR 2.45 2.35 2.25' 2.15' 2.05' C.
- a. 1.95 1.85 1.75' 1.65 1.55' 1.45' 1.35' 1.25' 1.15 3000 3500 2500 4000 1500 2000 2500 3000 3500 0 500 1000 Power (MWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPRP Limit Power M%) MCPRP Limit 100 1.49 100 1.47 60 1.63 63 1.63 25 2.15 25 2.17 25 2.20 25 2.22 0 2.70 0 2.70 Figure 5.7 EOC Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits for ATRIUM-SB Fuel
EMF-2277 PeM 5-2271 LaSalle Unit 1 Cycle 9 Page 5-18 Plant Transient Analysis 1.30 1.25 1.20 1.15 1.10 1.05 0.) 1.00
.J 0.95
,-I 0.90 0.85 0.0 0.75 0.70 0.65 0.60 2500 3000 3500 4000 0 500 1000 1500 2000 Power (MWIh) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) LHGRFACo Power (%) LHGRFACp Power (%)
100 0.94 100 0.94 0.90 60 0.90 63 0.66 25 0.67 25 0.66 25 0.67 25 0.66 0 0.67 0
Figure 5.8 EOC Turbine Bypass Valves Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 5-19 Plant Transient Analysis 2.75 2.65 2.55 2.45 2.35 2.25 2.15 2.05 I. 1.95 1.55 1.75 1.65 1.55 1.45 1.35 1.25 1.15 1500 2000 2500 3000 3500 4000 0 500 1000 Pow.r (MWIh) 3323 MWth Rated Power 3489 MWth Rated Power MCPRP Limit Power (%) MCPRP Limit Power (%)
100 1.53 100 1.51 1.65 60 1.65 63 2.17 25 2.15 25 2.22 25 2.20 25 2.70 0 2.70 0
Figure 5.9 EOC Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel
.. , ... t ri' leNr l tnn
EMF-2277 Revision I LaSalle Unit 1 Cycle 9 Plant Transient Analysis Page 5-20 2.75 2.65 2.55 2.45 2.35 2.25 2.15 2.05 a.
1.95 1.,5 1.75 1.65 1.55 1.45 1.35 1.25 1.15 3000 3500 4000 0 500 1000 1500 2000 2500 Power IMWIh) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPRP Limit Power 1%) MCPRP Limit 100 1.51 100 1.50 63 1.60 60 1.60 25 2.07 25 2.05 25 2.22 25 2.20 0 2.70 0 2.70 Figure 5.10 EOC Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel
EMF-2277 Revision I LaSalle Unit 1 Cycle 9 Page 5-21 Plant Transient Analysis 3000 3500 4000 1000 1500 2000 2500 0 500 Power (MWth) 3489 MWth Rated Power 3323 MWth Rated Power LHGRFACP Power (%)
Power (%) LHGRFACP 100 0.86 100 0.86 60 0.86 63 0.86 0.67 0.66 25 25 25 0.67 25 0.66 0 0.67 0 0.66 Figure 5.11 EOC Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel Siemens Power Corporation
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Pl~ni T r'nc:;pnt Analysis Page 5-22 Pinnt Trmnsient Analysis 0,
U 3000 -3500 4000 0 500 1000 1500 2000 2500 Power (MWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPRV Limit Power (%M MCPRP Limit 100 1.56 100 1.55 63 1.63 60 1.63 25 2.12 25 2.10 25 2.22 25 2.20 0 2.70 0 2.70 Figure 5.12 EOC Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel
t EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Plant Transipnt Analysis Page 5-23 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 3489 MWth Rated Power Power (%) MCPRP Limit Power (%) MCPRP Limit 100 1.56 100 1.54 84 1.63 80 1.63 84 1.65 80 1.65 25 2.16 25 2.15 25 2.22 25 2.20 0 2.70 0 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 LaSalle Unit 1 Cycle 9 Plant Transient Analysis Page 5-24 1.30 1.25 . Slow TCV Closure
-Power Uprale LHGRFACp 1.15
..... LHGRFACo Pre-Power Uprale 1.10 1.05 u0. 1.00 C: 0.95
-J 0.90
- C Ue 0.85 0.50 0.75 0.70 0.65 0.60 0 500 1000 1500 2000 2500 3000 3500 4000 Power IMWth) 3323 MWth Rated Power 3489 MWth Rated Power Po*wer (%) LHGRFACP Power (%) LHGRFACP LHGRFACp Power (%)
100 0.86 100 0.86 84 0.86 80 0.86 84 0.86 80 0.86 25 0.66 25 0.67 25 0.66 25 0.67 0 0.66 0 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
1 5-25 EMF-2277 EMF-2277 Revision LaSalle Unit 1 Cycle 9 Revision Page 1 Plant Transient Analysis Page 5-25 2.75 2.65 2.55 2.45 2.35 2.25 2.15 Z.
2.05 C. 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 Power 1%) MCPRP Limit Power (%) MCPRP Limit 100 1.60 100 1.58 84 1.67 80 1.67 84 1.69 80 1.69 25 2.16 25 2.15 25 2.22 25 2.20 0 2.70 0 2.70 S.
Figure 5.15 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel
-. ..... n.)...
. UA rnt*iln
.i'% ,tin
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Paqe 5-26 C
cc CL 0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWth) 3323 MWth Rated Power 3489 MWth Rated Power
.Power (%) MCPRP Limit Power (%) MCPRP Limit 100 1.56 100 1.54 84 1.63 80 1.63 84 1.65 80 1.65 25 2.38 25 2.35 25 2.38 25 2.35 0 2.85 0 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 LaSalle Unit 1 Cycle 9 Page 5-27 Plant Transient Analysis U) a'.
=,
0 500 1000 1500 2000 2500 3000 3500 4000 Power (MWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) LHGRFACP Power (%) LHGRFACP 100 0.B6 100 0.86 84 0.86 80 0.86 84 0.86 80 0.86 25 0.63 25 0.64 25 0.63 25 0.64 0 0.63 0 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 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 5-28 CL 0 S00 1000 1500 2000 2500 3000 3500 A000 Power (MWth) 3323 MWth Rated Power 3489 MWth Rated Power Power (%) MCPRP Limit Power (%) MCPRP Limit 100 1.60 100 1.58 84 1.67 80 1.67 84 1.69 80 1.69 25 2.38 25 2.35 25 2.38 25 2.35 0 2.85 0 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 LaSalle Unit 1 Cycle 9 Page 7-2 Plant Transient Analysis Table 7.1 ASME Overpressurization Analysis Results 102%P/105%F Peak Peak Maximum Maximum Vessel Pressure Dome Neutron Heat Lower Plenum Pressure Event Flux Flux (psig) S(psig)
(%rated) (%rated) 1320.26 1291.12 MS IV 425.43 135.28 1318.41 1287.56 70S.96 142.84 TCV 142.85 1318.41 1287.55 TSV 710.29 s'ininiS Pnwtir Coriorrntinn
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Page 7-3 I~lnt~ T;r~ncipnt Analvsis I1n- Trn~sian Analysis 0
Li I.-
W.
0 I
z 0d 0.
.0 .0 2.0 3.0 4.0 5.0 TIME, SECONDS "Figure 7.1 Overpressurization Event at 102/105 MSIV Closure Key Parameters Siomons Power Corporation
EMF-2277 Revision 1 LaSalle Unit 1 Cycle 9 Plant Trans*ient Analysis Page 7-4.
Plant Transient Analysis 0
N id 0
z w-J 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 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 7-5 a.
D:
V)
(/L z
0.
.0 Idnj
.0 1.0 2.0 3.0 4.0 5.0 TIME, SECONDS
=Figure 7.3 Overpressurization Event at 102/105 MSIV Closure Lower Plenum Pressure Siemens Power Corporation
EMF-2277 LaSalle Unit 1 Cycle 9 Revision 1 Plant Transient Analysis Page 7-6 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
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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 I VESSEL PRESS RISEP![ I) 2 SAFElY VALVE FLOW 3 RELIEF VALVE FLOW a DYDACC VAIE *i nu 19.10-.
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Figure 1. LRNBP from Rated Power, All TCV Fast Closure, Direct Scram, EOC-RPT LaSalle Unit I Cycle 9 November 1999
Administrative Technical Requirements - Appendix A LIC9 Reload Transient Analysis Results C
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Figure 2. LRNBP from Rated Power, One TCV Slow Closure(50%/second)/Three TCV Fast Closure, Flu>
Scram, EOC-RPT OOS LaSalle Unit I Cycle 9 November 1999
Administrative Technical Requirements - Appendix A LIC9 Reload Transient Analysis Results I NEUTRON FLUX I YES EL PRESS RISECPSI) 2 AVE SURFACE HEAT FLUX 2 SAF TY VALVE FLOW 5 COR INLET FLOW 3 REL EF VALVE FLOW
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Administrative Technical Requirements - Appendix A LIC9 Reload Transient Analysis Results
. NEU RON FLUX I VES 2 AVE SURFACE HEAT FLUX 2 SAF SELYElPRESS RISEIPSI) 5 COR INLET FLOW 3 REL [EF VALVE FLOW VALVE FLOW A QVD lee vii uC rl Iu a
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Figure 4. LRNBP from 50% Power, All TCV Closure at 19%/second, Pressure Scram LaSalle Unit I Cycle 9 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, IL60515-5701
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 Tel: (509) 375-8100 Richland, WA 99352 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.
..............-... .I.~. a IWIJAMLIL y DEG:01:045 Attachment Page A-2 Figures 1 and 2 present the revised base case MCPRp limits for the ATRIUM*-9B* 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
. *- - - fv 1%-8 08" lit o p i ary 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 NuclearStation 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 ReloadAnalysis, 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 OperatingDomain (EOD) and Equipment Out T -9B Fuel, Siemens Power Corporation,
' bj~~e196.
oflService (EOOS) Safety Analysis forATRlUMU
- 8. EMF-96-1 89 Revision 0, LaSalle Unit I Cycle 9 PrincipalTransientAnalysis 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.
. . .ý46
. . . . . . . . o *,omalA iim.. Fa I6 ,P*I 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 r-ir l l 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
/Flow ATRIUM-9B ATRIUM-9B GE9 Neutron Flux Heat Flux 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 / 10 5t 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 Attachment Page A-8 2.7 26 2.5 2.4 2Z3' 2.2 21!
2.1O=
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a- 1.95 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
r aup. ftasy DEG:01:045 Attachment Page A-9 2.7:
2.35 2Z25 2.15 2.05 e,
9I.95 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 Attachment Page A-10 1.30 1.20 1.10 CL 1.00 U
-' 0.90, 0.60 1000 1500 2000 2500 3000 3500 4000:.
0 500
.. Pow(".ft)
Power LHGRFACp
(%) Multiplier 100 1.00 80 0.98 60 0.94 25 0.67 25 0.67 0 0.67 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- KIC*hland, Inc. Proprietary DEG:01:045 Attachment Page A-11 2.95 2.85 2.75
.65 2.55 2.45 2.35 2.25 2.15 2.05-0 500 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; ?*iPvr matIiIsdiU, InC. rroprletary DEG:01:045 Attachment Page A-12 2.9*
2-W 2.7.=
2.65 255 2.45 2.35 2.25 06 2.15 UE 0 500 1000 1500 2000 2500 3000 3500 4000 PMWM jO Power MCPRp
(%) Limit 100 1.58 80 1.69 80 1.71 25 2.35 25 2.35 0 2.85 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 Attachment Page A-13 1
1*.
1.
1.1*
1.
1.*
0 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 Tel: (509) 375-8100 Richland, WA 99352 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 100°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 1 through_6.
_limits-and LHGlRFACpmultipliers provided-in.Figures_
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 PrincipalTransientAnalysis 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*
AA 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.
Framatome ANP, Inc. Proprietary DEG:01:148 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 t
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.
Framatome ANP, Inc. Proprietary DEG:01:148 Attachment Page A-7 2.95 2.85
- LRNB 2.75
- OLMCPR 2.65 2.55 2.45 2.35 2.25 2.15 a.
i.IL, 2.05 1.95 1.75 a 1.45 1.15 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
Framatome ANP, Inc. Proprietary DEG:01:148 Attachment Page A-8 2.95 285 2.75 2.65*
2.55*
2.45 2.35 2.25 2,15 "C.) 2.05 1.95 1.85 1.75 1.65 1.55 1.45 1.35 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 Attachment Page A-9 1.30 1.25
- LHIGRFACp 1.20 OL 1.00 U U.
S0.95
-J 0.90 a
0.85 0.80 0.75 0.70 a
0.65 0.60 10 20 30 40 50 60 70 80 90 100 .110 Power (% rated)
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
Framatome ANP, Inc. Proprietary DEG:01:148 Attachment Page A-10 2.85
- Slow TCV Closure
- LRNB No RPT 2.75
S 2.05 1.95 1.85 1.75 1.65 1.55 1.45 1.35 1.25 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
Framatome ANP, Inc. Proprietary Attachment DEG:01:148 Page A-1I 2.
- Saw TCV Closure 2.85
- LRNB No RPT 2 .75
.45 2.35 2-25 U
0j 1.85 1.75 1.65 1.55 1.45 1.35 1.25
- 1. 1;)
60 70 80 90 100 110 0 10 20 30 40 50 Power (* rated)
Power MCPRp
(%) Limit 100 1.61 80 1.69 80 1.71 25 2.35 25 2.35 0 2.85 Figure 5 FFTR/Coastdown Slow TCV Closure and/or No RPT Power-Dependent MCPR Limits for GE9 Fuel
Framatome ANP, Inc. Proprietary DEG:01:148 Attachment Page A-12 1.25
- Slow TCV Closure
- LRNB No RPT 1.20
-LHGRFACp 1.15 1.10 1.05
- 0. 1.00, S0.95 (9
-I
-' 0.90-0.85 0.80.
0.75 0.70 0.65
-A-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