ML023430481
| ML023430481 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 11/30/2002 |
| From: | Barnes G Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML023430481 (205) | |
Text
Technical Requirements Manual Appendix J (Amendment 51)
LaSalle Unit 2 Cycle 9A Core Operating Limits Report and Reload Transient Analysis Results November 2002
Technical Requirements Manual - Appendix J Section 1 LaSalle Unit 2 Cycle 9A Core Operating Limits Report November 2002
Technical Requirements Manual Appendix J Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Issuance of -Changes Summary November 2002 LaSalle Unit 2 Cycle 9A i
Technical Requirements Manual Appendix J -
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Table of Contents References ...................................................... il
- 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. Minimum 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 Withdrawal Block Instrumentation (3.3.2.1) ................. 4-1 4.1 Tech Spec Reference ................................................................................ 4-I 4.2 Description ................................................................................................. 4-I
- 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 2 Cycle 9A ii November 2002
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report References
- 1. Letter from D.M.Crutchfield to Ail Power Reactor Licensees and Applicants, Generic Letter 88-16; Concerning the Removal of Cycle-Specific Parameter Umits from Tech Specs, dated October 4, 1988.
- 2. LaSalle Unit 2 Cycle 9 Neutronlcs LicensinaFleoort (NLRI. NFM ID#0000115, October 2000.
- 3. LaSalle Unit 2 Cycle 9 Reload Analysis. EMF-2437, Revision 0. October 2000.
- 4. LaSalle Unit 2 Cycle 9 Plant Transient Analysis, EMF-2440, Revision 0, October 2000.
- 5. LOCA Break S ectrum Analysis fo LaSalle Units 1 and 2. EMF-2174(P), March 1999.
- 6. LaSalle LOCA-ECCS Analysis MAPLHGR Umits for ATRIUM-SB fuelW EMF-2175(P), March 1999.
- 7. LaSalle Extended Ooerating Domain (EOD) and Eouliment Out of Service (EOOS) Safety Analysis for ATRIUM-9B Fuel.
EMF-95-205(P), Rev. 2, June 1996.
and
- 8. ARTS Improvement Program analysis for Lasalle County Station Units I and 2, NEDC-31531P, December 1993 Supplement 1, June 1998 (Removal of Direct Scram Bypassed Umit).
0,
- 9. Lattice-Dependent MAPLHGR Relort for LaSalle County Station Unit 2 Reload 6 Cycle 7, 24A5162AA, Revision December 1994.
- 10. "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.
- 11. Evaluation of a Postulated Slow Turbine Control Valve Closure Event for LaSalle County Station. Units ! and 2. GE-NE 187-13-0792 Revision 2, July 1998.
- 12. Transient Analysis Evaluation for LaSalle 3 TCV Operation at Power Uprate and MELLLA Conditions, NFM:BSAkO0-025, R.W. Tsal to D.Bost, April 13,2000.
I
- 13. 'Updated Transient Analysis: Abnormal Start-up of an Idle Recirculation Loop for LaSalle County Nuclear Station. Units Unit 2 Cycle 8 Abnormal Idle Recirculation Loop Startup Analysis",
and 2", B33-00296-03P, March 1998 and "LaSalle DEG:99:070, D.Garber to R. Chin, March 8, 1999.
to
- 14. 'TIP Symmetry Testing", J.H. Riddle to R.Chin. January 20,1997 and 'TIP Symmetry Testing', DEG:99:085, D. Garber R. Chin. March 8, 1999
- 15. "POWERPLEX41 CMSS Startup Testing'. DEG:00:254, D. Garber to R. Chin, December 5, 2000.
- 16. "On-Site and Off-Site Reviews of the GE Turbine Control Valve Slow Closure Analysis, T.Rleck to G.Spedl. NFS:BSS:93 117. May 19, 1993.
April 6, 1995.
- 17. *LaSalle Units I and 2 Operating Limits with Multiple Equipment Out of Service (EOOS), NFS:BSA:95-024,
- 18. NFM Calculation No. BSA-L-99-07. "MAPFACf Thermal Limit Multiplier for 105% Maximum Core Flow.'
- 19. "CornEd GE9IGE10 LHGR Improvement Program" J11-03692-LHGR, Revision 1, February 2000.
Revision 1,
- 20. *LaSalle County Station Power Uprate Project, Task 201: Reactor Power/Flow Map, GE-NE-A1300384-07-01, September 1999.
October 2000.
- 21. "Evaluation of CBH Effects on Fresh Fuel for LaSalle Unit 2 Cycle 9", DEG:00:232, D.Garber to R. Chin, Uncallbrated
- 22. DEG:00:091, *Revised Measured Nodal Power Distribution Uncertainty for POWERPLEX Operation with LPRMs7. David Garber to Dr. R. J. Chin. April 5. 2000.
- 23. "POWERPLEX-II CMSS Startup Testing", DEG:00:256, D.Garber to R. Chin. December 6. 2000.
ioi November 2002 LaSalle Unit 2 Cycle 9A
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report
- 24. Reactor Stability Detect and Sugpress Solutions Licensing Basis Methodology for Reload Aoolications. NEDO-32465-P-A, August 1996.
- 25. ANFB Critical Power Correlation, ANF-1125(PXA) and Supplements I and 2. Advanced Nuclear Fuels Corporation, Apdl 1990.
- 26. Letter, Ashok C. Thadani (NRC) to R. A. Copeland (SPC), "Acceptance for Referencing of ULTRAFLOWTM Spacer on 9X9 IX/X BWR Fuel Design, July 28,1993.
- 27. Advanced Nuclear Fuels Corooratlon Critical Power Methodology for Boiling Water Reactors/Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing Effects/NRC Correspondence- XN-NF-524(PXA) Revision 2 and Supplement 1 Revision 2 , Supplement 2, Advanced Nuclear Fuels Corporation November 1990.
- 28. COTRANSA 2: A Computer Proaram for Boiling Water Reactor Transient Analysis, ANF-913(PXA), Volume 1, Revision 1 and Volume I Supplements 2, 3, and 4, Advanced Nuclear Fuels Corporation, August 1990.
- 29. HUXY: A Generalized Multirod Heatup Code with I0CFR50. Appendix K Heatuo OPtion, ANF-CC-33(PXA), Supplement 1 Revision 1; and Supplement 2, Advanced Nuclear Fuels Corporation, August 1986 and January 1991, respectively.
- 30. Advanced Nuclear Fuels Methodoloav for Boiling Water Reactors. XN-NF-80-19(PXA), Volume 1, Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990.
- 31. Exxon Nuclear Methodology for Boiling Water Reactors: Anplication of the ENC Methodology to BWR Reloads, XN-NF-80 19(P)(A), Volume 4, Revision 1, Exxon Nuclear Company, June 1986.
- 32. Exxon Nuclear Methodoloav for Boli[no Water Reactors THERMEX: Thermal Limits Methodology Summary Description.
XN-NF-80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Company. January 1987.
- 33. Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-85-67(PXA) Revision 1, Exxon Nuclear Company, September 1986.
- 34. Advanced Nuclear Fuels Comoratlon Generic Mechanical Desian for Advanced Nuclear Fuels Corooration 9X9-IX and 9X9 9X BWR Reload Fuel ANF-89-014(PXA), Revision 1 and Supplements I and 2, October 1991.
- 35. Volume I - STAIF - A Comouter Program foi BWR Stability Analysis in the Frequency Domain. Volume 2 - STAIF - A Computer Pro-ram for BWR Stability Analysis in the Frequency Domain. Code Qualification Report, EMF-CC-074(PXA),
Siemens Power Corporation, July 1994.
- 36. RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, XN-NF-81-58(PXA), Revision 2 Supplements I and 2.
Exxon Nuclear Company, March 1984.
- 37. XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis XN-NF-84-105(PXA), Volume 1 and Volume 1 Supplements 1 and 2; Volume I Supplement 4, Advanced Nuclear Fuels Corporation, February 1987 and June 1988, respectively.
- 38. Advanced Nuclear Fuels Corooration Methodology for Boiling Water Reactors EXEM BWR Evaluation Model. ANF-91 048(PXA), Advanced Nuclear Fuels Corporation, January 1993.
- 39. Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Deslin and Analysis. XN-NF-80-19(PXA)
Volume 1 and Supplements 1 and 2, Exxon Nuclear Company, Richland, WA 99352, March 1983.
- 40. Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF-79-71 (P)(A), Revision 2 Supplements 1, 2, and 3, Exxon Nuclear Company, March 1986.
LaSalle Unit 2 Cycle 9A iv November 2002
Technical Requirements Manual -,Appendix J L2C9A Core Operating Limits Report I Supplement 1,
- 41. Generic Mechanical Design C0iteria for BWR Fuel Deslans. ANF-89-98(PXA). Revision I and Revision Advanced Nuclear Fuels'Corporatlon. May 1995.
- 42. NEDE-2401 1-P-A, General Electdc Standard Aoolicatlon for Reactor Fuel. Rev. 14, June 2000.
November 1990.
- 43. Commonwealth Edison Topical Report NFSR-0085, Benchmark of BWR Nuclear Deslan Methods Revision 0.
Deslgn Methods - Quad
- 44. Commonwealth Edison Topical Report NFSR-0085, Supplement 1,-Benchmark of BWR Nuclear Cities Gamma Scan Comparisons, April 1991, Revision 0.
BWR Nuclear Deslan Methods
- 45. Commonwealth Edison Topical Report NFSR-0085, Supplement 2, Benchmark of Neutronic LUcensina Analyses, April 1991. Revision 0.
BWR Nuclear Design Methods
- 46. Commonwealth Edison Topical Report NFSR-0091, Benchmark of CASMO/MICROBURN respectively; SER letter dated March 22, Revision 0, Supplements 1 and 2, December 1991, March 1992. and May 1992, 1993.
Supplement 2, Siemens Power
- 47. BWR Jet Pump Model Revision for RELAX, ANF-91-048(PXA), Supplement 1 and Corporation, October 1997.
Supplement 1. Appendix C. Siemens
- 48. ANFB Critical Power Correlation Application for Coresident Fuel, EMF-1125(PXA),
Power Corporation, August 1997.
Uncertainties. ANF-1125(PXA).
- 49. ANFB Critical Power Correlation Determination of ATRIUM-gB Additive Constant Supplement 1, Appendix E, Siemens Power Corporation, September 1998.
2 Facility Operating Ucense. License
Corrected Fuel Thermal Conductivity'.
- 51. *LaSalle Unit 2 Cycle 9 Operating Limits for Proposed ITS Scram Times and DEG:01:046, D. Garber to R. Chin, March 22, 2001.
Anthony Giancatadno to Jeff
- 52. 'LaSalle Unit I and Unit 2 Rod Block Monitor COLR Setpolnt Change', NFM:MW:01-0106.
Nugent. April 3, 2001.
D. Garber to R. Chin, June
- 53. "Transmittal of Revised CBH Effects on Fresh Fuel for LaSalle Unit 2 Cycle 9%, DEG:01:090.
2001.
Speed and Exposure Limited to
- 54. 'LaSalle Unit 2 Cycle 9 Equipment Out-of-Servlce Operating Limits Using Nominal Scram 14,000 MWd/MTU'. DEG:02:009. D. Garber to F. W. TrIkur, January 10. 2002.
1, DEG:01:185, D. Garber to F. W.
- 55. 'Assessment of Continued Applicability of the CBH Study Documented In Reference Tdkur, November 13,2001.
E. Garberto F. W. Tdkur. May 31, 2002.
- 56. "Revised Control Blade History Study for LaSalle Unit 2 Cycle 9," DEG:02:100. D.
DEG:02:125, D. E. Garber to F. W.
- 57. 'LaSalle Unit 2 Cycle 9 Operating Umits for Cycle Extension to 19.300 MWd/MTU,"
"Trlkur,August 9. 2002.
0 and Supplement I(PXA)
- 58. RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model EMF-85-74 (P)(A) Revision and Supplement 2(PXA), February 1998.
V November 2002 LaSalle Unit 2 Cycle 9A
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report
- 59. "LaSalle Unit 2 Cycle 9 NSS Base Case and TBVOOS or FHOOS Operating Umits for Proposed ITS Scram Times With Corrected Fuel Thermal Conductivity.' DEG:01:076, D. Garber to R. Chin. May 15, 2001.
- 60. "Control Blade History Penalties for LaSalle Unit 2 Cycle 9 Following the October 2002 Plant Outage,. DEG:02:151, letter from D. Garber to F. Tdkur, October 25, 2002.
- 61. 'Licensing Evaluation for LaSalle 2 Cycle 9A, DEG:02:153, letter from D. Garber to F. Trlkur, October 30, 2002.
- 62. 'Transmittal of Ucensing Evaluation for LaSalle Unit 2 Cycle 9A," TODI NF0200155, Revision 0. October 30, 2002.
- 63. "LaSalle Unit 2 Cycle 9A Design Basis Loading Plan," TODI NF0200148, Revision 0, October 29, 2002.
LaSalle Unit 2 Cycle 9A vi November 2002
Technical Requirements Manual - Appendix J L2C9A CreeOperating Limits Report
- 1. Average Planar Linear Heat Generation Rate (APLHGR) (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 and 1.2-2. For Single Reactor Recirculation Loop Operation, the MAPLHGR limits in Tables 1.2-1 and 1.2-2 are multiplied by the MAPFAC multipliers provided in Figures 1.2-1 and 1.2-2. .
1.2.2 SPC Fuel The MAPLHGR Limit is the Lattice-Type MAPLHGR Limit. The Lattice-Type MAPLHGR limits are determined from the table given below:
Fuel Type . j. J, Cycle First Inserted SPCA9-381 B-1 3GZ7-80M 8 SPCA9-384B-11GZ6-BOM 9 SPC-A9-391 B-14G8.0-100M 9 SPC-A9-410B-19G8.0-100M SPC-A9-383B-1 6G8.0-i OM SPC-A9-396B-12GZ-100M 9 V
(References 2 and 3) -
Planar Average Exposure MAPLHGR (kW/ft)
(qWd-MTU) (all Siemens fuel r . types) -
0.0 13.5 20.0 13.5 61.1 9.39 (Referenhes 3 and 6)
For single loop operation, the MAPLHGR limits from the table above are multiplied by the MAPLHGR multiplier. The MAPLHGR multiplier for SPC fuel is 0.90. (References 3, 5 and 6)
LaSalle Unit 2 Cycle 9A 1-1 November 2002
I Technical Requirements Manual - Appendix J L2C9A 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-CECO (Reference 9 and 19)
Exposure Exposure Lattice-Type MAPLHGR (kW/ft)
(MWD/ST) (MWDIMT)
P8CWL071 P8CWL345 P8CWL362 P8CWL362 P8CWL345 P8CWL071 NOG 5G5.0/4G4.0 9G4.0 2G5.0/9G4.0 9G4.0 11GE 0 0 12.74 12.09 11.65 11.25 12.11 12.74 200 220.5 12.67 12.13 11.70 11.32 12.15 12.67 1000 1102.3 12.48 12.22 11.83 11.46 12.25 12.48 2000 2204.6 12.42 12.35 12.00 11.61 12.39 12.42 3000 3306.9 12.41 12.48 12.14 11.77 12.54 12.41 4000 4409.2 12.44 12.62 12.28 11.94 12.70 12.44 5000 5511.6 12.46 12.77 12.43 12.11 12.86 12.46 6000 6613.9 12.49 12.90 12.58 12.29 13.02 12.49 7000 7716.2 12.51 13.03 12.73 12.46 13.19 12.51 8000 8818.5 12.54 13.16 12.88 12.64 13.33 12.54 9000 9920.8 12.55 13.30 13.01 12.82 13.43 12.55 10000 11023.1 12.57 13.42 13.12 12.98 13.44 12.57 12500 13778.9 12.41 13.41 13.08 13.04 13.40 12.41 15000 16534.7 12.04 13.05 12.78 12.77 13.06 12.04 20000 22046.2 11.27 12.38 12.16- 12.16 12.40 11.27 25000 27557.8 10.49 11.74 11.51 11.51 11.76 10.49 27215.6 30000 12.314 12.314 12.314 12.314 12.314 12.314 48080.8 53000 10.800 10.800 10.800 10.800 10.800 10.800 58967.1 65000 6.000 6.000 6.000 6.000 6.000 6.000 Lattice No. 733 1817 1818 1819 1820 1821 LaSalle Unit 2 Cycle 9A 1-2 November 2002
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Table 1.2-2 Maximum Average Planar Linear-Heat Generation Rate (MAPLHGR) vs.
Average Planar Exposure for Fuel Type GE9B-P8CWB320-9GZ3-1 OOM-150-CECO (Reference 9 and 19)
Exposure Exposure Lattice-Type MAPLHGR (kWtft)
(MWD/ST) (MWDIMT) _ __ _
P8CWL071 P8CWL346 P8CWL358* P8CWL358 P8CWL346 P8CWL071
____ NOG 4G5.013G4.0 7G4.0 2G5.0/7G4.0 7G4.0 9GE2 0 0 - 12.74 .12.05 11.62 11.10 12.09 12.74 200 220.5 12.67 .12.09- 11.64 11.15 12.14 12.67' ;
1000 1102.3 -12.48 i 12.19, 11.73 11.27 12.25 12.48 2000 2204.6 12.42 12.32 11.86 11.44 12.39 12.42 3000 3306.9 - --. 12.41 --12.44--. . 11.99 11.62 . 12.53 12.41 4000 4409.2 12.44 12.57 12.13 11.80 12.67 12.44 5000 5511.6 12.46 12.70 12.27 11.96 12.81 12.46 6000 6613.9 12.49 12.83 - 12.42 12.09 12.89 12.49-7000 7716.2 12.51 12.97 12.54 12.23 12.98 12.51 8000 8818.5 12.54 13.07:- 12.62 12.37 13.07 12.54 9000 9920.8 12.55 13.15 12.70 12.51 13.15 12.55 10000 11023.1 12.57 13.20 - 12.77 - 12.66 13.22 12.57 12500 13778.9 12.41 13.19 12.70 12.67 13.20 12.41 15000 16534.7 12.04 12.89 -12.40 12.40 12.90 12.04 20000 22046.2 11.27 12.29 11.82 11.82 12.30 11.27 25000 27557.8 10.49 11.69- -11.25 11.25 11.70 10.49 27215.6 30000 12.314 12.314 12.314 12.314 12.314 12.314:-
48080.8 53000 -10.800 10.800 - 10.800 10.800 10.800 10.800 58967.1 65000 6.000 6.000, 6.000 6.000 6.000 6.000 Lattice No. 733 1812 1813 1814 1815 1816 E-LaSalle Unit 2 Cycle 9A 1-3 November 2002
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Figure 1.2-1 Power-Dependent SLO MAPLHGR Multipliers for GE Fuel (MAPFAC p)
(References 8 and 19) 1 I m a.
0.95 0.9 U[lliLM~fiLl J4iJJjJ[,* , .... . -Ii
-I C.
a.
0.85-0.8 0.75- poor--t llll tif A-4 1- -- .1 i...Lj.1.! . Lf I II For 25> P:
0.7- "NoThermal Limits Monitoring Required; IfOfficial 0.65 Monitoring Is Desired, the Equations for - 25% Power" May Be Extrapolated for 25 > P, provided the Official 0.6 monitoring is only performed with the TCV/TSV closure 0I IL 0.55 scrams and RPT enabled.
CL 0.5 For 25 -<P s: 100 0.45 0
C3 0.4 MAPFACp = 1.0+0.005224 (P-100) e.1 0.35 T-T-T--T-- I-' I' 1"'
-- '1 -"" -?-'-- t" -!-'-t- 4-4 T-,- -I-I- -T-i--
For 100 < P, MAPFACp = 1.00 0.3
,IL P = % Rated Core Thermal Power 0.25
- 0. 0.2- I..I----1--. l,-.,--l-,-,-.l---.+
i..--
0.15-0.1
-Ii 0.05 I. .].
Ii 0
25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Core Thermal Power (% Rated)
LaSalle Unit 2 Cycle 9A 1-4 November 2002
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Figure 1.2-2 Flow-Dependent SLO MAPLHGR Multiplier (MAPFAC F)for GE Fuel (References 8, 18, and 19) 1
-U 0.9 LL 0.8 U.
M
..u.J ,
0.7 0(2. 0.6 U)
U-0.
c* 0.5 CL
.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 2 Cycle 9A 1-5 November 2002
Technical, Requirements Manual -,Appendix J L2C9A Core Operating Limits Report
- 2. Minimum Critical Power Ratio (3.2.2) 2.1 Tech Spec
Reference:
2.2 Desrition is based on Prior to initial scram time testing fOrin bop6-ating cycle, the MCPR operating limit refer to the Technical Specification Scram Times. -For Technical Specification requirements Technical Specification table 3.1.4-1.
exposure of MCPR limits from BOC to Coastdown are api~licable up to a core average 3) used by SPC). (Reference 30,266.2 MWd/MTU (which is the licensing basis exposure of MCPR limits for Coastdown are applicable from a core average exposure 57).
core average exposure of 31,242.7 MWd/MTU (Reference 30,266.2 MWd/MTU to a 2.2.1 Manual Flow Control MCPR Limits is either The Governing MCPR Operating Limit while in Manual Flow Control flow 2.2.1.2, whichever is greater at any given power, determined from 2.2.1.1 or condition.
2.2.1.1 Power-Dependent MCPR (MCPRp)*
2.2.1.1.1 GE Fuel Table 2-1 gives the MCPRp limit as a function of core thermal power for'Technical Specifications Scram Speed (TSSS) and Nominal Scram Speed (NSS).
2.2.1.1.2 Siemens Fuel Table 2-2 gives the MCPRp limit as a function of core thermal power for, Technical Specifications Scram Speed (TSSS) and Nominal Scram Speed (NSS).
2.2.1.2 Flow-Dependent MCPR (MCPRF)
Table 2-3 gives the MCPRF limit as a function of flow.
2.2.2 Automatic Flow Control MCPR Limits Automatic Flow Control is not supported for L2C9A.
limits should be applied.
- For thermal limit monitoring at greater than 100%P, the 100% power MCPRp 2-1 November 2002 LaSalle Unit 2 Cycle 9A
I Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report 2.2.3 Nominal Scram Speeds To utilize the MCPR limits for Nominal Scram Speeds (NSS), the core average scram speed insertion times must be equal to or less than the following values (References 4, 59).
Notch Position Time (sec.)
45 0.380 39 0.680 25 1.680 05 2.680 LaSalle Unit 2 Cycle 9A 2-2 November 2002
Technical Requirements Manual -Appendix J L2C9A Core Operating Limits Report Table 2-1 MCPRpfor GE-Fuel (References 2, 3, 51,'56.57, and 59)
Percent Core Thermal Power' EOOS Combination 25 I0 LJU IOU 18O 1..!80 1100 No EOOS with TSSS (BOCý to 2.70 2.20 2.01 - 1.53 Coastdown) ,. __.. .. W 1.52 Single RR Loop only with TSSS (BOG' to 2.71 2.21 2.02 1.54 3)
Coastdown . S1.96 11.86 1 1.63 EOOS' with TSSS (BOC' to Coastdown 4) 2.85 2.35 2.24 EOOS 4ISingle RR Loop with TSSS (BOCG 2.86 2.36 2.25 3
to Coastdown ) -_
TBVOOS- or FHOOS3 with TSSS (BOC' 2.85 2.35 2.24 1.63 to Coastdown_)
TBVOOS ISingle RR Loop or 2.86 2.36 2.25 1.64 FHOOS 5/Slngle RR Loop with TSSS (BOCe to Coastdown*)
No EOOS with NSS (BOCG to 2.70 2.20 1.99 1.51 3 1.53 Coastdown )
Single RR Loop only with NSS (BOCW to 2.71 2.21 2.00 1.52 CoastdowrO)
No EOOS with TSSS (Coastdown; ) 2.70 2.20 2.01 1.53 Single RR Loop only with TSSS 2.71 2.21 2.02 1.54 (CoastdownO) 1.57 Feedwater Heaters OS with TSSS 2.74 2.24 2.24 1.57 (Coastdown3) _ _____
Feedwater Heaters OOS/Single RR Loop 2.75 2.25 2.25 1.58 3 ___
with TSSS (Coastdown )
Feedwater Heaters DOS/Turblne Bypass - 2.74- 2.24 2.24 1.64 Valves OOS with TSSS (Coastdown) "
- e. 4.za ia Feedwater Heaters OOS/Turblne Bypass 2.75 4.40-Valves OOS/Single RR Loop with TSSS TCV Slow ClosurelEOC
-L.,nUaSUUN Feedwater Heaters (Coastdown )
3 I RPT OOSI OOS with TSSS 1 2.74 2..24 - 2.24 1.97 1.87 1.74 TCV Slow ClosureIEOC RPT OOSI 2.75 2.25 2.25 Feedwater Heaters DOS/Single RR Loop 3
with TSSS (Coastdown )
more limiting value is used.
Values are interpolated between relevant power levels.- For operation at exactly 25% or 80% CTP, the 3489 MWt Is rated power.
2 BOC is defined as the beginning of Cycle 9A.
thermal limits are to be 3 Coastdown Is defined as occurring at a core average exposure of 30,266.2 MWd/MTU. The coastdown not provided In the COLR for and 31,242.7 MWd/MTU. Umits are applied for core average exposures between 30,266.2 MWd/MTU 4 cycle exposures beyond 31,242.7 MWd/MTU.
to coastdown, see the specific thermal Allowable EOOS conditions are listed In Section 5. For TBVOOS or FHOOS conditions prior limit set In Table 2-1.
MCPR limits may be used. Use the For TBVOOS or FHOOS conditions (with and without single RR loop), this less limiting set of bounding "EOOS" set for all other EOOS conditions allowed per Section 5.
I,., I I-;#0f An 2-3'-
Sv November 2002 L-0,.,"il: cz U J " l Y,* ,1;;-J*
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Table 2-2 MCPRpfor Siemens Fuel (References 2, 3, 21. 51, 53, 54, 55, 56, 57, 59, and 60)
For all Siemens fuel EXCEPT Fuel Types 26, 27, 28, 38, 41 and 42 (as listed in Reference 63)
Percent Core Thermal Power1 EOOS Combination 10 -1 25 25 1 60 Coastdown 3 1.41 No EOOS with) TSSS (BO toC 2.70 2.20 1.93 1.48 80 S1 8 100
.4 2 Single 4RR Loop only with TSSS (BOC' to 2.71 2.21 1.94 1.49 EOOS with3 TSSS (BOC' 1.0 Coastdown ) to Coastdown") _48 2.85 16 1.53 2.35 2.17 1.3 1.54 EOOS /Single RR Loop with TSSS (BOC' 2.86 2.36 2.18 S~1.43 TBVOOS' or FHOOS' with TSSS (BOC' 2.85 2.35 2.17 1.54 3
to Coastdown )
1.44 TBVOOS'/Single RR Loop or 2.86 2.36 2.18 1.55 FHOOS 5/SIngle RR Loop with TSSS Coastdown 3
)l (BOC to Coastdown 3 )____
2 No EOOS with NSS (BOC' to 270 _____
2.20 1.91
_____ 1.44 1.39 4
Single RR Loop only with NSS (BOC to 2.71 2.21 1.92 1.45 1 1.40 3
Coastdown ) _ _
No EOOS with TSSS (Coastdown") 2.70 2.20 1.93 1.48, 1.44 Single RR Loop only with TSSS 2.71 2.21 1.94 1.49 1.45 (Coastdown33)
(Coastdown _ _ 1-- _
Feedwater Heaters 00S with TSSS 2.70 2.20 2.17 ___ _
1.54 1.44 Feedwater Heaters O OS/Single RR Loop 2.71 2.21 2.18 1.55 1.45 with TSSS (Coastdown3 ) i__
Feedwater Heaters OOSiTurbine Bypass 2.70 2.20 2.17 1.60 1.46 Valves OOS with TSSS (Coastdown ) I Feedwater Heaters OOS/Turbine Bypass 2.71 2.21 2.18 .1.47 Valves OOS/Single RR Loop with TSSS (Coastdown3 1 1.60 TCV Slow ClosurelEOC RPT OOSI 2.70 2.20 2.17 Feedwater Heaters 3
OOS with TSSS I
(Coastdown ) ! I TCV Slow ClosureIEOC RPT OOS/ 2.71 2.21 2.18 Feedwater Heaters OOSISingle RR Loop with TSSS (Coastdown 3 ) 1 11.61 Table continues on next page.
LaSalle Unit 2 Cycle 9A 2-4 November 2002
Technical Reqcuirements Manual - Appendix J L2C9A Core Opperating Limits Report
...Table 2-2 (Continued)
MCPRp for"Siemens Fuel For ONLY Siemens Fuel Type 38 (as listed in Reference 63)
PercentfCore Thermal Power' EOOS Combination 0 1 25 --- Z:) 60 2.70 2.20 1.93 1.48 No EOOS with TSSS (BOC' to Coastdown3))
Single RR Loop only with TSSS (BOCý to 2.71 2.21 1.94 1.49 Coastdown33) .. S1.70 11.62 11.53 L..J. SUIr CIL on & f 4-1 RA ,!)4.* '1 17 l 1.71 1 1.63 1 .5 EOOS'Single RR Loop with TSSS (BOC" 2.86 2.36÷ 2.18 TBVOOS' or FHOOSW with TSSS (BOC' 2.85 2.35 2.17 1.54 to Coastdown3) 1.44 "TBVOOSI/Single RR Loop or 2.86 2.36 2.18 1.55 FHOOS 5ISingle RR Loop with TSSS :I- - - -
3 (BOC2 to Coastdown )
No EOOS with NSS (BOC' to 2.70 2.20 . 1.91 1.44 i1.39' Coastdown 3 ) 1.40 Single RR Loop only with NSS (BOC" to 2.71 2.21 1.92 1.45 Coastdown3 I___
1.44 No EOOS with TSSS (Coastdown*) 2.70 2.20 - I 1.93- 1.48 1.45 Single RR Loop only with TSSS 2.71 2.21 1.94 . 1.49 (Cos.dwn (Coastdowno) .. 1.44 Feedwater Heaters OOS with TSSS 2.70. 2.20 . 2.17 1.54 -
1.45 Feedwater Heaters OOS/SIngle RR Loop 2.71 . 2.21 2.18 1.55 with TSSS (Coastdown 31______ 1.46 Feedwater Heaters OOS/Turbine Bypass 2.70 - 2.20 2.17 1.60 ,
Valves OOS with TSSS (Coastdown?) ... ... 1.47 Feedwater Heaters OOS/Turbine Bypass 2.71 2.21 2.18 Valves OOS/Single RR Loop with TSSS '..
-1?
1.60 TCV Slow Closure/EOC RPT OOS/ 2.70 2.20 2.17 Feedwater Heaters OOS with TSSS . , .....
(Coastdown3 ) ,
TCV Slow Closure/EOC RPT OOS/ 2.71 - 2.21 2.18 Feedwater Heaters OOSISingle RR Loop ....
with TSSS (Coastdown )
3 "11.63 I .6 1 Table continues on next page.
2-5 November 2002 LaSalle Unit 2 Cycle 9A
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report
-Table 2-2 (Continued)
MCPRp for Siemens Fuel For ONLY Siemens Fuel Types 26 and 27 (as listed in Reference 63)
Percent Core Thermal Power1 EOOS Combination 0 25 25 604 180 1so 100 No EOOS with TSSS (BOO to Coastdown 3) 4 Single RR Loop only with TSSS (BOd2 to 2.70 2.71 f 2.20 2.21 1.93 1.94 1.48 1.49 1.41 1.42 Coastdowno) . . .. ,_ ____
EOOS. with TSSS (BOO4 to Coastdown4 ) 2.85 2.35 2.17 1.53 I.UU-0 tn I.Single IRK Loop WIfl 1155 C*n=fntrin3n Z.00 2.36 2.18 1.54 TBVOOSO or FHOOS" with TSSS (BO9C 2.85 2.35 2.17 1.54 1.43 to Coastdown 3)
TBVOOS'ISingle RR Loop or 2.86 2.36 2.18 1.55 1.44 FHOOS 5/Slngle RR Loop with TSSS (BOC2 to Coastdown 3 )
No EOOS with NSS (BOCO to 2.70 2.20 1.91 1.44 1.39 Coastdown 3) -____
Single RR Loop only with NSS (BOO4 to, 2.71 2.21 1.92 1.45 1.40 Coastdown 3) .. .. _ _..... ,
No EOOS with TSSS (Coastdown") 2.70 220 1.93 1.48 1.44 Single RR Loop only with TSSS 2.71 2.21 1.94 1.49 1.45 (Coastdown3 ) .
Feedwater Heaters OS with TSSS 2.70 2.20 2.17 1.54 1.44 (Coastdown,3 ) i_-_
Feedwater Heaters OOS/Single RR Loop 2.71 2.21 2.18 1.55 1.45 with TSSS (Coastdown3 ) _
Feedwater Heaters OOS/Turbine Bypass 2.70 2.20 2.17 1.60 1.46 Valves OOS with TSSS (Coastdown )
Feedwater Heaters OOSITurbine Bypass 2.71 2.21 2.18 1.47 Valves OOS/Single RR Loop with TSSS (Coastdown3 )
TCV Slow Closure/EOC RPT OOS/ 2.70 2.20 2.17 1.70 I 1.62 I 1.60 Feedwater Heaters OOS with TSSS (Coastdown 3 ) 1 TCV Slow Closure/EOC RPT OOS/ 2.71 2.21 2.18 1.71 1.63 1.61 Feedwater Heaters OOS/Slngle RR Loop with TSSS (Coastdown3) I Table continues on next page.
LaSalle Unit 2 Cycle 9A 2-6 November 2002
Technical Requiriements Manual - Appendix J L2C9A Core Operating Limits Report
-Table'2-2 (Continued)
MCPRp for Siemens Fuel For ONLY Siemens Fuel Type 28 (as listed in Reference 63) 1 PIr,'nt ernre The~rmal Power Percent Core Thermal PoweO 60 EOOS Combination 0 25 - 25 -
2.72 2.22 1.95 1.50 80 8 100 ..
No EOOS with TSSS (BOCz to - 1 3
Coastdown ) 1.44 Single RR Loop only with TSSS (BOC to 2.73 2.23 - 1.96 1.51 1.72 1.4 1.55 3
Coastdown )
ILUU5 withri I SSE) (60 to CoasdLuuw I EOOS ' /Single RR Loop with TSSS (BOCG 2.88 2.38 2.20 I 1.7- 165 1.56-I ¶,oast OWP I TBVOOS'or 3FHOOS ) with TSSS (BOC" 2.87 2.37 2.19.. 1.56 to Coastdown 1.46 "TBVOOS/JSingle 5
RR Loop or 2.88 2.38 2.20-- 1.57 FHOOS /Single RR Loop with TSSS (BOC2 to Coastdown-3 ) 1.41 No EOOS with NSS (BOC to 2.72 2.22 1.93 1.46 Coastdown 3 )
Single RR Loop only with NSS (BO to 2.73 2.23 1.94 1.47 Coastdown 3 ) - ~1.46 No EOOS with TSSS (Coastdown") 2.72 2.22 - 1.95 1.50 1.47 Single RR Loop only with TSSS 2.73 2.23 1.96 1.51 (Coastdown 3 ) 1.46 Feedwater Heaters OOS with TSSS
- 2.72 2.22 - " 2.19 . 1.56 Coastdown 3)
Feedwater Heaters OOS/Single RR Loop 2.73 2.23 - 2.20 1.57
- ~1.47_ :
with TSSS (Coastdow3).
Feedwater Heaters OOS/Turbine Bypass 2.72 2.22 ; 2.19,- 1.62-,
Valves OOS with TSSS (Coastdown I -I -1.49 Fe.dwater Heaters OOS/Turbine Bvyass 2.73 2.23 2.20 I 1.63 I
Valves OOS/Single RR Loop with TSSS I ,nt 31 1.2 16 1.62 TCV Slow ClosureIEOC RPT OOS/ 2.72 2.22. 2.19 OOS with TSSS Feedwater Heaters (Coastdown3 ) 1.7 1.5, 1.63 *-
TCV Slow Closure/EOC RPT OOS/ 2.73 2.23 2.20 Feedwater Heaters OOSISingle RR Loop 3
with TSSS (Coastdown )
Table continues on next page.
S'ovember 2002 LaSalle Unit 2 Cycle 9A 2-7
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Table 2-2 (Continued)
MCPRp for Siemens Fuel For ONLY Siemens Fuel Types 41 and 42 (as listed in Reference 63)
-.... Percent Core Thermal Power1 EOOS Combinabon 0 25 25 60 No EOOS with TSSS (B0C' to 2.74 2.24 1.97 1.52 Coastdown 3)
Single RR Loop only with TSSS (BOCz to 2.75 2.25 1.98 1.53 Coastdown 3) - ..
EOOS' with TSSS (BOCG to Coastdownj) 2.89 2.39 2.21 EOOS'/Single RR Loop with TSSS (BOC' 2.90 2.40 2.22 to Coastdown3 )
TBVOOS' or 3FHOOS" with TSSS (BOC' 2.89 2.39 2.21 1.58 to Coastdown )
TBVOOSO/Single RR Loop or 2.90 2.40 2.22 1.59 FHOOS 5/Slingle RR Loop with TSSS (BOC 2 to Coastdown3)
No EOOS with NSS (BOC' to 2.74 2.24 1.95 1.48 3
Coastdown )
Single RR Loop only with NSS (BO to 2.75 2.25 1.96 1.49 3
Coastdown° ) _.
No EOOS with TSSS (Coastdown") 274 2.24 1.97 1.52 Single RR Loop 3 only with TSSS 2.75 2.25 1.98 1.53 (Coastdown )-_
Feedwater Heaters OOS with TSSS 2.74 2.24 2.21 1.58 3
(Coastdown ) ... ..
Feedwater Heaters OOS/Single RR Loop 2.75 2.25 2.22 1.59 3
with TSSS (Coastdown )
Feedwater Heaters OOSITurbine Bypass 2.74 2.24 2.21 1.64 Valves OOS with TSSS (Coastdown7) I Feedwater Heaters OOS/Turbine Bypass 2.75 2.25 2.22 Valves DOSISingle RR Loop with TSSS 3
(Coastdown l TCV Slow Closure/EOC RPT 00S1 2.74 2.24 2.21 Feedwater Heaters OOS with TSSS (Coastdown3 )
TCV Slow Closure/EOC RPT DOSI 2.75 2.25 2.22 Feedwater Heaters OOS/Single RR Loop with TSSS (Coastdown 3)
'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.
2 BOC Is defined as the beginning of Cycle 9A.
3 Coastdown Is defined as occurring at a core average exposure of 30,266.2 MWd/MTU. The coastdown thermal limits are to be applied for core average exposures between 30,266.2 MWd/MTU and 31,242.7 MWd/MTU. Limits are not provided In the COLR for core average exposures beyond 31,242.7 MWd/MTU.
4 Allowable EOOS conditions are listed in Section 5. For TBVOOS or FHOOS conditions prior to coastdown, see the specific thermal limit set In Table 2-2.
5 For TBVOOS or FHOOS conditions (with and without single RR loop), this less limiting set of MCPR limits may be used. Use the bounding "EOOS"set for all other EOOS conditions allowed per Section 5.
LaSalle Unit 2 Cycle 9A 2-8 November 2002
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report
-Table 2-3 MCPRF for GE and Siemens Fuel (Reference 3) all EOOS scenarios.,
The MCPRF limits are applicable from BOC through coastdown and in 2-9 November 2002 LaSalle Unit 2 Cycle 9A
Technical Requirements Manual - Appendix J L2C9A 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 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 are determined from Figures 3.2-1 through 3.2-3. The following GE LHGR limits apply for the entire cycle exposure range: (References 2, 8, 10, 19 and 60)
- 1. "GE9B.P8CWB322.11GZ-100M-150-CECO (bundle 3861 in Reference 2)
Nodal Exposure (GWd/MT)--- LHGR Limit (KW/ft)
.0 .13.75 13.06 13.75 27.80 11.75 50.31 10.31 60.89 6.00
- 2. GE9B-P8CWB320-9GZ-100M-150-CECO (bundle 3860 in Reference 2)
'i Nodal Exposure (GWdlMT) LHGR Limit (KWlft) 0.00 14.25 12.14 14.25 26.19 12.18 48.16 10.80 59.93 6.00 3.2.2 Siemens Fuel The LHGR Limit is the product of the Steady-State LHGR Limit (given below) and the minimum of either the power dependent LHGR Factor*, LHGRFACp, or the flow dependent LHGR Factor, LHGRFACF. LHGRFACp is determined from Table 3-1.
LHGRFACF is determined from Table 3-2. SPC LHGRFAC multipliers from BOC to Coastdown are applicable up to a core average exposure of 30,266.2 MWd/MTU (which is the licensing basis exposure used by SPC) (References 3 and 59). SPC LHGRFAC multipliers for Coastdown are applicable for core average exposures between 30,266.2 MWdIMTU and 31,242.7 MWd/MTU (Reference 57).
For All Siemens Fuel EXCEPT Fuel Types 28,41, and 42 (References 3. 56,60, and 63)
Planar Average Exposure (GWd/MTU) LHGR limit (kW/ft) 0.0 14.4 15.0 14.4 61.1 8.32
- For thermal limit monitoring at greater than 100%P, the 100% power LHGRFACp limits should be applied.
,n f,,,,,A I.. a-1 v .
November 2002 Lasadlle Uill L- .. Y%,'V
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report For ONLY Siemens Fuel Type 28 (References 56, 63)
Planar Average Exposure (GWd/MTU) LHGR limit (kW/ft) 0.0 14.2 15.0 14.2 61.1 8.12 For ONLY Siemens Fuel Types 41 and 42 (Reference 60. 63)
Planar Average Exposure (GWd/MTU) LHGR limit (kW/ft)
"0.0 14.0
-15.0 14.0 61.1 7.92 LaSalle Unit 2 Cycle 9A 3-2 November 2002
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Figure 3.2-1 Power-Dependent LHGR Multipliers for GE Fuel (Formerly MAPFACp)
(References 8 and 19)
"4 K{L14I Ii)t I I
+
i 0.95 0.9 0.85 0.8 - I**I***IIII
- i -t TIITII((III U * . .I , .
0.75 IFor 25 > P.
C, No Thermal Urits Monitoring Required; IfOffidal
.- 0.7 "Monitoring Is Desired, the Equations for 1 25% Power I
May Be Extrapoiated for 25 > P. provided the Official 0.65 monitodring is only performed with the TCvWTSV dosure 0o. 0.6 scrims and RPT enabed.
C, 0.55 For25k Pk 100 0.5 0.45 L.HGRFACp = 1.0+0.005224 (P-100) 0 0.4 For 100 < P, LHGRFACp = 1.00 C
0.35 C P = % Rated Core Thermal Power 0.3 a1. I 0.25 0.2 0.15 I-L 11J 0.1 0.05 U
t 25 I I I I I I I I I I I I I I I I I I I I I I I IIl*ll=
30 35 IIt*im*
40 45 I I I I I I Ip I I I 50 55. 60 I I 'I I I 65 I I I 70 I I 75 I
80 85 I II 90 95 100 Core Thermal Power (%Rated)
? "3-
'-3 t , ° November 2002 LaSalle Unit 2 Cycle 9A
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Figure 3.2-2 Power-Dependent LHGR Multiplier for GE Fuel (TCV(s) Slow Closure) (formerly MAPFACp)
(References 11 and 19)
I 0.95
[L..
[
'-V K-0.9 0I. 0.85 IL 0.8 U..
0.75 Fir
-J 0.7 of
+/-
I 0.65 -1 -J 0.6 I I
]
I I
I For 25 > P: ,i 0.55 -- 1-- 4 --- 4-4 -- 4-- L-4-4-1 No Thermal Limits Monitoring Required; If Official 0.5 Monitoring is Desired, the Equations for 25 <
-J 0.45 P < 100 May Be Extrapolated for 25 > P I..--
0.4- For 25.< P < 100:
0.35 LHGRFACp = 1.0+0.008 (P-100) 1**-
0.3 For P > 100:
- a. 0.25 LHGRFACp = 1.00 0.2 0.15 [-I------IP = % Rated Core Thermal Power 0.1 -t- f .I I
- I.I H1-i 1-1 -1t-!]1--
. I-lllllll f 0.05 0
0 10 I . . I : I I . I . . I I I 40
. . I i I I t!-t1 I--It 20 30 50 60 70 80 90 100 Core Thermal Power (% Rated)
LaSalle Unit 2 Cycle 9A 3-4 November 2002
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Figure 3.2-3 Flow-Dependent LHGR Multiplier for GE Fuel (formerly MAPFAC F)
- (References 8, 13, 18, and 19) 0* -i ~EL LL.0.9--
. LL ~For 105% Maximum Attain'able Core Flow4
.,0.8 -
LH--------- = Th iiu fETE .
- a. .
,OR 0.6807x (WTIIO0) + 0.4672)
KtWT j~ =%Rated Core Flow' For Ab-rmal Idle Loop Startup; LHGRFAC f=
00..
rv0.6
.4- li I 0
S0.4I
,"0.G
!-1***
1
{**1** t I.-. - l i...t,- tIt i -
i I !- ti -
o.5'. , 65,,, ! 5.
- 70. 90 95 105 30 35 40 45 .50 ;55 60 65, " 7 5 80 85 90 95 100 105 Core Flow (% Rated)
I
- i C In rI OA 3-5 November 20 02 iL¢;',./~ll*/llb C. =,P tWlr.*'
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Table 3-1 LHGRFACp for Siemens Fuel (References 3, 51, 54. 57. and 59)
- -- r Percent I --
Core Thermal Power' EOOS Combination 10 25 I 40 100 No EOOS with TSSS (BOCO to 0.77 0.77 S1.00 Coastdown 3) __
Single RR Loop only with TSSS (BOG4 to 0.77 0.77 I 1.00 Coastdown 3)
EOOS' with TSSS (BO1C to Coastdown*) 0.67 0.67 10.89 EOOS'lSingle RR Loop with TSSS (BOCO 0.67 0.67 0.89 to Coastdown 3)
TBVOOS or FHOOSO with TSSS (BOC' 0.68 0.68 0.99 to Coastdown 3 )
TBVOOSV/Single RR Loop or 0.68 0.68 0.99 FHOOSSlSingle RR Loop with TSSS (BOC 2 to Coastdown 3)
No EOOS with NSS (BO=' to 0.78- 0.78 1.00 Coastdown 3 )
Single RR Loop 3
only with NSS (BO:' to 0.78 0.78 1.00 Coastdown )
No EOOS with TSSS (Coastdownj) 0.77 0.77 1.00 Single RR Loop only with TSSS 0.77 0.77 1.00 (Coastdown3 )
Feedwater Heaters OOS with TSSS 0.68 0.68 1.00 (Coastdown3)__
Feedwater Heaters OOS/Single RR Loop 0.68 0.68 1.00 with TSSS (Coastdown 3) _
Feedwater Heaters OOS/Turbine Bypass 0.68 0.68 0.97 Valves OOS with TSSS (Coastdown0)
Feedwater Heaters OOSITurblne Bypass 0.68 0.68 0.97 Valves OOSISingle RR Loop with TSSS 3
(Coastdown ) I _
TCV Slow Closure/EOC RPT OOSI 0.67 0.67 0.79 Feedwater Heaters OOS with TSSS (Coastdown 31 : _ _ : ___
TCV Slow Closure/EOC RPT 00S/ 0.67 0.67 0.79 0.79 Feedwater Heaters OOS/Single RR Loop with TSSS (Coastdown 3) I_ II
'Values are interpolated between relevant power levels. For operation at exactly 80% CTP, the more limiting value Is used. 3489 MWt is rated power.
2 BOC is defined as the beginning of Cycle gA.
3 Coastdown is defined as occumng at a core average exposure of 30,266.2 MWd/MTU. The coastdown thermal limits are to be applied for core average exposures between 30,266.2 MWd/MTU and 31,242.7 MWd/MTU. Umits are not provided In the COLR for 4 cycle exposures beyond 31,242.7 MWd/MTU.
Allowable EOOS conditions are listed in Section 5. For TBVOOS or FHOOS conditions prior to coastdown, see the specific thermal limit set In Table 3-1.
5 For TBVOOS or FHOOS conditions (with and without single RR loop), this less limiting set of LHGRFACp limits may be used. Use the bounding "EOOS"set for all other EOOS conditions allowed per Section 5.
LaSalle Unit 2 Cycle 9A 3-6 November 2002
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report Table 3-2 LHGRFACFfor Siemens Fuel (Reference 3)
Values Applicable for up to 105% Maximum Attainable Core Flow Flow (% rated) LHGRFACE 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 2 Cycle 9A 3-7 November 2002
Technical Requirements Manual - Appendix J L2C9A 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:
determined from the The Rod Block Monitor Upscale Instrumentation Setpoints are relationships shown below:
ROD BLOCK MONITOR UPSCALE TRIP FUNCTION TRIP SETPOINT ALLOWABL _E VALUE Two Recirculation Loop 0.66 W + 51%** 0.66 W + 54 Operation*
B.7%**
Single Recirculation Loop 0.66 W + 45.7%** 0.66 W + 4E Operation*
This setpoint may be lower/higher and will still comply with the RWE Analysis, because RWE is analyzed unblocked.
flow (W) of Clamped, with an allowable value not to exceed the allowable value for recirculation loop 100%.
4-1 November 2002 LaSalle Unit 2 Cycle 9A
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report
- 5. ' Allowed Modes of Operation (B 3.2.2, B 3.2.3) are as described below:
The Allowed Modes of Operation with combinations of Equipment Out-of-Service -OPERATING REGION 1 Standard MELLLA ICF7 Coastdowng Equipment Out of Service Options Yes Yes Yes Yes
,None Yes No3 Yes Yes Feedwater Heaters? (Reference 8)
Yes Noa N/A Yes Single RR Loop 10 (Reference 8)
Yes Yes Yes Yes 11 Turbine Bypass Valves (Reference 8)
Yes Yes Yes Yes EOC Recirculation Pump Trip (Reference 8)
Yes Yes Yes Yes TCV Slow Closure/EOC Recirculation Pump Trip (Referenceil)
Yes No3 Yes Yes TCV Slow ClosurelEOC Recirculation Pump Trip I Feedwater Heaters 2 (References 11, 16, and 17)
NO12 No12 No12 Yes Turbine Bypass Valves I Feedwater Heaters 25 (Reference 8)
Yes 4 No3 Yes4 Yes EOC Recirculation Pump Trip /
Feedwater Heaters 2 (Reference 8)
Yes Yes Yes No TCV Stuck Closedý (Reference 12) up to two TIP Machines OOS or the
- 1. Each EOOS condition may be combined with one SRV OOS, a refuel outage), a 20°F reduction in equivalent number of TIP channels (100% available at startup from and/or up to 50% of the LPRMs feedwater temperature (without Feedwater Heaters considered OOS),
out of service.
with Feedwater Heaters Out-of-Service.
- 2. Up to 100 0F Reduction in Feedwater Temperature Allowed an intentionally entered mode of operation Feedwater Heaters OOS may be an actual OOS condition, or to extend the cycle energy.
in MELLLA is supported by current
- 3. If operating with Feedwater Heaters Out-of-Service, operation transient analyses, but administratively prohibited due to core stability concerns. using the TCV Slow OOS is allowed
Closure/EOC Recirculation Pump Trip OOS/Feedwater Heaters this combination is not allowed.
- 5. Only when operating in coastdown, otherwise when less than 10.5 million Ibm/hr steam flow and when
- 6. Operation prior to coastdown is only allowed
<103%, and the MCFL setpoint 'a average position of 3 open TCVs is less than 50% open, with FCL TBVOOS or TCV Slow Closure.
120%. TCV Stuck Closed may be in combination with any EOOS except adjustment for the other EOOS(s) as If in combination with other EOOS(s), thermal limits may require designated in Sections 1, 2, and 3.
- 7. ICF is analyzed for up to 105% core flow.
MELLLA. The flow boundary for SLO at
- 8. The SLO boundary was not moved up with the incorporation of (Reference 20) uprated conditions remains the ELLLA boundary for pre-uprate conditions. (which is the licensing exposure of 30,266.2 MWdIMTU
- 9. Coastdown is defined to begin at a core average (Reference 3 and 56) basis exposure used by SPC). ICF is allowed during coastdown.
any of the EOOS options listed in this table.
- 10. Single loop operation is allowed with using the Feedwater Heaters is allowed during coastdown operation
- 11. Turbine Bypass Valves OOS OOS/Turbine Bypass OOS operating limits.
- 12. Operation in these regions is permitted during coastdown only.
5-1 November 2002 LaSalle Unit 2 Cycle 9A
Technical Requirements Manual - Appendix J L2C9A Core Operating Limits Report
- 6. , Traversing In-Core Probe System (3.2.1, 3.2.2, 3.2.3) 6.1 Tech Spec
Reference:
for recalibration of the LPRM Tech Spec Sections 3.2.1, 3.2.2., 3.2.3 for thermal limits require the TIP system detectors and monitoring thermal limits.
6.2 Descriotion
is used for When the traversing in-core probe (TIP) system (for the required measurement locations) be operable with the TIP system shall recalibration of the LPRM detectors and monitoring thermal limits, the following:
required measurement
- 1. movable detectors, drives and readout equipment to map the core in the locations, and location.
- 2. indexing equipment to allow all required detectors to be calibrated in a common applies for use of the SUBTIP F61lowing the first TIP set (required prior to BOC + 500 MWDIMT), the following methodology:
inoperable measurement location With one or more TIP measurement locations inoperable, the TIP data for an software system adjusted using may be replaced by data obtained from a 3-dimensional BWR core monitoring the previously calculated uncertainties, provided the following conditions are met:
operating cycle when the
- 1. All TIP traces have previously been obtained at least once in the current23) and reactor core was operating above 20% power, (References 14, 15 and exceed 42% (18
- 2. The total number of simulated channels (measurement locations) does not channels).
the above applicable monitoring or Otherwise, with the TIP system inoperable, suspend use of the system for calibration functions.
6.3 Bases
of equipment ensures that the The operability of the TIP system with the above specified minimum complement spatial neutron flux distribution of accurately represent the measurements obtained from use of this equipment internal to the core monitoring the reactor core. The normalization of the required detectors is performed software system.
calculated data which is Substitute TIP data, if needed, is 3-dimensional BWR core monitoring softwareSince uncertainty could be sets.
adjusted based on axial and radial factors calculated from previous TIP channels may be simulated to ensure introduced by the simulation and adjustment process, a maximum of 18 remain valid.
that the uncertainties assumed in the substitution process methodology I -C-1II- I Irnit ') 1"*%ýI OA 6_1 November 2002
Technical Requirements Manual - Appendix J Section 2 LaSalle Unit 2 Cycle 9A Reload Transient Analysis Results November 2002
Technical Requirements Manual - Appendix J L2C9A Reload Transient Analysis Results Table of Contents Preparer Document Attachment Neutronic Licensing Report 1I . Exelon Reload Analysis Report 2 Siemens Power Corporation Siemens Power Corporation Plant Transient Analysis 3
General Electric ARTS Improvement Program 4 Analysis, Supplement I (Excerpts)
TCV Slow Closure Analysis 5 General Electric (Excerpts)
LaSalle Unit 2 Cycle 9 Operating 6 Framatome ANP Limits for Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity LaSalle Unit 2 Cycle 9 Equipment Framatome ANP,'
7 Out-of-Service Operating Limits Using Nominal Scram Speed and Exposure Limited to 14,000 MWdIMTU LaSalle Unit 2 Cycle 9 Operating Framatome ANP a Limits for Cycle Extension to 19,300 MWdIMTU LaSalle Unit 2 Cycle 9 NSS Base Framatome ANP' 9 Case and TBVOOS or FHOOS Operating Limits for Proposed ITS Scram Times With Corrected Fuel Thermal Conductivity Transmittal of Licensing Evaluation 10 ExelonlFramatome ANP for LaSalle Unit 2 Cycle 9A November 2002 LaSalle Unit 2 Cycle 9A
Technical Requirements Manual - Appendix J L2C9A Reload Transient Analysis Results Attachment 1 LaSalle Unit 2 Cycle 9 Neutronics Licensing Report November 2002 LaSalle Unit 2 Cycle 9A
4 I ItcDGOO - O013.03 NUCLEAt FUEL MANAGEMENT TRANSMITTAL OF DESIGN INFORMATION
- SAFETY RELATED Originatig Oranizato NFM ID# NFM0000115 n NON-SAFETY RELATiD Nuclear Fud Management Sequence 0 REGULATORY RELATED Other (specify) Page Iof 21 Station: LaSalle Unit: 2 Cycle: 9 Generic:
To: Jeffery K. Nugent (IS)
Subject:
LaSalle Unit 2 Cycle 9 Neutrnics Icensing Report Minx-Yuan 1Hsiao,______________ _____
P Departmert Head . s Sidnatuc lDate Peter A. Weggeman &Z___L____Y_____
Reviewer :eiwetgaueDf Adelmo S. pallotta ______________
NFM Department Head .Approver's Signature iDate Status oflnfornmation: Verified oo Unverified Engineeing Judgement Action Tracking # for Method and Schedule of Verification for Unverified DESIGN INFORMATION:
Des*ription of Information: Provide the station and BSS group LaSalle Unit 2 Cycle 9 Neutronics Licensing Report (NLR).
Purpose of Infannation:
Seq. 0: Provide the station and BSS group LaSalle Unit 2 Cycle 9 Neutronics Licensing Report (NLR).
Source of Informtion: As referenced Supplemental Distribution: DannyBost (IS) John J. Reimer (IS)' Amy Goss (LS) Edward A. McVey Thomas J. Rausch R. W. Tsai Adelmo S. Pallotta Ming Y. Hsiao LaSalle Central File Downers Grove Central File
II IkIIIrIrannmn
'UI...LZ/V flJCL IV IWMtIN flAAk.AflflW* I -
-.1 '
TRANSMITTAL OF DESIGN INFORMATION I Seq. No.
NFM ID#
I Pop~e z of z,i n----n-aI NFMOOOOI 15 0
/
COMMONWEALTH EDISON COMPANY NUCLEAR FUEL SERVICES NEUTRONICS LICENSING REPORT for LaSalle Unit 2 Cycle 9 preparer- "'YN, c8.3I-oo reviewer PAUt -31Ii
I NUCLEAR FUEL MANAGEMENT NFM ID# NFMOO001 15 Seq. No. 0 TRANSMITTAL S. .. * . .. INFORMATION OF DESIGN . . ... . .. .... Page 3 of 21 .. . .
-Licensing Basis This document, in conjunction with the references 1,'2 and 4 in Section ViII provide the licensing basis for LaSalle Unit 2 Reload 8, Cycle 9.
Table of Contents L Nuclear Design Analysis LI Fuel Bundle Nuclear Design Analysis 1.2 Core Nuclear Design Analysis
.1.2.1 Core Configuration and Licensing Exposure Limits L2.2 Core Reactivity Characteristics HI. Control Rod Withdrawal Error III. Fuel Loading Error H1.1 Fuel Mislocation Error 111.2 Fuel Misrotation Error IV. Control Rod Drop Accident V. Loss of Feedwater Heating VI. Maximum Exposure Limit Compliance VII. Spent Fuel Pool and Fresh Fuel Vault Criticality Compliance VII.1 Fresh Fuel Vault Criticality Compliance VII.2 Li Spent Fuel Pool Criticality Compliance VII.3 L2 Spent Fuel Pool Criticality Compliance VHII. References preparer: "'n'1)4, 05`3f- reviewer
I S-NUCLEAR TRANSMITTAL OF FUEL MANAGEMENT DESIGN -
INFORMATION NFM No.
Seq. ID# 0NFM00001 15 Page 4 of 21 L ' Nuclear Design Analysis 1.1 Fuel Bundle NuclearDsin-A- alsis. ..-.... -
Assembly Average Enrichment (ATRIUM-9B), w/o U-235 SPCA9-391B-14G8.0-100M 3.91 SPCA9-410B-19G8.0-100M 4.10 SPCA9-383B-16G8.0-100M 3.83 SPCA9-396B-12GZ-IOOM 3.96 Axial Enrichment and Burnable Poison Distribution SPCA9-391B-14G8.0-100M Figure 1 SPCA9-410B-19G8.0-100M Figure 1 SPCA9-383B-16G8.0-100M Figure 2 SPCA9-396B-12GZ-100M Figure 2 Radial Enrichment and Burnable Poison Distribution SPCA9-4.53L-11G8.0-100M Figure 3 SPCA9-4.56L-12G8.0-100M Figure 4 SPCA9-4.21L-13G8.0-100M Figure 5 SPCA9-4.27L-12G8.0-100M Figure 6 SPCA9-3.96L-8G5.0-100M Figure 7 SPCA9-4.58L-8G6.0-100M Figure 8 SPCA9-4.58L-8G6.0/4G3.0-100M Figure 9 preparer: "1)'f, 3/'g0 reviewer (-f 3 ,
NUCLEAR FUEL MANAGEMENT NFM ID# NFM000115 I
"TRANSMfITTALOF DESGN INFORMATION Seq. No. -- -- 0 .........
Paae 5 of 21
,1.2 Core Nuclear Design Analysis 1.2.1 Core Configuration and Licensing Exposure Limits ycle " Nmber in Core
]Bundle TyDC LOaded GE9B-P8CWB322-11GZ-IOOM-150-CECO 7 84 7 "76 "GE9B-P8CWB320-9GZ-100M-150-CECO 8 128 SPCA9-381B-13GZ7-80M 8 -128'° SPCA9-384B-11GZ6-80M 9 40 SPCA9-391B-14G8.0-100M 9 120 SPCA9-410B-19G8.0-100M 9 132' SPCA9-383B-16G8.0-100M 9 56 SPCA9-396B-12GZ-1OOM Licensing Exposure Limits Core Cycle Value of Interest VloftesAverage
-..... Incremental Ex.
Exposure' Exposure (M__DnD OMWD/MY Nominal EOC 8 Exposure 27892 . 13750 Short EOC 8 Exposure 27392 13250 Minimum EOC 8 Energy for which C9 Neutronic Licensing Analyses are 27392 13250 Valid BOC 9 Exposure1790 (assuming nominal EOC 8 energy) 1790 BOC 9 Exposure 0 (assumingshort EOC 8energy) 11470 Nominal EOC 9 Exposure 29598 17800 1
(assumingnominalEOC 8 energy)
Core U0, Weights Cycle of Interest U0 2 Total Weight (MT)
.Cycle 8 135.11 Cycle 9 133.50 reviewer preparer: W"Y4 -
NUCLEAR FUEL MANAGEMENT NFM 1D# NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No. 0
.' Page 6 of 21 1.2.2 Core Reactivity Characteristics All values reported below are with zero xenon and are for 68"F moderator 7-1 xr e MCROBURN-B cold BOC best estimate K-effective bias is 1.004 at BOC. The shutdown margin calculations are based on the short EOC8 energy given in Section 1.2.1.
BOC Cold K-Effective, All Rods Out 1.11257 BOC Cold K-Effective All Rods In 0.95674 BOC Cold K-Effective, Strongest Rod Out 0.99360 BOC Shutdown Margin, % AK 1.040 Minimum Shutdown Margin, % AK 1.020 Reactivity Defect (R-value), % AK 0.020 Cycle Incremental Exposure Corresponding to Minimum Shutdown Margin R-Value (MWD/MTUT) 250 Standby Liquid Control System Shutdown Margin, Cold Condition, (% AK) 17.8 LaSalle station has upgraded its Standby Liquid Control System so that the B-10 enrichment has been increased from 18.9% to 45%. The above SBLC analysis assumes 660 ppm with the boron enriched to 45% B-10.
preparer: 7"'/Y6, .-I5-oo reviewer POW q.15.0"
.RANSMITTAL OF DESIGN INFORMATION- Seq. No. -D 0NF......
,NUCLEAR FUEL MANAGEMENT NFM 1D# NFM00001115 Page 7 of 21 IH. Control Rod Withdrawal Error The control rod withdrawal error event is analyzed at 100% of rated power, 100% of rated flow*
and unblocked conditions only.
Distance Withdrawn (ft) ACPR "12(Unblocked) - 0.30.
The design complies with the SPC 1% plastic strain and centerline melt criteria via conformance to the PAPT (Protection Against Power Transient) LHGR linmit. 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 the GE mechanical overpower, protection (MOP) criteria..
TIT. Fuel Loading Error The Fuel Loading Error, including fuel mislocation and misorientation, is classified as an accident. By demonstrating that the Fuel Loading Error meets the'more stringent Anticipated Operational Occurrence (AO0) 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 a'd hence off-site dose requirements are-met for the Fuel Loading Error.
JHI.1 Fuel Mislocation Error The following value bounds both the SPC and the co-resident GE fuel types.
- Event 'ACPR '
Mislocated Bundle 0.23 1H.2 Fuel Misrotation EVrrr' The following value bounds both the SPC and the co-resident GE fuel types.
-~Event ;- ALCPR Misoriented Bundle 0.15 preparer: r~H. ? /T-ivoO reviewer U . .00
I NUCLEAR FUEL MANAGEMENT NFM00001 15 11 TRANSMITTAL OF DESIGN INFORMATION I NFM Seq. ID#
No.
IPci*A 8 f 91 0
I IV.' Control Rod Drop Accident LaSalle is--abanked p-sito-6-wiihthdw-asequence-plant.-In order-to allow the site the option of inserting control rods using the simplified control rod sequence shown in Table 1, a control rod drop accident analysis was performed for the simplified sequence. The results from this simplified sequence analysis bound those where BPWS guidelines are followed. The results demonstrate that the simplified shutdown sequence meets the Technical Specification limit of 280 callg for a control rod drop accident. Therefore, the simplified sequence is valid for for control rod insertion for shutdown.
An adder of 0.32 %AK is incorporated in this analysis (for other than 00 to 48 control rod drops) to account for possible rod mispositioning errors as well as clumping effects.
Maximum Dropped Control Rod Worth, %AK 1.375 Doppler Coefficient, AkWk*F -9.50E-06 Effective Delayed Neutron Fraction used 0.0053 Four-Bundle Local Peaking Factor 1.281 Maximum Deposited Fuel Rod Enthalpy, (cal/g): 222 Number of Rods Greater than 170 cal/g 266 Note that the limit on maximum deposited fuel rod enthalpy is 280 cal/g and the limit on the number of rods greater than 170 cal/g (failed rods) is 770 for the GE 8x8 fuel and 850 for the SPC ATRIUM-9B fuel (in LaSalle UFSAR).
V. Loss of Feedwater Heatin!
The loss of feedwater heating event is analyzed at 100% of rated power for 81%, 100% and 105%
of rated flow and an assumed inlet temperature decrease of 145"F. The event was analyzed from BOC to EOC. The ACPR value reported below is bounding for both the SPC and the co-resident GE fuel types and all the analyzed flows.
Event ACPR Loss of Feedwater Heating 0.23 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 preparer: ),n)H, IO-q--,v reviewer re 9A ,s t.'-
NUCLEAR FUEL MANAGEMENT JNFM ID# NFMC1OO1 15 "TRANSMITrAL
' OF DESIGN INFORMATION I Seq. No.
~Pone 9 of 21, -_0 1% plastic strain criteria via conformance to the mechanical overpower protection (MOP) limit.
- The design does not meet the GE thermal overpoweiprotection' (TOP) criteria during a loss of feedwater heatg event; hlence; the LHGR values in the COLR for the affected lattice are adjusted ac5coidingly (lfe#rence 9,13and 14) as:f61ows.
GE9B-PSCWB322-11GZ-IOOM-150-CECO Bundle (Fuel Type 1) _
, i
- D LHGR Limits for L2C9
,* ° Nodal Exposure Nodal Exposure LHGR Limit 0 0 13.75 11.8459 13.06 13.75 25.2182 '27.80 11.75 45.6410 50.31 10.31 55.2370 60.89 6.00 GE9B-P8CWB320-9GZ-10OM-150-CECO Bundle (Fuel Type 2)
LHGR Limits for L2C9 .
"Nodal Exposure Nodal Exposure LHGR Limit (GWDAMS- (GW ftl*
0 0 , 14.25 11.0152 12.14 14.25 23.7593 26.19 12.18 43.6866 48.16 10.80 54.3675 59.93 6.00 VI. Maximum Exposure. Limit Compliance I Note that the following exposures are based on a nominal Cycle 8 EOC exposure of 13750 MWD/MT and -a-nominal Cycle 9 exposure of 17800 MWD/MT. If Cycle 9 reaches it's long window (approximately 500 MWD/MTU beyond the nominal Cycle 9 energy),'the exposure limits will still be met --,
GE9B GE9B ATRrUM -9 ATRIUM-9B Exposure ,Projected .. Limit : Projected
-- ", Limit*
Peak Batch -... 39989 - 42000 -36794 . -- -NA
45399 - -NA 39460 - 48000 Peik Assembly Peak Rod NA- .- NA- - - -- 43243 -- 55000 Peak Pellet 62595 65000 54918 , 66000
- 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 GWDIMT for Peak Fuel Assembly (no change), 50.0 GWD/MT for Peak Fuel Rod and 60.0 GWD/MT for Peak Fuel Pellet.
preparer:. ?7"IJ, -I-s'c' reviewer PA* 0(
C*,.0"(
SNUCLEAR FUEL MANAGEMENT - NFM ID# NFM0(01 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No. 0 1 1 Page 10of 21 VIIL Spent Fuel Pool and Fresh Fuel Vault Criticality Compliance For the L2C9 reload, there are four new SPC ATRIUM-9B assembly types consisting of seven unique enriched lattices, as identified in 1.1 Fuel Bundle Nuclear Design Analysis.
V1I.1 Fresh Fuel Vault Criticality Compliance The fuel storage vault criticality analysis that is detailed in Reference 5 remains valid for the above lattices. All the new (ATRIUM-9B) assemblies comply with the fresh fuel vault criticality limiks, 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% Gd2O3.
Note that the new fuel vault is a moderation-controlled area which implies that hydrogenous materials will be limited within the new fuel storage array. IAdministrative controls as generally defined in GE SIL No. 152. (dated March 31,1976) must be incorporated for the area.
VII.2 Li Spent Fuel Pool Criticality Compliance The LaSalle Unit I spent fuel pool criticality analysis that is detailed in Reference 6 remains valid for the above lattices. All the new (ATRIUM-9B) assemblies comply with the spent fuel pool criticality limits, i.e., all lattices have an enrichment of less than 4.60 wt % U-235 and a gadolinia content that is greater than 8 rods at 3.0 wt% Gd2O3 .
VI.3 L2 Spent Fuel Pool Criticality Compliance The LaSalle Unit 2 spent fuel pool criticality analysis that is detailed in Reference 7 remains valid for the above lattices. As shown below, all the new (ATRIUM-9B) assemblies comply with the LaSalle Unit 2 spent fuel pool criticality limit of k-eff < 0.95.
Lattice Type Maximum Maximum in-Rack Spent Fuel Pool k-inf* k-eff** k-eff Limit SPCA9-4.21L-13G8.0-100M 1.169 < 0.85 0.95 SPCA9-4.27L-12G8.0-100M 1.180 < 0.85 0.95 SPCA9-4.53L- 11G8.0-100M 1.192 < 0.85 0.95 SPCA9-4.56L-12G8.0-100M 1.187, < 0.85 0.95 SPCA9-3.96L-8G5.0-100M 1.231 < 0.86 0.95 SPCA9-4.58L-8G6.0/4G3.0-100M 1.233. < 0.86 0.95 SPCA9-4.58L-8G6.0-100M 1.236 < 0.86 0.95
- From 68 OF, uncontrolled CASMO-3G results.
- From Figure 6.1 of Reference 7.
preparer: -T't* ?, (S o re-v00 reviewer otj,, ,(S. 0 0
NUCLEAR FUEL MANAGEMENT NFM ID# NFM0000115 TRANSMITTAL OF DESIGN INFORMATION Seq. No. 0
,- -.. . .Pagellof2l VIII. References Z. 1. "LaSalle Unit 2 Cycle 9 Reload Analysis". Siemens Power QCoporationi EMF-2437,. Latest Revision.
- 2. "LaSalle Unit 2 Cycle 9 Plant Transient Analysis", Siemens Power Corporation, EMF-2440, Latest Revision.
- 3. ,' LaSalle 2 cycle 9 Core Design," ND0T NFMOO00056 Seq. 1, April 7, 2000 and 7L2C9 FLLP."
BNDL:00-005, Revision 0, 4/7/2000...
- 4. Commonwealth Edison, Nuclear Fuel Services, NFSR-0091, "Be, chmark of CASMO/MJCROBURN BWR Nuclear Design Methods", as supplemented and approved.
- 5. "Criticality Safety Analysis for ATRIUM-9B Fuel, LaSalle Units 1 and 2 New Fuel Storage Vault,"
Siemens PoweCrporatibn, EMF-95-134(P), December 1995. [ND1T 960089, Rev. 0]
- 6. "Criticality Safety Analysis for ATRIUM-9B Fuel, LaSalle'Unit 1 Spent Fuel Storage Pool (BORAL Rack)," Siemens Power Corporation, EMF-96-117(P), Apiil 1996. [NDIT 960087, Rev. 0]
- 7. "Criticality Safety Analysis for ATRIUM-9B Fuel, LaSalle'Unit 2 Spent Fuel Storage Pool (Boraflex Rack)," Siemens Power-Corporation, EMF-95-088(P), February 1996. [NDIT 960088, Rev. 0]
- 8. "L2C9 Standby Liquid Control System Worth Calculations," BNDL:00-028, Revision 0, July 14, 2000." , , .
- 9. "L2C9 Loss of Feedwater Heating Licensing Analysis," BNDL:00--024, Re'1sion 0, Juily13, 2000."
- 11. 'L2C9 Rod Withdrawal Error MOP/TOP Analysis," BNDL:00-023, Revision 0, August 17, 2000.
!do '_ Magi Caclain" BND:07°2
- 12. "La*Safle Unit 2 Cycle 9 Neutronid Licefising Shutdown Margin Calculation," BNDL:O0-032, Revision 0, August 17, 2000.
-13. "LaSalle 2 Cycle 9 LFWH TOP Violation and LHGR Limrit Calculation," Letter NFM:BND:00-050, July 13,2000. . .
- 14. "LaSalle 2 Cycle 9 GE9 Curve Adjustment forLFWH TOP Violation-," GE Letter KF-00-063, August 24,2000 *--. ,:,
- 15. "IaSýille2 Cycle 9 LFWH TOP Violation and LHGR Limit Calculaiion," Letter NFM:BND:00-050, July 13, 2000.
- 16. "L2C9 Mislocation Licensing Analysis," BNDL:00-025, September 2000.
- 17. "L2C9 Bundle Misorientation Analysis," BNDL'00-030, September 2000.
preparer: 1"7"'H, '-3!-a" reviewer ( .31 -00
NUCLEAR FUEL MANAGEMENT - NFM ID# NFM0000115 TRANSMITTAL OF DESIGN INFORMATION Seq. No. 0 Pace 12 of 21 Table 1
--- -L2C9 Simplifi-lShi-tdow-n-Sequ eýý -- -..
Shutdown From an Al Sequence' Insertion Rod Group* (Bank) Comments**__
7 or 8 48-00 Either Group 7 or 8 may be inserted first.
10 48-00 Groups 7 and 8 must be fully inserted prior to inserting any Group 10 rod.
9 48-00 Group 10 must be fully inserted prior to inserting any Group 9 rod.
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 48-00 Group 4 must be fully inserted prior to inserting any Group 3 rod.
2 48-00 Group 3 must be fully inserted prior to inserting any Group 2
__ _ _ rod.p 8 7 48-00 Group 2 must be fully inserted prior to inserting any Group 7 rod.
Shutdown from an A2 Sequenc Insertion Rod Group* (Bank)- GCommentsu 9 or 10 48-00 Either Group 9 or 10 may be inserted first.
8 48-00 Groups 9 and ,10 must be fully inserted prior to inserting any I Group 8 rod.
7 48-00 Group 8 must be fully inserted prior to inserting any Group 7 rod..
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 Growp 4 rod.
3 48-00 Group 4 must be fully inserted prior to inserting any Group 3 rod.
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.
- Group definitions are from LAP-100-13 Revision 21.
- The standard BPWS rules concerning out-of-service rods apply to the shutdown sequences.
preparer, H.Y-/-O_ reviewer pAL)
T 1.6.Oa
NUCLEAR FUEL MANAGEMENT NFM ID# NFM0000115 TRANSMITTAL OF DESIGN INFORMATION " Seq. No. "0 Paoe 13 of 21 FT16 FT,17
-... 40o-Bundles--.-- - --120 Bundles ii i NaturalU 11" 'Natural U 4.53 w/o See Figure 3'-- 4.53 w/o 11G8.0 36" 11G8.0 84" 1OO~
4.27 w/o.~ See*Figure 6 12G8.0 60" See Figure 4 -* '4.56 w/o 36 See Figure 5 6w 3.91 w/6' 4.10 w/o.
SPCA9-391B-14G8.0-100M SPCA9-410B-19G8.0-100M Figure 1. L2C9 Bundle Design (Fuel Types 16 and 17) preparer: 9?liH"*'- 341-coe3 reviewefp4W 4,31,00
NUCLEAR FUEL MANAGEMENT - NFM ID# NFM00001 15 TRANSMITAL OF DESIGN INFORMATIONI Seq. No 0I A O1 EPone 14 of 210 FT18 FT19 132 Bundles 56 Bundles 11 " Natural U 'i" 74 See Figure 6 3.96 w/o 72"1 8G5.0 42" See Figure 7 See Figure 8-* 4.58 w/o 8G6.0 24" See Figure 9 0 60" 4.58 w/6 4- See Figure 5 12GZ 66" 6" Natural U 6"1 3.83 w/o 3.96 w/o SPCA9-383B-16G8.0-100M SPCA9-396B-12GZ-100M Figure 2. L2C9 Bundle Design (Fuel Types 18 and 19) preparer: m4"/yj31-0c, reviewer Qfat, 31-o
NUCLEAR FUEL MANAGEMENT NFM ID# NFM0I* 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No. 0
'I IPaae 15 of 21
" "1 * -' *--- 3 5 " 3 - 21
-4.00 4.40 4.70 4.95 4.70 4.40 3.60 3.00 3.00 3.60 2 4 4 41 4 2 4.9S 4.95 4.95 4.95 3.60 3.60, 0 4 4 4, 4 4 4 3 3 4 4.95 4.9i 4.9 4.95 4.95 4.95 4.40 4.40 4.95
'** 4.705 ..... 4.954 4.95 4 "-" -" -. . ."*: *-*"' 4.' -04.954 !4.954 , "
4 " 4 4- 4 4
- 4.70 4.95 44.S9 4.95 4.95 45 44 4.40 4.9540 4.95 - 4.95 4.9 "34.0 , 4.9S 4.95 4.95 4.95
- 4.95 4.90
. 4.403.
.4 4 4 43 1 2 3 4 4 4 3 2 1 4.40 4.95 S - 4.95 4.95 4.40 3.60 " 3.00 3.00 -3.60 -
TYPE # ENR GD 1 4 . .3.00. 0 28 3.60 0 .
3 8 44.40.'. 0 4 .37 Z, 4.95 2 0 5 4- 14.70-" 0 ,
06.~.i 0 (,-
7 0 0 8 11 4.40 8.00 9 0 0.00 0 Figure 3. SPCA9-4.53L-1lG8.0-100M Lattice Enrichment Distribuiion
,/-],
preparer:, ?'"3W1O reviewer M'UTJ ,3/1,O0
NUCLEAR FUEL MANAGEMENT NFM ID# NFM00001 15 TRANSMITTAL. OF DESIGN INFORMATION Seq. No. a IASMTA POEe 16 of 21 1 Rods (4) 3.00 wlo U-235 2 Rods (12) 4.00 wlo U-235 3 Rods (8) 4.70 wlo U-235 4 Rods (36) 4.95 wlo U-235 G1 Rods (8) 4.20 w/o U-235+8.0 w/o Gd203 G2 Rods (4) 4.70 wlo U-235+8.0 w/o Gd203 Figure 4. SPCA9-4.56L-12GS.0-100M Lattice Enrichment Distribution preparer: g7hJ, &8-3)- 0o reviewer jPv/W ,j'1.00
INUCLEAR FUEL MANAGEMENT NFM ID# NFMOOOO1 15 Seq. No. 0
' '= -- .. OF
, TRANSMITTAL . ..INFORMATION
. DESIGN ... .... ... ... ... Page 17 of 21 TYPE ENR GD 1 4 2.60. 0 2- 8 3.20 0 3 14 4.00 0 4 31 4.70, 0 5 2 4.40 0 6 0 0 7 0 0 8 13 4.40, 8.00 9 0 0.00 0 Figure 5. SPCA9-4.21L-13G8.0-100M Lattice Enrichment Distribution reviewer &AW *3*3.049 preparer:
I NUCLEAR FUEL MANAGEMENT NFM ID# NFM0000 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No. 0 1 OF DIPGSe 18 of 21 TYPE # ENR GD 1 4 2.60 0 2 8 3.20 0 3 8 4.00 0 4 36 4.70 0 5 4 4.40 0 6 0 0 7 0 0 a 12 4.40 8.00 9 0 0.00 0 Figure 6. SPCA9-4.27L-12G 8.0-100M Lattice Enrichment Distribution preparer: W-11), P-/-31 reviewer r NO g,316
I'frM IL1* i'JIIVILAAAJI I~mlIBII..J NUCLEAR FUEL MANAGEMENT I -0IU I Seq. No..-
I
-"4 TRANSMITTAL OF DESIGN INFORMATION - NFMIM IL)
Pooe 19 of 21
-43
_iI I 2
4, 3 24 221 3 4.40 4.40 3.80 2.60 4.40 3.40
_ 2.60 3.40 3.80 4 S2 2 3.40 2 2 4.40 3.40 4.40 4.40 3.40 . 4 4 '4 . 4 4 41 3.80 3 4 .4 4.40 S4.40* 3.80 4.4O 4.40 4.40.
3.80 4.40 4.40 Ti_____t A
~-, .4
-m 4 Li :4 .4 "4 "4.40 "4.40
-4.40 4
4.40
- 33.
4 4.4 4.40
.4 4.40 ......
I 4 4.40
-. 4
""4.40 4
4 "4.40 P
4
-4 34D 4 N 4 4.40 4.40 4A0 E4. 4.4 i 1" 1 S...
_____ .. . i i 4 3-F .4 4
- 4 4 4 4 4.40 3.80 3 4.
4.40 4.40 4.40 4.4O 4.40 4.40 3.80 2 "2
.4A 4 2 ~ 4 4.40 3:40 "~3.40" 2 4.40 4.40 3.40 3.40
-3A0I 4 3 3, 4 2 1 2 4 4.40 3.80 3.40 2.60 3.80 4.40 S2.60 3.4O -4.40-I 1. I 1 Rods (4) 2.60 w/o Y-235 Rods (12)" 3.40 wo U-235 2
Rods (8) 3.80 wlo U-235 3
A4 Rods (40) 4.40 w/o U-235 GI Rods (8) 3.40 w/o U-235+5.0 wlo Gd203 Figure 7. SPCA9.3.96L-8G5.0-004 M Lattice Enrichment Distribution reviewer f"yk ,31.00 preparer: "71H I-3,-O0 6'
SNUCLEAR FUEL MANAGEMENT NFM ID# NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No. 0 Pooe 20 of 21 1 Rods.(4) 3.00 wlo U-235 2 Rods (12) 4.00 w/o U-235 3 Rods (8) 4.70 w/o U-235 4" Rods (40) 4.95 w/o U-235 G1 Rods (8) 4.20 w/o U-235+6.0 w/o Gd203 Figure 8. SPCA9-4.58L-8G6.0-100M Lattice Enrichment Distribution preparer. F-3)- 0 0 reviewerrAL,)
193(-0o6
NUCLEAR FUEL MANAGEMENT NFM ID# NFMOOOO 115 TRANSMITTAL OF DESIGN INFORMATION Seq. No. 0
, Pace 21 of 21
(.
1 Rods (4) 3.00 w/o U-235 2 Rods (8) 4.00 w/o U-235 3 Rods (8) I 4.70 wlo U-235 4 Rods (40) 4.95 wlo U-235 G1 Rods (8) 4.20 wlo U-235+6.0 w/o Gd203 G2 Rods (4) 4.00 wlo U-2d5+3.0 w/o Gd203 Figure 9. SPCA9-4.58L-8G6.0/4G3.0-100M Lattice Enrichment Distribution iH' *3/-00re preparer: '7-I rvee PAA)
Technical Requirements Manual - Appendix J L2C9A Reload Transient Analysis Results It h Attachment 2 LaSalle Unit 2 Cycle 9 Reload Analysis Report LaSalle Unit 2 Cycle 9A November 2002
.SIEMENS EMF-2437 Revision 0 LaSalle Unit 2 Cycle 9 Reload Analysis October 2000 Siemens Power Corporation Nuclear Division
Siemens Power Corporation ISSUED INSPC ON-LINE DOCUMENT SYSTEJA DATE: 11) Z~ EMF-2437 Revision 0
,' Is
-LaSalle Unit 2 Cycle 9 Reload Analysis Prepared:
SDate J. M. Haun, Engineer BWV,'Neutronics Prepared: Date r.B.McBum ,Engineer B afe alysis ,-:-
Prepared: Date A. White, Engineer Product Mechanic Engineering Concurred: 4 A Iate H. D.' C4%.Manager Date Produ~ctL~censing- -_ *::/;*.
Concurred: /D/te D. J. Denver, Manager, ' - Date Commercial Operations Approved: 4a Date Approved: It) /10 3/OL M. E. Garrett, Manager/ Date Safety Analysis Approved: 10-03-tM T. M. Howe, Manager Date Product Mechanical Engineering
/sp
Customer Disclaimer Important Notice Regarding the Contents and Use of This Document PleaseRead Carefully Siemens Power Corporation's warranties and representations concerning the subject matter of this document the agreement between Siemens Power Corporation are those set forth in Customer pursuant to which this document Is issued. and the except as otherwise expressly provided in such Accordingly, Siemens Power Corporation nor any person acting agreement, neither on its behalf:
- a. makes any warranty or representation; express or implied, with respect to the accuracy, completeness, or usefulness the information contained in this document, or that of any information, apparatus, method, or process the use of disclosed in this document will not infringe privately owned rights; or
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EMF-2437 Revision 0 LaSalle Unit 2 Cycle 9 Page Hi M 1* AAn I li Nature of Changes Item Page Description and Justification
- 1. All This is a new document.
I,
EMF-2437 Revision 0 LaSalle Unit 2 Cycle 9 Page iii Reload Analysis
- -Contents 1-I 1.0 Introduction ..................................................................................................
............. 2-1 2.0 Fuel Mechanical Design Analysis ...........................
............... 3-1 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 Centedine Temperature ..........................
3.2.5 Bypass Flow ................ *...................................................................... 3-1 3.3 MCPR Fuel Cladding Integrity Safety Limit (SLMCPR) .................................... 3-1 Coolant Thermodynamic Condition .................................................... 3-1 3.3.1 3.3.2 Design Basis Radial Power Distribution ............................................. 3-2 3.3.3 Design Basis Local Power Distribution ............................................... 3-2 3-2 3.4 Licensing Power and Exposure Shape ............................................ :................
......................................... 4-1 4.0 Nuclear Design Analysis ................. * ........................... 4-1 4.1 Fuel Bundle Nuclear DesignAnyis ..............................
Aayi............................
4-2 4.2 Core Nuclear Design Analysis ......................................................................
4.2.1 Core Configuration ........................................................................ 4-2 4.2.2 Core Reactivity Characteristics .......................... 4-2 Core Hydrodynamic Stability ........................... 4-2 4.2.4
. . . . ............................. 5-1 5.0 Anticipated Operational Occurrences Analysis of Plant Transients at Rated Conditions ..................... 5-1 5.1 5-1 5.2 Analysis for Reduced Flow Operation ..............................................................
Analysis for Reduced Power Operation ............................................................ 5-2 5.3 ASME Overpressurization Analysis ............................. 5-2 5.4 Control Rod Withdrawal Error ......................................................................... 5-2 5.5 5.6 Fuel Loading Error ....................................................................................... 5-2 Determination of Thermal Margins ................................................................... 5-2 5.7 6.0 Postulated Accidents ......................... 6-1 6.1 Loss-of-Coolant Accident............ ................................
6.1.1 Break Location Spectrum .............................. 6-1 6.1.2 Break Size Spectrum ......... 6-1
.2.
6 Contr 6.1.3 l MAPLHGR HGdro Accidents.....
Analyses .......................................................................
...... ................... .....--.- ................ 6-1 6-1 6.2 Control Rod Drop Accident ....................................
Spent Fuel Cask Drop Accident, ................. .............. 6-1 6.3
.......................................... 7-1 7.0 Technical Specifications Limiting Safety System Settings....................................................................... 7-1 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 ......... ..................... o
............... 7-2
LaSalle Unit 2 Cycle 9 EMF-2437 Reload Analysis Revision 0 Page iv 8.0 Methodology References ........................................................................................ 8-1 9.0 Additional References ........... ........... .......................................... 9-1 Siemens Power Corporation
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page v Tables 1.1 EOD and EOOS Operating Conditions ........................................................................ 1-2 3.1 Licensing Basis Core Average Axial Power Profile and Licensing Axial Exposure Ratio ............................................................................................................ 3-3 4.1 Neutronic Design Values ............................................................................................. 4-4 5.1 EOC Base Case and EOOS MCPRp Limits and LHGRFACp Multipliers for TSSS Insertion Times .............................................................................................. 5-4 5.2 EOC Base Case MCPRp Limits and LHGRFACp Multipliers for NSS Insertion Times ............................................................................................................ 5-6 5.3 Coastdown Operation Base Case and EOOS MCPR, Limits and LHGRFACp Multipliers for TSSS Insertion Times ........................................................ 5-7 5.4 FFTR/Coastdown Operation Base Case and EOOS MCPRI Limits and LHGRFACp Multipliers for TSSS Insertion Times ......................................................... 5-9 Figures 3.1 Radial Power Distribution for SLMCPR Determination ................................................. 3-4 3.2 LaSalle Unit 2 Cycle 9 Safety Limit Locýl Peaking Factors SPCA9-391B-14G8.0-100M With Channel Bow ......................................................... 3-5 3.3 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-410B-19G8.0-100M With Channel Bow .......................................................... 3-6 3.4 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-383B-16G8.0-100M With Channel Bow............................ 3-7 3.5 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-396B-12GZ-100M With Channel Bow ............................................................. 3-8 4.1 LaSalle Unit 2 Cycle 9 Reference Loading Map ........................................................... 4-5 5.1 Flow-Dependent MCPR Limits for Manual Flow Control Mode ................................... 5-11 5.2 Flow Dependent LHGR Multipliers for ATRIUM-9B Fuel ............................................. 5-12 5.3 EOC Base Case Power-Dependent MCPR Limits for ATRUM-9B Fuel - TSSS Insertion Times ...................................................................................... 5-13 5A EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel - TSSS Insertion Times ............... .................. ........... 5-14 5.5 EOC Base Case Power-Dependent MCPR Limits for ATRUM-9B Fuel- NSS Insertion Times ................................ 5-15 5.6 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel- NSS Insertion Times ........................................................................................................... 5-16 5.7 Starting Control Rod Pattern for Control Rod Withdrawal Analysis ........................... 5-17 7.1 Protection Against Power Transient LHGR Limit for ATRIUM-9B Fuel ......................... 7-3
EMF-2437 Revision 0 LaSalle Unit 2 Cyle 9 Page vi Reload Analysis Nomenclature AO0 abnormal operational occurrence BOC beginning of cycle EFPH effective full power hours EOC end of cycle EOD extended operating domain EOFP end of full power EOOS equipment out of service FFTR final feedwater temperature reduction FHOOS feedwater heater out of service FWCF feedwater controller failure ICA interim corrective actions ICF increased core flow LFWH loss of feedwater heating iJ linear heat generation rate ilI LHGR fl LHGRFAC LHGR multiplier i i
LOCA loss of coolant accident LPRM local power range monitor LRNB load rejection no bypass MAPFAC MAPLHGR multiplier MAPLHGR maximum average planar linear heat generation rate MCPR minimum critical power ratio MELLLA maximum extended load line limit analysis 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
i LaSalle Unit 2 Cycle 9 EMF-2437 Reload Analysis Revision 0 Page vii TSSS technical specification scram speed UFSAR updated final safety analysis report ACPR change in critical power ratio Siemens Power Corporation
.EMF-2437 Revision 0 Page 1-1 LaSalle Unit 2 Cycle 9 Reload Analysis 1.0 Introduction by Siemens Power Corporation This report provides the results of the analysis performed 9 reload for LaSalle Unit 2. This (SPC) as part of the reload analysis in support of the Cycle topical Report XN-NF-80-19(P)(A),
report is intended to be used in conjunction with the SPC to BWR Reloads, which describes Volume 4, Revision 1, Application of the ENC Methodology methodology used for those the analyses performed in support of this reload, identifies the in this report are the same as analyses, and provides a generic reference list. Section numbers 4, Revision 1. Methodology corresponding section numbers in XN-NF-80-19(P)(A), Volume 4, Revision 1, is referenced in used in this report which supersedes XN-NF-80-19(P)(A), Volume d cuments, Section 8.0. The NRC Technical Limitations presented in the methodology satisfied by these analyses.
including the documents referenced in Section 8.0, have been are described elsewhere.
Analyses performed by Commonwealth Edison Company (CoinEd) or the appropriate This document alone does not necessarily identify the limiting events must be determined in operating limits for Cycle 9. The limiting events and operating limits conjunction with results from CornEd analyses.:
348 unirradiated and 256 The Cycle 9 core consists of a total of 764 fuel assemblies, including
-The reference core irradiated ATRIUMm-9B assemblies arnd 160'irradiated GE9gassemblies.
configuration is described in Section 4.2.
The design and safety analyses reported in this document were based on the design and operating cycle. The operational assumptions in effect for LaSalle Unit 2 during the previous The extended effects of channel bow-are explicitly accounted for in the safety limit analysis.
presented in Table operating domain (EOD) and equipment out of service (EOOS) conditions 1.1 are supported.
ATRIUM is a trademark of Siemens.
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 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)
Coastdown Final Feedwater Temperature Reduction (FFTR)
FFTR/Coastdown 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 (100% available at startup)
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.
Siemens Power Corporation
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 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 2 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.
EMF-2437
. . .Revision 0 LaSalle Unit 2 Cycle 9 Page 3-1 Reload Analysis 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 2 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 2 core are provided in Reference 9.1, Figure 4.2.
3.2.3 Fuel Centerline Temperature ApplicableReport ATRIUM-9B Reference 9.1,
.- Figure 3-.3 3.2.5 Bypass Flow Calculated Bypass Flow 14.8 MIb/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 e1ati -
Reference 9.3 Single-Loop Operation° 1.12 3.3.1- Coolant Thermodynarnic Co'ndition Thermal Power (at SLMCPR) 5167.29 MWt Feedwater Flow Rate (at SLMCPR) 22.4 Mlbm/hr Core Exit Pressure (at Rated Conditions) 1031.35 psia Feedwater Temperature 426.5"F Includes the effects of channel bow, up to 2 TIPOOS (or the equivalent number of TIP channels), a 2500 EFPH LPRM calibration interval, cycle startup with uncalibrated LPRMs (BOC to 500 MWd/MTU). and up to 50% of the LPRMs out of service.
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 3-2 7
3.32 Design 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 Design Basis Local Power Distribution Figures 3.2, 3.3, 3.4 and 3.5 show the local power peaking factors used in the MCPR Fuel Cladding Integrity Safety Limit analysis.
SPCA9-391 B-1 4G8.0-1 OOM Figure 3.2 SPCA9-41OB-19G8.0-100M Figure 3.3 SPCA9-383B-1 6G8.0-1 OOM Figure 3.4 SPCA9-396B-12GZ-100M Figure 3.5 3.4 Licensing Power and Exposure Shape The licensing axial power profile used by SPC for the plant transient analyses bounds the projected end of full power (EOFP) axial power profile. The conservative licensing axial power profile as well as the corresponding axial exposure ratio are given in Table 3.1. Future projected Cycle 9 power profiles are considered to be in compliance when the EOFP normalized power generated in the bottom of the core is greater than the licensing axial power profile at the given state conditions when the comparison is made over the bottom third of the core height.
Siemens Power Corporation
EMF-2437 LaSalle Unit 2 Cycte 9 Revision 0 Page 3-3 Dalmorl Analysis Table 3.1 Licensing Basis Core Average Axial Power Profile and Licensing Axial Exposure Ratio State Conditions for Power Shape Evaluation Power, MWt 3489.00 Core Pressure, psia 1020.00 Inlet Subcooling, Btu/ibm 18.20 Flow, Mlb/hr 108.50 Licensing Axial Power Profile 1* Node Power Top 25 0211 24 0.417 23 0.967 "22 1.207 S, 21 1.371 S* : *
- 20 1.445 19 1.454 18 1.428 1< j 17 1.384 S15 1.346
-1.299
'14 1.248 13 1.199 1.151 r , " 1.102 10 1.053 9 1.002 8 0.944 7 0.887 6 0.835 0.796 0.770 3 0.726 2 0.583 Bottom I 0.177 Licensing Axial Exposure Ratio (EOFP)
Average Bottom Sft/12 ft = 1.098
LaSalle Unit 2 Cycle 9 EMF-2437 Reload Analysis Revision 0 lib Page 3-4 200 175 150 a)
"o 125 C
m o 100 L
a)
E 75 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 Radial Power Peaking Figure 3.1 Radial Power Distribution for SLMCPR Determination Siemens Power Corporation
EMF-2437 Revision 0 LaSalle Un*It 2 cycle 9 Page 3-5 Reload Analysis C ontr:ol Rod Corner; 0
n t* 1.052 1.045-" -1.088 1.088' 1.104-- 1.079 1.068 1.013 1.005 r
0 I' 0.991 1.045 -0.951 1.019 0.996- '-0.852 0.986 -0.998 0.914 R
0O d* -1.088 1.019, -1.001 '1.059 -1.089 1.051- 0.982 0.981 1.027 C
0 1.088 0.996,1 1.059 - - 0.905 0.957 1.050 r
n Internal .....
e Water- '1.068- 0.807 1.035 1.104 0.852 -1.089 r +
Channel 1.079 0.986 -1.051 . " 1.025 0.942 1.039 1.068 0.998- 0.982-, '0.905 1.068 -A 1.025 ,0.811 0.954 1.005
-1.013 0.914 -0.981 - 0.957- <0.807 - 0.942 0.954 0.874 --0.957 I I 1.005. '0.991 :1.027- -1.050-- -1.035 1.039- 1.005 '0.957 1-0.956 Figure 3.2 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-391B-14G8.0-100M With Channel Bow
LaSalle Unit 2 Cycle 9 EMF-2437 Revision 0 Reload Analysis Page 3-6 Control Rod Corner 0
n t 1.058 1.049 1.092 1.091 1.107 1.082 1.072 1.017 1.010 r
0 I 1.049 0.945 1.020 0.996 0.843 0.987 0.998 0.906 0.995 R
4.
0 T 1*
d 1.092 1.020 1.002 1.061 1.090 1.052 0.981 0.980 1.030 C
0 1.091 0.996 1.061 0.894 0.955 1.053 n Internal r 1.053 1.107 0.843 I 0.955 I. -
r B 1.090 Water -
1.067 0.797 1.036 S~Channel 1.082 0.987 1.052 1.024 0.941 1.041 1.072 0.998- 0.981 0.894 1.067 1.024 0.800 0.952 1.007 1.017 0.906 0.980 0.955 0.797 0.941 0.952 0.865 0.960 1.010 0.995 1.030 1.053 1.036 1.041 1.007 0.960 0.960 Figure 3.3 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-410B-19GS.0-100M With Channel Bow Siemens Power Corporation
EMF-2437 Revision 0 LaSalle Unit 2 Cycle 9 Page 3-7 Reload Analysis C o6n t r o Rod - Co6rnr R -..*°EMF-2437 0 S....
PageRevision 0
3-7 n 0.970 "
t .1.017 1.017 1.068 1.083 - -1.107 1.074 - 1.048 - 0.985 r
0
'1.017 0.986 1.024- .1.000 0.885 0.992 1.004 0.956 " 0.965 R
0 1.068 1.024 0.890 1.063 1.091 - 1.055- 0.990 0.989-- 1.009 d
C 1.083 1.000 1.063 10.944 10.966 1.055 0
r Internal - "
n i e 1.040 r 1.107-- 0.885. 1.091 Water 1.074 0.846
- -Channel 1.074 -0.992 -1.055 .. .... 5 1.043 1.048 1.004 0.990. _0.944 .- 1.074 1.032- 0.850 0.964 0.988 0.985 0.956 0.989 0.966 *0.846 0.951 -0.964 0.916 0.932
'0.970 ,0.965 1.009 1.055 .1.040 1.043 0.988 -0.932 -0.924 Figure 3.4 LaSalle Unit 2 Cycleb9 Safety Limit Local Peaking Factors SPCA9-383B-16G8.0O1DOM With Channel Bow-
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 3-8 C ontrol Rod Corner 0
n t 1.025 1.058 1.062 1.117 1.100 1.108 1.043 1.026 0.979 r
"*o I
1.058 0.934 1.018 0.852 1.003 0.845 0.999 0.903 1.005 R
0 d 1.062 1.018 1.003 1.067 1.092 1.058" 0.984 0.983 1.006 C
0 1.117 0.852 1.067 1.046 0.823 1.056 r
n Internal e
r 1.100 1.003 1.092 Water 1.072 0.968 1.039 I Channel 1.108 0.845 1.058- 1.038 0.816 1.046 1.043 0.999 0.984 1.046 1.072 1.038 0.965 0.963 0.986 1.026 0.903 0.983 0.823 0.968 0.816 0.963 0.873 0.973 0.979 1.005 1.006 1.056 1.039 1.046 0.986 0.973 0.933 Figure 3.5 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-396B-12GZ-100M With Channel Bow Siemens Power Corporation
- I EMF-2437 Revision 0 LaSalle Unit 2 Cycle 9 Page 4-1 4.0 Nuiclear Design Aialnysis "
4.1 Fuel Bundle NuclearDesign Analysis The detailed fuel bundle design information for the fresh ATRIUM*-gB fuel to be loaded in, LaSalle Unit 2 Cycle 9 is provided in References 9.1 and 9.12. The following summary provides the appropriate cross-references.
Assembly Average Enrichment (ATRIUM-9B fuel) ,
3.91 wt%
SPCA9-391 B-14G8.0-1 0DM (FT16)
(FT17) 4.10 Wt%f SPCA9-41OB-19G8.0-100M -3.83 wt%
SPCA9-383B-16G8.0-100M (FT18)
(FT19) S3.96 wt%
SPCA9-396B-12GZ-100M Radial Enrichment Distribution SPCA9-4.56L-12G8.0-100M Ref. 9.12 Figure B.19 Ref. 9.1 "FigureD.1 SPCA9-4.21L-13G8.0-1 00M "Figure D.2 SPCA9-4.27L-12G8.0-100M Ref. 9.1 Ref. 9.1 Figure D.3 SPCA9-4.53L-11G8.0-100M Figure B.122 SPCA9-3.96L-8G5.0-1OOM Ref. 9.12
- Ref.., 9.12 'Figure B.140 SPCA9.4.58L-8G6.014G3.0-10OM: Figure B.157 SPCA9-4.58L-8G6.0-100M' Ref. 9.12 Axial Enrichment Distribution Ref. 9.1 Figures 5.1-5.4 Ref. 9.1 Figures 5.1-5.4 Burnable Absorber Distribution Non-Fueled Rods Ref. 9.1 Figures 5.1-5.4 Neutronic Design Parameters Table 4.1 Fuel Storage 1, LaSalle New Fuel Storage Vault . Reference 9.4
. The LSB-2 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.
LaSall e Unit 1 Spent Fuel Storage Pooi (BORAL Racks) Reference 9.5 The LSB-2 Reload Batch'fuel designrs meet the fuel design limitations defined in Table 2.1 of Reference 9.5 and therefore can be safely stored in the pool.
LaSalle Unit 2 Cycle 9 EMF-2437 Reload Analysis Revision 0 Page 4-2 LaSalle Unit 2 Spent Fuel Storage Pool (Boraflex Racks)
Reference 9.6 The LSB-2 Reload Batch fuel designs can be safely stored as long as the fuel assembly reactivity limitations defined in Reference 9.6 are met.
< CoinEd has responsibility to confirm that fuel meets reactivity limitations. >
4.2 Core NuclearDesign Analysis 4.2.1 Core Configuration Figure 4.1 Core Exposure at EOC8, MWd/MTU 27,893.9 (nominal value)
Core Exposure at BOC9, MWd/MTU" 11,808.0 (from nominal EOC8)
Core Exposure at EOC9, MWd/MTU 30,266.2.
(licensing basis to EOFP)
NOTE: Analyses in this report are applicable for EOFP up to a core exposure of 30,266.2 MWd/MTU.
< Cycle 9 short window exposure to be determined by ComEd.>
4.2.2 Core Reactivity Characteristics
< This data is to be furnished by ComEd. >
4.2.4 Core Hydrodynamic Stability Reference 8.7 LaSalle Unit 2 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:
Siemens Power Corporation
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 4-3 Provides trending information to qualitativiely compare the stability from cycle to cycle.
Provides decay ratio sensitivities to rod line and flov changes near the ICA're-gions.
- 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 2 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 determnine the highest expected decay ratio throughout the cycle. The results from this analysis are shown below. ,
Power/Flow Maximum Maximum MY Global Regional 30.1/26.6 0.59 0.53; 31.6/29.2 P.40 0.50
- t 61.9/45.0 0.50 0.88 73.6150.0 0.52 0.95 78.2/60.0 0.33 0.63 82.4/60.0 0.36 0.72 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 MVWt as rated. % flow is based on 108.5 Mlb/hr as rated.
Siemens Power Corporation
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437
- , Revision Page 4 0 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 18.2 Moderator Temperature, OF 548.8 Channel Thickness, inch 0.100 Fuel Assembly Pitch, inch 6.0 Wide Water Gap Thickness, inch 0.261 Narrow Water Gap Thickness, inch 0.261 Control Rod Data*
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 B4C, %TD 70 The control rod data represents original equipment control blades at LaSalle and were used in the neutronic calculations.
Siemens Power Corporation
EMF-2437 Revision 0 LaSalle Unit 2 Cycle 9 Page 4-5 Reload Analysis 9111315.1719212325 2729 31 33 35 373941 43 45 47 49 51 53 55 57 59 J: I 5 7 I:
60 1 1 1 1 58 19 19 19 19 1519 19 19 18 1 2 14 19 19 19 15 171141417 56 161Z11719 Z1192 1 9 54 151 1jj1518 15 1 14 18 17 18 18 14 18 14 41 71 71 1 1865 52 17 18 18 17 14 18 15 18 14 17 14 1 1179 19 17 14 52 1414 17115 15 15 15 15 1716 188189 1 18 [,4 15 18 22 19 50 1 14 1815417 17 14 14 18 18 18 18 142 2 14 48 46 17 1414 17 18118 14 18 14114 1817717 117 15 22 44 14 18 18 14 18 18 18 18 14 14 17 17 15 18 19 14 1 42 I1 18 14 14 18 15 17 14 18 1 18 17 17 17 19 2 1 15 161815714718481815 16 151 15[2"2 15 40 38 1 19 17 14 5 17 15 14 17 14 18 15 165116 F 15 1516 1522 1615181718 17 14 15 17417 1817144 181517119 18716 1522 16 16 16 16 14 15 16 15 18 1818 17 14 181151 36 34 1 19 *71/Z_ 18 1 jJ18 151618 14 17 1411751791 14 *!*_16 18 15 15 18 16 16161642411 1518 14 17 1991517 12 159 1 32 17 17 1 15 16 18 14 1414 18 18 15 15 17 18 16 15 15 15 14 151 18 14 2F1 1516 15 2 14 18 4 15 1811 14119 1 30 151'2215 119 1 518 175 17 1 21*9 17 814r"T*"1718 18 16 18 15 15 18 1616 15 18 149 17 14 29 17 15 28 219L*15 152 14j18 142 15416 16 16 16 16 14 181 18 12 16 15181817 18 15421 17 15 517 151-14 17 14 19 18 16 2 1 26 2 191711517 181471417156 '16 15 15 1616*
24 1 141918 15,171714181818 6151 1 8 15 2Lj 15
- 14 181 1 2 1511 15 19 2 22 91 1 21 2 61 18 14 14 18 15 17 14 14 18 18 18 17 15 14 17 19 2 2 1717 15 17 1814:1814 20 18 14 18 18 14 16 2 19 17 14 1 17 1851 18 17[1414117 14 14 1 14 17 17 17 2 1 1781421.2j.14 185!* ý18 17118 1_14 17 19 2/
12 15 15 15 15 10 17 18 18 17 15 8 2 151917151716 15
- 6 17i*4 17 16F17 1.5.17 19 15 2_. 2 14 19 19 19 15 4 19 19 19 19 15 19 19 19, 14 2 1 1 1 1 112217 5 2
Fuel Number Load Bundle Name of Bundles Typ 1 GE9B-P8CWB322-1 1GZ-10DM-150 84 7 76 7 2 GE9B-P8CWB32D-9GZ-10DM-15O SPCAg-381 B-13GZ7-80M 128 8 14 8 15 SPCA9-384 B-11GZ6-80M 128 40 9 16 SPCA9-391B-14 G8.0-1 0DM 120 9 17 SPCA9-410B-1 9G8.0-1 0DM 9 18 SPCA9-383B-1 6G8.0-1 0DM 132 56 9 19 SPCA9-396B-12GZ-1 0DM Figure 4.1 LaSalle Unit 2 Cycle 9 Reference Loading Map
EMF-2437 LaSalle Unft 2 Cycle 9 Revision 0 Reload Analysis Page 5-1 5.0 Anticipated Operational Occurrences Applicable Disposition of Events Reference 9.7 5.1 Analysis of Plant Transientsat 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-9BI/GE9 LRNB* "TSSS 422 127 1218 0.30/0.40 FWCF" TSSS 298 123 1176 0.25/0.31 LRNB" NSS 380 124 1211 0.28/0.37 FWCF NSS 263 120 1169 .0.23/0.29 tJ f LFWHt t t 5.2 Analysis for Reduced Flow Operation Reference 9.3 Limiting Transient: Slow Flow Excursion MCPRf Manual Flow Control - ATRIUM-9B and GE9 Fuel Figure 5.1 LHGRFAC- ATRIUM-9B Fuel Figure 5.2 MAPFACf- GE9 Fuel - 1 MCPR, and LHGRFACf results are applicable at all Cycle 9 exposures and in all EOD and EOOS scenarios presented in Table,1.1.
Based on 100%P/105%F conditions.
I This data to be furnished by CornEd.
Siemens PowerComoration
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis
ýben I.')_
5.3 Analysis for Reduced Power Operation Reference 9.3 Limiting Transient Load Rejection No Bypass (LRNB)
Feedwater Controller Failure (FWCF)
MCFRp Base Case Operation Tables 5.1-5.4 LHGRFACp Base Case Operation' Figures 5.3-5.6 Tables 5.1-5.4 MCPR?, EOOS Conditions Tables 5.1-5.4 LHGRFACp, EOOS Conditions" Tables 5.1-5.4 MAPFACp - All Operating Conditions" <To be furnished by CornEd.>
5A ASME OverpressurizationAnalysis Reference 9.3 Limiting Event MSIV Closure Worst Single Failure Valve Position Scram Maximum Vessel Pressure (Lower Plenum) 1346 psig Maximum Steam Dome Pressure 1320 psig 5.5 ControlRod WithdrawalError 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 CornEd. >
5.7 Determinationof 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 the MCPR and LHGR limits at reduced flow (MCPR, Figure LHGRFACp values presented are applicable to SPC fuel. GE MAPFACp limits will continue to be applied to GE9 fuel at off-rated power.
Siemens Power Corporation
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 MIAMlývl Anwluet Page 5-3 5.1; LHGRFACI, 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). Umits are presented for base case operation and the EOD arid EOOS scenarios presented in Table 1.1.
The results presented are based on the anialyses performed by SPC. As indicated above, the final Cycle 9 MCPR operating limits need to be established in conjunction with the results from CornEd analyses.
II.
LaSalle Unit 2 Cycle 9 EMF-2437
- Revision 0 Reload Analysis k,H, Page 5*.
Table 5.1 EOC Base Case and EOOS MCPRp Limits and LHGRFACp Multipliers for TSSS Insertion Times EOOS/EOD Power -ATRIUM-9B Fuel GE9 Fuel Condition (% rated) lvi'-m LMltNt-ALp MCPRp 0 2.70 0.78 2.70 Base 25 2.20 0.78 2.20 case 25 1.91 operation 0.78 1.99 60 1.46 1.00 1.52 100 1.41 1.00 1.51 0 2.85 0.69 2.85
-eeawater 25 I
2.35 0.69 2.35 heaters out-of-service 25 2.14 0.69 2.22 (FHOOS) 60 1.51 0.97 1.57 100 1.41 1.00 1.51 0 2.71 0.78 2.71 Single-loop 25 2.21 0.78 2.21 operation 25 1.92 0.78 2.00 (SLO) 60 1.47 1.00 1.53 100 1.42 1.00 1.52 0 2.70 0.76 2.70 Turbine 25 2.20 0.76 2.20 bypass valves out-of-service 25 1.98 0.76 2.08 (TBVOOS) 60 1.52 0.97 1.62 100 1.43 0.97 1.52 1.43 0.99 Siemens Power Corporation
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 5-5 I
Table 5.1 EOC Base Case and EOOS MCPR, Limits and LHGRFACp Multipliers for TSSS Insertion Times.
- (Continued)
EOOS i EOD Power ATRIUM-9B Fuel GE9 Fuel Condition (% rated)* MCPRp LHGRFACp MCpRM 0 2.70 0.78 2.70 Recirculation 25 2.20 0.78 2.20 pumpnpp 25 1.91 0.78 1.99 out-of-service (RPT) 60 1.51 0.89 1.61 100 1.51 0.89 1.61 0 2.70 0.70 2.70 Turbine control 25 2.20 0.70 2.20 valve (TCV) 25 2.10 0.70 2.10 slow closure ANDIOR 80 1.69 0.86 1.95 no RPT 80 " 1.61 0.89 1.84 100 1.53 0.89 1.63 0 2.65 0.68 2.85 TCV 25 2.35 0.68 2.35 slow closure/ 25 2.14 0.68 2.22 FHOOS 80 1.69 0.86 1.95 AND/OR no RPT 80 1.61 0.89 1.84 100 1.53 0.89 1.63 0 2.60 0.40 2.60 Idle 25 " 2.60 0A0 2.60 loop 25 2.60 0A0 2.60 startup 60 2.60 0.40 2.60 100 2.60 0.40 2.60 Siemens PowarComoratimn
LaSalle Unit 2 Cycle 9 EMF-2437
- Revision 0 Reload Analysis
%",I I . "Page5-6
'Table 5.2- EOC Base Case MCPRt Limits and LHGRFACp Multipliers for NSS Insertion Times
,Y Siemens Power Corporaton
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 0 1 A~sAn hilul Page 5-7 T'iable 5.3 Coastdown Operation Base Case and EOOS MCPR, Limits and LHGRFACp Multipliers for TSSS Insertion Times
.' ATRIUM-9B Fuel GE9 Fuel EOo S EOD. -,(% ;Power--,
rated)-, MCPRp LHGRFACp MCPRp.
'Condition 0 "2.70 0.75 2.70 Coastdown 25 2.20 :0.75 2.20 base case 25 '2.05 0.75 2.05 operaton 60 1.48 0.99 1.54 100 1.42 1.00 1.52 0 2.71 0.75 -2.71 25 2.21 0.75 2.21 single-loop with Coastdown 25 2.06 0.75 2.06 operation 60 1.49 0.99 1.55 o100o 1.43 1.00 1.53 0 -2.70 0.73- 2.70 Coastdown with turbine 25 2.20 0.73 2.20 bypass valves 25 2.05 0.73 2.15 out-of-service 60 1.55 0.97 1.64 (TBVOOS) 100 1.44 0.99 1.53 0 2.70 0.75 2.70 Coastdown with 25 2.20 0.75 2.20 recirculation pump trip 25 ! 2.05 0.75 2.05 out-of-service 60 1.55 0.88 1.67 (no RPT) 100 ,;1.55 0.88 1.67 Siemens Power Corporation
LaSalle Unit 2 Cycle 9 EMF-2437 Revision 0 Reload Analysis Page 5-8 Table 5.3 Coastdown Operation Base Case and EOOS MCPRp Limits and LHGRFACP Multipliers for TSSS Insertion Times (Continued)
EOOS / EOD Power ATRIUNl-9B Fuel GE9 Fuel Fuel Condition (% rated) MCPRP I-9BLHGRFACp MCPR,F 0 2.70 0.68 2.70 Coastdown with 25 2.20 0.68 turbine control 2.20 0.68 valve (TcV) 25 2.15 2.15 slow closure 80 1.70 0.85 AND/OR 1.96 0.88 no RPT 80 1.62 1.85 0.88 100 1.55 1.67 0.40 0 2.60 2.60 0.40 2.60 Coastdown with 25 2.60 idle loop 0.40 2.60 25 2.60 startup 60 2.60 0.40 2.60 0.40 2.60 100 2.60 0.40 2.60 Siemens Power Corporation
EMF-2437 LaSafle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 5-9 Table 5.4 FFTR/Coastdown Operation Base Case and EOOS MCPRP Limits and LHGRFACp Multipliers for TSSS Insertion Times EOOSE1 Powe - ATRIUM-9B Fuel GE9 Fuel Condition - (% rated) MCPRN LHGRFACp MCPRN 0 '2.85 0.65 2.85 FFTR/coastdown 25 2.35 '0.65 2.35 base case 25 _ 2.30 0.65 2.30 operation 60 *1.56 0.97 1.59 100.. 1.42 1.00 1.52 0 2.86 . 0.65 2.86 FFTR/coastdown ""Fic-dn25
- . 2 2.36 0.65 2.36 with single-loop 251 2.31 0.65 2.31 operation 60 1.57 0.97 1.60 100 ,- 1.43 1.00 1.53 2.85 . . .0.65 2.85 S0 FFTRPcoastdown 25 2.35 0.65 2.35 with turbine bypass valves 25 ' 2.30 0.65 2.30 out-of-service (TBVOOS) 60 1.57 0.97 1.64 100 1.44 0.99 1.53 0 2.85 0.65 2.85 FFTR/coastdown 25 $ 2.35 0.65 2.35 with recirculation pump trip 25 2.30 0.65 2.30 out-of-service 60 1.56 0.88 1.67 (no RPT) 100 1.55 0.88 1.67
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis III vageb-4U Table 5.4 FFTR/Coastdown Operation Base Case and EOOS MCPRý Limits and LHGRFACp Multipliers for TSSS Insertion Times (Continued)
EOOS I EOD Power ATRIUM-9B Fuel GE9 Fuel Condition (% rated) MCPI, LHGRFACp MCPRN 0 2.85 0.65 2.85 FFTR/coastdown 25 2.35 0.65 2.35 with turbine control valve (TCV) 25 2.30 0.65 2.30 slow closure 80 1.70 0.85 1.96 AND/OR no RPT 80 1.62 0.88 1.85 100 1.55 0.88 1.67 0 2.60 2.60 0.40 FFTR/coastdown 25 2.60 0.40 2.60 with idle 25 loop 2.60 0.40 2.60 startup 60 2.60 0.40 2.60 0.40 100 2.60 2.60 0.4 260 Siemens Power Corporation
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Pame 5-11 0 10 20 30 40 60 60 70 so 90 100 110 F"o (%of Rated)
MCPRtGE9 Flow MCPRI (penalty
(%'of rated) ATRIUM-9B included) 0 1.60 1.66 30 ý1.60 1.66 105 1.11 1.11 Figure 5.1 Flow-Dependent MCPR Limits for Manual FiowControl Mode 4* O
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 5-12 "IýI 41I 1I-1 1.1-1 0.9-0.8-3 0.7 0.6-16 2o 3So 40 0 60o 7o a8o 9o 16o 110 Percent of Rated Flow Flow (% rated) LHGRFAC, 0 0.69 30 0.69 76 1.00 105 1.00 Figure 5.2 Flow Dependent LHGR Multipliers for ATRIUM-9B Fuel Siemens Power Corporation
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 "Reload Analysis Page 5-13 a.
0 10 20 30 40 50 60 70 80 s0 100 110 Power (% of Rated)
Power MCPRp
. (%) L"Umit 100- 1A1
-- 60 -1.46
- ,25 *... -- -1.91 25 2.20.
- 0 ' 2.70 Figure 5.3 -EOC Base Case Power:Dependent MCPR Limits for ATRUM-gB Fuel -- TSSS Insertion Times S1L 1M
La;Reo Anal22 Cyci LaSalle.Unit Cyclee 99Revision EMF-2437 0
Reload Analysis Page 5-14 2M5 2.55 I2..35 50 60 70 so 90 100 110 Power (%of Rpvtm Figure 5.4 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel - TSSS Insertion Times Siernmrn Pnwvir wa,,.,
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Paoe 5-15 K
0 10 2D 30 40 50 6o 7To o 90 100 110 POW,?Or% of RatOd Power MCPRp
___%) .... Limit "100 . 1.39 60.... 1.44 25____
___ 1.89 25 -.... 2.20 0-2.
Figure 5.5ý EOC Base Case iPower-Dependent MCPR Limits for ATRUM-9B Fuel - NSS Insertion Times Siemens Power Comoratlon
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 4 IN Reload Analysis vaVOO-10 a
K 0 10 20 30 40 8o so 70 80 90 100 110 Pows. (% of Rated)
Power MCPRp
(%) Limit 100 1.48 60 1.51 25 1.97 25 2.20 0 2.70 Figure 5.6 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel - NSS Insertion Times Siemens Power Corporation
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Pame 5-17 Pae -1
< This data is to be furnished by ComEd. >
Figure 5.7 Starting Control Rod Pattern for Control Rod Withdrawal Analysis
. . . .. EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 6-1 6.0 Postuiated Accidents 6.1 Loss-of-CoolantAccident 6.1.1 Break Location Spectrum Reference 9.8 6.1.2 Break Size Spectrum Reference 9.8, 6.1.3 MAPLHGR Analyses -
The MAPLHGR limits presented in Reference 9.9 are valid for LaSalle Unit 2 ATRIUM-9B (LSB
- 2) fuel for Cycle 9 operation.
Limiting Break: 1.1 fe 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 1810°F and 0.70% 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%. LOC/theatup analysis results for LaSalle ATRIUM-9B are presented below (Reference 9.11):
Maximum PCT, Peak Local Metal-Water Reaction (OF) (%)
ATRIUM-9B Fuel 1825 0.79" The maximum core wide metal-water reaction is < 0.16%.
6.2 ControlRod DropAccident
< This data is to be furnished by CoinEd. >
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.
0-- fý-4L-
LaSalle Unit 2 Cycle 9 EMF-2437 Revision 0 Reload Analysis Page 6-2 15.7.5. The analysis is assumed to occur 360 days following shutdown of the reactor, and It Is assumed that all 32 fuel assemblies in the cask completely fail as a result of the accident.
Because the accident is assumed not to occur sooner than 360 days following shutdown of the reactor, the source term for the accident will be very low due to fission product decay.
Hence, the commensurate radiological whole-body and thyroid doses will be very low. The results of this analysis demonstrate that spent fuel cask drop accidents involving SPC ATRIUM fuel win not exceed the established radiological whole-body and thyroid dose limits which are a small fraction of the 10 CFR 100 limits for radiological exposures.
Siemens Power Corporation
-EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 7-1 7.0 Technical Specifications 7.1 Limiting Safety System Settings 7.1.1 MCPR Fuel Cladding Inteqrty Safety Limit MCPR Safety Limit (all fuel) - two-loop oper~ation 1.11*
MCPR Safety Limit (all fuel) - single-loop operation 1.12" 7.1.2 -Steam Dome Pressure Safety Limit Pressure-Safety Umit - 1325 psig 7.2 Limiting Conditions for Operation 7.2.1 Average Planar Linear Heat Generation Rate Reference 9.9 ATRIUM-9B Fuel GE9 Fuel MAPLHGR Limits MAPLHGR Limits Average Planar Exposure MAPLHGR (GWd/MTU) -. (kW/ft) :< To be furnished by ComEd. >
0.0135 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 Rated Conditions MCPR Limit t Flow Dependent MCPR Limits:
Manual Flow Control SFigure 5.1 Includes the effects of channel bow,, up to 2 TIPOOS (or the equivalent number of TIP channels), a 2500 EFPH LPRM calibration interval, cycle startup with uncalibrated LPRMs (BOC to 500 MWd/MTU) and up to 50% of the LPRMs out of service.
1 This data is to be furnished by CornEd.
-Siemens Power Corporation
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Page 7-2 Page
ýI 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 ATRIUM-9B Fuel Steady-State LHGR Limits GE9 Fuel Steady-State LHGR Limits Average Planar Exposure LHGR (GWdjMTU) (kW/ft) < To be furnished by ComEd.>
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.
LHGRFACt 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 7-*
Reference 9.10.
LHGRFAC Multipliers for Off-Rated Conditions
- ATRIUM-gB Fuel:
LHGRFAC, Figure 5.2 LHGRFACp Tables 5.1-5.4 MAPFAC Multipliers for Off-Rated Conditions
- GE9 Fuel:
MAPFAC1
"<To be furnished by ComEd.>
MAPFACp
"<To be furnished by CornEd. >
Siemens Power Corporation
EMF-2437 LbSalle Unit 2 Cycle 9 Revision
. Pawe 7-30 Reload Analysis Iae-
'79 1 20 (o.19.4) (¶5.19.4) 18 16 14 12
~10 8
6 4
21
- I
£ I I ________________________________________________________
I I I 0 5 10 15 20 215 g0 5 40 45 50 55 60 65 70 Average Planar Exposure, GWd/.ITU Figure 7.1 Protection Against Power Transient LHGR Limit for ATRIUM-9B Fuel
EMF-2437 LaSalle Uni2 Cycle 9 Revision 0 Reload Analysis Page 8-1 8.0 Methodology References See XN-NF-80-1 9(P)(A) Volume 4 Revision 1 for a complete bibliography.
8.1 ANF-913(P)(A) Volume I Revision I and Volume I 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 Supplements I and 2, ANF CriticalPowerMethodology for Boiling Water Reactors,Advanced Nuclear Fuels Corporation, November 1990.
8.3 ANF-1 125(P)(A) and ANF-1 125(P)(A), Supplements 1 and 2, ANFB CriticalPower Correlation,Advanced Nuclear Fuels Corporation, April 1990.
8A EMF-1 125(P)(A), Supplement I Appendix C, ANFB CriticalPower Correlation Application for Co-ResidentFuel, Siemens Power Corporation, August 1997.
8.5 ANF-1 125(1)(A), Supplement I Appendix E, ANFB CriticalPower Correlation DeterminationofATRIUM-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 Program for BWR StabilityAnalysis in the FrequencyDomain, and Volume 2, STAIF- A ComputerProgramfor BWR Stability Analysis in the FrequencyDomain - Code Qualification Report, Siemens Power Corporation, July 1994.
Siemens Power Comoration
LaSalle Unlt2 Cyce9 EMF-2437 "ReloadAnalysis Revision 0 Page 9-1 9.0 Additional References 9.1 EMF-2404(P) Revision 1, Fuel Design Report for LaSalle Unit 2 Cycle 9 ATRlUM"h-9B FuelAssemblies, Siemens Power Corporation, September 2000.
9.2 ANF-89-014(P)(A) Revision I and Supplements I and 2, Advanced NuclearFuels CorporationGeneric MechanicalDesign for Advanced Nuclear Fuels 9x9-1X and 9x9-9X BIWR Reload Fuel,Advanced Nuclear Fuels Corporation, October 1991.
9.3 EMF-2440 Revision 0, LaSalle Unit 2 Cycle 9 Plant TransientAnalysis, Siemens Power Corporation, October 2000.
9.4 EMF-95-134(P), CriticalitySafetyAnalysis for ATRlUMI,-gB Fuel, LaSalle Units I and2 New Fuel Storage Vault, Siemens Power Corporation, December 1995.
9.5 EMF-96-117(P) Revision 0, CriticalitySafety Analysis for A TRlUM.-9B Fuel,LaSalle Unit I Spent Fuel Storage Pool(BORAL Rack), Siemens Power Corporation, April 1996.
9.6 EMF-95-088(P) Revision 0, CriticalitySafety Analysis forATRIUM`,m .gB Fuel, LaSalle Unit 2 Spent Fuel StoragePool (Boraflex Rack), Siemens Power Corporation, February 1996.
9.7 EMF-95-205(P) Revision 2, LaSalle Extended OperatingDomain (EOD)and Equipment Out of Service (EOOS)SafetyAnalysis forATRIUM.-9B Fuel, Siemens Power Corporation, June 1996.
9.8 EMF-2174(P), LOCA Break Spectrum Analysis for LaSalle Units I and 2, Siemens Power Corporation, March 1999.
9.9 EMF-2175(P), LaSalle LOCA-ECCS Analysis MAPLHGR Limits forATRIUM*.-9B Fuel, Siemens Power Corporation, March 1999.
9.10 Letter, D. E. Garber (SPC) to R. J. Chin (CornEd), "LaSalle Unit 2 Cycle 9 Transient Power History for Confirming Mechanical Limits for GE9 Fuel." DEG:00:1 85, August 3,
2000.
9.11 Letter, D. E. Garber (SPC) to R. J. Chin (CornEd), "10 CFR 50.46 Reporting for the LaSalle Units," DEG:00:203, August 29, 2000.
9.12 EMF-2249(P) Revision 1, Fuel Design Report for LaSalle Unit I Cycle 9 ATRIUM'm-9B FuelAssemblies, Siemens Power Corporation, September 1999.
Siemens PowerCorporaton
LaSalle Unit 2 Cycle 9 EMF-2437 Revision 0 V. Wca 0 yo a Distribution D. G. Carr, 23 D.E. Garber, 38 (9) hI'l M.E. Garrett, 23 J. K. Haun, 34 D.B. McBumey, 23 Notification List (e-mail notification)
O. C. Brown J. A. White P.D. Wimpy
Technical Requirements Manual - Appendix J L2C9A Reload Transient Analysis Results Attachment 3 LaSalle Unit 2 Cycle 9 Plant Transient Analysis LaSalle Unit 2 Cycle 9A November 2002
SIEMENS EMF-2440 Revision 0 LaSalle Unit 2 Cycle 9 Plant Transient Analysis i,
October 2000 Siemens Power Corporation Nuclear Division
Siemens Power Corporation
!.030ED INq SHG r-u; DOCUMENT SYSTEM EMF-2440
- --- ----. DATME: LQ Lr 2Q -_Re-vision 0 LaSalle Unit 2 Cycle 9
,II,Plant Transient Analysis Prepared:
D. B. McBumey, Engineer Date BWR Safety Analysis Reviewed: CO- Av- 2-O0 D. G. Carr, Team Leader "Date BWR Safety Analysis Concurred: JA Da, H.XuCUret, M.-nager Date P uct Licensing Approved:.
0..C-. Brown. IManaý6//_-,- , -Date BWR Neutronics Approved: /0/3616 M. E. Garrett,'Manage Date
-Safety Analysis Approved:
D.-J. IenveirKanager Date Commercial Operations paj
Customer Disclaimer
""I I Important Notice Regarding the Contents and Use of This Document Please Read Carefully Siemens Power Corporation's warranties and representations concerning the subject matter of this document are the agreement between Siemens Power Corporationthose set forth In Customer pursuant to which this document Is issued. and the except as otherwise expressly provided In such agreement, Accordingly, Siemens Power Corporation nor any person acting neither on Its behalf.
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EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Page i Nature of Changes Item Page Description and Justification I. All This is a new document.
Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page ii Contents 1.0 Introduction ............................................................ .............................................. 11
- 2.0 Summary ...................... *:............................... ................................................... 2-1
"-3.0 -Transient Analysis for Thermal Margin - Base Case operation ...................... 3-1 3.1 System Transients ....................................... 3-1 3.1.1 Load Rejection No Bypass ............................. 3-2 3.1.2 Feedwater Controller Failure ........................... 3-3 3.1.3 Loss-of-Feedwater Heating .............. ...... ..............3-4 3.2 MCPR Safety Limit ................................................ 3-4 3.3 Power-Dependent MCPR and LHGR Limits ..................................................... 3-6 3.4 Flow-Dependent MCPR and LHGR Umits ........................... ........................... 3-6 3.5 Nuclear Instrument Response .......................................................................... 3-7 4.0 Transient Analysis for Thermal Margin - Extended Operating Domain ......................... 4-1 4.1 Increased Core Flow ........................................................................................ 4-1 4.2 Coastdown Analysis .................................................................................... 4-1 4.3 Combined Final Feedwater Temperature Reduction/Coastdown ...................... 4-2 5.0 Transient Analysis for Thermal Margin ' Equipment Out-of-Service ............................. 5-1 5.1 Feedwater Heaters Out-of-ServIce (FHOOS) ................................................... 5-1 5.2 Single-Loop Operation (SLO) ............................................................................ 5-2 5.2.1 Base Case Operation ......................................................................... 5-2 5.2.2 Idle Loop Startup ........................................................................... 5-2 5.3 Turbine Bypass Valves Out-of-Service (TBVOOS) ........................................... 5-2 5.4 Recirculation Pump Trip Out-of-Service (No RPT) ........................................... 5-3 5.5 Slow Closure of the Turbine Control Valve ....................................................... 5-3 5.6 Combined FHOOS/TCV Slow Closure andlor No RPT ................................. 5-4 6.0 Transient Analysis for Thermal Margin - EODIEOOS Combinations ............................ 6-1 6.1 Coastdown With EOOS ......................... ......... ...6-1 6.1.1 Coastdown With Feedwater Heaters Out-of-Service .......................... 6-1 6.1.2 Coastdown With One Recirculation Loop ........................................... 6-1 6.1.3 Coastdown With TBV0OS ................................................................. 6-2 6.1.4 Coastdown With No RPT ................................................................... 6-2 6.1.5 Coastdown With Slow Closure of the Turbine Control Valve .............................................. 6-2 6.2 Combined FFTR/Coastdown With EOOS ......................................................... 6-3 6.2.1 Combined FFTRJCoastdown With One Recirculation Loop .............................. ................................................................. 6-3 6.2.2 Combined FFTRPCoastdown With TBVOOS ................... 6-3 6.2.3 Combined FFTR/Coastdown With No RPT ........................................ 6-4 6.2.4 Combined FFTRPCoastdown With Slow Closure of the Turbine Control Valve ................................................................... 6-4 Siemens Power Corporation
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 Page iri Contents (Continued) 7.0 Maximum Overpressurization Analysis ........................................................................ 7-1 7.1 Design Basis ................... ........... ............................................ 7-1 7.2 Pressurization Transients ............................ . ..... 7-1 8.0 References ....................................
"Appendix A Power-Dependent LHGR Limit Generation .......................... A-1 Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0
.' Plant Transient Analysis Page iv Tables 1.1 EOD and EOOS Operating Conditions ................................................................... 1-3 2.1 EOC Base Case and EOOS MCPPE* Limits and LHGRFACp Multipliers for TSSS Insertion Times .................................................................................................. 2-3 2.2 EOC Base Case MCPRP Limits and LHGRFACp Multipliers for NSS Insertion Times ............................................................................................................ 2-5 2.3 Coastdown Operation Base Case and EOOS MCPR, Limits and LHGRFACp Multipliers for TSSS Insertion Times ......................................................... 2-6 2A FFTRJCoastdown Operation Base Case and EOOS MCPRp Limits and LHGRFACp Multipliers for TSSS Insertion Times ......................................................... 2-8 3.1 LaSalle Unit 2 Plant Conditions at Rated Power and Flow .......................... :................ 3-9 3.2 Scram Speed Insertion Times ........................................ 3-10 3.3 EOC Base Case LRNB Transient Results .................................................................. 3-11 3.4 EOC Base Case FWCF Transient Results ................................................................. 3-12 3.5 Input for MCPR Safety Limit Analysis ........................................................................ 3-13 3.6 Flow-Dependent MCPR Results ................................................................................ 3-14 4.1 Coastdown Operation Transient Results ...................................................................... 4-3 4.2 FFTRPCoastdown Operation Transient Results ........................................................... 4-4 5.1 EOC Feedwater Heater Out-of-Service Analysis Results ............................................. 5-5 5.2 Abnormal Recirculation Loop Startup Analysis Results ................................................ 5-6 5.3 EOC Turbine Bypass Valves Out-of-Service Analysis Results ..................................... 5-7 5.4 EOC Recirculation Pump Trip Out-of-Service Analysis Results ................................... 5-8 5.5 EOC Turbine Control Valve Slow Closure Analysis Results ....................................... 5-9 5.6 EOC Recirculation Pump Trip and Feedwater Heater Out-of-Service Analysis Results ....................................................................................................... 5-10 6.1 Coastdown Turbine Bypass Valves Out-of-Service Analysis Results ........................... 6-5 6.2 Coastdown Recirculation Pump Trip Out-of-Service Analysis Results ......................... 6-6 6.3 Coastdown Turbine Control Valve Slow Closure Analysis Results ............................... 6-7 6.4 FFTRlCoastdown Turbine Bypass Vales Out-of-Service Analysis Results ................. 6-8 6.5 FFTRPCoastdown Recirculation Pump Trip Out-of-Service Analysis Results ....................................................................................................................... 6-9 6.6 FFTR/Coastdown Turbine Control Valve Slow Closure Analysis Results ................... 6-10 7.1 ASME Overpressurization Analysis Results 102%P/105%F ........................................ 7-2 Siemens Power Corporation
'LaSalle Unit 2 Cycle 9 Revision 0
'*' Plant Transient Analysis Page v Figures 1.1 " LaSalle County Nuclear Station Power Flow Map ....................................... 1-4 "2.1 Flow-Dependent MCPR Limits for Manual Flow Control Mode .......... .... 2-10
,2.2 Flow-Dependent LHGRFAC Multipliers for ATRIUM-gB Fuel.......... ......... .............. 2-11 3.1 EOC Load Rejection No Bypass at 1001105 -TSSS Key Parameters................3-15 3.2 EOC Load Rejection No Bypass at 1001105 -TSSS Vessel Water Level .................. 3-16 3.3 EOC Load RejectionNo Bypass at 100/105 -"TSSS Do-nme Pressure.......... 3-17 "3.4 EOC Feedwater Controller Failure at 100/105 -TSSS Key Parameters.................... 3-18 3.5 EOC Feedwater Controller Failure at 100/105- TSSS Vessel Water Level ................... ...... .............. .. ....... .................. 3-1g 1.........9......
3.6 EOC Feedwater Controller Failure at 100/105- TSSS Dome Pressure ......... 3-20 "3.7 Radial Power Distribution forSLMCPR Determination ............. ý3-21 3.8 LaSalle Unit 2'Cycle 9 Safety Limit Local Peaking Factors SPCA9-39'"B-14G8.0-10DM With Channel Bow (Assembly Exposure of 18,000 MWd/MTU) ................ ............ 3-22 3.9 'LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-410B-19G8.0-100M With Channel Bow (Assembly Exop sure*of 17,500 MWd/MT ) ............................................................. 3-23 3.10 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking'Factors SPCA9-383B-:16G8.0-10DM With Channel Bow
....... (Assembly Exposure of 17,500 MWd/MTU).... . .......................... 3-24 3.11" LaSalle Unit 2 Cyle 9 Safety Limi ýocal Peaking Factors-SPCAg9396B-12GZ-10DM With Channel Bow (Assembly Exposure of 15,000 MWdMTU) .:............. ...... ...... ....... 3-25 3.12 EOC Base Case Power-Dependent MCPR Lmits for ATRUM-9B Fuel
-. TSSS Insertion limes.... ...................... ................. 3-26 "3.13 TSSS EOC Base Case Power-Dependent MCPR Insertion ,T..me....Limits for GE9 Fuel- 3-27
. In eto .Times........................................ . .. _.................... ........... ... ... . 3-27 3.14 EOC Base Case Power-Dependent MCPR Limits forATRUM-9B Fuel NSS Insertion Times.................. o . ... ........ ; ....................... 3-28 3.15- EOC Base Case Power-pen'dient MCPR Umits for GE9 Fuel- "
NSS Insertion Times .................................. o ................... 3-29 3.16 EOC Base Case Power-Dependent LHGR Multipliers for ATRUM-9B .
""Fuel-TSSS Insertion ie ............ o.......... ............... 3-30 3.17 EOC Base Case Power-Dependent LHGR Multipliers for ATRUM-9B Fuel - NSS Inseition imes .. ............. ....................................................... .3-31 4.1 Coastdown Power-Dependent MCPR Limits for ATRUM-9B Fuel ................................ 4-5 4.2 Coastdown Power-Dependent LHGR Multipliers for ATRUM-9B Fuel ......................... 4-6 4.3 Coastdown Power-Dependent MCPR Limits for GE9 Fuel .......................................... 4-7 4.4 FFTRlCoastdown Power-Dependent MCPR Limits for ATRUM-9B Fuel ...................... 4-8 4.5 FFTR/Coastdown Base Case Power-Dependent LHGR Multipliers for ATRUM-9B Fuel .................................................................................................... 4.9 4.6 FFTR/Coastdown Power-Dependent MCPR Lmits for GES Fuel .............................. 4-10
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis
- ,, Revision Page vi0 Figures (Continued) 5.1 EOC Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel ..................................................................................................
5.2 EOC Feedwater Heaters Out-of-Service Power-Dependent LHGR 5-11 Multipliers for ATRIUM-91B Fuel .................................................................................
, 5.3 EOC Feedwater Heaters Out-of-Service Power-Dependent MCPR Lmits 5-12 for GE9 Fuel ....................
Abnormal Idle Recirculation Loop Startup Power-De
.......... ......... 513 p *endentM*CPR Limitso ee for ATRI UM-B Fuel.............. .............................
5.5 Abnormal Idle Recirculation Loop Startup Power-Dependent 5-14 Multipliers for ATRIUM-gB Fuel LHGR
................-0.........................
5.5 Abnormal Idle Recirculation Loop Startup Power-Dependent 5.&1 for GEr MCPR Limits Fuel..... ....... ..--.......... ..........
5.7 Turbine Bypass Valves Out-of-Service Power-Dependent nEOC o........... 5-16 Limits for ATRIGUM-B Fuel.'.....' MCPR 5.8 EOC Turbine Bypass Valves Out-of-Service Power-Dependent ...................................... 5-17 Multipliers for ATRIUM-gB Fuel LHGR o o .......... 5-17 5.9 EOC Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits for GE Fuel T ................................................
5.10 EOC Recirculation Pump Trip Out-of-Service Power-Dependent MCPR 5-19 Limits for ATRIUM-eB Fuel............ .............. ............... 5-20 5.11 EOC Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-GB Fuel ......................
5.12 EOC Recirculation Pump Trip Out-of-Service Power-Dependent MCPR 5-21 Limits for GE Fuel.......................................................................
5.13 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip 5-2 Out-of-Service Power-Dependent MCPR Limits for ATRIUM-GB Fuel.......................
5.14 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip 5-23 Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-GB Fuel
............. 5-24 5.15 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GET Fuel ................ 5-25 5.16 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip and Feedwater Heaters Out-of-Service Power-Dependent MCPR Limit for ATRIUM--B Fuel... .....................................
Fuel 5.17 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip 5-2 and Feedwater Heaters Out-of-Service Power-Dependent LHGR m Multipliers for ATRIUM-B Fuel........ ................................
5.18 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip 5-27 and Feedwater Heaters Out-of-Service Power-Dependent MCPR Limi for GE9 Fuel ..................................................
5-28 Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0
' Plant Transient Analysis Page vii Figures (Continued) 6.1 Coastdown Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel;;..:......::.............::.. ....................................... 6-11 6.2 Coastdown Turbine Bypass Valves Out-of-Service'Power-Dependent
- 6. LHGR Multipliers for ATRIUM-9B Fuel .............. ............. ...................... 6-12 6.3- Coastdown Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel .............. *....................... 6-13 6.4 Coastdown Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel ............................................................................. 6-14 6.5 Coastdown Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-gB Fuel ....................................................................... 6-15 6.6 Coastdown Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel ....................................................................................... 6-16 6.7 Coastdown Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel ..................................................................................................... 6-17 6.8 Coastdown Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel .......................... .......... ................. 6-18 6.9 Coastdown Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel .................. 6-19 6.10 FFTRlCoastdown Turbine Bypass Valves Out-of-Service Power Dependent MCPR Limits for ATRIUM-gB Fuel ......................................................... 6-20 6.11 FFTR/Coastdown Turbine Bypass Valves Out-of-Service Power Dependent LHGR Multipliers for ATRIUM-9B Fuel .................................................... 6-21 6.12 FFTR/Coastdown Turbine Bypass Valves Out-of-Service Power Dependent MCPR Limits for GE9 Fuel ...................................................................... 6-22 6.13 FFTRJCoastdown Recirculation Pump Trip Out-of-Service Power Dependent MCPR Limits for ATRIUM-9B Fuel .......................................................... 6-23 6.14 FFTRPCoastdown Recirculation Pump Trip Out-of-Service Power Dependent LHGR Multipliers for ATRIUM-9B Fuel ................................................... 6-24 6.15 FFTR/Coastdown Recirculation Pumi Trip Out-of-Service Power Dependent MCPR Limits for GE9 Fuel ................................................................... 6-25 6.16 FFTR/Coastdown Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel ................................................... ...................................... 6-26 6.17 FFTR/Coastdown Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel ................................................................................. 6-27 6.18 FFTR/Coastdown Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel .............................................................................................................. 6-28
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Page visi Figures (Continued) 7.1 Overpressurization Event at 102/105 - MSIV Closure Key Parameters .......................... 7-3 7.2 Overpressurization Event a-t 2/:1)5 SiVCl6soil-iVessel Water Level ..................... 7-4 7.3 Overpressurization Event at 102/105 - MSIV Closure Lower-Plenum Pressure............ .................................................... 7......75 7.4 Overpressurization Event at 102/105 - MSIV Closure Dome Pressure ...................... 7-6
,,,,7.5 Overpressurizatibn Event at 102/105 - MSIV Closure Safety/Relief Valve Flow Rates ................................................................................................................... 7-7 Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Paae ix P it1 ilt lBEic ~i i /r "llI Olil.
Nomenclature
=r ÷*[
AO0 anticipated operational occurrence CornEd Commonwealth Edison Company CPR critical power ratio EFPH effective full power hours EOC end of cycle EOD extended operating domain EOFP end of full power EOOS equipment out-of-service FFTR final feedwater temperature reduction FHOOS feedwater heater out-of-service FWCF feedwater controller failure HFR heat flux ratio ICF increased core flow L2C9 LaSalle Unit 2 Cycle 9 LFWH loss-of-feedwater heating LHGR linear heat generation rate LHGRFACI flow-dependent linear heat generation rate factors LHGRFACý power-dependent linear' heat generation rate factors LHGROL linear heat generation rate operating limit LPRM local power range monitor LRNB generator load rejection with no bypass MCPR minimum critical power ratio MCPRi flow-dependent minimum critical power ratio MCPRp power-dependent minimum critical power ratio MELLLA maximum extended load line limit analysis MFC manual flow control MSIV main steam isolation valve NSS nominal scram speed PAPT protection against power transient RPT recirculation pump trip SLMCPR safety limit MCPR SLO single-loop operation SPC Siemens Power Corporation SRV safety/relief valve.
SRVOOS safety/relief valve out-of-service SSLHGR steady-state LHGR Siemens Power Convration
LaSalle Unit 2Cycle 9 EMF-2440 Plant Transient Analysis Revision 0
'I"Pln a al rnietAayi ni yl Page x Nomenclature (Continued)
TBVOOS turbine bypass valve out-of-service TCV turbine control valve TIP traversing incore probe TIPOOS tip machine(s) out-of-service
TSSS technical specification scram speed TSV turbine stop valve "T-NB turbine trip with no bypass ACPR change in critical power ratio Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 - Revision 0
, Plant Transient Analysis Page 1-1 1.0 Introduction This report pre-enhts results of the plant transient analyses performed by Siemens Power Corporation (SPC) as part of the reload~sifetj anaiyses to support LaSalle Unit 2 Cycle 9 (L2C9) operation. The Cycle 9 core contains 348 fresh ATRIUMTM-B* assemblies, 256 previously loaded ATRIUM-9B assemblies and 160 previously loaded GE9 assemblies. Those portions of the reload safety analysis for which Commonwealth Edison Company (CornEd) has responsibility are presented elsewhere. The appropriate operating limits for Cycle 9 operation must be determined in conjunction with results from CoinEd analyses. The-scope of the transient analyses performed by SPC is presented in Reference 1.
The analyses reported in this document were performed using the plant transient analysis methodology approved by the Nuclear Regulatory Commission (NRC) for generic application to boiling water reactors (Reference 2). The transient analyses were performed in accordance with the NRC technical limitations as stated in the methodology (References 3-7). Parameters for the transient analyses are documented in Reference 8.
The Cycle 9 transient analysis consists of the calculation of the limiting transients identified in Reference 9 to support base case operationt for the power/flow map presented in Figure 1.1.
Results are also presented to support operation in the extended operating domain (EOD) and equipment out-of-service (EOOS) scenarios identified in Table 1.1. The analysis results are used to establish operating limits to protect against fuel failures. Minimum critical power ratio (MCPR) limits are established to protect the fuel from overheating during normal operation and anticipated operational occurrences (AOOs). Power-dependent MCPR (MCPRp) limits are required in order to provide the necessary protection during operation at reduced power. Flow dependent MCPR (MCPR1 ) limits provide protection against fuel failures during flow excursions initiated at reduced flow. Cycle 9 power- and flow-dependent MCPR limits are presented to protect both ATRIUM-9B and GE9 fuel.
Protection against violating the linear heat generation rate (LHGR) limits at rated and off-rated conditions is provided through the application of power- and flow-dependent LHGR factors "ATRIUM is a trademark of Siemens.
Base case operation is defined as two-loop operation within the standard operating domain, including the ICF and MELLLA regions, with allequipment in-service.
Siemens Power Corporation
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis
,, Revision Page 1-20
'(LHGRFACP and LHGRFACI, respectively). These factors or multipliers are applied directly to the stea~dy-state LHGR limit to ensure that the LHGR does not exceed the protection againsL power transient (PAPT)limit-durin9 p-sturated-AOO,. Cydie 9 power- and flow-dependent LHGR multipliers are presented forATRIUM-SB fuel.
Results of analyses that demonstrate compliance with the ASME Boiler and Pressure Vessel Code overpressurization limit are presented.
The results of the plant transient analyses are used in a subsequent reload analysis report (Reference 15) along with core and accident analysis results to justify plant operating limits and set points.
'I Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 1-3 Table 1.1 EOD and EOOS
-Operating Conditions Extended Operating Domain (EOD) Conditions Increased core flow Maximum extended load line limit'analysis (MELLLA)
Coastdown Final feedwater temperature reduction (FFTR)
Combined FFTR/coastdown 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 machines out-of-serviceorthe equivalent number of TIP channels (100% available at startup)
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 channels) and/or up to 50% of the LPRMs out-of-service is supported.
Dwmr
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 01 Page 1-4 110 I!'
100 I
"V 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 I Flow Map Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 . . .Revision 0.
Plant Transient Analysis Page 2-1 2.0 Summary e*..deterdmnaiojl..oQl*thermalllimlts(MCPR limits and LHGRFAC multi*piers) for*LaSalle Unit 2 Cycle 9 is based on analyses of the limiting'operational transients identified in Reference 9. Although the Reference 9 conclusions are based on I 8month cycles, the limiting operational transients identified remain valid for 24-month cycles. The transients evaluated are the generator load rejection with no bypass (LRNB). feedwater controller failure to maximum demand (FWCF) and loss-of-feedwater heating (LFWH). Thermal limits identified for Cycle 9 operation include both MCPR limits and LHGRFAC multipliers. The MCPR operating limits are established so that less than 0.1% of the fuel rods in the core are expected to experience boiling transition during an AOO initiated from rated or off-rated conditions and are based on a two-loop operation MCPR safety limit of 1.11. LHGRFAC multipliers are applied directly to the LHGR limits at reduced power and/or flow conditions to protect against fuel melting and overstraining of the cladding during an AOO. Operating limits are established to support both base case operation and the EOOS scenarios presented in Table 1.1. Operating limits are also established for the EOD and combined EOD/EOOS conditions presented in Table 1.1.
I, Base case MCPRp limits and LHGRFACp multipliers are based on results presented in Section 3.0. Results presented in Sections 4.0-6.0 are used to establish the operating limits for operation in the EOD, EOOS, and combined EODIEOOS scenarios.
Cycle 9 MCPRp limits and LHGRFACp multipliers for ATRIUM-9B fuel and MCPRp limits for GE9 fuel that support base case operation and operation in the EOD, EOOS and combined EODIEOOS scenarios are presented in Tables1,t 2.1-2.4. Tables 2.1 and 2.2 present base case limits and multipliers for Technical Specfitatioins scram speed (TSSS) insertion times and nominal scram speed (NSS) insertion times, respectively. Table 2.3 presents the limits and multipliers for coastdown operation. The combined FFTRlcoastdown limits and multipliers are identified in Table 2.4.
MCPRf limits for both ATRIUM-9B and GE9 that protect against fuel failures during a slow flow excursion event in manual flow control are presented in Figure 2.1. Automatic flow control is not supported for L2C9. The GE9 MCPRf limits include the effect of applying the MCPR penalty described in Reference 10. The MCPRf limits presented are applicable for all EOD and EOOS conditions presented in Table 1.1.
Siemens PowerCorporation
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440 Page 2-2 The Cycle 9 LHGRFACj multipliers for the ATRIUM-9B fuel are presented in Figure 2.2 and are applicable in all the EOD and EOOS scenarios presented in Table 1.1. 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% clad strain and centerline melt criteria for GE9 fuel is discussed in Reference 19.
Ih II Ii The results of the maximum overpressurization analyses show that the requirements of the ASME code regarding overpressure protection are met for Cycle 9. The analysis shows that the dome pressure limit of 1325 psig is not exceeded and the vessel pressure does not exceed the limit of 1375 psig.
Siemens Power Corporation
EMF-2440
-LaSalle Unit 2 Cycle 9 Revision 0 N-' Plant Transient Analysis Paoe 2-3 Paoe 2-3 Table 2.1 EOC Base Case and EOOS MCPRp Limits and' LHGRFAC, Multipliers for TSSS Insertion Timese h.
S*EOOS S...E..
I EOD Pwer
'Pow'er'- ATRIUM-9B Fuel GE9 Fuel Condition (% rated) MCPRv LHGRFACp MCPRp
- 0 2.70 0.78 2.70 Base'- 25 2.20 0.78 2.20 case 1.91 0.78 1.99 operation 60 1.46, 1.00 1.52
_______ 100. 1.41, 1.00 1.51 0 2.85 0.69 -2.85 Feedwater heaters 25;.- 2.35 0.69 2.35 hat-ers 25 - 2.14 0.69 2.22 out-of-service (FHOOS) o60 1.51 0.97 1.57 oo _ _ _
- 100 1.41 1.00 -1.51.
0 2.71 0.78 2.71 Single-loop '25 2.21 0.78 2.21 operation 25 1.92 0.78 2.00 (SLO) 60 1.47 1.00 1.53 100 1.42 1.00 1.52 0 2.70 0.76 2.70.
Turbine 25 L 2.20 0.76 2.20 bypass; valves b v25 1.98 0.76 2.08 out-of-service (TBVOOS) - 60 1.52 0.97 1.62
_ 100 1.43 0.99 1.52 Limifs support operation with any combination of I SRVOO, u'p to 2 TIPOOS (or the equivalent number of TIP channels), up to a 20VF re'duction 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.
Sierrns Pawmr frnnratinn
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 2-4 Table 2.1 EOC Base Case and EOOS MCPRp Limits and LHGRFACp Multipliers for TSSS Insertion Timesr (Continued)
EOOS i EOD Power ATRIUM-9B Fuel . GE9 Fuel Condition (% rated) MCPRp LHGRFACp MCPRp 0 2.70 0.78 2.70 Recirculation 25 2.20 0.78 2.20 pump trip 25 1.91 0.78 1.99 out-of-service (no RPT) 60 1.51 0.89 1.61 100 1.51 0.89 1.61 0 2.70 0.70 2.70 Turbine control 25 2.20 0.70 2.20 valve (TCV) 25 2.10 0.70 2.10 slow closure 80 1.69 0.86 1.95 AND/OR no RPT 80 1.61 0.89 1.84 100 1.53 0.89 1.63 0 2.85 0.68 2.85 TCV 25 2.35 0.68 2.35 slow closure/ 25 2.14 0.68 2.22 FIHOOS 80 1.69 0.86 1.95 AND/OR no RPT 80 1.61 0.89 1.84 100 1.53 0.89 1.63 0 2.60 0.40 2.60 Idle 25 2.60 0.40 2.60 loop 25 2.60 0.40 2.60 startup 60 2.60 0.40 2.60 100 2.60 0.40 2.60 Umits support operation with any combination of I SRVOOS, up to 2 TIPOOS (or the equivalent' number of TIP channels), up to a 20OF reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of seivice in the standard, ICF, and MELLLA regions of the power/flow map.
Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Pae 2-5 Table 2.2 -EOC Base Case MCPRI Limits and LHGRFACp Multipliers for NSS Insertion Timese Y
ATRIUM-98 Fuel ' GE9 Fuel S*EOOS ! EOD
" 'Power .9 Condition - -
(% rated) -:MCPRp LHGRFACý MCPRp 0 2.70 0.79 2.70 Base 25 2.20 0.79 2.20 case 25 1.89 0.79 1.97 operation 60 1.44 1.00 1.51
_..._..._ 10OX-... 1.39"- 1 1.00 1.48 I',
J ----.- ---
Limits support operation with any combination of 1 SRVOOS, up to 2 TIPOOS (or th equrilent' number of TIP channels). up to a 20OF reduction in feedwater temperature (except for ,onditions with FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF. and MELLLA regions of the power/flow map.
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis D0 02_
Revsin p
- .,,.q
,-,*=
r; Table 2.3 Coastdown Operation Base Case and EOOS MCPRp Limits and LHGRFACP Multipliers for TSSS Insertion Times*
I, I Limits support operation with any combination of I SRVOOS, up to 2 TIP-:OOS (or number of TIP channels), up to a 20OF reduction the equivalent i1 feedwater, 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-2440 LaSalle-Unit 2 Cycle 9 '...... ... Revision 0 Plant Transient Analysis Page 2-7 Table 2.3 Coastdown Operation Base Case and EOOS MCPRp Limits and LHGRFACp Multipliers for TSSS Insertion Times' (Continued)
".-;EOOS - Power: ...
Power *ATRIUM-9B Fuel GE9 Fuel Condition (%rated) MCPRp LHGRFACý MCPRP 20.270 0.68 2.70
.Coastdown with '25 2.20 0.68 2.20 turbine control Ivalve (TCV) 25 2.15 0.68 2.15 slow closure . _ 1.70 .--- 0.85 .1.96.
ANDORP 80s 1.62 0.88 1.85 noRFPT
" i "_"100
_ 1.55 0.88- 1.67
.0'_ 2.60 0.40 2.60 C0oastdown w 0 25 2.60 0.40 2.60 idle loop . 25 2.60 0.40 2.60
-startup.. . strtp60 2.60 0.40 2.60
__ 100-. 2.60 0.40 2.60 Io-.
- Limits support operation with any combination of I SRVOOS, up to 2 TIPOOS (or the equivalem number of TIP channels), up to a 20"F reduction in feedwater temperature, and up to 50% of the LPRMs out of service in the standard, ICF, and MELLLA regions of the power/flow map.
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis Page 2-8 Table 2.4 FFTRJCoastdown Operation Base Case and EOOS MCPRp Limits and LHGRFACp Multipliers for TSSS Insertion Timese
- Limits support operation with any combination of I SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), 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-2440 LaSalle Unit 2 Cycle 9 Revision 0
,' Plant Transient Analysis Paoe 2-9 Table 2.4 FFTRJCoastdown Operation Base Case and
... .... EOOS MCPRp Limits annd LHGRFACp Multipliemr h,
for TSSS Insertion Times" (Continued)
EOOS I EOD Power - ATRIUM-9B Fuel GE9 Fuel Condition (% rated) MCPRp LHGRFACý MCPR, "10 2.85 0.65 2.85 FFTR/coastdown 25 2.35 0.65 2.35 with turbine control valve (TCV) 25 2.30 0.65 2.30 slow closure 80 1.70 0.85 1.96 AND/OR **
noDR/ 80 1.62 0.88 1.85 100 1.55 0.88 1.67 0 2.60 0.40 2.60 FFTR/coastdown 25 - 2.60 0.40- 2.60 with idle 25 2.60 0.40 2.60 loop startup 60 2.60 0.40 2.60 1 100 2.60 0.40 2.60 "Limitssupport operation with any combination of 1 SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), and up to 50% of the LPRMs out of service Inthe standard, ICF, and MELLLA regions of the power/flow map.
Siemens Power Comoration
LaSalle Unit 2 Cycle 9 EMF-2440 Revison 0 Plant Transient Analysis I Page 2-10 I
so so Figure 2.1 Flow-Dependent MCPR Limits for Manual Flow Control Mode Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Dight "Tr~ne~cntAn~lv~i_ Page 2-11
" .iili " I J4P*
- I I
! I 1 1
0.9-C-)
0.8.
3 0.7 0.62 nAI 1 6 4 i i 0 10 20 30 40 50 60 70 so 90 100 110 Percent of Rated Flow Flow
(% rated) 't LHGRFAC?
0 0.69 30 0.69 76 1.00 105 . 1.00 Figure 2.2 Flow-Dependent LHGRFAC Multipliers for ATRIUM-9B Fuel
- ':--mm I*.*o /'%*.*,-*lk*.,m
L IEMF-2440
'LaSalle Unit 2 Cycle Plant Transient 9 Analysis ... .Page Revision3-10 3.0 Transient Analysis for Thermal Margin - Base Case Operation:
This section describes the analyses performed to determine the power- and flow-dependent MCPR and LHGR operating limits for base case operation at LaSalle Unit 2 Cycle 9.
COTRANSA2 (Reference 4), XCOBRA-T (Reference 11),,XCOBRA (Reference 7) and I, CASMO-3G/MICROBURN-B (Reference 3) are the major codes used in the thermal limits analyses as described in SPC's THERMEX methodology report (Reference 7) and neutronics methodology report (Reference 3). COTRANSA2 is a system transient simulation code, which includes an axial one-dimensional neutronics model that captures the effects'bf axial power shifts associated with the system transients. XCOBRA-T is a transient thermal-hydraulics code used in the analysis of thermal margins for the limiting fuel assembly.'XCOBRA is used in' steady-state analyses. The ANFB criticalWpo6we correlation (Reference 6) Is used to evaluate the thermal margin of the fuel assemblies. Calculations have been performed to demonstrate the applicability of the ANFB critical power correlation to GE9 fuel at LaSalle using the Reference 12 methodology. Fuel pellet-to-cladding gap conductance values are based on RODEX2 (Reference 13) calculations for the LaSalle Unit 2 Cycle 9 core configuration.
3.1 System Transients System transient calculations have been performed to establish thermal limits to support L2C9 operation. Reference 9 identifies the potential limiting events that need to be evaluated on a cycle-specific basis. The potentially limiting transients for which SPC has analysis responsiblity are the LRNB and FWCF events. Other transient events are either bound by the consequences of one of the limiting transients, or are part of ComEd's analysis responsibility.
Reactor plant parameters for the system translient analyses are shown in Table 3.1 for the 100%
power/I 00% flow conditions. Additional plant parameters used in the analyses are presented in Reference 8. Analyses have been performed to0determine power-dependent MCPR and LHGR limits that protect operation throughout the poweir/flow domain depicted in Figure 1.1. At LaSalle, direct scram and recirculation pump high- to low-speed transfer on turbine stop valve (TSV) and turbine control valve (TCV) position are bypassed at power levels less than 25% of rated. Reference 14 indicates that MCPR and LHGR limits need to be monitored at power levels greater than or equal to 25% of rated.-Asia resubi, all analyses used to establish base case MCPR* limits and LHGRFACp multipliers are performed with both direct scram and RPT operable for power levels at or above 25% of rated.
Siemens Power Corporation
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 Page 3-2 "Thelimiting exposure for rated power pressurization transients is at end of full power (EOFP) when the control rods are fully withdrawn. Off-rated power analyses were performed at earlier cycle exposures to ensure that the operating limits provide the necessary protection.
All pressurization transients assumed only the 11 highest set point safety relief valves (SRVs)
,were operable, consistent with the discussion in Section 7. In order to support operation with I SRV out-of-service, the pressurization transient analyses were performed with the lowest set point SRV out-of-service, which makes a total of 10 SRVs available.
The term, recirculation pump trip (RPT), is used synonymously with recirculation pump high- to low-speed transfer as it applies to pressurization transients. During the high- to low-speed transfer, the recirculation pumps trip off line and coast When they reach the low-speed setting, the pumps reengage at the low speed. The time it takes for the pumps to coast to the low-speed condition is much longer than the duration of the pressurization transients.
Therefore, a recirculation pump trip has the same effect on pressurization transients as a recirculation pump high- to low-speed transfer.
Reductions in feedwater temperature of less than 20°F from the nominal feedwater temperature are considered base case operation, not an EOOS condition. As discussed in Reference 9, the reduced feedwater temperature is limiting for FWCF transients. As a result, the base case FWCF results are based on a 20oF reduction in feedwater temperature.
The results of the system pressurization transients are sensitive to the scram speed used in the calculations. To take advantage of scram speeds faster than the TSSS insertion times presented in Reference 14 scram speed-specifit MCPRp limits and LHGRFACp multipliers are provided. The NSS insertion times used in the analyses reported are presented in Reference 8 and reproduced in Table 3.2. The NSS MCPRp limits and LHGRFACp multipliers can only be applied if the scram speed surveillance tests meet the NSS insertion times. System transient analyses were performed to establish MCPRp limits and LHGRFACp multipliers for base case operation for both NSS and TSSS insertion times.
3.1.1 Load Reieýtion NoBypass The load rejection causes a fast closure of the turbine control valve. The resulting compression, wave travels through the steam lines into the vessel and creates a rapid pressurization. The Siemens Power Corporatin
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-3 increase in pressure causes a decrease in core void, which in turn causes a rapid increase in power. The fast closure of the turbine control valve also causes a reactor scram and a recirculation pump_ higl- to low-speed transfer which helps mitigate the pressurization effects.
Turbine bypass system operation, which also mitigates the consequencesof the event, is not credited. The excursion of the core power due to the void collapse is terminated primarily by the reactor scram and revoiding of the core. The analysis assumes 3-element feedwater level:
control; however, manual- or single-element feedwater level control will not significantly affect thermal limit or pressure results.
The generator load rejection without turbine bypass system (LRNB) is a more limiting transient than the turbine trip no bypass (TTNB) transient The initial position of the TCV is such that It closes faster than the turbine stop valve. This more than makes up for any differences in the scram signal delayp between the two events. This has been demonstrated in calculations that support the Referenhce 9conclusiontfiat the TTNB event is bound by the LRNB event.
LRNB analyses were performed for several power/flow conditions to support generation of thb thermal limits. Table 3.3 presents the LRNB transient results for both TSSS and NSS insertion times for Cycle 9. For illustration, Figures 3.t-3.3 are presented to show the responses of , ..
various reactor and plant parameters during the LRNB event initiated at 100% of rated power and 105% of rated core flow with TSSS insertion times.
3.1.2 Feedwater Controller Failure- Q '
Theincrease in feedwater flow due to a failure of the feedwater control system to maximum, demand results in an increase in the water level and a decrease in the coolant temperature at the core inlet. The increase in core inlet subcooling causes an increase in core power. As the feedwater flow continues at maximum demand, the water level will continue to rise and eventually reaches the high water, level trip'set point. The initial waterlevel is conservatively assumed to b at the lower level operating range at 30 inches above instrument zero to delay the high level trip and maximize the -coreinlet subcooling that results from the FWCF. The high water level trip causes the turbine stop valvesto close in order to prevent damage to the turbine from excessive liquid inventory in the steam line. The valve closures create a compression wave that travels to the'cbre causing a.o9d c,ollapse and subsequent rapid power excursion. The .
closure of.the turbine valves initiates-a reactor scram and a recirculation pump high- to low speed transfer. In addition, the turbine bypass valves are assumed operable and provide some
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440
,,v'Page Revision340 pressure relief. The-core power excursion is mitigated in part by the pressure relief, but the primary mechanisms for termination of the event are reactor scram and revoiding of the core.
FWCF analyses were performed for several power/flow conditions to support generation of the thermal limits. Table 3.4 presents the base case FWCF transient results for both TSSS and
,NSS insertion times for, Cycle 9. For illustration, Figures 3.4-3.6 are presented to show the responses of various reactor and plant parameters during the FWCF event initiated at 100% of rated power and 105% of rated core flow with TSSS insertion times.
3.1.3 Loss-of-Feedwater Heating CornEd has the analysis responsibility for the loss-of-feedwater heating (LFVVH) event at, rated conditions. At reactor power levels less than rated, the LFWH event is less limiting than the LFWH event at rated conditions for the following reasons:
- At lower power/flow conditions with other core conditions such as control rod patterns and exposure unchanged, the initial MCPR is higher than the MCPR at rated power and flow. This results in additional MCPR margin to the MCPR safety limit.
0 The possible change in feedwater temperature during an LFWVH event decreases as the reactor power decreases.
3.2 MCPR Safety Umit The MCPR safety limit is defined as the minimum value of the critical power ratio at which the fuel can be operated, with the expected number of rods in boiling transition not exceeding 0.1%
of the fuel rods in the core. The MCPR safety limit for all fuel in the LaSalle Unit 2 Cycle 9 core was determined using the methodology described in Reference
- 5. The effects of channel bow on core limits are determined using a statistical procedure.
The mean channel bow is determined from the exposure of the fuel channels and measured channel bow data.
CASMO-3G is used to determine the effect on the local peaking factor distribution. Once the channel bow effects on the local peaking factors are determined, the impact on the core limits is determined in the MCPR safety limit analysis. Further discussion of how the effects of channel bow are accounted for is presented in Reference 5. The main input parameters and uncertainties used in the safety limit analysis are listed in Table 3.5. The radial power uncertainty includes the effects of up to 2 TIPOOS or the equivalent number of.TIP channels (100% available at'startup), up to 50% of the LPRMs out-of-service, and an LPRM calibration interval of 2500 EFPH as discussed in References 16 and
- 24. The channel bow local peaking-Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 .Revision 0 Plant Transient Analysis Page 3-5 uncertainty Is a function of the nominal and bowed local peaking factors and the standard odeviation of the -measured bow data. ... ............ .. .
The determinationof the safety limit explicitly includes the effects: of channel bow and relies on the following assumptions: .
- Cycle 9 will not contain channels used for more than one fuel bundle lifetime.
- The channel exposure at discharge will not exceed 48,000 MWd/MTU based on the fuel bundle average exposure.
- The Cycle 9 core contains all CarTech-supplied channels.
Analyses were performed with input parameters (including the radial power and local peaking factor distributions) consistent with each exposure step in the design basis step-through. The analysis that proddiced the highest'numb5er of rods in boiling transition corresponds to a Cycle 9 exposure of 15,000 MWd/MTU. The radial power distribution corresponding to a Cycle 9 exposure of 15,000 MWd/MTU is shown in Figure 3.7. Eight fuel types were represented in the LaSalle Unit 2 Cycle 9 safety limit analysis: four SPC ATRIUM-9B fuel types loaded in Cycle 9 (SPCA9-3918-I4G8.0-10DM, SPCA9-4*10B-1 9G8.0-10DM, SPCA9-383B-16G8.0-1 DOM, and*
SPCA9-396B-12GZ-1 0DM); two ATRIUM-98 fuel types loaded in Cycle 8 (SPCA9-381B-13GZ7 80M and SPCA9-384B-1 1GZ6-80M); and two GE9 fuel types loaded in Cycle 7 (GE9B "P8CWB322-1 GZ-100M-150 and GE9B-P8CWB320-9GZ-10DM-15O).
The local power peaking factors, including the effects of channel bow, at 70% void and assembly exposures consistent with a Cycle69- exposure of 15,000 MWd/MTU are presentedin Figures 3.8 through 3.11 for the Cyclee9 SPC ATRIUM-9B fuel.-The bowed local peaking factor
.data used in the MCPR safety limit analysis fdrlfuel type SPCA -391B-14G..0-100M is at an assembly average exposure of 18,000 MWdMu. The data for fuel types SPCA941 09 19G.0-1 0DM and SPCA9-383B-16G8.O-1 0DM is at an assembly average exposure of 17,5D0 MWd/MTU. The data is at an assembly average exposure of 15,000 MWd/MTU for fuel type SPCA9-396B-I2GZ-'l0M. :
The results of the analysis support a tio-loop 'operation MCPR safety limit of 1.11 and a single loop operation MCPR safety limit of 1.12 for all fuel types in the Cycle 9 core. These results are applicable for all EOD and EOOS conditions presented in Table 1.1 and support 'Startupwith uncalibrated LPRMs for an exposure range of BOC to 500 MWd/MTU.
Siemens Power Corporation
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440 Page 3vo 3.3 Power-DependentMCPR andLHGR Limits Figures 3.12 and 3.13 present the-base case operation TSSS ATRIUM-gB and GE9 MCPRp limits for Cycle 9. Figures 3.14 and 3.15 present the ATRIUM-gB and GE9 MCPRp limits for base case operation with NSS insertion times. The limits are based on the ,CPR results from t!he limiting system transient analyses discussed above and a MCPR safety limit of 1.11.
Relative to the TSSS MCPRp limits, using the faster NSS insertion times provide lower MCPRp limits.
The pressurization transient analyses provide the necessary information to determine appropriate multipliers on the fuel design LHGR limit for ATRIUM-9B fuel to support off-rated power operation. Application of the LHGRFACp multipliers to the steady-state LHGR limit ensures that the LHGR during AO0s initiated at reduced power does not exceed the PAPT limits. The method used to calculate the LHGRFACp multipliers is presented in Appendix A. The results of the LRNB and FWCF analyses discussed above were used to determine the base case LHGRFACp multipliers. The base case ATRIUM-gB LHGRFACp multipliers for Cycle 9 TSSS and NSS insertion times are presented in Figures 3.16 and 3.17, respectively.
3.4 Flow-DependentMCPR and LHGR Limits Flow-dependent MCPR and LHGR limits are established to support operation at off-rated core flow conditions. The limits are based on the CPR and heat flux changes experienced by the fuel during slow flow excursions. The slow flow excursion event assumes a failure of the recirculation flow control system such that the core flow increases slowly to the maximum flow physically attainable by the equipment. An uncontrolled increase in flow creates the potential for a significant increase in core power and heat flux. A conservatively steep flow run-up path was determined starting at a low-power/low-flow state point of 58.1 %P/30%F increasing to the high power/high-flow state point of 124.2%P/105%F.
MCPRr limits are determined for the manual flow control (MFC) mode of operation for both ATRIUM-9B and GE9 fuel. XCOBRA is used to calculate the change in critical power ratio during a two-loop flow run-up to the maximum flow rate. The MCPP* limit is set so that the increase in core power resulting from the maximum increase in core flow is such that the MCPR safety limit of 1.11 is not violated. Calculations were performed for several initial flow rates to Siemens Power Corporation
EMF-2440 "LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page_3-7 "determinethe corresponding MCPR values that put the limiting assembly on the MCPR safety
- limit at the high-flow' condition at the end of the flow excursion.
Results of the MFC flow run-up analysis are presented in Table 3.6 for both the ATRIUM-9B hand GE9 fuel. MCPRt I!mits that provide the required protection during MFC operation are presented in Figure 2.1. The Cycle 9 MCPRf limits were established such that they support base case operation and operation in the EOD, EOOS, and combined EOD/EOOS scenarios. The MCPRf limits.are valid for all exposure conditions during Cycle 9. Since a low- to high-speed pump upshift is required to attain high-flow rates, for initial core flows less than 30% of rated, the limit is conservatively set equal to the 30% flow value. The MCPRf penalty described in Reference 10 has been applied to the GE9 MCPRr limits shown in Figure 2.1. The penalty is a function of core flow with a value of 0.0 at 100% of rated and increases linearly to 0.05 at 40%
of rated. The penalty continues to increase to 30% of rated core flow where a penalty of 0.06 Is applied.
SPC has performed LHGRFAC1 analyses with the CASMO-3G/MICROBURN-B core simulator codes. The analysis assumes that the recirculation flow increases slowly along the limiting rod line to the maximum flow physically attainable by the equipment. A series of flow excursion analyses were performed at several exposures throughout the cycle starting from different initial power/flow conditions. Xenon is assumed to remain constant during the event. The LHGRFACr multipliers were established to ensure that the LHGR during the flow run-up does not violate the PAPT LHGR limit. Since a low- to high-speed pump upshift is required to attain high-flow rates, for initial core flows less than 30% of rated, the LHGRFACI multiplier is conservatively set equal to the 30% flow value. The LHGRFACI values as a function of core flow for the ATRIUM-9B fuel are presented in Figure 2.2. The Cycle 9 LHGRFAC1 multipliers were established to support base case operation and operation in the EOD, EOOS, and combined EODIEOOS scenarios for all Cycle 9 exposure conditions.
3.5 NuclearInstrument Response The impact of loading ATRIUM-9B fuel into the LaSalle core will not affect the nuclear instrument response. The neutron lifetime is an important parameter affecting the time response of the incore detectors. The neutron lifetime is a function of the nuclear and mechanical design of the fuel assembly, the in-channel void fraction, and the fuel exposure. The neutron lifetimes are similar for the SPC and GE LaSalle fuel with typical values of 39(10"6 to 40(1 0") seconds
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis
,Page38 Revision 0 for the ATRIUM-9B lattices and 41(106) to 43(10") seconds for the GE9 lattices as calculated with the CASMO-3G code at core average void and exposure conditions. Therefore, the neutron lifetimes for a full core of ATRIUM-9B fuel, a mixed core of ATRIUM-WB and GE9 fuel, and a full core of GE9 fuel are essentially equivalent.
Siemens Power Corpoation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-9
-Ta*. i-- LaSalle Unit 2 Plant Conditions at Rated Power and Flow Reactor thermal power 3489 MWt
---Total core flow .108.5 Mlbm/hr Core active flow 93.7 Mlbmlhr Core bypass flow*- 14.8 Mlbm/hr
-Core inlet enthalpy . 523.9 Btullbm Vessel pressures Steam dome 1001 psia Core exit (upper-plenfum)- 1013 psia Lower-plenum 1038. psia
'-Turbine pressure 948 psia Feedwater I steam flow 15.145 Mlbm/hr Feedwater enthalpy 406.6 Btullbrn Recirculating pump flow 15.83 Mlbm/hr (per pump)
Core average gap 1162 Btu/hr-ft2 -- F coefficient (EOC)
Includes water channel flow.
Siemens PowerCorporation
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 Ii, Page 3-10
--... Table 3Z Scram Speed InsertionTimes-- - -.-...
Ii,
" 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 slightly conservative results (Reference 22).
Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Paae 3-11
--- rable-3.3=EOC-Base-Case-LRN BT--Tnsient Results Peak Peak Power/ ATRIUM-9B ATRIUM-9B GE9 Neutron Flux Heat Flux Flow ACPR LHGRFACý ----ACPR (% rated) (% rated)"
TSSS Insertion Times 100/105 '0.30 1.01 0.40 422 127 1001100 0.29 1.01 0.39 . 431 128' 100/81 0.28 1.01 0.38 ,_437 126 80/105 0.29 1.04 0.39 . ... 324 100 80/57.2 0.29 1.50.39 265 9 601105 0.27 1.06 0.36 245- . 73 60/35.1s -0.17 '11.13, 0.21 . 96 .... 63 40/105 0.23' -1.13'; 0.27 100'. 46" 25/105 0.17- 1.22-' 0.19 44- 27' NSS Insertion Times 100/ 105 1 0.28 "1.02 :J-0.37 -,380 J 124 100/81 0.22 .1.03 , *' 0.30 358 120-
+ 4 J 80/105 "0.27 1.04, -0.36 3 02 . .
80/57.2 0.20 1.09 : 0, 26 218 __90 60/-105 0.26 1.07 0.35 . 236 73 60/35.1 0.13 1.18 0.14 .- 76 - 60 40/105 -0.20 . 1.14 0.27 115,_ 47 25/105 0..015 1.22 0.17 42' 27" The analysis results are from an earlier cycle exposure. The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.
Siemens Power Comoration
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analw~i* Revision 0 Plant
............... -' Page 3-12 Table 3.4 EOC Base Case FWCF Transient Results Peak Neutron Peak II II Power/ ATRIUM-9B ATRIUM-9B GEG, Flow Flux Heat Flux ACPR LHGRFAC;, ACPR (% rated) (% rated)
TSSS Insertion Times NSS Insertion Times
- The analysis results are from an earlier cycle exposure. The
,ACPR and LHGRFACp results are conservatively used to establish the thermal limits.
Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-13
-Table-3.51 nputlforMePR--S'Mlty-Ltitlhlysis I,
Fuel-Related Uncertainties, Source Statistical Parameter -Document Treatment ANFB correlation*
ATRIUM-98 Reference 17 Convoluted GE9 Reference 12' Convoluted Radial power References 16 and 21" Convoluted Local peaking factor - Reference 5 - Convoluted Assembly flow rate (mixed core) - Reference 5 Convoluted Channel bow local peaking - Function of nominal and bowed local Convoluted peaking and standard deviation of bow data (see Reference 18)
"NominalValues and" Plant Measurement Uncertainties Uncertainty (%) Statistical Parameter Value (Reference 8) Treatment Feedwater flow ratet (Mlbm/hr) ,22.4 1.76 Convoluted Feedwater temperature (OF) .426.5 0.76 Convoluted Core pressure (psia) 1031.35 0.50 Convoluted Total core flow (Mlbm/hr) 113.9 2.50 Convoluted Core powerl (MWth) 5167.29
- Additive constant uncertainties values are used.
1 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
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis
'II Page 3-14 Table 3.6 Flow-Dependent MCPR Results Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycie 9 Revision 0 Plant Transient Analysis Page 3-15 5Ow0 CORE POWER IHEAT FLUX CORE FLOW 400D STEAM FLOW FEED FLOW 0
w 9-1.0 2.0,'O TME SECONDS Figure 3.1 EOC Load Rejection No Bypass at 1001105 -TSSS Key Parameters
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis
~1,41 IFLC11a.-1 pII 0
it N
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W THME. SECONDS Figure 3.2 EOC Load Rejection No Bypass at 1001105 - TSSS Vessel Water Level Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0
'" Plant Transient Analysis Page 3-17 12M.O0 M
w a1.
in
,aJ 0
0 I. I
.0 10 2.0 4.0 TIME SECONS Figure 3.3 EOC L.oatd Rejection No Bypass at 1001105 - TSSS Dome Pressure
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis
,III-" Page 3-18 0
9 0
I zw IL
-lowo, T 1WO TIME SECONDS Figure 3.4 EOC Feedwater Controller Failure at 1001105 - TSSS Key Parameters Siemens Power Corporation
- EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-19 0.
N SI-i U) 0 W
'_z
-J U>
I (J
{/n wI I SCN TME, SECONDS Figure 3.56 EOC FeedwaterController Failure at 1001105 - TSSS Vessel Water Level Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Plant Transient Analysis Revision 0 rae-.-z
, 44 01.
'IL U,
U) 0.
i~i 00
.O 15.0
.
- TIE, SECONDS Figure 3.6 EOC Feedwater Controller Failure at 1001105 - TSSS Dome Pressure Siemens Power Corporation
EMF-2440 SLaSalle Unit 2 Cycle 9 Revision 0 n*.. Ir~e=#A1'hi Paoe 3-21 rlJilIM 10l"0 l4 ',11l~
Ii I
- " -200 175
'150
- " 25 C.
o6* 1()0 a)
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-.0.ý'Rd.2
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-. o.0 1.1 -1.2-13 1.4 ' 5 1 Radial Power Peaking Figu re3.7 Radial Power Distribution for
-*. SLMCPR Determination, -
Clmým. Dýr r~rwraf6-.n
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Mkic. %ý-
C 0%t ontrol Rod Corner n
t 1.052 1.045 1.088 1.088 1.104 1.079 1.068 1.013 r 1.005 1.045 0.951 1.019 0.996 0.852 0.986 0.998 0.914 0.991 F
C C 1.088 1.019 1.001 1.059 1.089 1.051 0.982 0.981 1.027 C
0 1.088 0.*99 1.059 0.905 0.957 1.050 r
n Internal e
r 1.104 0.852 1.089 Water 1.068 0.807 1.035 Channel 1.079 0.986 1.051 1.025 0.942 1.039 1.068 0.998 0.982 0.905 1.068 1.025 0.811 0.954 1.005 1.013 0.914 0.981 0.957 0.807 0.942 0.954 0.874 0.957 1.005 0.991 1.027 1.050 1:035 1.039 1.005 0.957 0.958 Figure 3.8 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-391B-14G8.0.100M With Channel Bow (Assembly Exposure of 18,000 MWdIMTU)
Siemens Power Corporation
S- EMF-2440
- , LaSalle Unit 2 Cycle 9 .. Revision 0 Plant Transient Analysis Page 3-23 Cont rol-R o d Corner 0
h II I n t 1.058 1.049 1.092 1.091 1.107- 1.082 1.072 1.017 1.010 r
0 1.049 0.945 1.020 0.996 0.843 0.987 0.998 - 0.906 0.995 R
0 d 1.092 1.020 1.002 1.061 1.090 1.052 - 0.981- 0.980 1.030 C
0 1.091 0.1996 1.061 0.894 --0.955 -1.053 r
n Internal _
e r 1.107 0.843 1.090 -Water 1.067 -0.797 1.036 Channel 1.082 0.987; 1.052 -1.024 0.941 1.041 1.072 0.998 0.981 0.894 1 1.067- -1.024 0.800 0.952- -1.007 1.017 0.906 0.980 0.955 0.797. 0.941.- -0.952 -0.865 0.960 1.010- 0.995 1.030 1.053 1.036 1.041 -1.007-- 0.960 -0.960
,Figure 3.9 LýaSalle Uniti2 Cycle 9 "
Safety'Limit Loc6al Peaking Factors SPCA9-410B-19G8.0-iOM With Channel Bow (Assembly Exposure of 17,500 MWd/MTU)
.q;rmrn_ Pnwur Cenmrntainn
LaSalle' Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis
'Page3-24
- 9 C ont ro I Rod Corner 0
n t
r 0
I R
0 d
C a
r n
e r
Figure 3.10 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCAg-383B-16G8.0-100M With Channel Bow (Assembly Exposure of 17,500 MWdIMTU)
Siemens Power Corporation
EMF-2440
- LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-25 ContArol* Rod Co rn er V
'*1 h n t 1.025 -1.058 1.062 1.117 1.100 1.108 1.043 1.026 0.979 r
0 I 1.058 0.934 1.018 0.852 1.003 0.845 0.999 0.903' 1.005 R
0 d 1.062 1.018 1.003 1.067 1.092 1.058 0.984 0.983 1.006 C
0 1.117 0.'852 1.067 1.046 0.823 1.056 r
n Internal e
r 1.100 1.003 1.092 Water 1.072 0.968 1.039 Channel 1.108 0.845 1.058 1.038 0.816 1.046 1.043 0.999 0.984 1.046 -1.072 1.038 0.965 0.963 0.986 1.026 0.903 0.983 0.823 0.968* :;*0.816 0.963 0.873 0.973 0.979 1.005 1.006 -1.056' 1.'039" 1.046 0.986 0.973 0.933 Figure 3.11 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-396B-12GZ-O10M With Channel Bow (Assembly Exposure of 15,000 MWdMTU)
Siemens Power Comoration
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 It Plant Transient Analysis Page 3-26 2.75 265 2M if s0 60 70 80 s0 100 110 POKM of RaWOd Figure 3.12 EOC Base Case Power-Dependent MCPR Limits for ATRUM-SB Fuel - TSSS Insertion Times Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0
,' Plant Transient Analysis Page 3-27 II h 2.15 06 1o056 1.195 1.75 I1S 0 -10 20 -- 30 40 50 s0 70 w g0 10w 110
'PaOr(6ofM d Pd Figure 3.13 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel -TSSS Insertion limes Siemens Power Comoration
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 Page 3-28 i,
s0 6o 70 s0 so 100 110 Po*wr (%of RaFs Figure 3.14 EOC Base Case Power-Dependent MCPR Limits for ATRUM-9B Fuel - NSS Insertion Times Siemens Powwr r".,Yp~44..
EMF-2440.
LaSalle Unit 2 Cycle 9 Revision 0 I
01 a
4I I"II& I EVEI1 it8I *F'*
rM
- '1 Paoe 3-29 Ii I,
0 10 20 30 40 - 50 O0 70 30 so 100 110 Po@r (%*o fR udW Power MCPRP
.(%) . Limit
'100 1.48 s60 o 1.51
-25 1.97
,25 2.20 0 2.70 Figure 3.15 EOC Base Case Power-Depiedent MCPR Limits for GE9 Fuel - NSS Insertion Times
.. I
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 33U
,I, IM 1.
IUL 0 10 20 30 40 "S "50 70 8D go 100 110 Pr.,r.(i Figure 3.16 EOC Base Case Power-Dependent LHGR Multipliers for ATRUM-9B Fuel - TSSS Insertion Times Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-31 1JhA 1
U 0 uSW II It 125' 0 SLGFVVC p 1.20' 1.15 0 U
1.10 U S
- 0. 1.00' 0.80' 0.75 0.70
- 1 0.8I 0 Am 0 10 20 30 40 50 60 70 s0 90 100 110 P r (NNQ Power LHGRFACp
(%) Multiplier 100 1.00 60 1.00 25 0.79 25 ' 0.79 0 0.79 Figure 3.17 EOC Base Case Power-Dependent LHGR Multipliers for ATRUM-9B Fuel - NSS Insertion Times Siemens Power Comomfion
EMF-2440 LaSalle Unlt2Cycleg Revision 0
-Plant Transient Analysis Page 4-1 4.0 Transient Analysis for Thermal Margin - Extended Operating Domain This section describes the development of the MCPR and LHGR limits to support operation in the following extended operating domains:
, , Increased core flow (ICF) to 105% of rated flow.
- Power coastdown to 40% of rated power.- - I
- Final feedwater temperature reduction (OFMR) of up to 100*F and with ICF. Since FFTR is typically used in connection with coastdown, analyses were performed to support combined FFTRlcoastdown operation.
Results of the limiting transient analyses are used to determine appropriate MCPRp limits and LHGRFACp multipliers for ATRIUM-9B and GE9 fuel to support operation in the EOD scenarios.
MCPRp limits are established for both ATRIUM-9B and GE9 fuel while LHGRFAC. multipliers are only established for the ATRIUM-9B fuel.--
As discussed in Reference 9, the MCPR safetylimit analysis for the base case remains valid for operation in the EODs discussed below. Also, the flow-dependent MCPR and LHGR analyses described in Section 3.4 were performed such that the results are applicable for all the EODs.
4.1 IncreasedCore Flow' The base case analyses presented in Section 3.0 were performed to support operation.in the powertflow domain presented in Figure 1.1, which includes operation in the ICF region. The coastdown and combined FFTRlcoastdown analyses are performed in conjunction with ICF to conservatively maximize the exposure at which a given power level can be attained. As a result, the analyses performed support operation in the ICF extended operating domain for all exposures.
4.2 CoastdownAnalysis Coastdown analyses were performed to ensure that appropriate MCPRp limits and LHGRFACý multipliers are applied to support coastdown operation. The analyses were performedfor coastdown operation to 40% of rated power using a conservative coastdown rate equivalent to a 10% decrease in rated power per 1000 MWd/MTU increase in exposure. An additional 1000 MWd/MTU was added to the EOFP exposure prior to the start of coastdown to provide operation support for operation at up to 10% of rated power above the equilibrium xenon coastdown power level. The MCPRp limits and LHGRFACp multipliers are based on results of Siemens Power Cor*ration
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 Page 4-2 LRNB and FWCF analyses. The analyses were performed at cycle exposures consistent with the assumed coastdown rate. This corresponds to the highest exposure at which the power can be obtained. The base case'coastdown ,CPRs for both the ATRIUM-9B and GE9 fuel as well as the ATRIUM-SB LHGRFACp results are presented in Table 4.1 for the indicated power/flow conditions. The ATRIUM-9B MCPRp limits and LHGRFACp multipliers for coastdown operation are presented in Figures 4.1 and 4.2. The GE9 coastdown MCPRF, limits are presented in Figure 4.3.
4.3 Combined Final FeedwaterTemperatureReduction/Coastdown Analyses were performed to support FFTR with thermal coastdown to ensure that appropriate MCPRp limits and LHGRFAC' multipliers are established. The combined FFTRlcoastdown analysis used a I 00 0F feedwater temperature reduction applied at EOFP to extend full thermal power operation. The coastdown exposure extension discussed in Section-4.2 (1000 MWdIMTU to support operation at up to 10% of rated power above the equilibrium xenon power level) was then applied. LRNB and FWCF analyses were performed to establish MCPRp limits and LHGRFACp multipliers. The Cycle 9 FFTRlcoastdown ACPR results for both -ATRIUM-SBand GE9 fuel as well as the LHGRFACp results are presented in Table 4.2 for the indicated power flow conditions. The ATRIUM-9B MCPRp limits and LHGRFACp multipliers for combined FFTR/coastdown operation are presented in Fidures 4.4 and 4.5. The GE9 coastdown MCPRp limits are presented in Figure 4.6.
Siemens Powor *"*wnnmtia.
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 4-3 Table 4.1-Coastdown Operation Transient Results Power/ Flow - - ATRIUM GES
(% rated I Event % rated) --ACPR- LHGRFAC, ACPR LRNB 1001105 -- 0.31-- 1.00 0.41 LRNB 80/105 - -... 0.32 1.00 0.35 LRNB 60 1 105 0.31 0.99 0.35 LRNB 40 i 105 ... 0.31 0.96 0.31 SLRNB 25 /105 0.19- 1.13 0.19 FWCF 100/105 -- 0.26 1.08 -0.32 FWCF , 80/105 . 0.29 1.08 0.31.
60 / 105- 0.34 --1.08 S...FWCF 0.36
..FWCF 40/105 . .. 0.44 1.12 0.44 FWCF -25 /105 ...... 0.86 1.08 0.88 Siemens PowerCororation
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 NJ Plant Transient Analvsis Page 4-4 Table 4.2 FFTR/Coastdown Operation Transient Results "
Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 4-5 1,II M
1.55 1.45
- _1.255 1.15 0 10 20 30 " 4o so so 70 s0 90 1o0 110 Pow*r %*of RaMd)
Power - MCPRp
(%) .Limit
.100 1.42 60 1.48 25 ' 2.05 25 2.20 0 2.70 Figure 4.1 Coastdown Power-Dependent MCPR Limits for ATRUM-9B Fuel Siemens Power Comoration
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis rat4-o Pag
" Ii 0
CL 0 1o 20 30 40 so 00 70 go go 100 11l Pow, Iof
% Rsb, Figure 4.2 Coastdown Power-Dependent LHGR Multipliers for ATRUM-9B Fuel Siemens Power Corporation
EMF-2440
%I LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 4-7 I,I h h I
S 10 20 30 40 0 s0 70 0 0 3*o 100 110 Povm (%of Rawd Power- MCPRp,
- (%) Limit.
'100 --1.52
'60 1.54
-25 -2.05 "25 2.20 0 2.70 Figure 4.3 Coastdown Power-Delpndent MCPR Limits for GE9 Fuel QI.Mmne pr%&~ f^ýMtLý
LaSalle Unit 2 Cycle g EMF-2440 Revision 0 Plant Transient Analysis Page II 0 10 40 s 70 s s0 1W 110 POWN 1%of i.-,
Figure 4.4 FFTR/Coastdown Power-Dependent MCPR Limits for ATRUM-9B Fuel Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Pame 4-9 I,
1.20 1.10 CL1.00 a 10 20 30 40 s0 s0 70 60 s0 10 110 pamt,*fR itmd)
Power, LHGRFACý
(%) ,Multiplier
'100 , 1.00
-560 0.97 25 0.65 25 0.65 0 0.65 Figure 4.5 FFTRiCoastdown Base Case Power-Dependent LHGR Multiplierm for ATRUM-9B Fuel
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 41 Page 4-10 II 0 10 20 30 40 s0 0 70 80 so 100 110 Pow"(% ofRaMOM Figure 4.6 FFTR/Coastdown Power-Dependent MCPR Limits for GE9 Fuel Siemens Power Corporation
EMF-2440
,' LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 5-1 5.0 Transient Analysis for Thermal Margin - Equipment Out-of-Service This section describes the development of the MCPR and LHGR operating limits to support operation with the following EOOS scenarios:
, " Feedwater heaters out-of-service (FHOOS) - 100°F feedwater temperature reduction.
"" '1 recirculation pump loop (SLO). *
- Turbine bypass system out-of-service (TBVOOS).
- Recirculation pump trip out-of-service (No RPT).
- Slow closure of I or more turbine control valves. .
Operation with 1 SRV out-of-service, up to 2 TIPOOS (or the equivalent number of TIP channels) and up to 50% of the LPRMs out-of-service is supported by the base case thermal limits presented in Section 3.0. No further discussion for these EOOS scenarios is presented in this section., The EOOS analyses presented in this section also include the same -EOOS scenarios protected by the base case limits.
Results of the limiting transient analyses are used to establish appropriate MCPRI limits and LHGRFAC, multipliers to support operation in the EOOS scenarios. All EOOS analyses were performed with TSSS insertion times.
As discussed in Reference 9, the base casie MCPR safety limit for two-loop operation remains applicable for operation in the EOOS scenarios discussed below with the exception of single loop operation- Also, the flow-dependent MCPR and LHGR analyses described in Section 3.4 were performed such that the results are applicable in all the EDOS scenarios.
5.1 FeedwaterHeaters Out-of-Service (FHOOS)
The FHOOS scenario assumes a 100OF reduction in the feedwater temperature. Operation with FHOOS is similar to operation with FFTR except that the reduction in feedwater temperature due to FHOOS can occur'at any time during the 'ycle".The effect of the reduced feedwater temperature is an increase in the core subcooling Which can change the power shape and core void fraction: While the LRNB event is less severe due to the decrease in steam flow, the FWCF event can-get worse due to the increase in core inlet subcooling. FWCF analyses were performed for Cycle'9 to determine thermal limits to'support operation with FHOOS. The ACPR and LHGRFAC, results used to develop the EOC operating limits with FHOOS are presented in Table 5.1. The EOC MCPRp limits and LHGRFACp multipliers for ATRIUM-9B fuel for FHOOS
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 Page 5-2 operation are presented in Figures 5.1 and 5.2, and the EOC FHOOS GE9 MCPRp limits are presented in Figure 5.3.
5.2 Single-Loop Operation(SLO) 5.2.1 Base Case Operation The impact of SLO at LaSalle on thermal limits was presented in Reference 9. The only impact is on the MCPR safety limit. As presented in Section 3.2, the single-loop operation safety limit Is 0.01 greater than the two-loop operating limit (1.12 compared to 1.11). The base case ACPRs and LHGRFACp multipliers remain applicable. The net result is an increase to the base case MCPRp limits of 0.01 as a result of the increase in the MCPR safety limit.
5.2.2 Idle Loop Startup The MCPRp limits and LHGRFACp multipliers for the startup of an idle recirculation pump are based on the results of the abnormal startup of the idle recirculation loop analysis and thi SLO MCPR safety limit analysis. As discussed in Section 3.2, the single-loop operation safety limit Is 1.12 or 0.01 higher than the two-loop operation Iimit. The process used for the abnormal startup of the idle recirculation loop analysis for-L2C9 is'presented in Reference 20. The responses of the system parameters for the L2C9 analysis are consistent with those presented in Reference
- 20. The Reference 20 results demonstrated that.,the lowest power (35%P/47%F) conditions provide conservative results. Subsequently, the L2C9 analyses were performed at 35%P/47%F.
The limiting exposure Ias'determined to be BOC. The ACPR and LHGRFACY results for the abnormal staitup of the idle recirculation loop are, presented in Table 5.2. Figures 5.4 and 5.5 present the ATRIUM-9B MCPRp limits and LHGRFACp multipliers for idle loop startup. The GE9 MCPRp limits for idle loop startup are presented in Figure 5.6.
5.3 Turbine Bypass Valves Out-of-Service (TB VOOS)
The effect of operation with TBVOOS is a reduction in the system pressure relief capacity, which makes the pressurization events more severe. While the base case LRNB event is analyzed assuming the turbine bypass system ot-of-service, operation with TBVOOS has an effect on the FWCF evenL The FWCF event was evaluated for LaSalle Unit 2 Cycle 9 to support operation with TBVOOS. The ACPR and LHGRFACP results used to develop the EOC operating limits with TBVOOS are presented in Table 5.3. The EOC MCPRp limits and LHGRFACp Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 "PlantTransient Analysis Page 5-3 multipliers'for ATRIUM-9B fuel for TBVOOS operation are presented in Figures 5.7 and 5.8, and the EOC TBVOOS GE9 MCPRp limits are presented in Figure 5.9.
5.4 RecirculationPump Trip Out-of-Service (No RP)
This section summarizes the development of the thermal limits to support operation with the EOC RPT inoperable. When RPT is inoperable, no credit for tripping the recirculation pump on TSV position or TCV fast closure is assumed. The function of the RPT feature is to reduce the severity of the core power excursion caused by the pressurization transient. The RPT accomplishes this by helping revoid the core, thereby reducing the magnitude of the reactivity insertion resulting from the pressurization transient. Failure of the RPT feature can result in higher operating limits because of the higher positive reactivity in the core at the time of control rod insertion.
Analyses were performed for LRNB and FWCF events assuming no RPT. The ACPR and LHGRFACp results used to develop the EOC operating limits with no RPT are presented in Table 5.4. The EOC MCPRp limits and LHGRFACp multipliers for ATRiUM-9B fuel for operation with no RPT are presented in Figures 5.10 and 5.1 1, and the EOC no RPT GE9 MCPRP limits are presented in Figure 5.12.
5.5 Slow Closure of the Turbine Control Valve LRNB analyses were performed to evaluate the impact of a TCV slow closure. Analyses were performed closing 3 valves in the normal fast closure mode and 1 valve in 2.0 seconds. Results provided in Reference 23 demonstrate that pirforminrg the analyses with 1 TCV closing In 2.0 seconds protects ,operation with up to 4 TCVs -closingslowly. Sensitivity analyses below 80% power have shown that the pressure relief provided by all 4 TCVs closing slowly can be sufficient to preclude the high-flux scram set point frorm being exceeded. Therefore, credit for high-flux scram is not taken for analyses at 80% power and below. The 80% power TCV slow closure analyses were performed both with and without high-flux scram credited. The ACPR and LHGRFACp results of the analyses performed are presented in Table 5.5.
The MCPRp limits and LHGRFACp multipliers are established with a step change at 80% power.
At 80% power, the lower-bound MCPR, limits and upper-bound LHGRFACI, multipliers are based on the analyses which credit high-flux scram; the upper-bound MCPRp limits and lower bound LHGRFACp multipliers are based on analyses which do not credit high-flux scram. While
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 Page 5.4 th'e TCV slow closure analysis is performed without RPT on valve position, it does not necessarily bound the LRNB no RPT or FWCF no RPT events at all power levels because the slow closing TCV provides some pressure relief until it completely closes. Therefore, the MCPRp limits and LHGRFACp multipliers for the TCV slow closure EOOS scenario are established using the limiting of the no RPT results reported in Section 5.4 and the TCV slow closure results.
The EOC MCPRp limits and LHGRFACp multipliers for ATRIUM-9B fuel for operation with TCV slow closure are presented in Figures 5.13 and 5.14 and the EOC TCV slow closure GE9 MCPRp limits are presented in Figure 5.15. The limits presented in Figures 5.13 through 5.15 protect the scenario of all 4 TCVs closing slowly.
5.6 CombIned FHOOSITCV Slow Closure and/or No RPT MCPRp limits and LHGRFACp multipliers were established to support operation with FHOOS, TCV slow closure and/or no RPT. The TCV slow closure ACPR and LHGRFACp results with FHOOS become less limiting than the TCV slow closure event with nominal feedwater temperature since the initial steam flow with FHOOS is lower and produces a less severe pressurization event. Subsequently, no TCV slow closure with FHOOS analyses were performed. The TCV slow closure results with nominal feedwater temperature are considered in determining the combined FHOOSrIcv slow closure and/or no RPT MCPRp limits and LHGRFACp multipliers. The limits were developed based on the limiting of either the TCV slow closure analysis results discussed in Section 5.5 or the analyses with both FHOOS and no RPT presented in Table 5.6.
The EOC MCPRp limits and LHGRFACp multiplieis for ATRIUM-9B fuel with FHOOSfTCV slow closure and/or no RPT are presented in Figures 5.16 and 5.17, and the EOC GE9 MCPRp limits for the same EOOS scenario are presented in Figure 5.18. The limits presented in Figures 5.16 through 5.18 protect the scenario of all 4 TCVs closing slowly.
Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 5-5 Table 5.I EOC Feedwater Heater ht Out-of-Service Analysis Results
-Power/Flo w1 ATRIUM GE9
(% ratedI Event % rated) LACPR LHGRFACý , CPR FWCF 100/ 105 0.26 1.0o* 0.31 FWCF 100/81 0.23 1.11 0.28 FWCF 8D 1105 0.30 1.03* 0.36 FWCF 60/ 105 0.40* 0.97* 0.46*
FWCF 40/105 0.62" 0.87' 0.69' FWCF 25/105 1.03 0.69" 1.11' The analysis results presented are from an eadier cycle exposure. The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.
Siemens Power Comoration
EMF-2440 LaSalle Unit 2 Cycle 9 Plant Revision 0 Transient Analvsis 4 Plant~ri~
I~~ Trnietnavi Pae Table 5.2 Abnormal Recirculation Loop Startup Analysis Results II Power I Flow FCV ATRIUM-9B
(% rated /
% rate*Position F nV
%ACPR LHGRFACp 35/47 27% open 1.46t 0.421 "A ,CPR results forATRIUM-gB fuel are conservatively t The analysis results presented applicable for GE9 fuel.
are from an earlier cycle exposure. The ACPR and LHGRFAC, results are conservatively used to establish the thermal limits.
Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 5-7 Table 5.3 EOC Turbine Bypass Valves Out-of-Service Analysis Results II Power I Flow ATRIUM -GE9
(% ratedI Event % rated) -,CPR LHG.RFACp &ICPR FWCF 100/105 -0.32 1.02 '0.41 FWCF- 160/81 0.31 0.99- 0.41 FWCF 80/105 t0.35 1.00' 0.45 FWCF 80/57.2 0.31 1.05 0.41 FWCF 601-105 0.41* 0.97' 0.51 FWCF 60 /35.1 0.15 1.14 0.25 FWCF 40/105 0.58' 0.90" 0.668 "FWCF 25 / 105 0.87. 0.76*' 0.97 The analysis results presented are from an earlier cycle exposure. The -ACPR and LHGRFAC4 results are conservatively used to establish the thermal limits.
RIamane PnsAmr rtjnrr*mmfn
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analvsis Pln ............
An -'-s Page 5-8.
Table 5.4 EOC Recirculation Pump Trip Out-of-Service Analysis Results Power / Flow ATRIUM GE9
(% rated I Event % rated) ACPR LHGRFACý ACPR LRNB 100/105 0.40 0.89 0.50 LRNB 100/81 0.32 0.91 0.47 LRNB 80/105 0.35 0.94 0.47 LRNB 80/57.2 0.30 0.97 0.44 LRNB 601105 0.32 0.99 0.44 FWCF 100/105 0.31 0.97 0.40 FWCF 100/81 0.26 0.99 0.35 FWCF 80/105 0.33 1.00" 0.43 FWCF 60 /105 0.38 0.97- 0.48 FWCF 40/105 0.510 0.91' 0.59*
FWCF 25/105 0.78' 0.79* 0.87*
" The analysis results presented are from an earlier cycle exposure. The
,CPR and LHGRFACp results are conservatively used to establish the thermal limits.
Siemens Power Corporation
EMF-2440
-NiLaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Pace 5-g Paae 5-9 Table 5.5 EOC Turbine Control Valve Slow Closure Analysis Results II Slow Power I Flow ATRIUM-9B GE9 Valve (%rated/
Event Characteristics % rated) ACPR LHGRFAC, .ACPR LRNB 1 TCV Closing at 2.0 sec 100/ 105' 0.42 0.93 0.52 LRNiB4 , 1 TCV closing at 2.0 sec 100181" 0.33 0.97 0.49 LRNB 1 TCV closing at 2.0 sec 80/ 105* 0.40 0.96 0.49 LRNB I TCV closing at 2.0 sec 801 57.2 0.50 0.97 0.73 11RNB.
TCV closing at 2.0i sec s80105 0.52* 0.86W 0.62 LRNB 1TCV closing at 2.0 sec 80/57.2t 0.58 0.92 0.84 LRNB 1 TCV closing at 2.0 sec 60/ 105? 0.61* 0.83* 0.71S LRNB I TCV closing at 2.0 sec 601 3 5 .1t 0.63* 0.940 0.86 LRNB 1 TCV closing at 2.0 sec 40 / 105t 0.78 0.77* 0.84 LRNB 1 TCV closing at 2.0 sec 25/105? 0.99 0.70e 0.97*
t Scram initiated by high-neutron flux.
Scram initiated by high dome pressure, 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 Comoration
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis Page 5-10 Table 5.6 EOC ReclrculatJon Pump Trip and Feedwater Heater Out-of-Service Analysis Results I,1,
- The analysis results presented are from an earlier cycle exposure.
The ACPR and LHGRFACý results are conservatively used to establish the thermal limits.
Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 5-11 2.5 ]
£:9 0 10 20 30 40 W0 I 70 60 90 100 110 Rg.4 dur(%c Power MCPRP S(%) .... . Limit
. .*1 00 -~-* - . .. ..1 ;41 S.
60- 1.51
-25 -..-- ..2.14
__25 2.35 0 2.85 Figure 5.1 EOC Feedwatei Heaters Out-of-Service Power-Dependent MCPR Limits for ATRIUM-gB Fuel
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 II . *1 I II Ii C.
5 0 10 20 30 40 '50 60 70 s0 s0 100 110 Power (% of Rpod)
.. Figure 5.2 EOC Feedwater Heaters Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel Siemens Power Corporation
EMF-2440
--LaSalle Unit 2 Cycle 9 Revision 0 II' Plant Transient Analysis Page 5-13 2.95
,, 2.75' '
2M5 2.45 2.35 2.25'
- 2.15' 02.o5' IM5.
1.65 1AS 1.35 0 10 20 30 40 70 s0o0 100 110 PO.f (%of Raod)
Power MCPRP
(%) Umit 100 1.51 60 - 1.57 25' 2.22 25-- 2.35 0 2.85 Figure 5.3 EOC Feeciwater Heaters Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel Siemens Power Corporation
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis Page 5-u 1-5 £ 2255 215 li I
131S 1.45,
'E"IcP 1.25 I 1.15 0 10 20 30 40 50 60 70 s0 . 90 100 110 PawW Mf aRaMd Figure 5.4 Abnormal Idle Recirculation Loop Startup Power-Dependent MCPR Limits for ATRIUM-9B Fuel Siemens Power Corporation
EMF-2440 Revision 0 LaSalle Unit*. 2 Cycle 9
- ri
"'.,,,.=,P4 ,*.h 1 * . Paoe 5-15 ra1ilL I ¢II Ia *I V rI "0 saIyaO 1.20 1,20 k9e Loop 1.15 - LHGRFAQP ýýq 1.10 1.00' 0.U5 O. 0.110' 0.70 0.65 0.600 0.45 *
- 0.40' OM 35 0 10 20 30 40 10 60 70 s0 90 100 110 Power (% of RatMO)
Power LHGRFACp Jo(_)_........ Multiplier
-100 0.40
'60 0.40
"-25-- 0.40 25-- 0.40 0 0.40 Figure 5.5 Abnormal Idle Recirculation Loop Startup Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel
'--,*,=De psar i"_*,vraftin
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis iPage 5-16 Y
Ii CL 0 10 20 30 40 SO so 70 so so lW 110 POW CA of Ratbd)
Figure 5.6 Abnormal Idle Recirculation Loop Startup Power-Dependent MCPR Limits for GE9 Fuel Siemens Power Corporation
EMF-2440 Revision 0 LaSalle Unit 2 Cycle 9 Page 5-17 Plant Transient Analysis 2lS IN I
0 10 .20 30 40 50 OD 70 so 90 1M 110 Powr (%d RoOM Power- MCPRp
(%) Limit 100 1.43
---60 .. . 1.52
.825 1.8
-'25 2.20, 0 2.70 Figure 5.7 -EOC Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits for ATRIUM-SB Fuel CL--. 0-, rýM#Lý
EMF-2440 LaSalle Unit 2 Cycle 9' Revision 0 Plant Transient Analysis Page 5-18 I
0 10 20 30 40 S0 W0 70 0o 90 100 110 Po"Tr (%of Ratad)
- Figure 5.8 EOC Turbine Bypass Valves Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel Siemens Power Corporation
- EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 5-19 275 2.65
- 1 h I 0 10 20 30 40o 60o o 70 so 90 100 110 Power (%odf 4 Power MCPR,
(%) Limit 100 1.52
__ -1.62
__0 25-* 2.08 25 2.20 0 2.70 Figure 5.9 EOC Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel Siemens Power Comoration
I LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis
.Page 5-20 0 1o 0 30 40 5 O 7 a s m 117 POW.Wr (%ofpa Figure 5.10 EOC Recirculatlon Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel Siemens Power Corporation
EMF-2440 Revision 0
," LaSalle Unit 2 Cycle 9 Page 5-21
-lantl i ans li~ it etno y*a" 1-h II
¶20 1.20 I LRN8 FIJcF
- FWACP I 1.15" 1.10" 1.05 a." 1.00 0 0 U
U U
0 a
S0.85' MgOO
- 0.30 a 0.75 0.70' 0.65 1 L
0 10 20 30 40 s0 s0 70 80 90 100 110
,Poamr (M Power- LHGRFACp
+(%) Multiplier oo :.0.89 .
++60 ... . 0.89 25 0.78 =.
+25 0.78 0 0.78 Figure 5.11 EOC Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-SB Fuel I
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analysis Revision 0 in Plant APage Transien 5-22 1850 U POWM (%of Paea Figure 5.12 EOC Recirculatdon Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 "PlantTransient Analysis Page 5-23 Ii 1.5 1.75 1.65 1.55 1.35' "125 1.15 0 10 20 30 40 so so 70 Wo O0 10 110 P~omr(%* rtmd)
Power MCPRp
(%) t .. Limit
-100 - 1.53 S... 80 .. .. .1.61
'80 1.69 25 2.10 "25 - 2.20 0 2.70 Figure 5.13 EOC Turbine.Control Valve Slow Closur-e andlor Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel
I LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis Page 52 0 10 20 30 40 so o0 70 80 90 100 110 Figure 5.14 EOC Turbine Control Valve Slow Closure andlor Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel Siemens Power Corporation
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 "PlantTransient Analysis Page 5-25 II 0 10 2 0 40 so 60 70 W s0 100 110 PwAr(% d Ramd Figure 5.15 EOC Turbine Control Valve Slow Closure and/or Recirculation Pumpl Trip Out-of-Service Power-Dependeht MCPR Limits for GE9 Fuel Siemens Power Corporation
LaSalle Unit 2 Cycle 9 EMF-2440 Plant Transient Analvsis Revision 0 Plant A Transiet s IPage 5-26 2.95 25*'
'I
'I 2.75' FOF tbNo RP, R-oos I -* OI*C 2Z5 215~
0 1.75 a 2 9
14A5 a
U 0 S
U 1.15 0 10 W . 3 40 so 0 0 10 110 POWN (%d. "4 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-SB Fuel Siemens Power Corporation
ZEMF-2440
,, LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 5-27
.1.30 II 125
- SIOW TCV SFV*C tikmn%4%
No RPT FHOOS 1.20 - -1 LHGRFACp Tf F 1.15 1.10 1..
a 0.800 a II 0
a U
a 0.80" 0.75 U
0.65-,
nI Rfn 0 10 20 30 40 50 G0 70 80 90 100 110 PC~w(%of pift Power LHGRFACp
(%) .Multiplier 100 0.89
-80 0.89 80 0.86
'25 " 0.68 25 0.68 0 0.68 Figure 5.17 EOC Turbine Control Valve Slow Closure andlor Recirculation Pump Trip and Feedwater Heaters Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel QIMM.ne Dýr
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Plant Transient Analysis
. ae -o rage 2.85 S 2.75' C.
i o-10 0 10 2D 30 40 80 s0 70 80 go 1W 10 POW.r(% of abd Figure 5.18 EOC Turbine Control Valve Slow Closure andlor Recirculation Pump Trip and Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel Siemens Power Corporation