ML022940532
| ML022940532 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/15/2002 |
| From: | Barnes G Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| -RFPFR | |
| Download: ML022940532 (164) | |
Text
Technical Requirements Manual - Appendix J Section 1 LaSalle Unit 2 Cycle 9 Core Operating Limits Report August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Issuance of Changes Summary LaSalle Unit 2 Cycle 9 Affected Affected Summary of Changes Date Section Pages All All Original Issue (Cycle 9) 11/00 References; 6 iii; 6-1 Revised Requirements for Use of SUBTIP Methodology 12/00 All All ITS changes, RBM trip setpoint and allowable value 5/01 changes, TIP symmetry Chi-Squared testing, incorporated results of revised thermal limits with correct thermal conductivities, and other necessary administrative changes Table of ii, v, 2-3 Incorporate revised MCPR operating limits for ATRIUM-9B 8/01
- Contents, fuel due to schedule changes and changes in the target References, 2 rod patterns.
References, v, 2-1. 2-2, Defined nominal scram speed (NSS) requirements and 2/02 Section 2.2, 2-3, 2-4, added EOOS operating limits for NSS. Added Attachment Table 2-1, 2-5,3-5.
- 7.
Table 2-2, Table 3-1, Section 2, Table of Contents,.
References, iii, iv, v, Incorporate revised MCPR operating limits and revised 8/02 Section 2.2, 2-1,2-3, LHGR limits due to changes in target rod patterns.
Table 2-1, 2-4, 2-5, Table 2-2, 2-6, 2-7, Incorporate revised MCPR operating limits and revised Section 3.2.2, 3-1, 3-2, LHGR limits for operation between core average Section 5 3-6, 5-1 exposures of 30,266.2 MWd/MTU and 31,242.7 MWd/MTU.
Removed Nominal Scram Speed MCPR operating limits.
1 1_ Other necessary administrative changes.
August 2002
Technical Requirements Manual
- Appendix J Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Table of Contents References.........................................................................................................................
ii
- 1.
Average Planar Linear Heat Generation Rate (3.2.1)..........................................
1-1 1.1 Tech Spec Reference................................................................................
1-1 1.2 Description.................................................................................................
1-1
- 2.
M inim um Critical Power Ratio (3.2.2)..................................................................
2-1 2.1 Tech Spec Reference................................................................................
2-1 2.2 Description.................................................................................................
2-1
- 3.
Linear Heat Generation Rate (3.2.3)....................................................................
3-1 3.1 Tech Spec Reference.......................................
3-1 3.2 Description.................................................................................................
3-1
- 4.
Control Rod Withdrawal Block Instrumentation (3.3.2.1)..................
4-1 4.1 Tech Spec Reference................................................................................
4-1 4.2 Description 4-1
- 5.
Allowed M odes of Operation (B 3.2.2, B 3.2.3)....................................................
5-1 7
- 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 9 ii August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report References 1
Letter from D. M. Crutchfield to All Power Reactor Licensees and Applicants, Generic Letter 88-16; Concerning the Removal of Cycle-Specific Parameter Limits from Tech Specs, dated October 4, 1988.
- 2.
LaSalle Unit 2 Cycle 9 Neutronics Licensing Report (NLR), 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 Spectrum Analysis for LaSalle Units I and 2, EMF-2174(P), March 1999.
- 6.
LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM-9B fuel, EMF-2175(P), March 1999.
- 7.
LaSalle Extended Operating Domain (EOD) and Equipment Out of Service (EOOS) Safety Analysis for ATRIUM-9B Fuel, EMF-95-205(P), Rev. 2, June 1996.
- 8.
ARTS Improvement Program analysis for LaSalle County Station Units 1 and 2, NEDC-31531P, December 1993 and Supplement 1, June 1998 (Removal of Direct Scram Bypassed Limit).
- 9.
Lattice-Dependent MAPLHGR Report for LaSalle County Station Unit 2 Reload 6 Cycle 7, 24A5162AA, Revision 0, 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 1 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 BSA'00-025, R.W. Tsai to D. Bost, April 13, 2000.
- 13.
"Updated Transient Analysis: Abnormal Start-up of an Idle Recirculation Loop for LaSalle County Nuclear Station, Units 1 and 2", B33-00296-03P, March 1998 and "LaSalle Unit 2 Cycle 8 Abnormal Idle Recirculation Loop Startup Analysis%,
DEG.99 070, D. Garber to R. Chin, March 8, 1999.
- 14.
"TIP Symmetry Testing", J H. Riddle to R. Chin, January 20, 1997 and "TIP Symmetry Testing", DEG.99 085, D. Garber to R. Chin, March 8, 1999
- 15.
"POWERPLEX-II 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 Rieck to G.Spedl, NFSBSS 93 117, May 19, 1993.
- 17.
"LaSalle Units 1 and 2 Operating Limits with Multiple Equipment Out of Service (EOOS)", NFS:BSA'95-024, April 6, 1995.
18 NFM Calculation No. BSA-L-99-07, *MAPFACf Thermal Limit Multiplier for 105% Maximum Core Flow."
- 19.
"ComEd GE9/GE10 LHGR Improvement Program" J1 1-03692-LHGR, Revision 1, February 2000.
- 20.
'LaSalle County Station Power Uprate Project", Task 201: Reactor Power/Flow Map, GE-NE-A1300384-07-01, Revision 1, September 1999.
- 21.
"Evaluation of CBH Effects on Fresh Fuel for LaSalle Unit 2 Cycle 9", DEG:00.232, D. Garber to R. Chin, October 2000.
- 22.
DEG.00.091, "Revised Measured Nodal Power Distribution Uncertainty for POWERPLEX Operation with Uncalibrated LPRMs", 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.
LaSalle Unit 2 Cycle 9 iii August 2002
Technical Requirements Manual -'Appendix J L2C9 Core Operating Limits Report
- 24.
Reactor Stability Detect and Suppress Solutions Licensing Basis Methodoloay for Reload Applications, NEDO-32465-P-A, August 1996.
- 25.
ANFB Critical Power Correlation, ANF-1125(P)(A) and Supplements I and 2. Advanced Nuclear Fuels Corporation, April 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 Corporation Critical Power Methodoloqy for Boiling Water Reactors/Advanced Nuclear Fuels Corporation Cntical Power Methodology for Boiling Water Reactors' Methodology for Analysis of Assembly Channel Bowing Effects/NRC Correspondence, XN-NF-524(P)(A) Revision 2 and Supplement 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation November 1990.
- 28.
COTRANSA 2" A Computer Program for Boiling Water Reactor Transient Analysis, ANF-913(P)(A), Volume 1, Revision I and Volume 1 Supplements 2, 3, and 4, Advanced Nuclear Fuels Corporation, August 1990.
- 29.
HUXY: A Generalized Multirod Heatup Code with 10CFR50, Appendix K Heatup Option, ANF-CC-33(P)(A), Supplemerit 1 Revision 1; and Supplement 2, Advanced Nuclear Fuels Corporation, August 1986 and January 1991, respectively.
- 30.
Advanced Nuclear Fuels Methodology for Boiling Water Reactors, XN-NF-80-19(P)(A), Volume 1, Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990
- 31.
Exxon Nuclear Methodology for Boiling Water Reactors' Application of the ENC Methodology to BWR Reloads, XN-NF-80 19(P)(A), Volume 4, Revision 1, Exxon Nuclear Company, June 1986.
- 32.
Exxon Nuclear Methodology for Boiling Water Reactors THERMEX-Thermal Limits Methodology Summary Description, XN-NF-80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Company, January 1987.
- 33.
Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-85-67(P)(A) Revision 1, Exxon Nuclear Company. September 1986.
- 34.
Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9X9-IX and 9X9 9X BWR Reload Fuel, ANF-89-014(P)(A), Revision 1 and Supplements 1 and 2, October 1991.
- 35.
Volume 1 - STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain. Volume 2 - STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain, Code Oualification Report, EMF-CC-074(P)(A),
Siemens Power Corporation, July 1994.
- 36.
RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, XN-NF-81-58(P)(A), 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(P)(A), Volume 1 and Volume 1 Supplements 1 and 2; Volume 1 Supplement 4, Advanced Nuclear Fuels Corporation, February 1987 and June 1988, respectively.
- 38.
Advanced Nuclear Fuels-Co*pioation Methodology-for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91 048(P)(A), Advanced Nuclear Fuels Corporation. January 1993.
- 39.
Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, XN-NF-80-19(P)(A)
Volume 1 and Supplements 1 and 2, Exxon Nuclear Company, Richland, WA 99352, March 1983.
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 9 iv
'August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report
- 41.
Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A), Revision 1 and Revision 1 Supplement 1, Advanced Nuclear Fuels Corporation. May 1995.
- 42.
NEDE-2401 1-P-A, General Electnc Standard Application for Reactor Fuel, Rev. 14. June 2000.
43 Commonwealth Edison Topical Report NFSR-0085, Benchmark of BWR Nuclear Design Methods, November 1990.
Revision 0
- 44.
Commonwealth Edison Topical Report NFSR-0085, Supplement 1, Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma Scan Comparisons, Apnl 1991, Revision 0.
- 45.
Commonwealth Edison Topical Report NFSR-0085, Supplement 2, Benchmark of BWR Nuclear Design Methods Neutronic Licensing Analyses, April 1991, Revision 0.
- 46.
Commonwealth Edison Topical Report NFSR-0091, Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods, Revision 0, Supplements 1 and 2, December 1991, March 1992, and May 1992, respectively; SER letter dated March 22, 1993.
- 47.
BWR Jet Pumo Model Revision for RELAX, ANF-91-048(P)(A), Supplement 1 and Supplement 2, Siemens Power Corporation, October 1997.
- 48.
ANFB Critical Power Correlation Application for Coresident Fuel, EMF-1 125(P)(A), Supplement 1, Appendix C, Siemens Power Corporation, August 1997.
- 49.
ANFB Critical Power Correlation Determination of ATRIUM-9B Additive Constant Uncertainties, ANF-1125(P)(A),
Supplement 1, Appendix E, Siemens Power Corporation, September 1998.
50 Exelon Generation Company LLC, Docket No. 50-374, LaSalle County Station, Unit 2 Facility Operating License, License No. NPF-18.
- 51.
"LaSalle Unit 2 Cycle 9 Operating Limits for Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity',
DEG.01:046, D. Garber to R. Chin, March 22, 2001.
- 52.
'LaSalle Unit 1 and Unit 2 Rod Block Monitor COLR Setpoint Change", NFM.MW 01-0106, Anthony Giancatanno to Jeff Nugent, April 3, 2001.
- 53.
"Transmittal of Revised CBH Effects on Fresh Fuel for LaSalle Unit 2 Cycle 9", DEG 01 090, D Garber to R. Chin, June 2001.
- 54.
"LaSalle Unit 2 Cycle 9 Equipment Out-of-Service Operating Limits Using Nominal Scram Speed and Exposure Limited to 14,000 MWd/MTU", DEG 02.009, D. Garber to F. W. Trikur, January 10, 2002.
- 55.
"Assessment of Continued Applicability of the CBH Study Documented in Reference 1", DEG:01:185, D Garber to F. W Tnkur, November 13, 2001.
- 56.
"Revised Control Blade History Study for LaSalle Unit 2 Cycle 9," DEG 02:100, D. E. Garber to F. W. Tnkur, May 31, 2002.
- 57.
"LaSalle Unit 2 Cycle 9 Operating Limits for Cycle Extension to 19,300 MWd/MTU," DEG 02:125, D E. Garber to F W.
Tnkur, August 9, 2002.
.8.
RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model, EMF-85-74 (P)(A) Revision 0 and Supplement 1(P)(A) and Supplement 2(P)(A), February 1998.
LaSalle Unit 2 Cycle 9 V
August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating 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:==
For operation without a full TIP set from BOC to 500 MWd/MT a penalty of 11.01% must be applied to all APLHGR limits.
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 SPCA9-381B-13GZ7-80M SPCA9-384B-11IGZ6-80M SPC-A9-391B-14G8.0-100M SPC-A9-41 0B-I 9G8.0-1 00M SPC-A9-383B-16G8.0-100M SPC-A9-396B-12GZ-100M (References 2 and 3)
Planar Average Exposure (GWd/MTU).
0.0 20.0 61.1 Cycle First Inserted 8 8 9
'-9 (References 3 and 6) 9 9
MNAP'LHGR (kW/ft)
(all Siemens fuel types) 13.5 13.5 9.39 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 9 1-1 August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Table 1.2-1 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) vs.
Average Planar Exposure for Fuel Type GE9B-P8CWB322-1 1 GZ-1 OOM-1 50-CECO (Reference 9 and 19)
Exposure Exposure Lattice-Type MAPLHGR (kW/ft)
MW D/ST)
,(MWD/MT)
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 9 1-2 August 2002
Technical Requirements Manual - Appendix J L2C9 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-1 50-CECO (Reference 9 and 19)
Exposure Exposure Lattice-Type MAPLHGR (kW/ft)
(MWDIST)
(MWD/MT)
I f
P8CWL071 P8CWL346 P8CWL358 P8CWL358 P8CWL346 P8CWL071 NOG 4G5.0/3G4.0 7G4.0 2G5.017G4.0 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 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 LaSalle Unit 2 Cycle 9 1-3 August 2002
I.1 For 25 > P:
No Thermal Limits Monitoring Required; If Official Monitoring is Desired, the Equations for > 25% Power May Be Extrapolated for 25 > P, provided the Official TCV/TSV closure J I 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Core Thermal Power (% Rated)
L Unit 2 Cycle 9 Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Figure 1.2-1 Power-Dependent SLO MAPLHGR Multipliers for GE Fuel (MAPFAC p)
(References 8 and 19) 1 0.95 0.9 0.85 0.8 0.75 0.7 0.65 0.6 0.55 0.5 0.45 0.4 0.35 0.3 0.25 02 0.15 01 0.05 0
- a.
ILL
- a.
75 2
- 0.
- 0.
0
_j U)
CL a,
- 0.
I 0.
monitoring is only performed with the scrams and RPT enabled.
For 25 _< P < 100 MAPFACp = 1.0+0.005224 (P-100)
For 100 < P, MAPFACp = 1.00 P = % Rated Core Thermal Power La,'
1-4 August 2002
C 1
-U 0.9 C.,
, 0.8 0
Ia O0.7 03
-j CL CL 0= 0.5" C=.
0 0
u- 0.4 0.3 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 Core Flow (% Rated) 1-5 August 2002
(
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Figure 1.2-2 Flow-Dependent SLO MAPLHGR Multiplier (MAPFAC F) for GE Fuel (References 8, 18, and 19)
(1 LaSalle Unit 2 Cycle 9
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report
- 2.
Minimum Critical Power Ratio (3.2.2) 2.1 Tech Spec
Reference:
2.2
==
Description:==
Prior to initial scram time testing for an operating cycle, the MCPR operating limit is based on the Technical Specification Scram Times. For Technical Specification requirements refer to Technical Specification table 3.1.4-1.
TIP Symmetry Chi-squared testing shall be performed prior to reaching 500 MWd/MTU to validate the MCPR calculation.
MCPR limits from BOC to Coastdown are applicable up to a core average exposure of 30,266.2 MWd/MTU (which is the licensing basis exposure used by SPC). (Reference 3)
MCPR limits for Coastdown are applicable from a core average exposure of 30,266.2 MWd/MTU to a core average exposure of 31,242.7 MWd/MTU (Reference 57).
2.2.1 Manual Flow Control MCPR Limits The Governing MCPR Operating Limit while in Manual Flow Control is either determined from 2.2.1.1 or 2.2.1.2, whichever is greater at any given power, flow condition.
2.2.1.1 Power-Dependent MCPR (MCPRp)*
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).
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).
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 L2C9.
- For thermal limit monitoring at greater than 100%P, the 100% power MCPRp limits should be applied.
LaSalle Unit 2 Cycle 9 2-1 August 2002
Technical Requirements Manual -Appendix J L2C9 Core Operating Limits Report 2.2.3 Nominal Scram Speeds Nominal Scram Speeds (NSS) ire not supported for L2C9.
LaSalle Unit 2 Cycle 9 2-2 August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Table 2-1 MCPRp for GE Fuel (References 2, 3, 51, 56, and 57)
Percent Core Thermal Power' EOOS Combination 1l0 I
25 No EOOS with TSSS (BOC to Last 2.70 2.20 2.01 1.53 Sequence Exchange 2)_
Single RR Loop only with TSSS (BOC to 2.71 2.21 2.02 1 54 Last Sequence Exchange 2)
EOOS4 with TSSS (BOC to Last 2.85 2 35 2.24 Sequence Exchange2 )
EOOS/ISingle RR Loop with TSSS (BOC 2 86 2.36 2.25 to Last Sequence Exchange 2)--.
No EOOS with TSSS (Last Sequence 2.70 2.20 201 1.53 Exchange 2 to Coastdown 3)
Single RR Loop only with TSSS (Last 2.71 2.21 2 02 1.54 Sequence Exchange 2 to Coastdown 3)
EOOS' with TSSS (Last Sequence 2.85 2.35 2 24 Exchange2 to Coastdown3)
EOOS'/Single RR Loop with TSSS (Last 2 86 2.36 2.25 Sequence Exchange 2 to Coastdown 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 (Coastdown 3 )
Feedwater Heaters OOS with TSSS 274 2.24 2.24 1.57 (Coastdown 3 )
Feedwater Heaters OOSISingle RR Loop 2.75 2.25 2.25 1.58 with TSSS (Coastdown 3) 1 Feedwater Heaters OOS/Turbine By ?ass 2.74 2.24 Z24 1.64 Valves OOS with TSSS (Coastdown )
I I
Feedwater Heaters OOSlTurbine Bypass Valves OOS/Single RR Loop with TSSS (Coastdown31 2.75 2.25 2.25 TCV Slow Closure/EOC RPT OOSI 2.74 2.24 2 24 Feedwater Heaters OOS with TSSS (Coastdown3) 1 TCV Slow Closure/EOC RPT OOSI 2 75 2.25 2.25 Feedwater Heaters OOS/Single RR Loop with TSSS (Coastdown3) 100 1.51 1.52 1.63 1.64 1.51 1.52 1 63 1.64 1.52 1.53 1 52 1 53 1.53 1.54 1.73 1.74
'Values are interpolated between relevant power levels. For operation at exactly 25% or 80% CTP, the more 2,limiting value is used. 3489 MWt is rated power.
Last Sequence Exchange' is defined as the Al to A2 sequence exchange that occurs at approximately 15,600 MWd/MTU cycle exposure.
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 cycle exposures beyond 31,242.7 MWd/MTU.
Allowable EOOS conditions are listed in Section 5.
LaSalle Unit 2 Cycle 9 25 160 2-3 August2002 180 180 1
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Table 2-2 MCPRp for Siemens Fuel (References 2, 3, 21, 51, 53, 54, 55, 56. and 57)
For all Siemens fuel EXCEPT Fuel Type 18 in 10B cell locations and Fuel Types 16, 17, and 18 in Al (7A, 7B, 7C, 8A, and 8B) cell locations Percent Core Thternmal Power' EOOS Combination 0
25 25 60 80 80, 100 No EOOS with TSSS (BOC to Last 2.70 2.20 1.93 1.48 1.41 Sequence Exchange2 )
W_
____1 Single RR Loop only with TSSS (BOC to 2.71 2.21 1.94 1.49 1.42 Last Sequence Exchange?)
EOOS' with TSSS (BOC to Last 2.85 2.35 2.17 1.7 1
1.53 Sequence Exchange 2 )
E00SIngle RR Loop with TSSS (BOC 2.86 2.36 2.18 1.54 to Last Sequence Exchange2)
No EOOS with TSSS (Last Sequence 2.70 2.20 1.93 1.48 1.41 Exchange2 to Coastdown_)_
Single RR Loop only with TSSS (Last S*.ouenc*. F~xchanae 2 to Coastdown3) 2.71 2.21 1.94 1.49 EOOS' with TSSS (Last Sequence 2.85 2.35 2.17 Exchange? to Coastdown 3)
EOOSxSangl e RR Loop with TSSS ast 2.86
-2.36 2.18 Sequence Exchanoe2 to Coastdown3 )
No EOOS with TSSS (Coastdowni) 2.70 2.20 1.93
=
.
Single RR Loop only with TSSS 2.71 2.21 1.94 1.49 (Coastdown
- 3)
Feedwater Heaters OOS with TSSS 2.70 2.20 2.17 1.54 (Coastdown3)
I Feedwater Heaters OOS/Single RR Loop 2.71 2.21 2.18 1.55 with TSSS (Coastdown3)
Feedwater Heaters OOS/Turbine Bypass 2.70 2.20 2.17 1.60 Valves OOS with TSSS (Coastdown*)
I__
Feedwater Heaters OOSfTurbine Bypass Valves OOS/Single RR Loop with TSSS frCan*tdown 3l,
2.71 2.21 2.18 "TCV Slow Closure/EOC RPT OOSI 2.70 2.20 2.17 Feedwater Heaters OOS with TSSS -..
(Coastdown3)
2.71 2.21 2.18 Feedwater Heaters 00S/Single RR Loop I
with TSSS (Coastdown 3 )
1 1
1.42 1.53 1.54 1.44 1.45 1 44 1.45" 1.46 1.47 1.60 1.61 Table continues on next page.
LaSalle Unit 2 Cycle 9 I
I 2-4 August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Table 2-2 (Continued)
MCPRpfor Siemens Fuel For ONLY Siemens Fuel Type 18 in 10B cell locations Percent Core Thermal Power1 EOOS Combination 0
25 25 60 80 80 100 No ECOS with TSSS (BOC to Last 2.74 2.24 1.97 1.52 1.45 Sequence Exchange 2)
Single RR Loop only with TSSS (BOO to 2.75 2 25 1 98 1 53 1.46 Last Sequence Exchange 2) 1 a
n EOOS'with TSSS (BOC to Last Sequence 2.89 2.39 2.21 1.74 1.66 1.57 Exchange 2)_____
-UUS /Single RR Loop witn i S (OUu to Last Sequence Exchance 21 2.4U No EOOS with TSSS (Last Sequence 2.70 2 20 1.93 1.48 Exchange 2 to Coastdown 3 )
Single RR Loop only with TSSS (Last 2.71 2.21 1.94 1 49 Sequence Exchange 2 to Coastdown3)
EOOS' with TSSS (Last Sequence 2.85 2.35 2 17 Exchange 2 to Coastdown3)
E EOOS'lSingle RR Loop with TSSS (Last 2.86 2 36 2.18 Sequence Exchange2 to Coastdown 1 No EOOS with TSSS (Coastdown")
2.70 2 20 1 93 1 48 Single RR Loop only with TSSS 2.71 2.21 1.94 1 49 (Coastdown 3)
Feedwater Heaters OOS with TSSS 2.70 2.20 2.17 1.54 (Coastdown 3)
Feedwater Heaters OOS/Single RR Loop 2.71 2.21 2 18 1.55 with TSSS (Coastdown3 )
Feedwater Heaters OOS/Turbine Byass 2.70 220 2.17 1.60 Valves OOS with TSSS (Coastdown )
I I
Feedwater Heaters OOS/Turbine Bypass Valves DOS/Single RR Loop with TSSS (Coastdown3 )
2.71 2.21 2.18 I
I TCV Slow Closure/EOC RPT COS/
2.70 2.20 2.17 Feedwater Heaters OOS with TSSS (Coastdown3 )
1 TCV Slow Closure/EOC RPT OOS/
2.71 2.21 2.18 Feedwater Heaters DOS/Single RR Loop with TSSS (Coastdown 3 )
1 1
1.54 1 44 1.45 1.44 1.45 1.46 1.47 1 60 1.61 Table continues on next page.
LaSalle Unit 2 Cycle 9 1.58 S1.41 1142 1 53 I
2.90 2.22 I
2-5 August 2002 1.75 1.67
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Table 2-2 (Continued)
MCPRp for Siemens Fuel For ONLY Siemens Fuel Type 16 and 17 in Al (7A, 7C, and 8B) cell locations Percent Core Thermal Power' EOOS Combination 0
25 25 60 80 80 100 No EOOS with TSSS (BOC to Last 2.73 2.23 1.96 1.51 1.44 Sequence Exchange 2)y-M O
M Single RR Loop only with TSSS (BOC to 1 2.74 2 24 1.97 1.52 1.45 Last Sequence Exchange 2)_1 EOOS* with TSSS (BOC to Last 2.88 2.38 2.20 1
1 1.56 Sequence Exchange2 )
EOOS'/Single RR Loop with TSSS (BOC 2.89 2.39 2.21 1.57 to Last Sequence Exchange 2)
No EOOS with TSSS (Last Sequence 270 2.20 1.93 1.48 1.41 Exchange 2 to Coastdown3)
EmI_-_I Single RR Loop only with TSSS (Last 2.71 2.21 1.94 1.49 1.42 Sequence Exchange 2 to Coastdown3) )_
EOOS4 with TSSS (Last Sequence 2.85 2.35 2.17 1.70 1.62 1.53 Exchange to Coastdown3)......
EOOS'/Single RR Loop with TSSS q.ast Sequence Exchange2 to Coasldown) 2.66 2.36 2.18 I1.63 No EOOS with TSSS (Coastdown")
2.70 2.20 1.93 1 48 Single RR Loop only with 'SSS 2.71 2.21 1.94 1.49 (Coastdown 3)
Feedwater Heaters QOS with TSSS 2.70 2.20 2.17 1.54
( Coastdown3)
Feedwater Heaters OOSISingle RR Loop 2.71 2.21 2.18 1.55 with TSSS (Coastdown
- 3) -
I I
Feedwater Heaters OOS/Turbine By'ass 270 2.20 2.17 1.60 Valves 0OS with TSSS (CoastdownT)
I, Feedwater Heaters OOS/Turbine Bypass Valves OOS/Single RR Loop with TSSS (Coastdown 3) 2.71 2.21 2.18 TCV Slow Closure/EOC RPT 0OS/
2.70 2.20 2.17 Feedwater Heaters 0OS with TSSS (Coastdown 3) )
"TCV Slow ClosureIEOC RPT 0OS/
2.71 2.21 2.18 Feedwater Heaters OOSSingle RR Loop with TSSS (Coastdown3) 1.54 1.44 1.45 1.44 1.45 1.46 1.47 1.60 1.71 1.63 1.61 Table continues on next page.
LaSalle Unit 2 Cycle 9 S~1.711 I
2-6 August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Table 2-2 (Continued)
MCPRp for Siemens Fuel For ONLY Siemens Fuel Type 18 in Al (7A, 7B, and 8A) cell locations Percent Core Thermal Power' EOOS Combination 0
25 No EOOS with TSSS (BOC to Last 2 73 2.23 1 96 1.51 Sequence Exchange 2)
Single RR Loop only with TSSS (BOC to 274 2.24 1 97 1.52 Last Sequence Exchange 2 )
EOOS" with TSSS (BOC to Last 2.88 2 38 2 20 Sequence Exchange
- 2)
EOOS 4/Single RR Loop with TSSS (BOC 2 89 2 39 221 to Last Sequence Exchange)
No EOOS with TSSS (Last Sequence 2.72 2.22 1.95 1.50 Exchange 2 to Coastdown 3)
Single RR Loop only with TSSS (Last 2.73 2.23 1.96 1.51 Sequence Exchange 2 to Coastdown
- 3)
EOOS' with TSSS (Last Sequence 287 2 37 2.19 Exchange2 to Coastdown 3)
_j EOOS4 /Single RR Loop with TSSS (Last 2.88 2 38 2.20 Sequence Exchange 2 to Coastdown No EOOS with TSSS (Coastdown')
2.72 2.22 1.95 1.50 Single RR Loop only with TSSS 2.73 2.23 1.96 1.51 (Coastdown 3)
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 with TSSS (Coastdown3)
Feedwater Heaters OOS/Turbine Byýass 2.72 2 22 2 19 1 62 Valves OOS with TSSS (Coastdown )
I Feedwater Heaters OOS/rurbine Bypass Valves OOS/Sngle RR Loop with TSSS 2.73 2.23 2 20 80 80 100 1 44 1.45 1.73 1.65 1 1.56 1.57 1.43 1 44 1.55 I
TCV Slow Closure/EOC RPT OOSI 2.72 2.22 2.19 Feedwater Heaters OOS with TSSS (Coastdown3)
2.73 2.23 2 20 Feedwater Heaters OOS/Single RR Loop with TSSS (Coastdown 3) 1 56 1 46 1.47 1.46 1 47 1.48 1.49 1 62 1 Values are interpolated between relevant power levels. For operation at exactly 25% or 80% CTP, the more 2limiting value is used. 3489 MWt is rated power.
'Last Sequence Exchange' is defined as the Al to A2 sequence exchange that occurs at approximately 15,600 MWd/MTU cycle exposure.
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 cycle exposures beyond 31,242.7 MWd/MTU.
Allowable EOOS conditions are listed in Section 5.
LaSalle Unit 2 Cycle 9 1.73 1 165 1 163 25 1 60 2-7 August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Table 2-3 MCPRF for GE and Siemens Fuel (Reference 3)
MCPRF limits for 105% Maximum Attainable Core Flow Flow (% rated)
MCPRF ATRIUM-9B MCPRE GE9 0
1.60 1.66 30 1.60 1.66 105 1.11 1.11 The MCPRF limits are applicable from BOC through coastdown and in all EOOS scenarios.
LaSalle Unit 2 Cycle 9 2-8 August 2002.
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report
- 3.
Linear Heat Generation Rate (3.2.3) 3.1 Tech Spec
Reference:
3.2
==
Description:==
For operation without a full TIP set from BOC to 500 MWd/MT a penalty of 11 01% must be applied to all LHGR limits.
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 and 19)
- 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)
Nodal Exposure (GWd/MT)
LHGR Limit (KW/ft) 0.00 14.25 12.14 14.25 2619 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) (Reference 3)
SPC LHGRFAC multipliers for Coastdown are applicable for core average exposures between 30,266.2 MWd/MTU and 31,242.7 MWd/MTU (Reference 57).
For All Siemens Fuel EXCEPT Fuel Type 18 in Al (7A, 7B, and 8A) cell locations (Reference 3)
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.
LaSalle Unit 2 Cycle 9
_______________________[_________
3-1 August2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report For ONLY Siemens Fuel Type 18 in Al (7A, 78, and 8A) cell locations Prior to the Last Sequence Exchange at -15,600 MWd/MTU cycle exposure (Reference 3) ° Planar Average Exposure (GWd/MTU)
LHGR limit (kW/ft) 0.0
.14.4 15.0 14.4 61.1 8.32 For ONLY Siemens Fuel Type 18 in Al (7A, 7B, and 8A) cell locations Following the Last Sequence Exchange at -15,600 MWdIMTU cycle exposure (Reference 56)
Planar Average Exposure (GWd/MTU)
LHGR limit (kWlft) 0.0 14.2 15.0
-14.2 61.1 8.12 LaSalle Unit 2 Cycle 9 August 2002 3-2
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Figure 3.2-1 Power-Dependent LHGR Multipliers for GE Fuel (Formerly MAPFACp)
(References 8 and 19) 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Core Thermal Power (% Rated)
August 2002 LaSa Jnit 2 Cycle 9 1
0.95 0.9 0.85 0.8 0.75 0.7 0.65 0.6 0.55 0.5 0.45 0.4 0.35 0.3 0.25 0.2 0.15 0.1 0.05 0
L.L CL
-J 0o
- a.
°D I
0.9
_1 "a) 0.
(
Technical Requirements Manual - Appendix J L2C9 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) 100 August 2002 LaSalle Unit 2 Cycle 9 C
- 0.
C.)
LL CL 0
C.
°,
4 C
C=
o.
1 0.95 0.9 0.85 0.8 0.75 0.7 0.65 0.6 0.55 0.5 0.45 0.4 0.35 0.3 0.25 0.2 0.15 0.1 0.05 0
0 10 20 30 40 50 60 70 80 90 Core Thermal Power (% Rated) 3-4
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Figure 3.2-3 Flow-Dependent LHGR Multiplier for GE Fuel (formerly MAPFAC F)
(References 8, 13, 18, and 19) 1
-" 0.9 U.,
- a.
0~
_>,08 C
0.7
- 0.
t 0.6
-1.
CL
- 0 E 0.5 S0.4 0.3 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 Core Flow (% Rated)
LaSalht t 2 Cycle 9 I
i F
I i
ii
'I II l
ii F
ji
- I' I
I Ii 1
i F
I I,i
' i I;
I i
I [
F,
i, j
I I:'
',For 105% Maximum Attainable Core Flow LHGRFACf = The Minimum of EITHER 10 OR {0.6807 x (WT/100) + 0 46721 SWT= % Rated Core Flow For Abnormal Idle Loop Startup, LHGRFACf =
0.40 FI F
F A
I*
P I
I F
F F
I F
'F K,
I I
F August 2002 3-*'
Technical Requirements Manual'- Appendix J L2C9 Core Operating Limits Report Table 3-1 LHGRFACp for Siemens Fuel (References 3, 51, 54, and 57)
EOOS Combination l0 Percent Core Thermal 125 No EOOS with TSSS (BOC to Last 0.77 0.77 Sequence Exchange 2 )
Single RR Loop only with TSSS (BOG to 0.77 0.77 Last Sequence Exchange2)
EOOS' with TSSS (BOG to Last 067 067 Sequence Exchange 2)
EOOS'lSingle RR Loop with TSSS (BOC 0.67 0.67 to Last Seqeunce Exchange2)
No EOOS with TSSS (Last Sequence 0.77 0.77 Exchange to Coastdown3)
Single RR Loop only with TSSS (Last 0.77 0.77 Sequence Exchange 2 to Coastdown 3)
EOOS' with TSSS (Last Sequence 067 0.67 Exchange 2 to Coastdown3)
EOOS'/Single RR Loop with TSSS (Last 0.67 067 Sequence Exchange2 to Coastdown)-t No EOOS with TSSS (Coastdown')
077 0.77 Single RR Loop only with TSSS 0.77 077 (Coastdown 3 )
Feedwater Heaters OOS with TSSS 0.68 0 68 (Coastdown
- 3)
Feedwater Heaters OOSISingle RR Loop 0.68 068 with TSSS (Coastdown3 )
Feedwater Heaters OOS/Turbine Bypass 0.68 0.68 Valves OOS with TSSS (Coastdown )
Feedwater Heaters OOS/Turbine Bypass 068 0.68 Valves OOS/Single RR Loop with TSSS (L
Coastdown 3)
TCV Slow Closure/EOC RPT OOSI Feedwater Heaters OOS with TSSS fr*'r:trnwn 3
0 67 0.67 "TCV Slow Closure/EOC RPT 0OS 0.67 0.67 0.79 Feedwater Heaters OOSISingle RR Loop with TSSS (Coastdown3)
I
'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,Last Sequence Exchange' is defined as the Al to A2 sequence exchange that occurs at approximately 15,600 MWd/MTU cycle exposure.
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 cycle exposures beyond 31,242.7 MWd/MTU.
Allowable EOOS conditions are listed in Section 5.
LaSalle Unit 2 Cycle 9
.1 August 2002 3-6
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report Table 3-2 LHGRFACF for Siemens Fuel (Reference 3)
Values Applicable for up to 105% Maximum Attainable Core Flow Flow (% rated)
LHGRFACF ATRIUM-9B 0
0.69 30 0.69 76 1.00 105 1.00 These LHGRFACf multipliers apply from BOC through coastdown and in all EOOS scenarios.
LaSalle Unit 2 Cycle 9 3-7 August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report
- 4.
Control Rod Withdrawal Block Instrumentation (3.3.2.1) 4.1 Tech Spec
Reference:
Tech Spec Table 3.3.2.1-1.
4.2
==
Description:==
The Rod Block Monitor Upscale Instrumentation Setpoints are determined from the relationships shown below:
ROD BLOCK MONITOR UPSCALE TRIP FUNCTION Two Recirculation Loop Operation*
Single Recirculation Loop Operation*
TRIP SETPOINT 0.66 W + 51%**
0.66 W + 45.7%**
ALLOWABLE VALUE 0.66 W + 54%**
0.66 W + 48.7%**
This setpoint may be lower/higher and will still comply with the RWE Analysis, because RWE-is analyzed unblocked.
Clamped, with an allowable value not to exceed the allowable value for recirculation loop flow (W) of 100%.
LaSalle Unit 2 Cycle 9 August 2002 4-1
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report
- 5.
Allowed Modes of Operation (B 3.2.2, B 3.2.3)
The Allowed Modes of Operation with combinations of Equipment Out-of-Service are as described below:
OPERATING REGION Equipment Out of Service Options' Standard MELLLA ICF7 Coastdown9 None Yes Yes Yes Yes Feedwater Heaters 2 (Reference 8)
Yes NO 3 Yes Yes Single RR Loop10 (Reference 8)
Yes No8 N/A Yes 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 Tnp (Referencell)
Yes Yes Yes Yes TCV Slow Closure/EOC Recirculation Pump Trip I Yes No3 Yes Yes Feedwater Heaters2 (References 11, 16. and 17)
Turbine Bypass Valves I Feedwater Heaters2.5 (Reference 8)
NO12 No12 No12 Yes EOC Recirculation Pump Trip /
Yes 4 No3 Yes4 Yes Feedwater Heaters 2 (Reference 8)
TCV Stuck Closed6 (Reference 12)
Yes Yes Yes No
- 1. Each EOOS condition may be combined with one SRV OOS, up to two TIP Machines OOS or the equivalent number of TIP channels (100% available at startup from a refuel outage), a 20°F reduction in feedwater temperature (without Feedwater Heaters considered OOS), cycle startup with uncalibrated LPRMs (BOC to 500 MWd/MTU), and/or up to 50% of the LPRMs out of service.
- 2.
Up to 100OF Reduction in Feedwater Temperature Allowed with Feedwater Heaters Out-of-Service.
Feedwater Heaters COS may be an actual 0OS condition, or an intentionally entered mode of operation to extend the cycle energy.
- 3.
If operating with Feedwater Heaters Out-of-Service, operation in MELLLA is supported by current transient analyses, but administratively prohibited due to core stability concerns.
- 4.
EOC Recirculation Pump Trip OOS/Feedwater Heaters OOS is allowed using the TCV Slow Closure/EOC Recirculation Pump Trip OOS/Feedwater Heaters OOS operating limits.
- 5.
Only when operating in coastdown, otherwise this combination is not allowed.
- 6.
Operation prior to coastdown is only allowed when less than 10.5 million Ibm/hr steam flow and when average position of 3 open TCVs is less than 50% open, with FCL <103%, and the MCFL setpoint >
120%. TCV Stuck Closed may be in combination with any EOOS except TBVOOS or TCV Slow Closure.
If in combination with other EOOS(s), thermal limits may require adjustment for the other EOOS(s) as designated in Sections 1, 2, and 3.
- 7. ICF is analyzed for up to 105% core flow.
- 8. The SLO boundary was not moved up with the incorporation of MELLLA. The flow boundary for SLO at uprated conditions remains the ELLLA boundary for pre-uprate conditions. (Reference 20)
- 9. Coastdown is defined to begin at a core average exposure of 30,266.2 MWd/MTU (which is the licensing basis exposure used by SPC). ICF is allowed during coastdown. (Reference 3 and 57)
- 10. Single loop operation is allowed with any of the EOOS options listed in this table.
- 11. Turbine Bypass Valves OOS is allowed during coastdown operation using the Feedwater Heaters OOS/Turbine Bypass OOS operating limits.
- 12. Operation in these regions is permitted during coastdown only.
LaSalle Unit 2 Cycle 9 5-1 August 2002
Technical Requirements Manual - Appendix J L2C9 Core Operating Limits Report
- 6.
Traversing In-Core Probe System (3.2.1, 3.2.2, 3.2.3)
\\_>
6.1 Tech Spec Reference Tech Spec Sections 3.2.1, 3.2.2, 3.2 3 for thermal limits require the TIP system for recalibration of the LPRM detectors and monitoring thermal limits.
6.2 Description When the traversing in-core probe (TIP) system (for the required measurement locations) is used for recalibration of the LPRM detectors and monitoring thermal limits, the TIP system shall be operable with the following:
- 1.
movable detectors, drives and readout equipment to map the core in the required measurement locations, and
- 2.
indexing equipment to allow all required detectors to be calibrated in a common location.
For BOC to BOC + 500 MWD/MT, cycle analyses support thermal limit monitoring without the use of the TIPs.
Following the first TIP set (required prior to BOC + 500 MWD/MT), the following applies for use of the SUBTIP methodology:
With one or more TIP measurement locations inoperable, the TIP -data for an inoperable measurement location may be replaced by data obtained from a 3-dimensional BWR core monitoring software system adjusted using the previously calculated uncertainties, provided the following conditions are met:
- 1.
All TIP traces have previously been obtained at least once in the current operating cycle when the reactor core was operating above 20% power, (References 14, 15 and 23) and
- 2.
The total number of simulated channels (measurement locations) does not exceed 42% (18 channels).
Otherwise, with the TIP system inoperable, suspend use of the system for the above applicable monitoring or calibration functions.
6.3 Bases The operability of the TIP system with the above specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core. The normalization of the required detectors is performed internal to the core monitoring software system.
Substitute TIP data, if needed, is 3-dimensional BWR core monitoring software calculated data which is adjusted based on axial and radial factors calculated from previous TIP sets. Since uncertainty could be introduced by the simulation and adjustment process, a maximum of 18 channels may be simulated to ensure that the uncertainties assumed in the substitution process methodology remain valid.
LaSalle Unit 2 Cycle 9 6-1 August 2002
Technical Requirements Manual - Appendix J Section 2 LaSalle Unit 2 Cycle 9 Reload Transient Analysis Results August 2002
Technical Requireme'nts Manual -:Appendix J L2C9 Reload Transient Analysis Results
Table of Contents Attachment Preparer ComEd Siemens Power Corporation Siemens Power Corporation General Electric General Electric Framatome ANP Framatome ANP Framatome ANP Document Neutronics Licensing Report Reload Analysis Report Plant Transient Analysis ARTS Improvement Program Analysis, Supplement I (Excerpts)
TCV Slow Closure Analysis (Excerpts)
LaSalle Unit 2 Cycle 9 Operating Limits for Proposed ITS Scram Times and Corrected Fuel Thermal Conductivity LaSalle Unit 2 Cycle 9 Equipment Out-of-Service Operating Limits Using Nominal Scram Speed and Exposure Limited to 14,000 MWdlMTU LaSalle Unit 2 Cycle 9 Operating Limits for Cycle Extension to 19,300 MWd/MTU LaSalle Unit 2 Cycle 9 I
2 3
4 5
6 7
8 August 2002
Technical Requirements Manual - Appendix J L2C9 Reload Transient Analysis Results LaSalle Unit 2 Cycle 9 Neutronics Licensing Report LaSalle Unit 2 Cycle 9
___j August 2002
I Do G0 001300 NUCLEAR FUEL MANAGEMENT TRANSMITTAL OF DESIGN INFORMATION r-SAFETY RELATED Oiginating Organization NFM ID#
NFMOOOOI IS EO NON-SAFETY RELATED ED Nuclear Fuel Management Sequence 0
o REGULATORY RELATED 0 Other (specify)
PaeI of 21 Station:
LaSalle Unit 2
Cycle:
9 Generic:
To: Jeffery K. Nugent (IS)
Subject:
LaSalle Unit 2 Cycle 9 Neutoics Licensing Report Ming-Yuan Hsiao 2!:aSy 4f
+/-4"iZ4
__________0 Pparer
's Situr Date Peter A. Werrenian
-.0 Reviewer Reyiewe'i M
Datm Adelmo S. pallotta NFM Department Head Approver's Signatur Date Status of Information:
E0 Verified o Unverified o Engineering Judgement Action Tracking # for Method and Schedule of Verification for Unverified DESIGN INFORMATION:
Description of Information: Provide the station and BSS group LaSalle Unit 2 Cycle 9 Neutronics Licensing Report (NLR).
Purpose of Information:
Seq. 0: Provide the'station and BSS group LaSalle Unit 2 Cycle 9 Neutronics Licensing Report (NLR).
Source of Information: As rfcrenced Supplemental Distribution:
Danny Bost (LS)
John i. Reimer (LS)
Amy Goss (.IS)
Edward A-McVey Thomas J. Rauselh R. W. Tsai Adelmo S. Pallotta Ming Y. Hsiao LaSalle Central File Downers Grove Central File
NUCLEAR FUEL MANAGEMENT
'NFM ID#
NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 1Page 2 of 21 COMMONWEALTH EDISON COMPANY NUCLEAR FUEL SERVICES NEUTRONICS LICENSING REPORT for LaSalle Unit 2 Cycle 9 preparer: "'Y//
8 1-0oo reviewer pq(j
NUCLEAR FUEL MANAGEMENT NFM ID#
NFMOOOO115 TRANSMITTAL OF DESIGN INFORMATION
-Seq.
No.
0 I
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 I.
Nuclear Design Analysis 1.1 Fuel Bundle Nuclear Design Analysis 1.2 Core Nuclear Design Analysis 1.2.1 Core Configuration and Licensing Exposure Limits 1.2.2 Core Reactivity Characteristics II.
Control Rod Withdrawal Error II.
Fuel Loading Error M.T.1 Fuel Mislocation Error 11M.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.l Fresh Fuel Vault Criticality Compliance VII.2 Li Spent Fuel Pool Criticality Compliance VII.3 12 Spent ýFuel Pool Criticality Compliance VMI.
References preparer:
--,n/14, 6-:3/-oo reviewer QAw
.3t-4
NUCLEAR FUEL MANAGEMENT NFM ID#
NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Page 4 of 21
- 1.
Nuclear Design Analysis 1.1 Fuel Bundle Nuclear Design Analysis 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-I2GZ-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-I IG8.0-100M Figure 3 SPCA9-4.56L-12G8.0-100M Figure 4 SPCA94.21L-13G8.0-IOOM 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-l00M Figure 9 preparr:
-r"yl,
&-31-voi reviewer Q7iJ*
3I
NUCLEAR FUEL MANAGEMENT NFM ID#
NFMOOOO1 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Page 5 of 21 1.2 Core Nuclear Desien Analysis 1.2.1 Core Configuration and Licensing Exposure Limits Bundle Type a,
a, GE9B-P8CWB322-1IGZ-100M-150-CECO GE9B-P8CWB320-9GZ-100M-150-CECO SPCA9-381B-13GZ7-80M SPCA9-384B-11GZ6-80M SPCA9-391B-14G8.0-100M SPCA9-41OB-19G8.0-100M SPCA9-383B-16G8.0-100M SPCA9-396B-12GZ-IOOM Cycle Loaded 7
7 8
8
-9 9
9 9
Number in Core 84 76 128 128 40 120 132 56 Licensing Exposure Limits Core Cycle Value of Interest Average Incremental Exposure Exposure (M`WD/Mr)
(MWD/MT)
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 Exposure 11799 0
(assuming nominal EOC 8 energy)
BOC 9 Exposure 11470 0
(assuming short EOC 8 energy)
Nominal EOC 9 Exposure 29598 17800 (assuming nominal EOC 8 energy)
Core UO Weights Cycle of Interest U0 2 Total Weight (MT)
Cycle 8 135.11 Cycle 9 133.50 preparer: ?tWyl4, Y-i-,I reviewer
-P
NUCLEAR FUEL MANAGEMENTI NFM ID#
NFMD0001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 1Pagie 6 of 21 1.2.2 Core Reactivity Characteristics All values reported below are with zero xenon and are for 68TF moderator temperature. The MICROBURN-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/MTU) 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: ?"'16;, '-I5-oo reviewerp ftw, q. 6. v'
NUCLEAR FUEL MANAGEMENT
.NFM ID#
NFMOOOO1 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Page 7 of 21
",, I II.
Control Rod Withdrawal Error The control rod withdrawal error event'is analyzed at 100%' of rated power, 100% of rated flow and unblocked conditions only...
Distance Withdrawn (ft) 12 (Unblocked)
ACPR 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 limits. The design complies with the GE centerline melt criteria via conformance to the GE thermal overpower protection (TOP) criteria. The design complies with the GE 1% plastic strain criteria via conformance to the GE mechanical overpower protection (MOP) criteria..
IH.
Fuel Loading Error The Fuel Loading Error, including fuel mislocation and mnisorientation, is classified as an accident. By demonstrating that the Fuel Loading Error meets the more stringent Anticipated Operational Occurrence (AOO) requirements, the offsite dose requirement is assured to be met.
Because the events listed below result in a ACPR value that is less than that of the limiting transient, the AOO requirements and hence off-site dose requirements are met for the Fuel Loading Error.
1MA.1 Fuel Mislocation Error The following value bounds both the SPC and the co-resident GE fuel types.
Event Mislocated Bundle ACPR 0.23' M1.2 Fuel Misrotation Error The following value bounds both the SPC and the co-resident GE fuel types.
Event Misoriented Bundle ACPR 0.15 preparer: *y/H, 7"1-oO rei w rP U5' 1 0 1 o
NUCLEAR FUEL MANAGEMENT NFM ID#
NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 1Page 8 of 21 IV.
Control Rod Drop Accident LaSalle is a banked position withdrawal sequence 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 cal/g 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 Doppler Coefficient, Ak/kPF Effective Delayed Neutron Fraction used Four-Bundle Local Peaking Factor Maximum Deposited Fuel Rod Enthalpy, (cal/g)
Number of Rods Greater than 170 callg 1.375
-9.50E-06 0.0053 1.281 222 266 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 ofFeedwater HeatinE 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 I % plastic strain and centerline melt criteria via conformance to the PAPT (Protection Against Power Transient) LHGR limits. The design complies with the GE preparer: MrY/H, /0-f--0Z>
reviewer 9 -W 1 -f -0o0
NUCLEAR FUEL MANAGEMENT NFM ID#
NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Poae 9 of 21 1% plastic strain criteria via conformance' to the mechanical overpower protection (MOP) limit.
The design does not meet the GE thermal overpower protection (TOP) criteria during a loss of feedwater heating event; hence, the LHGR values in the COLR for the affected lattice are adjusted accordingly (References 9, 13 and 14) as follows: "
GE9B-P8CWB322-1IGZ-100M-150-CECO Bundle (Fuel Type 1)
LHGR Limits for L2C9 Nodal Exposure Nodal Exposure LHGR Limit (GWD/ST)
(GWD/MT) 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-100M-150-CECO Bundle (Fuel Type 2)
LHGR Limits for L2C9 Nodal Exposure Nodal Exposure LHGR Limit (GWD/ST)
(GWD/MT) 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 Note that the following exposures are based on a nominal Cycle 8 EOC exposure of 13750 MWD/MT and a nominalCycle 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 ATRIUM-9B ATRIUM-9B
..Exiosure Pr6j&ted Limit--
Projected Limit*
Pak Batch
-.- 39989.42000-36794
____NA__
Peak Assembly
-,45399
-NA 39460 48000 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 GWD/MT for Peak Fuel Assembly (no change), 50.0 GWD/MT for Peak Fuel Rod and 60.0 GWD/MT for Peak Fuel Pellet.
preparer: "71//-J, 7n
,rei reviewer
NUCLEAR FUEL MANAGEMENT NFM ID#
NFM00001 15 TRANSMITTIAL OF DESIGN INFORMATION Seq. No.
0 Poae 10 of 21 VII. 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.
VII.1 Fresh Fuel Vault Criticality Compliance The fuel storage vault criticality analysis that is detailed in Reference 5 remains valid for the above lattices. All the new (ATRIUM-9B) assemblies comply with the fresh fuel vault criticality limits, i.e., all lattices have an enrichment of less than 5.00 wt % U-235 and a gadolinia content that is greater than 6 rods at 3.0 wt% Gd20 3.
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. Administrative controls as generally defined in GE SIL No. 152 (dated March 31,1976) must be incorporated for the area.
VIL.2 L. Spent Fuel Pool Criticality Compliance The LaSalle Unit 1 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% Gd20 3.
VII.3 L2 Spent Fuel Pool Criticality Compliance The LaSalle Unit 2 spent fuel pool criticality analysis that is detailed in Reference 7 remains valid for the above lattices. As shown below, all the new (ATRIUM-9B) assemblies comply with the LaSalle Unit 2 spent fuel pool criticality limit of k-eff < 0.95.
Lattice Type 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 *F, uncontrolled CASMO-3G results.
- From Figure 6.1 of Reference 7.
preparer: 7" Y t'i, r-e wr-o reviewer otjJ,.,-..
. 0
NUCLEAR FUEL MANAGEMENT NFM ID#
NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0
-Poqe 11of21 VIII. References
- 1. "LaSalle Unit 2 Cycle 9 Reload Analysis". Siemens Power Corporation, 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," ND1T NFM0000056 Seq. 1, April 7, 2000 and "L2C9 FLLP,"
BNDL:00-005, Revision 0, 4/7/2000.
- 4. 'Commonwealth Edison, Nuclear Fuel Services, NFSR-0091, "Benchmark of CASMOJMICROBURN "BWR Nuclear Design Methods", as supplemented and approved.
- 5.
"Criticality Safety Analysis for ATRIUM-9B Fuel, LaSalle"Units 1 and 2 New Fuel Storage Vault,"
"Siemens Power Corporation, EMF-95-134(P), December 1995. [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), April 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, Revision 0, July 13, 2000.
- 11. "L2C9 Rod Withdrawal Error MOP/TOP Analysis," BNDL:00-023, Revision 0, August 17, 2000.
- 12. "LaSAlle Unit 2 Cycle 9 Neutronic-Licensing Shutdown Margin Calculation," BNDL:00-032, Revision 0, August 17, 2000.
- 13. "LaSalle 2 Cycle 9 LFWH TOP Violation and LHGR Limit Calculation," Letter NFM:BND:00-050, Julyi13,'2000.
- 14. "LaSalle 2 Cycle 9 GE9 Curve Adjustment for LFWH TOP Violation," GE Letter KF-00-063, August 24,2000.
- 15. "'LaSalle 2 Cycle 9LFWH TOP Violation and LHGR Limit Calculation," Letter NFM:BND:00-050, July 13, 2000.
- 16. "L2C9 Mislocation Licensing nalysis," BNDL:00-025,_Septermber 2000.'"
17." "L2C9 Bundle Misorientation A"alysis," BNDL:00-030, September 2000.
"preparer:
9 -H x31-oo reviewer POW 3
-ID
NUCLEAR FUEL MANAGEMENT NFM ID#
NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Page 12 of 21 Table 1 L2C9 Simplified Shutdown Sequence 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.
1 48-00 Group 2 must be fully inserted prior to inserting any Group I rod.
Shutdown from an A2 Sequence Insertion Rod Group*
(Bank)
Comments**
9 or 10 48-00 Either Group 9 or 10 may be inserted first.
Groups 9 and 10 must be fully inserted prior to inserting any 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 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.
1 48-00 Group 2 must be fully inserted prior to inserting any Group 1 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.?
revi 48-00 ewer PAO 8
NUCLEAR FUEL MANAGEMENT TRANSMITTAL OF DESIGN INFORMATION NFM ID#
Seq. No.
Poae 13 of 21 FT 16 40 Bundles Natural U 4.53 w/o 11G8.0 4.27 w/o 12G8.0 4.21 w/o 13G8.0 11"
"-See Figure 3 0
36" See Figure 6 60" See Figure 4 36" See Figure 5 Natural U
, 6" 3.91 w/o o SPCA9-391B-14G8.0-100M FT 17 120 Bundles Natural U 4.53 w/o 11G8.0
- 4.56 w/o 12G8.0 11 "1 84" 48" Natural U 6"
4.10 w/o SPCA9-410B-19G8.0-100M Figure 1. L2C9 Bundle Design (Fuel Types 16 and 17) preparer.
MH93-c reviewer P4LJ - 1,31,00 NFM00001 15 0
I
-I
NUCLEAR FUEL MANAGEMENT NFM ID#
NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Page 14 of 21 FT18 132 Bundles FT19 56 Bundles Natural U 4.27 w/o 12G8.0 72" 60" 6"
4.21 w/o 13G8.0 Natural U 11 of See Figure 6 See Figure 7 See Figure 8
,0 See Figure 9
,0 See Figure 5 Natural U 3.96 wlo 8G5.0 4.58 w/o 8G6.0 4.58 w/o 12GZ Natural U 3.83 w/o 3.96 w/o SPCA9-396B-12GZ-100M Figure 2. L2C9 Bundle Design (Fuel Types 18 and 19) preparer:
p ar r
rei-31e-w reviewer J.3-11"1 42" 24" 66" 6"1 SPCA9-383B-16G8.0-100M
NUCLEAR FUEL MANAGEMENT NFM ID#
NFMO00O1 15 TRANSMITTAL OF DESIGN INFORMATION Se~q. No.
0 Page 15 of 21 TYPE 1
2 3
4 5
-6 7
8 9
- ENR 4
3.00 a-,
3.6o 8
4.40 37 4.95 4, -
4.70 0
0 11 0
4.40 0.00 GD 0
0
'0 0
0 0
8.00 0
Figure 3. SPCA9-4.53L-11G8.0-100M Lattice Enrichment Distribution preparer 771W, 6,-I/'o reviewer ('&tJ 0.3.oo
/
SNUCLEAR FUEL MANAGEMENT NFM ID#
NFMODD01 15 TRANSMITrAL OF DESIGN INFORMATION Seq. No.
0 Pooe 16 of 21 1
Rods (4) 2 Rods (12) 3 Rods (8) 4 Rods (36)
G1 Rods (8)
G2 Rods (4) 3.00 w/o U-235 4.00 w/o U-235 4.70 w/o U-235 4.95 w/o U-235 4.20 w/o U-235+8.0 w/o Gd203 4.70 w/o U-235+8.0 wlo Gd203 Figure 4. SPCA9-4.56L-12G8.0-100M Lattice Enrichment Distribution preparer:
rfft-/, &-3)-oo reviewer P~kl 9,31- 00 1
2 3
4 4
4 3
2 1
S 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 2
4 G
S 2
244 2
2 4.00 4.00 4.95 A:70-4.95 4.00 4.00 4.0 40 4.95 4.95 4.95 4.95 4.95 4~04.70 3
4 4
4
-'-.o 4
4
'3 4.70 4.95 4.95 4.95S 4.95 4.95 4.95 4
G24 4
4 4
4.95 4.95 4.95 4.95 4.95 4.95 474 4
4 4
4 4.95 4.95 4.95 4.95 4.95 4.95 4
4 4
4 4
4.7 4.00 4.00 4.95 4.95 4.90 4.00 4.00 1
2 3
4 4
4 4
2 3
3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00
NUCLEAR FUEL MANAGEMENT NFM ID#
NFMOO001 15 "TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Pope 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-l00M Lattice Enrichment Distribution preparer "'*Yd 6-riroc reviewer PAC,.,
ý.31-00
NUCLEAR FUEL MANAGEMENT NFM ID#
NFM0000115 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 1
1 Page 18 of 21 TYPE ENR GD 1
4 2.60 0
2 8
3.20 0
3 a
4.00 0
4 36 4.70 0
5 4
4.40 0
6 0
0 7
0 0
8 12 4.40 8.00 9
0 0.00 0
Figure 6. SPCA9-4.27L-12G8.0-100M Lattice Enrichment Distribution preparer:
W'//J, 1
r L-reviewer
- d*f.6
NUCLEAR FUEL MANAGEMENT NFM ID#
NFMOO001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Page 19 of 21 2 24 4,
4' 3
2 1
2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 2
2 4
4G
'G 4,
2 2
"3.40 3.40 4.40
,.o 4.40 30 4.40, 3.4 3.4D 3
4
.4 4
4,
,4 4
4 3
3.80 4.40 4.40 4.40 4.4OA 4.40 4.40 4.40 3.80 S4 4
i"'
440
.4 0
4.40 4.40 "
4.40 4
4 4
4.40 4.40 4.40
.440 4.40 4.40 i
IV4 4
4 4
4 4
4 Nt 4.40 3.40 4.40 4.40 4
4.40 3
4 4
4 3.80 4.40 4.40 4.4 4.40 4.4 4.40 4.40 3.80 2 2 3.40 3.40 4.40 U3.40 4.40 34 4:40 3.40 3.40 12 3
4 3
2 1
2.60 3.40 3.80 4.40 4.40 4.40 3.80 3.40 2.60 1
2 3
4 G1 Rods (4)
Rods (12)
Rods (8)
Rods (40)
Rods (8) 2.60 wlo U-235 3.40 w/oVU-235 3.80 wlo U-235 4.40 wo"U-235 3.40 w/o U-235+5.0 wlo Gd203 Figure 7. SPCA9-3.96L-8G5.0-100M Lattice Enrichment Distribution preparer: "
/-W y
'-0 1
-oD reviewer Piht
,,' 1-0 &
NUCLEAR FUEL MANAGEMENT NFMV ID#
NFMV000015 IS TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Page 20 of 21 1
2 3
4 GI Rods (4)
Rods (12)
Rods (8)
Rods (40)
Rods (8) 3.00 w/o U-235 4.00 w/o U-235 4.70 w/o U-235 4.95 w/o U-235 4.20 w/o U-235+6.0 w/o Gd203 Figure 8. SPCA9-4.58L-8G6.0-100M Lattice Enrichment Distribution preparer: m Yi d?-3)*- 0 reviewer [-
NUCLEAR FUEL MANAGEMENT NFM ID#
NFM00001 15 TRANSMITTAL OF DESIGN INFORMATION Seq. No.
0 Page 21 of 21 2
3 4
4
.4 3
2 3.00 4.00 4.70 4.95 4.95 4.95 4.70 4.00 3.00 4§ 4
j 4
_G
-V 2
4.00 ON 4.95 42 4.95 V~~0 4.95S 0
4.00 3
4 4
4 4
4 4
4 3
4.70 495 4.95 4.95 4.95 4.95 4.95 4.95 4.70 44.90 495 4.95 4.95 4
4 4
4 4
4 4.95 4.95 4.95 a
4H.9ne 4.95 4.95 4.95 S 4.95 4.20
- 4.
45
-2.
4.95 495 4.9
.9 4.70 4.95 4.95 4.9 4-95__ L
.95 4.95 4.95 4.70
.
2 4.00 420 1
2 3
4 4
4 3
2 1
3.00 4.00 4.70 4.9S 4.95 4.95 4.70 4.00 3.00 2
4.00 4
4.95 1
2 3
4 GI G2 Rods (4)
Rods (5)
Rods (8)
Rods (40)
Rods (8)
Rods (4) 4 4.95 Gi
4.20
3.00 w/o U-235 4.00 wlo U-235 4.70 w/o U-235 4.95 w/o U-235 4.20 w/o U-235+6.0 wlo Gd203 4.00 wlo VU-235+3.0 w/o Gd203
'v--
Figure 9. SPCA9-4.58L-8G6.0/4G3.0-100M Lattice Enrichment Distribution preparer:
1 )H' I?/-1-oe 4
4.95 02
00
[
i!
reviewer PA d
Technical Requirements Manual - Appendix J L2C9 Reload Transient Analysis Results LaSalle Unit 2 Cycle 9 Reload Analysis Report LaSalle Unit 2 Cycle 9 August 2002
SIEMENS LaSalle Unit 2 Cycle 9 Reload Analysis October 2000 EMF-2437 Revision 0 Siemens Power Corporation Nuclear Division
Siemens Power Corporation ISSUED IN SPC ON-LiNE DOCUMENT SYSTEM DATE:
0 LaSalle Unit 2 Cycle 9 Reload Analysis Prepared:
J. M. Haun, Engineer BWR Neutronics Prepared:
2mrD S
- r. B. McBumpy, Engineer
- k Prepared:
Concurred:
Concurred:
Approved:
Approved:
Approved:
Product Mechani Engineering H. D. C~t Manager Pro,ý censing Date Ita2. Icw Date Date Date D. J. Denver, Manager Commercial Operations "finD~r.%~ /0g/LA,
- & o Cý
- 0. C. Brown, Madr 6U BWR Neutronics M. E. Garrett, Managwr Safety Analysis T. M. Howe, Manager Product Mechanical Engineering Date Date Date 10-03-a Date Siemens Power Corporation EMF-2437 Revision 0
/sp
Customer Disclaimer 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 those set forth in the agreement between Siemens Power Corporation and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Siemens Power Corporation nor any person acting on Its behalf:
- a.
makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or
- b.
assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this document.
The information contained herein is for the sole use of the Customer.
In order to avoid impairment of rights of Siemens Power Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such information until so authorized in writing by Siemens Power Corporation or until after six (6) months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement. No rights or licenses in or to any patents are implied by the furnishing of this document
LaSalle Unit 2 Cycle 9 Reload Analysis Nature of Changes Item Page Description and Justification
- 1.
AJI This is a new document clftftc 9n...r r'^PUWý*iý EMF-2437 Revision 0 Paae ii
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page iii Contents 1.0 Introduction................................-......,..........
1-1 2.0 Fuel Mechanical Design Analysis................................................................................
2-1 3.0 Thermal-Hydraulic Design Analysis.............................................................................
3-1 3.2 Hydraulic Characterization...............................................................................
3-1 3.2.1 Hydraulic Compatibility.......................................................................
3-1 3.2.3 Fuel Centerline Temperature............................................................
3-1 3.2.5 Bypass Flow.......................................................................................
3-1 3.3 MCPR Fuel Cladding Integrity Safety Limit (SLMCPR)....................................
3-1 3.3.1 Coolant Therm odynamic Condition....................................................
3-1 3.3.2 Design Basis Radial Power Distribution.............................................
3-2 3.3.3 Design Basis Local Power Distribution...............................................
3-2 3.4 Licensing Power and Exposure Shape...........................
3-2 4.0 Nuclear Design Analysis..............................................................................................
4-1 4.1 Fuel Bundle Nuclear Design Analysis..............................................................
4-1 4.2 Core Nuclear Design Analysis..........................................................................
4-2 4.2.1 Core Configuration.............................................................................
4-2 4.2.2 Core Reactivity Characteristics..........................................................
4-2 4.2.4 Core Hydrodynamic Stability.............................................................
4-2 5.0 Anticipated Operational Occurrences..........................................................................
5-1 5.1 Analysis of Plant Transients at Rated Conditions.........................................
5-1 5.2 Analysis for Reduced Flow Operation..............................................................
5-1 5.3 Analysis for Reduced Power Operation............................................................
5-2 5.4 ASME Overpressurization Analysis.................................................................
5-2 5.5 Control Rod W ithdrawal Error..........................................................................
5-2 5.6 Fuel Loading Error......................................................................................
5-2 5.7 Determ ination of Therm al Margins...................................................................
5-2 6.0 Postulated Accidents...................................................................................................
6-1 6.1 Loss-of-Coolant Accident.................................................................................
6-1 6.1.1
, Break Location Spectrum..................................................................
6-1 6.1.2 Break Size Spectrum.........................................................................
6-1 6.1.3 MAPLHGR Analyses..........................................................................
6-1 6.2 Control Rod Drop Accident..............................................................................
6-1 6.3 Spent Fuel Cask Drop Accident.......................................................................
6-1 7.0 Technical Specifications..............................................................................................
7-1 7.1 Limiting Safety System Settings.......................................................................
7-1 7.1.1 MCPR Fuel Cladding Integrity Safety Limit........................................
7-1 7.1.2 Steam Dome Pressure Safety Limit...................................................
7-1 7.2 Limiting Conditions for Operation................................................................
7-1 7.2.1 Average Planar Linear Heat Generation Rate....................................
7-1 7.2.2 Minimum Critical Power Ratio............................................................
7-1 7.2.3 Linear Heat Generation Rate...........................
7-2
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 V~i m Methodology References.............................................................................................
8-1 Additional References.................................................................................................
9-1 Siemens Power Comoration 8.0 9.0
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 MCPRP Limits and LHGRFACp Multipliers for TSSS Insertion Times.......................................................
5-7 5.4 FFTRPCoastdown Operation Base Case and EOOS MCPR% 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 Local 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 5.4 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
LaSalle Unit 2 Cycle 9 Reload Analysis Nomenclature abnormal operational occurrence beginning of cycle effective full power hours end of cycle extended operating domain end of full power equipment out of service final feedwater temperature reduction feedwater heater out of service feedwater controller failure interim corrective actions increased core flow LFWH LHGR LHGRFAC LOCA LPRM LRNB MAPFAC MAPLHGR MCPR MELLLA MSIV NSS PAPT PCT RPT SLMCPR SLO SPC SRVOOS TBVOOS TCV TIP TIPOOS loss of feedwater heating linear heat generation rate LHGR multiplier loss of coolant accident local power range monitor load rejection no bypass MAPLHGR multiplier maximum average planar linear heat generation rate minimum critical power ratio maximum extended load line limit analysis main steam isolation valve nominal scram speed protection against power transient peak clad temperature recirculation pump trip safety limit minimum critical power ratio single-loop operation Siemens Power Corporation safety/relief valve out of service turbine bypass valves out of service turbine control valve traversing in-core probe traversing in-core probe out of service Siemens Power Corporation EMF-2437 Revision 0 Page vi Page vi AOO BOC EFPH EOC EOD EOFP EOOS FFTR FHOOS FWCF ICA ICF
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 Page vii technical specification scram speed updated final safety analysis report change in critical power ratio TSSS UFSAR ACPR e
14
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 1-1 1.0 Introduction This report provides the results of the analysis performed by Siemens Power Corporation (SPC) as part of the reload analysis in support of the Cycle 9 reload for LaSalle Unit 2. This report is intended to be used in conjunction with the SPC topical Report XN-NF-80-19(P)(A),
Volume 4, Revision 1, Application of the ENC Methodology to BWR Reloads, which describes the analyses performed in support of this reload, identifies the methodology used for those analyses, and provides a generic reference list. Section numbers in this report are the same as corresponding section numbers in XN-NF-80-19(P)(A), Volume 4, Revision 1. Methodology used in this report which supersedes XN-NF-80-19(P)(A), Volume 4, Revision 1, is referenced in Section 8.0. The NRC Technical Limitations presented in the methodology documents, including the documents referenced in Section 8.0, have been satisfied by these analyses.
Analyses performed by Commonwealth Edison Company (CornEd) are described elsewhere.
This document alone does not necessarily identify the limiting events or the appropriate operating limits for Cycle 9. The limiting events and operating limits must be determined in conjunction with results from ComEd analyses.
The Cycle 9 core consists of a total of 764 fuel assemblies, including 348 unirradiated and 256 irradiated ATRIUM'-9B" assemblies and 160 irradiated GE9 assemblies. The reference core configuration is described in Section 4.2.
The design and safety analyses reported in this document were based on the design and operational assumptions in effect for LaSalle Unit 2 during the previous operating cycle. The effects of channel bow are explicitly accounted for in the safety limit analysis. The extended operating domain (EOD) and equipment out of service (EOOS) conditions presented in Table 1.1 are supported.
ATRIUM is a trademark of Siemens.
Siemens Power Corporation
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 Feedwate.r Temperature Reduction (FFTR)
FFTRJCoastdown 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 1 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 CorDoration
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 72.3. 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.
LaSalle Unit 2 Cycle 9 EMF-2437 Revision 0 Page 3-1 3.0 Thermal-Hydraulic Design Analysis 3.2 Hydraulic Characterization 3.2.1 Hydraulic Compatibility Component hydraulic resistances for the fuel types in the LaSalle Unit 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 Applicable Report ATRIUM-9B Reference 9.1, Figure 3.3 3.2.5 - Bypass Flow Calculated Bypass Flow at 100%P/100%F (includes water channel flow) 14.8 Mlb/hr Reference 9.3 3.3 MCPR Fuel Cladding Integrity Safety Limit (SLMCPR)
Two-Loop Operation'.
Single-Loop Operation"
-1.11
- ' 1.12 Reference 9.3 Coolant Thermodynamic Condition Thermal Power (at SLMCPR)
Feedwater Flow Rate (at SLMCPR)
Core Exit Pressure (at Rated Conditions)
Feedwater Temperature 5167.29 MWt 22.4 Mlbm/hr 1031.35psia 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.
3.3.1
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 InDo a q-9 3.3.2 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 Umit analysis.
SPCA9-391B-14GB.0-100M SPCA9-410B-19G8.0-1OOM SPCA9-383B-1 6G8.0-1 0DM SPCA9-396B-12GZ-100M Figure 3.2 Figure 3.3 Figure 3.4 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 CorDoration
SLaSalie Unit 2 Cycle 9 Reload Analysis Table 3.1 Licensing Basis Core Average Axial Power Profile and Table 3.1 Licensing Basis Core Average Axial Power Profile and Licensing Axial Exposure Ratio State Conditions for Power Shape Evaluation Power, MWt Core Pressure, psia Inlet Subcooling, Btu/Ibm Flow, Mlb/hr EMF-2437 Revision 0 Page 3-3 3489.00 1020.00 18.20 108.50 Licensing Axial Power Profile Node Power Top 25 24 23 22 21 20 19 18 17 16 15 14 13 12 "11 10 8
7 6
5 4
2 Bottom I 0.211 0.417 0.967 1207 1.371 1.445 1.454 1.428 1.384 1.346 1.299 1.248 1.199 1.151 1.102 1.053 1.002 0.944 0.887 0.835 0.796 0.770 0.726 0.583 0.177 Licensing Axial Exposure Ratio (EOFP)
Average Bottom 8ft12 ft = 1.098 Riprmn PnrwApr t'rvnnmuw#i
LaSalle Unit 2 Cycle 9 Reload Analysis 200 175 150 U) 5 125 Io 00 E 75 z
50 25 0
EMF-2437 Revision 0 Paoe 3-4
.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 Rinmrnon PnwAmr.nmnwatifr
LaSalle Unit 2 Cycle 9 PpInmrd Anahic~
EMF-2437 Revision 0 Page 3-5 Control Rod Corner 0
n t
r 0
R 0
d C
0 r
n e
r
-Figure 3.2 LaSalle Unit 2 Cycle,9,Safety Limit Local Peaking Factors SPCA9-391B-14G8.D-IDOM With Channel Bow 1.052 1.045 1.088 o1.0881 1.104 11.079 1.068 1.013 1.005 1.045 0.951
-1.019-0.996, 0.852, 0.986 0.998' 0.914.
0.991 1.088 1.019 1.001 1.059 1.089, 1.051
ý0.982 0.9811 1.027 1.088 0.996 1.059
' 0.905 0.957 1.050 Internal 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 0.956
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 Page 3-6 C
0 n
t r
0 R
0 d
C 0
r n
e r
Figure 3.3 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-41 OB-19G8.0-1 0DM With Channel Bow ontrol Rod Corner 1.058 1.049 1.092 1.091 1.107 1.082 1.072 1.017 1.010 1.049 0.945 1.020 0.996 0.843 0.987 0.998 0.906 0.995 1.092 1.020 1.002 1.061 1.090 1.052 0.981 0.980 1.030 1.091 0.996 1.061 0.894 0.955 1.053 Internal 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 1.067 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
LaSalle Unit 2 Cycle 9 Reload Analysis ntr EMF-2437
-Revision 0 Page 3-7 ol Rod Corner C
0 n
A r
0 R
0 d
C 0
r n
e r
Figure 3.4 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-383B-16GS.0-100M With Channel Bow 0
1.017 1.017 1.068 1.083 1.107 1.074 1.048.
0.985 0.970 1.017 0.986 1.024 1.000 0.885 0.992 1.004 0.956 0.965 1.068 1.024
'0.890 1.063 1.091 1.055 -0.990 0.9B9 1.009 1.083 1.000 1.063 0.944 0.966 1.055
-Internal 1.107ý 0.885 1.091 i Water 1.074:
0.846-1.040 Channel.
1.074 0.992 1.055
-1.032 0.951 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
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 Page 3-8 ntrol Rod Corner C
0 n
t r
0 R
0 d
C 0
r n
e r
Figure 3.5 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-396B-12GZ-100M With Channel Bow 0
1.025 1.058 1.062 1.117 1.100 1.108 1.043 1.026 0.979 1.058 0.934 1.018 0.852 1.003 0.845 0.999 0.903 1.005 1.062 1.018 1.003 1.067 1.092 1.058 0.984 0.983 1.006 1.117 0.852 1.067 1.046 0.823 1.056 Internal I
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
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 4-1 4.0 Nuclear Design Analysis 4.1 Fuel Bundle Nuclear Design Analysis "The detailed fuel bundle design information forthe fresh ATRIUM7M-9B 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)
SPCA9-391B-14GB.0-100M (FT16) 3.91 wt%
SPCA9-410B-19G8.0-100M (FT17) 4.10 wt%
SPCA9-383B-16G8.0-1 00M (FT18) 3.83 wt%
SPCA9-396B-12GZ-100M (FF19) 3.96 wt%
Radial Enrichment Distribution SPCA9-4.56L-12GS.0-100M Ref. 9.12 Figure B.19 SPCA9-4.21L-13G8.0-100M Ref. 9.1
'Figure D.1 SPCA9-4.27L-12G8.0-1OOM Ref. 9.1 Figure D.2 SPCA9-4.53L-11G8.0-1OOM Ref. 9.1 Figure D.3 SPCA9-3.96L-8G5.0-1 00M Ref. 9.12 Figure B.122 SPCA9-4.58L-8G6.0/4G3.0-100M Ref. 9.12 Figure B.140 SPCA9-4.58L-8G6.0-1 0DM Ref. 9.12 Figure B.157 Axial Enrichment Distribution Ref. 9.1 Figures 5.1-5.4 Burnable Absorber Distribution Ref. 9.1 Figures 5.1-5.4 Non-Fueled Rods Ref. 9.1 Figures 5.1-5.4 Neutronic Design Parameters Table 4.1 Fuel Storage 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 thýerefore7can be safely stored in the vault.
LaSalle Unit I Spent Fuel Storage Pool (BORAL Racks),
Reference 9.5 The LSB-2 Reload Batch fuel designs meet the fuel design limitations defined in Table 2.1 -of Reference 9.5 and therefore can be isafely stored in the pool.
Ripnrnr Pnwmr Comomtin
LaSalle Unit 2 Cycle 9 Reload Analysis 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.
< CornEd has responsibility to confirm that fuel meets reactivity limitations. >
4.2 Core Nuclear Design Analysis 4.2.1 Core Confiquration Core Exposure at EOC8, MWd/MTU (nominal value)
Core Exposure at BOC9, MWd/MTU (from nominal EOC8)
Core Exposure at EOC9, MWd/MTU (licensing basis to EOFP)
Figure 4.1 27,893.9 11,808.0 30,266.2 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 CornEd. >
4.2.2 Core Reactivity Characteristics
< This data is to be furnished by CornEd. >
4.2.4 Core Hydrodynamic Stability Reference 8.7 LaSalle Unit 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 Revision 0 Page 4-2 LaSalle Unit 2 Cycle 9 Reload Analysis
LaSalle Unid 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 Page 4-3 Provides trending information to qualitatively compare the stability from cycle to cycle.
Provides decay ratio sensitivities to rod line and flow changes near the ICA regions.
Allows ComEd 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 determine the highest expected decay ratio throughout the cycle. The results from this analysis are shown below.
Power/Flow Maximum Maximum
'Global Regional 30.1/26.6 0.59 0.53 31.6/29.2 0.40
- 0.50 61.9145.0 0.50 0.88 73.6150.0 0.52 0.95 78.2/60.0 0.33 0.63 82.4160.0 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: % p*over is based on 3489 MWt as rated. % flow is based on 108.5 Mlb/hr as rated.
- SiemensPower Corxorati
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 4-4 Table 4.1 Neutronic Design Values Number of Fuel Assemblies 764 Rated Thermal Power, MWt 3489 Rated Core Flow, Mlbm/hr 108.5 Core Inlet Subcooling, Btu/lbm 18.2 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 B4C 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 Comoration
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 4-5 J: 1 3 5 7 9 11 13 15 17 1921 2325272931 333537 39 41 43 45 47 49 51 53 55 57 59 60 2 2 1 1 1 1 1 1 1 1 2 2 1 58
_-214191919 15 191919-1915 191919142 56 2 2 15 191715171617141417167-151719152 2
54 1 19 17 18,14 1 151815 I 1 15 1815 1 14 18 17 19 1 52 1 14 17 17 151515181417181817141815 1517 1714 1 50 2 114 1 14 17 17 17 2 "17 17 15 15 15 15 17 17 2 17 17 1714 1 14 1 2 1411 17 1 7 144 4
2 91714 11817 8118414 214 181 5*.1 181718 1114 1719 2 46 2 1517 1717 18614 14181141818S1711414 171811814 1814414 181717 1715 2 44 1 14 19 18 15 17 17,14 14 1818 1818 14 18 I14 18 18 18 18 1414 1717 1518 19 14 1 42 2 1917 1415 1718 181814 14 1715 1814 14 18151714 1441818 1817715514417719922 38 8
40 2 -19155115 2 1 1418 14 21416s15 2 2 1516114 2 1418114112 2151151192 38 1 1911715 18 1715 181817142156151516 15121417 18 181517 181151719 1 36 1 15 16 18 14 17 18 18 18-15 16 15 14 16 16 16 16 14 15 16-15 18 18 18 17 14 18 1615 1 34 1 19 1715 1715 1417 14 1815 a16618 1515 181616 15 181417 14 1517 151719 1 32 1191411815214181421516152 2151615214181421518114191 30 1 19 14,1,8 152 1418 1412 1516152 2 15161152 14 1814 2 1518 11 14*19 1 28 1 19 17 15 17 15 14 17 14 18 15 16 16 18 15 15 18 16 16 1518 14 17 14 15 17 15 17 19 1 26 1-151618 14 17 118 18 15 16,15 14 16 16 16 1614 15 16 15 18 18 18 17 14 18 1615 1 24 1 1911715118 17J!ý5181[i11j115i78jj78i~19 22 2 1911
-152111842146152 2 11614-2 14 18 14 1 2 151 15 19 2 20 21 9 17 1415 1718181814 141 7 15 1814 141 8 15 7 1 14 188817 1 51417 19 2 18 1 14191815 171714 14 18 18 18 1814 1 18 11418181818 14 1417 17 15181914 1 16 2 215 1717 171814114181141818S171141417 18[18F17418-14114 181717 1715 2 14 2 19 1714 117 1811 1518 142 2 14 18~15161711 11441771922 1
I 12 2 1 14 1 14 1717 17 217 171515 15 1517 17-2.171717 1441 1441 22 10 1 1417171 15*15 18 1417181817141815 15 15 171714 1 8
1 19 17 1814711l518 151,1 15 18 5
14 18-17 19 1 Fuel Number Load Bundle Name of Bundles Cd 1
GE9B-P8CWB322-11GZ-100M-150 84 7
2 GE9B-P8CWB320-9GZ-1ODM-150 76 7
14 SPCA1-381B-13GZ7-80M 128 8
15 SPCA2-384B-11GZ6-80M 128 8
16 SPCA9-391B-14GS.0-100M 40 9
17 SPCA2-410B-19G8.0-100M 120 9
18 SPCA1-383B-16G1.0-100M
¶32 9
19 SPCA9-396B-12GZ-1O0M 56 9
Figure 4.1 LaSalle Unit 2 Cycle 9 Reference Loading Map
LaSalle Unit 2 Cycle 9 Reload Analysis 5.0 Anticipated Operational Occurrences Applicable Disposition of Events 5.1 Analysis of Plant Transients at Rated Conditions Limiting Transients:
Reference 9.7 Reference 9.3 Load Rejection No Bypass (LRNB)
Feedwater Controller Failure (FWCF)
Loss of Feedwater Heating (LFWH)
Transient LRNB" FWCF" LRNB" FWCF" LFWHt S cram Speed "TSSS TSSS NSS NSS Peak Neutron Flux
(% Rated) 422 298 380 263 Peak Heat Flux
(% Rated) 127 123 124 120 t
Peak Lower Plenum Pressure (psig) 1218 1176 1211 1169 "1
5.2 Analysis for Reduced Flow Operation Limiting Transient: Slow Flow Excursion MCPRr Manual Flow Control -
ATRIUM-9B and GE9 Fuel LHGRFAC-- ATRIUM-9B Fuel MAPFACf-- GE9 Fuel ACPR ATRIUM-9B/GE9 0.30/0.40 02510.31 0.2810.37 0.2310.29 "t
Reference 9.3 Figure 5.1 Figure 5.2 MCPRf and LHGRFACI 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.
S This data to be furnished by ComEd.
Rs*immat Pet~rfwnrrwh~uir EMF-2437 Revision 0 Paqe 5-1 Paq 5-r.
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 PnriFL9 5.3 Analysis for Reduced Power Operation Limiting Transient Load Rejection No Bypass (LRNB)
Feedwater Controller Failure (FWCF)-
MCPRp Base Case Operation LHGRFACý Base Case Operation" MCPRp, EOOS Conditions LHGRFACp, EOOS Conditions" MAPFACý -
All Operating Conditions*
5A ASME Overpressurization Analysis Limiting Event Worst Single Failure Maximum Vessel Pressure (Lower Plenum)
Maximum Steam Dome Pressure 5.5 Control Rod Withdrawal Error Starting Control Pattern for Analysis Reference 9.3 Tables 5.1-5A Figures 5.3-5.6 Tables 5.1-5A Tables 5.1-5A Tables 5.1-5A
<To be furnished by CornEd.>
Reference 9.3 MSIV Closure Valve Position Scram 1346 psig 1320 psig Figure 5.7
< This data is to be furnished by ComEd. >
5.6 Fuel Loading Error
< This data is to be furnished by CornEd. >
5.7 Determination of Thermal Margins The results of the analyses presented in Sections 5.1-5.3 are used for the determination of the operating limit. Section 5.1 provides the results of analyses at rated conditions. Section 5.2 provides for the determination of 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 C")-,
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 5-3 5.1; LHGRFACf, Figure 5.2 ). Section 5.3 provides for the determination of the MCPR and LHGR limits at conditions of reduced power (Figures 5.3-5.6, Tables 5.1-5.4). Limits are presented for base case operation and the EOD and EOOS scenarios presented in Table 1.1.
The results presented are based on the analyses performed by SPC. As indicated above, the final Cycle 9 MCPR operating limits need to be established in conjunction with the results from CornEd analyses.
Siemens PowerCooratioan
LaSalle Unit 2 Cycle 9 Reload Annlysis Table 5.1 EOC Base Case and EOOS MCPR, Limits and LHGRFAC, Multipliers for TSSS In-sertion Times
'Power.
ATRIUM-9B Fuel GE9 Fuel SCondition
- (% rated)
MCPi*
LHGRFACý MCPRv
-0 2.70 0.78 2.70 Base 25 220 0.78 2.20 case 25 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 25 2.35 0.69 2.35 heaters 25 2.14 0.69 222 out-of-service (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 0.99 1.52 EMF-2437 Revision 0 Pane 5-4 R
l a
mi..
el-21-
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 Page 5-5 Table 5.1 EOC Base Case and EOOS MCPRi Limits and LHGRFACp Multipliers for TSSS Insertion Times (Continued)
EOOS / 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 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
2.85 0.68 2.85 TCV 25 2.35 0.68 2.35 slow closure/
25 2.14 0.68 2.22 FHOOS 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
2.60 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 Siemens Power Comoratkin
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 Page 5-6 STable 5.2 EOC Base Case MCPRI Limits and LHGRFACp-Multipliers for NSS Insertion Times EOOSIEOD Power ATRIUM-9B Fuel GE9 Fuel Condition-(%Wrated)
MCPRP LHGRFACp 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 1.51 100 1.39 1.00 1.48 Siemens Power Covoration I
LaSalle Unit 2 Cycle 9 Reload Analysis Table 5.3 Coastdown Operation Base Case and EOOS MCPRp Limits and LHGRFACp Multipliers for TSSS Insertion Times EQOS (EOD Power ATRIUM-9B Fuel GE9 Fuel Condition
(% rated)
MCPRp LHGRFACp MCPR*
0 2.70 0.75 2.70 Coastdown 25 2.20 0.75 2.20 base case 25 2.05 0.75 2.05 operation 60 1.48 0.99 1.54 100 1.42 1.00 1.52 0
2.71 0.75 2.71 Coastdown with 25 2.21 0.75 2.21 single-loop 25 2.06 0.75 2.06 operation 60 1.49 0.99 1.55 100 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 (TBVOOS) 60 1.55 0.97 1.64 100 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) 1 100 1.55 0.88 1.67 Siemens Power Corporaton EMF-2437 Revision 0 Paae 5-7 Paoe 5-7
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 1 Page 5-8 Table 5.3 Coastdown Operation Base Case and EOOS MCPRP Limits and LHGRFACp Multipliers for TSSS Insertion Times
-(Continued)
- EOOS I EOD
,Power ATRIUM-9B Fuel GE9 Fuel Condition
(% rated)
MCPR*
LHGRFACp MCPRP 0
2.70 0.68 2.70 Coastdown with 25' 2.20
-0.68-2.20 turbine control valve (TCV) 25 2.15 0.68 2.15 slow closure 80 1.70 0.85 1.96 ANDIOR no RPT 80 1.62
_.0.88 1.85
_100 1.55 0.88 1.67 0
2.60 0.40' 2.60 Coastdown with 25 2.60 0.40 2.60 idle loop 25 2.60 0.40 2.60 startup 60 2.60 0.40 2.60 100 2.60 0.40 2.60 Siemens Power Comoration
LaSalle Unit 2 Cycle 9 Reload Analysis Table 5.4 FFTR/Coastdown Operation Base Case and EOOS MCPRI Limits and LHGRFACp Multipliers for TSSS Insertion Times EOOS 1 EOD Power ATRIUM-9B Fuel GE9 Fuel Condition
(% rated)
MCPR*
LHGRFACp MCPR9 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 FFTRPcoastdown 25 2.36 0.65 2.36 with single-loop 25 2.31 0.65 2.31 operation 60 1.57 0.97 1.60 100 1.43 1.00 1.53 0
2.85 0.65 2.85 FFTR/coastdown 25 2.35 0.65 2.35 with turbine bypass valves 25 2.30 0.65 2.30 out-of-service 60 1.5 0.97 1.64 (TBBVOOS) 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 Revision 0 Page 5-9
)
KJ
LaSalle Unit 2 Cycle 9 Reload Analysis Table 5.4 FFTR/Coastdown Operation Base Case and EOOS MCPRP Limits and LHGRFAC. Multipliers for TSSS Insertion Times (Continued)
EOOS/EOD Power ATRIUM-96 Fuel GE9 Fuel Condition
(% rated)
MCPRP LHGRFACp MCPRP 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 868 no RPT 80 1.62 0.88 1.85 100 1.55 0.88 1.67 0
2.60 0A0 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 2.60 100 2.60 0.40 2.60 EMF-2437 Revision 0 Page 5-10 R
l a
at si CL--
Dýr e"rWWWMNý
LaSalle Unit 2 Cycle 9 Reload Analysis 0
10 20 30 40 50 60 70 80 90 100 110 Flow (% of Ratme)
MCPRf GE9 Flow MCPRP (penalty
(% of rated)
ATRIUM-9g included) 0 1.60 1.66 30 1.60 1.66 105 1.11 1.11 Figure 5.1 Flow-Dependent MCPR Limits for Manual Flow Control Mode
EMF-2437 Revision 0 Pame 5-11 Paqe5-11 U
K
- a.
U z I
I
LaSalle Unit 2 Cycle 9 Reload Analysis U-EMF-2437 Revision 0 Pame 5-12 Percent of Rated Flow Flow (% rated)
-0 30 76 "105
- LHGRFAC, 0.69 0.69 1.00 1.00 Figure 5.2 Flow Dependenit LHGR Multipliers 'for ATRIUM-9B Fuel
LaSalle Unit 2 Cycle 9 Reload Analysis EMF-2437 Revision 0 PameS-13 0
10 20 30 40 50 60 70 s0 90 100 110 Power (% of Ratwd)
I Power MCPR:
(%)
mit 100 1.41 60 1.46 25 1.91 25 2.20 0
2.70 Figure 5.3 EOC Base Case Power-Dependent MCPR Limits for ATRUM-9B Fuel - TSSS Insertion Times A.
U z
\\
4
LaSalle Unit 2 Cycle 9 IoRle'nad Analvi..
EMF-2437 Revision 0
--Pa-oe 5-14 0
10 20 320 40 so so 70 s0 90 100 110 Power (C%
of Ra.eW)
Power MCPRp
(%)
Limit 100
-1.51 60..
1.52
§25 1.99 25 2.20 0
2.70 Figure 5.4 EOC Base Caise Power-Dependent MCPR Limits for GE9 Fuel - TSSS Insertion Times
- a.
- a.
- a.
U a
LaSalle Unit 2 Cycle 9 Reload Analysis CL a.
Z75 2M' 2Z45 2.35 2.25 2.15 2.05 1.95 115.
1.75 1.65, 1.55 1.45' 1.35 125 1.15 EMF-2437 Revision 0 Paae 5-15 Pace 5-15 0
10 20 30 40 50 60 70 s0 90 100 110 Ponr (% of Rated)
Power MCPRF
(%)
Limit 100 1.39 60 1.44 25 1.89 25 2.20 0
2.70 Figure 5.5 EOC Base Case Power-Dependent MCPR Limits for ATRUM-9B Fuel - NSS Insertion Times Siemens Power Corporation
EMF-2437 Revision 0 PaNe 5-16 LaSaUe Unit 2 Cycle 9 SReload Analysis 0E eU 0
10 20 30 40 50 60 70 s0 90 100 110 P=*w*v(% aof fm )
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 C"un.,
MEWU irrWWrhrVWr R
l a
is..
LaSalle Unit 2 Cycle 9 Reload Analysis
< This data is to be furnished by ComEd. >
Figure 5.7 Starting Control Rod Pattern for Control Rod Withdrawal Analysis EMF-2437 Revision 0 Paoe 5-17 PaQe 5-17
LaSale Unit 2 Cycle 9 W Woua r~iwla7yan EMF-2437 Revision 0 Paie 6-1 6.0 Postulated Accidents 6.1 Loss-of-Coolant Accident 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 If 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%. LOCA/heatup analysis results for LaSalle ATRIUM-9B are presented below (Reference 9.11):
Maximum PCT (OF)
Peak Local Metal-Water Reaction
(%)
ATRIUM-9B Fuel 1825 0.79" The maximum core wide metal-water reaction is < 0.16%.
6.2 Control Rod Drop Accident
< This data is to be furnished by CornEd. >
6.3 Spent Fuel Cask Drop Accident The radiological consequences of a spent fuel cask drop accident have been evaluated for SPC ATRIUM fuel designs in conformance with the analysis described in the LSCS UFSAR Section The peak local metal water reaction result is consistent with the limiting PCT analysis results reported in Reference 9.11.
EMF-2437 LaSalle Unit 2 Cycle 9 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 will not exceed the established radiological whole-body and thyroid dose limits which are a small fraction of the 10 CFR 100 limits for radiological exposures.
Siemens Power Corporation
LaSalle Unit 2 Cycle 9
- l0lnne Analurit 7.0 Technical Specifications 7.1 Limiting Safety System Settings 7.1.1 MCPR Fuel Cladding Inteqrity Safety Limit MCPR Safety Limit (all fuel) -
two-loop operation MCPR Safety Limit (all fuel) -
single-loop operation 7.1.2 Steam Dome Pressure Safety Limit Pressure Safety Limit 7.2 Limiting Conditions for Operation "7*1 AvPrnne Pl-nnr Linear Het Generation Rate ATRIUM-9B Fuel MAPLHGR Limits Average Planar Exposure (GWd/MTU) 0.0 20.0 61.1 MAPLHGR (kWlft) 1.1 1 1.12" 1325 psig Reference 9.9 GE9 Fuel MAPLHGR Limits
< To be furnished by CornEd. >
13.5 13.5 9.39 Single Loop Operation MAPLHGR Multiplier for SPC Fuel is 0.90 7.22 Minimum Critical Power Ratio Rated Conditions MCPR Limit Reference 9.9 Flow Dependent MCPR Limits:
Manual Flow Control Figure 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.
This data is to be furnished by CornEd.
EMF-2437 Revision 0 Page 7-1
- *e I
LaSalle Unit 2 Cycle 9 Reload Analysis Power Dependent MCPR Limits:
Base Case Operation - TSSS Insertion Times Base Case Operation - NSS Insertion Times EOD and EOOS Operation 7.2.3 Linear Heat Generation Rate ATRIUM-9B Fuel Steady-State LHGR Limits Average Planar Exposure LHGR (GWdjMTU)
(kW/ft) 0.0 14.4 15.0 14.4 61.1 8.32 Figures 5.3 & 5.4 Figures 5.5 & 5.6 Tables 5.1-5.4 Reference 9.1 GE9 Fuel Steady-State LHGR Limits
< To be furnished by ComEd.>
The protection against power transient (PAPT) linear heat generation rate curve for ATRIUM--i) fuel is identified in Reference 9.1 and is presented here as Figure 7.1 for convenience.
LHGRFACf and LHGRFACý multipliers are applied directly to the steady-state LHGR limits at reduced power, reduced flow and/or EOD/EOOS conditions to ensure the PAPT LHGR limits are not violated during an AOO. Comparison of the Cycle 9 nodal power histories for the rated power pressurization transients with the approved bounding curves to show compliance with the 1% strain criteria for GE9 fuel is discussed in Reference 9.10.
LHGRFAC Multipliers for Off-Rated Conditions - ATRIUM-9B Fuel:
LHGRFAC1
- LHGRFAC, Figure 5.2 Tables 5.1-5.4 MAPFAC Multipliers for Off-Rated Conditions - GE9 Fuel:
- MAPFAC, MAPFAC,
< To be furnished by CornEd. >
< To be furnished by ComEd.>
EMF-2437 Revision 0 On-rab 7-"1 Vn 7.
ki
.mmane Demmr (jr.wms#**,
LaSalle Unit 2 Cycle 9 Reload Analysis 22 20 18 16 14 12 10-8 6
4 2-fl
- .1 0
5 b
10 A5 2
253 354
-4
ý 5
50o Average Planr E.xposure',GWd/vITU Figure 7.1 Protection Against Power Transient LHGR Limit for ATRIUM-gB Fuel (019.A)
(15,19.4) 3 EMF-2437 Revision 0 Page 7-3 70 65 55 60
ý 0
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 8-1 8.0 Methodology References See XN-NF-80-19(P)(A) Volume 4 Revision 1 for a complete bibliography.
8.1 ANF-913(P)(A) Volume I Revision I and Volume 1 Supplements 2, 3 and 4, COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.
- 82.
ANF-524(P)(A) Revision 2 and Supplements 1 and 2, ANF Critical Power Methodology for Boiling Water Reactors, Advanced Nuclear Fuels Corporation, November 1990.
8.3 ANF-1 125(P)(A) and ANF-1 125(P)(A), Supplements I and 2, ANFB Critical Power Correlation, Advanced Nuclear Fuels Corporation, April 1990.
8.4 EMF-1 125(P)(A), Supplement 1 Appendix C, ANFB Critical Power Correlation Application for Co-Resident Fuel, Siemens Power Corporation, August 1997.
8.5 ANF-1 125(P)(A), Supplement 1 Appendix E, ANFB Critical Power Correlation Determination of ATRIUM-9B Additive Constant Uncertainties, Siemens Power Corporation, September 1998.
8.6 XN-NF-80-19(P)(A) Volume I Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Methodology for Boiling Water Reactors:
G Benchmark Results for CASMO-3GIMICROBURN-B Calculation Methodology, Advanced Nuclear Fuels Corporation, November 1990.
8.7 EMF-CC-074(P)(A) Volume 1, STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain, and Volume 2, STAIF - A Computer Program for BWR Stability Analysis in the Frequency Domain - Code Qualification Report, Siemens Power Corporation, July 1994.
EMF-2437 LaSalle Unit 2 Cycle 9 Revision 0 Reload Analysis Page 9-1
-9.0 Additional References 9.1 EMF-2404(P) Revision 1, Fuel Design Report for LaSalle Unit 2 Cycle 9 ATRIUMh-g9B Fuel Assemblies, Siemens Power Corporation, September 2000.
9.2 ANF-89-014(P)(A) Revision -I and Supplements 1 and 2, Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels 9x9-IX and 9x9-9X BWR Reload Fuel, Advanced Nuclear Fuels Corporation, October 1991.
9.3 EMF-2440 Revision 0, LaSalle Unit 2 Cycle 9 Plant Transient Analysis, Siemens Power Corporation, October 2000.
9.4 EMF-95-134(P), Criticality Safety Analysis for ATRIUMTu-9B Fuel, LaSalle Units I and 2 New Fuel Storage Vault, Siemens Power Corporation, December 1995.
9.5 EMF-96-117(P) Revision 0, Criticality'Safety Analysis for ATRIUM"-9B Fuel, LaSalle Unit I Spent Fuel Storage Pool (BORAL Rack), Siemens Power Corporation, April 1996.
9.6 EMF-95-088(P) Revision 0, Criticality Safety Analysis for ATRIUMIr-9B Fuel, LaSalle Unit 2 Spent Fuel Storage Pool (Boraflex Rack), Siemens Power Corporation, February 1996.
9.7 EMF-95-205(P) Revision 2, LaSalle Extended Operating Domain (EOD) and Equipment Out of Service (EOOS) Safety Analysis for ATRIUMT h-9B Fuel, Siemens Power Corporation, June 1996.
9.8 EMF-2174(P), LOCA Break Spectrum Analysis for LaSalle Units 1 and 2, Siemens Power Corporation, March 1999.
9.9 EMF-2175(P), LaSalle LOCA-ECCS Analysis MAPLHGR Limits for ATRIUM'r-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 1 Cycle 9 ATRIUMDA-9B Fuel Assemblies, Siemens Power Corporation, September 1999.
Siemens Power Corporation
LaSalle Unit 2 Cycle 9 Reload Analysis K-)
Distribution D. G.
D.E.
M.E.
J.M.
D.B.
Carr, 23 Garber, 38 (9)
Garrett, 23 Haun, 34 McBumey, 23 Notification List (e-mail notification)
O.C. Brown J. A. White P.D. Wimpy a---
EMF-2437 Revision 0
Technical Requirements Manual - Appendix J L2C9 Reload Transient Analysis Results LaSalle Unit 2 Cycle 9 Plant Transient Analysis LaSalle Unit 2 Cycle 9 A'ugust 2002
SIEMENS LaSalle Unit 2 Cycle 9 Plant Transient Analysis October 2000 EMF-2440 Revision 0
-V Siemens Power Corporation Nuclear Division
Siemens Power Corporation DOCUMENT SYSTEM DATE:
LaSalle Unit 2 Cycle 9 Plant Transient Analysis D. B. McBurney, Engineer BWR Safety Analysis D. G. Carr, Team Leader BWR Safety Analysis Pr*uct Licensing Approved:
')LMJT O iIi w O. C. Brown, Manai BWR Neutronics M. E."Garrett, Manage,
'I Safety`Analysis o,.
D. J.
lynver,rranager Commercial Operations Date Iat-e Date Date
/0/3/6 Date' Date Date EMF-2440 Revision 0 Prepared:
Reviewed:
Concurred:
Approved:
Approved:
paj
,4- &B
Customer Disclaimer 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 those set forth in the agreement between Siemens Power Corporation and the Customer pursuant to which this document is issued. Accordingly, except as otherwise expressly provided in such agreement, neither Siemens Power Corporation nor any person acting on its behalf:.
- a.
makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights; or
- b.
assumes any liabilities with respect to the use of. or for damages resulting from the use of, any information, apparatus, method, or process disclosed in this document The information contained herein is for the sole use of the Customer.
In order to avoid impairment of rights of Siemens Power Corporation in patents or inventions which may be included in the information contained in this document, the recipient, by its acceptance of this document, agrees not to publish or make public use (in the patent use of the term) of such infoirmation until so authorized in writing by Siemens Power Coiporation or until after six (6)_months following termination or expiration of the aforesaid Agreement and any extension thereof, unless expressly provided in the Agreement No rights or licenses in or to any patents are implied by the furnishing of this document
LaSalle Unit 2 Cycle 9 DI0#- nneiant Anliv Nature of Changes Item Page Description and Justification
- 1.
All This is a new document.
Siemens PowerCorporation EMF-2440 Revision 0 Page i
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page ii Contents 1.0 Introduction..................................................................................................................
1-1 2.0 Summary.....................................................................................................................
2-1 3.0 Transient Analysis for Therm al 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 Limits...................................................
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...........................
r "
5.1 Feedwater Heaters Out-of-Service (FHOOS)..................................................
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 FHOOSrTCV Slow Closure and/or No RPT.....................................
5-4 6.0 Transient Analysis for Thermal Margin - EODIEOOS Combinations........................ 6-1 6.1 Coastdown W ith EOOS....................................................................................
6-1 6.1.1 Coastdown With Feedwater Heaters Out-of-Service..........................
6-1 6.1.2 Coastdown W ith One Recirculation Loop...........................................
6-1 6.1.3 Coastdown W ith TBVOOS.................................................................
6-2 6.1.4 Coastdown W ith No RPT..............................................................
6-2 6.1.5 Coastdown With Slow Closure of the Turbine Control Valve..................................................................................................
6-2 6.2 Combined FFTRlCoastdown W ith EOOS.........................................................
6-3 6.2.1 Combined FFTR/Coastdown With One Recirculation Loop...................................................................................................
6-3 6.2.2 Combined FFTRlCoastdown With TBVOOS......................................
6-3 6.2.3 Combined FFTRPCoastdown With No RPT.........................................
6-4 6.2.4 Combined FFTR/Coastdown With Slow Closure of the Turbine Control Valve........................................................................
- J
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440 Revision 0 Paqe iii Contents (Continued) 7.0 Maximum Overpressurization Analysis;;;...............
7-1 7.1 Design Basis....................................................................................................
7-1 7.2 Pressurization Transients....................................
........................................ 7-1 8.0 References 8-1 Appendix A Power-Dependent LHGR Limit Generation...................................................
A-I 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 MCPRp 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 MCPRp Limits and LHGRFACý Multipliers for TSSS Insertion Times.........................................................
2-6 2.4 FFTRPCoastdown 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 FFTRJCoastdown Operation Transient Results..........................................................
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 FFTR/Coastdown Turbine Bypass Valves Out-of-Service Analysis Results................. 6-8 6.5 FFTR/Coastdown Recirculation Pump Trip Out-of-Service Analysis Results........................................................................................................................
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
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page v Figures 1.1 LaSalle County Nuclear Station Power I 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-9B Fuel.....................................
2-11 3.1 EOC Load Rejection No Bypass at 1001105 -TSSS Key Parameters.......................
3-15 3.2 EOC Load Rejection No0Bypass at 100/105 -TSSS Vessel Water Level.................. 3-16 3.3 EOC Load Rejection No Bypass at 100/1105 -TSSS Dome Pressure........................
3-17 3.4 EOC Feedwater Controller Failure at 100/105 - TSSS Key Parameters.................... 3-18 3.5 EOC Feedwater Controller Failure at 1001105 - TSSS Vessel Water Level......
........... ;................................................................................. 3-19 3.6 EOC Feedwater Controller Failure at 100/105-TSSS Dome Pressure.................... 3-20 3.7 Radial Power" Distribution for SLMCPR'Determination............
3-21 3.8 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-391B-14GB.O-1 0DM With Channel Bow (Assembly Exposure of 18,O00MWd/MTU)..............................................................
3-22 3.9 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-410B-1i9G8.0-1 0DM With Channel Bow (AssemblyExposure of 17,500 MWd/MTU)...............................................................
3-23.
3.10 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-383B-16G8.0-1 0DM With Channel Bow (Assembly Exposure of 17,500 MWd/MTU).............................................................
3-24 3.11 LaSalle Unit 2 Cycle 9 Safety Limit L*cal Peaking Factors SPCA9-396B-12GZ-1 0DM With Channel Bow (Assembly Exposure of 15,000 MWdMTU)..............................
325 3.12 EOC Base Case Power-Dependent MCPR Limits for ATRUM-9B Fuel TSSS Insertion Times................
....................................................... 3-26 3.13 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel
'TSSS Insertion Times.................................................................... 3-27 3.14 EOC Base Case Power-Dependent MCPR Limits for ATRUM-9B Fuel NSS Insertion Times..................................................................
3-28 3.15' :EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel NSS Insertion Times.......... ;......................................
....... 3-29 3.16 -EOC'Base Case Power-Depende~nt LHGR Multipliers for ATRUM-gB Fuel TSSS Insertion Times..
................. 3-30 3.17 EOC Base Case Power-Dependent LHGR Multipliers for ATRUM-9B
- Fuel-NSS Insertion Times....
. 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 Limbts for GE9 Fuel..........................................
4-7 4.4 FFTRlCoastdown Power-Dependent MCPR Limits for ATRUM-9B Fuel.................. 4-8 4.5 FFTRlCoastdown Base Case Power-Dependent LHGR Multipliers for ATRUM-9B Fuel..........................................................................................................
4-9 4.6 FFTRlCoastdown Power-Dependent MCPR Limits for GE9 Fuel..............................
4-10
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page vi Figures (Continued) 5.1 EOC Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel..................................................................................................
5-11 5.2 EOC Feedwater Heaters Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel.................................................................................
5-12 5.3 EOC Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel..............................................................................................................
5-13 5.4 Abnormal Idle Recirculation Loop Startup Power-Dependent MCPR Limits for ATRIUM-gB Fuel.............................................................................................
5-14 5.5 Abnormal Idle Recirculation Loop Startup Power-Dependent LHGR Multipliers for ATRIUM-gB Fuel.................................................................................
5-15 5.6 Abnormal Idle Recirculation Loop Startup Power-Dependent MCPR Limits for GE9 Fuel ;............................................................................................................
5-16 5.7 EOC Turbine Bypass Valves Out-of-Service Power-Dependent MCPR Limits forATRIUM-gB Fuel.......................................
............................................ 5-17 5.8 EOC Turbine Bypass Valves Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-gB Fuel.................................................................................
5-18 5.9 EOC Turbine Bypiss Valves Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel..................................................
- ................................................ 5-19 5.10 EOC Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-SB Fuel........................................................................................
5.11 EOC Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel.................................................................................
5-21 5.12 EOC Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel......
5-22 5.13 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel................... 5-23 5.14 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel.................. 5-24 5.15 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel....................................
5-25 5.16 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip and Feedwateri Heaiters Out-of-Service Power-Dependent MCPR Limits for ATRIUM-9B Fuel...............................................................................................
5-26 5.17 EOC Turbine 'Control Valve Slow Closure and/or Recirculation Pump Trip and Feedwater Heaters Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-SB Fuel.............................................................................
5-27 5.18 EOC Turbine Control Valve Slow Closure and/or Recirculation Pump Trip and Feedwater Heaters Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel..............................................................................................................
5-28
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 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-9B 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 FFTRiCoastdown Turbine Bypass Valves Out-of-Service Power Dependent MCPR Limits for ATRIUM-9B Fuel..........................................................
6-20 6.11 FFTRPCoastdown 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 FFTR/Coastdown Recirculation Pump Trip Out-of-Service Power Dependent MCPR Limits for ATRIUM-9B Fuel..........................................................
6-23 6.14 FFTRlCoastdown Recirculation Pump Trip Out-of-Service Power Dependent LHGR Multipliers for ATRIUM-9B Fuel....................................................
6-24 6.15 FFTR/Coastdown Recirculation Pump Trip Out-of-Service Power Dependent MCPR Limits for GE9 Fuel......................................................................
6-25 6.16 FFTRlCoastdown 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 FFTRPCoastdown Turbine Control Valve Slow Closure and/oi Recirculation Pump Trip Out-of-Service Power-Dependent LHGR Multipliers for ATRIUM-9B Fuel.................................................................................
6-27 6.18 FFTRlCoastdown Turbine Control Valve Slow Closure and/or Recirculation Pump Trip Out-of-Service Power-Dependent MCPR Limits for GE9 Fuel
.................................................. 6-28 EA
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page viii Figures (Continued) 7.1 Overpressurization Event at 1021105 - MSIV Closure Key Parameters.......................... 7-3 7.2 Overpressurization Event at 102/105 - MSIV Closure Vessel Water Level..................... 7-4 7.3 Overpressurization Event at 102/105 - MSIV Closure Lower-Plenum Pressure..............
7-5 7.4 Overpressunzation Event at 1021105 - MSIV Closure Dome Pressure........................... 7-6 7.5 Overpressurization Event at 1021105 - MSIV Closure Safety/Relief Valve Flow Rates...................................................................................................................
7-7
LaSalle Unit 2 Cycle 9 ri@ILIIO anII
~Iayu AOO CornEd CPR EFPH EOC EOD EOFP EOOS FHOOS FWCF HFR ICF L2C9 LFWH LHGR LHGRFACf LHGRFACý LHGROL LPRM LRNB MCPR MCPPr MCPRp MELLLA MFC MSIV NSS PAPT RPT SLMCPR SLO SPC SRV SRVOOS SSLHGR Nomenclature anticipated operational occurrence Commonwealth Edison Company critical power ratio effective full power hours end of cycle extended operating domain end of full power equipment out-of-service final feedwater temperature reduction feedwater heater out-of-service feedwater controller failure heat flux ratio increased core flow LaSalle Unit 2 Cycle 9 loss-of-feedwater heating linear heat generation rate flow-dependent linear heat generation rate factors power-dependent linear heat generation rate factors linear heat generation rate operating limit local power range monitor generator load rejection with no bypass minimum critical power ratio flow-dependent minimum critical power ratio power-dependent minimum critical power ratio maximum extended load line limit analysis manual flow control main steam isolation valve nominal scram speed protection against power transient recirculation pump trip safety limit MCPR single-loop operation Siemens Power Corporation safety/relief valve.
safety/relief valve out-of-service steady-state LHGR EMF-2440 Revision 0 Page ix
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis 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
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 1-1 1.0 Introduction This report presents results of the plant transient analyses performed by Siemens Power Corporation (SPC) as part of the reload safety analyses to support LaSalle; Unit 2 Cycle 9 (L2C9) operation. The Cycle 9 core contains 348 fresh ATRIUMTm-9gB 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 CornEd 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 operation' 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 (MCPRt) 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.
I Base case operation is defined as two-loop operation within the standard operating domain, including the ICF and MELLLA regions, with all equipment in-service.
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 1-2 (LHGRFACp. and LHGRFAC,, respectively). These factors or multipliers are applied directly to the steady-state LHGR limit to ensure that the LHGR does not exceed the protection against power transient (PAPT) limit during postulated AO0s. Cycle 9 power-and flow-dependent LHGR multipliers are presented for ATRIUM-9B 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.
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 liinit analysis (MELLLA)
Coastdown Final feedwater temperature reduction (FFTR)
Combined FFTR/coastdown Equipment Out-of-Service (EOOS) Conditions*
Feedwater heaters cut-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-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 channels) and/or up to 500A of the LPRMs out-of-service is supported.
LaSalle Unit 2 Cycle 9 Plant Transient Analysis 110 100 CL V.
EMF-2440 Revision 0 Page 1-4 i) 4 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 The determination of the thermal limits (MCPR limits and LHGRFAC multipliers) 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 18-month 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 pbwer 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 EODIEOOS conditions presented in Table 1.1.
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 EOD/EOOS scenarios are presented in Tables 2.1-2.4. Tables 2.1 and 2.2 present base case limits and multipliers for Technical Specifications 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 FFTR/coastdown 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 MCPRr 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.
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 2-2 The Cycle 9 LHGRFACf 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.
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.
.A~man Pnr.r f.nrutinn
LaSalle Unit 2 Cycle 9 Din *mnaifl *nm c.
EMF-2440 Revision 0 Paoe 2-3 Table 2.1 EOC Base Case and EOOS MCPRp Limits and LHGRFACP Multipliers for TSSS Insertion Times*
EOOS I EOD Power ATRIUM-9B Fuel-GE9 Fuel Condition
(% rated)
- MCPP*
LHGRFAC*
MCPRp 0
2.70 0.78 2.70 Base 25 2.20 0.78 2.20 case 25 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 25 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 25 1.98 0.76 2.08 out-of-service (TBVOOS) 60 1.52 0.97 1.62 "100 1.43 0.99 1.52 Limits support operation with any combination of I SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), up to a 20F reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF, and MELLLA regions of the power/flow map.
Z=
LaSalle Unit 2 Cycle 9 Plant Transient Analysis Table 2.1 EOC Base Case and EOOS MCPRp Limits and LHGRFACp Multipliers for TSSS Insertion Times (Continued)
EoosIEOD Power ATRIUM-9B Fuel GE9 Fuel Condition
(% rated)
MCPRp
- LHGRFAC, 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 AND/OR 80 1.69 0.86 1.95 no RPT 80 1.61 0.89 1.84 100 1.53 0.89 1.63 0
2.85 0.68 2.85 TCV 25 2.35 0.68 2.35 slow closure/
25 2.14 0.68 2.22 FHOOS AND/OR 80 1.69 0.86 1.95 no RPT 80 1.61 0.89 1.84 100 1.53 0.89 1.63 0
2.60 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 iUmits support operation with any combination of 1 SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), up to a 20"F reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF, and MELLLA regions of the power/flow map.
EMF-2440 Revision 0 Paoe 2-4 S.
e 2 -4 I
LaSalle Unit 2 Cycle 9 ian-i I
adIl~lIl i-,I*
b I'uy" EMF-2440 Revision 0 Paoe 2-5 Table 2.2 EOC Base Case MCPRp Limits and LHGRFACp Multipliers for NSS Insertion Times*
ATRIUM-9B Fuel GE9 Fuel Condition
(% rated)
MCPRi LHGRFACý MCPR1 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 1.51 100 1.39 1.00 1.48 LUmits support operation with any combination of 1 SRVOOS, up to 2 TIPOOS (or the equivalent number of TIP channels), up to a 20*,F reduction in feedwater temperature (except for conditions with FHOOS), and up to 50% of the LPRMs out of service in the standard, ICF, and MELLLA regions of the powerlfiow map.
Siemens Pomr Copom
LaSalle Unit 2 Cycle 9 DI 1ft*%
Trr,!,neion# Anniue
- o;c
- e.
EMF-2440 Revision 0 Paoe 2-6 K>
Table 2.3 Coastdown Operation Base Case and EOOS MCPRp Limits and LHGRFACp Multipliers for TSSS Insertion Times*
EQOS EOD Power ATRIUM-9B Fuel GE9 Fuel Condition
(% rated)
MCPR9 LHGRFACP MCPRp 0
2.70 0.75 2.70 Coastdown 25 2.20 0.75 2.20.
base case 25 2.05 0.75 2.05 operation 60 1.48 0.99 1.54 100 1.42 1.00 1.52 0
2.71 0.75 2.71 Coastdown with 25 2.21 0.75 2.21 single-loop 25 2.06 0.75 2.06 operation 60 1.49 0.99 1.55 100 1.43 1.00 1.53 0
2.70 0.73 2.70 Coastdown with 25 2.20 0.73 2.20 turbine bypass valves 25 2.05 0.73 2.15 out-of-service 60 1.55 0.97 1.64 100 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 0.68 1.67 Limits support operation with any combination of I SRVOOS, up to 2,TIPOOS (or the equivalent number of TIP channels), up to a 20*F reduction in feedwater, and up to 50% of the LPRMs out of service in the standard, ICF, and MELLLA regions of the powertflow map.
I i~ll I~lltl~lL
~
lilOIZ2
LaSalle Unit 2 Cycle 9 EMF-2440 Revision 0 Page 2-7 Table 2.3 Coastdown Operation BaseCase and EOOS MCPRP Limits and LHGRFACp Multipliers for TSSS Insertion Timese (Continued)
EOOSEOD
-Power ATRIUM-9B Fuel -
GE9 Fuel Condition
(% rated)'
MCPRp LHGRFACv MCPRp 0
2.70 0.68 2.70 Coastdown with 25 2.20 0.68 2.20 turbine control valve (TCV) 25 2.15 0.68 2.15 slow closure 80 1.70 0.85 1.96 AND/OR 80 1.62 0.88 1.85 no RPT 100 1.55 0.88 1.67 0
2.60 0.40 2.60 Coastdown with 25 2.60
-0.40 2.60 idle 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 20F 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.
dL=
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440 Revision 0 Paoe 2-8 Q-)
Table 2.4. FFTRlCoastdown Operation Base Case and EOOS MCPR, Limits and LHGRFACp Multipliers for TSSS Insertion Timesr EOOS I EOD Power ATRIUM-9B Fuel GE9 Fuel Condition
(% rated)
MCPP4 LHGRFACý MCPRP 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 25 2.36 0.65 2.36 with single-loop 25 2.31 0.65 2.31 operation 60 1.57 0.97 1.60 100 1.43 1.00 1.53 0
2.85 0.65 2.85 FFTR/coastdown 25 2.35 0.65 2.35 with turbine bypass valves 25 2.30 0.65 2.30 out-of-service 60 1.57 0.97 1.64 (TB'VOOS) 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) 1 100 1.55 0.88 1.67 ULmits 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.
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440 Revision 0 Page 2-9 Table 2.4 FFTRlCoastdown Operation Base Case and EOOS MCPRP Limits and LHGRFAC, Multipliers for TSSS Insertion Times*
(Continued)
EOOS/EOD Power ATRIUM-9B Fuel GE9 Fuel Condition
(% rated)
MCPR%
LHGRFACj MCPFR 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 ANDlOR noDRPT 80 1.62 0.88 1.85 100 1.55 0.88 1.67 0
2.60 0.40 2.60 FFTRPcoastdown 25 2.60 0.40 2.60 with idle 25 2.60 0.40 2.60 loop startup 60 2.60 0.40 2.60 100 2.60 0A0 2.60 SLimits 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 m the standard, ICF, and MELLLA regions of the power/flow map.
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440 Revision 0 Pane 2-10 0
10 20 SO 40 so go 70 so go 100 110 Plo M%
of ROOM MCPRrGE9 Flow MCPRf (penalty
(% of rated)
ATRIUM-9B included) 0 1.60 1.66 30 1.60 1.66 105 1.11 1.11 Figure 2.1 Flow-Dependent MCPR Limits for Manual Flow Control Mode e:-
ml-K
LaSalle Unit 2 Cycle 9 Plant Transient Analysis U.
C, 63 40 50 60
_7O Percent of Rated flow Flow
(% rated)
'LHGRFACI 0'
- 0.69 30
'0.69 76 o
105 1.0D0 Figure 2.2 Flow-Dependent LHGRFAC Multipliers for ATRIUM-9B Fuel Siemens Power Comoration EMF-2440 Revision 0 Page 2-11
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-1 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 CASMO-3G/MICROBURN-B (Reference 3) are the major codes used in the thermal limits analyses as described in SPC's THERMEX methodology repoit (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 of 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 critical power 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 responsibility 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 transient analyses are shown in Table 3.1 for the 100%
power/1 00% flow conditions. Additional plant parameters used in the analyses are presented in Reference 8. Analyses have been performed to determine power-dependent MCPR and LHGR limits that protect operation throughout the power/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. As a result, all analyses used to establish base case e
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 CaDoration
I EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-2 The limiting 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 1 SRV out-of-service, the pressurization tran'sient 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 20°F 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-specific MCPRk 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 LHGRFACn multipliers for base case operation for both NSS and TSSS insertion times.
3.1.1 Load Reiection No Bypass 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
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 high-to low-speed transfer which helps mitigate the pressurization effects.
Turbine bypass system operation, which also mitigates the consequences of 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 delays between the two events. This has been demonstrated in calculations that support the Reference 9 conclusion that the TTNB event is bound by the LRNB event.
LRNB analyses were performed for several powerfilow conditions to support generation of.'
thermal limits. Table 3.3 presents the LRNB transient results for both TSSS and NSS inserJrtn times for Cycle 9. For illustration, Figures 3.1-3.3 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 The increase 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 water level is conservatively assumed to be at the lower level operating range at 30 inches above instrument zero to delay the high level trip and maximize the core inlet subcooling that results from the FWCF. The high water level trip causes the turbine stop valves to 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 core causing a void collapse and subsequent rapid power excursion. The closure of the turbine valves initiates a reactor scram and a recirculation pump high-to Ioii-f speed transfer. In addition, the turbine bypass valves are assumed operable and provide some
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-4 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 34 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 CoinEd has the analysis responsibility for the loss-of-feedwater heating (LFWH) 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 powertflow 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.
The possible change in feedwater temperature'during an LFWH event decreases as the reactor power decreases.
3.2 MCPR Safety Limit 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 discussionfof how the effects of channel bow are accounted for is presented in Reference 5. The main input parameters and uncer'tainties used in the safety limit analysis'aie listed in Table 3.5. The radial power uncertainty includes the effects of up to 2 TIPOOS or the equivalent nuriber 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
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-5
<9 uncertainty is a function of the nominal and bowed local peaking factors and the standard deviation of the measured bow data.
The determination of 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 produced the highest number 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 0 (SPCA9-391B-14G8.0-10OM, SPCA9-410B-19G8.0-100M, SPCA9-383B-16G8.0-1OOM, a.n SPCA9-396B-12GZ-1 00M); two ATRIUM-9B fuel types loaded in Cycle 8 (SPCA9-381 B-1 3GZ7 80M and SPCA9-384B-11GZ6-80M); and two GE9 fuel types loaded in Cycle 7 (GE9B P8CWB322-11 GZ-10DM-1 50 and GE9B-P8CWB320-9GZ-1 0DM-150).
The local power peaking factors, including the effects of channel bow, at 70% void and assembly exposures consistent with a Cycle 9 exposure of 15,000 MWd/MTU are presented in Figures 3.8 through 3.11 for the Cycle 9 SPC ATRIUM-9B fuel. The bowed local peaking factor data used in the MCPR safety limit analysis for fuel type SPCA9-391B-14G8.0-100M is at an assembly average exposure of 18,000 MWd/MTU. The data for fuel types SPCA9-410B 19G8.0-100M and SPCA9-383B-16G8.0-100M is at an assembly average exposure of 17,500 MWdIMTU. The data is at an assembly average exposure of 15,000 MWd/MTU for fuel type SPCAg-396B-12GZ-100M.
The results of the analysis support a two-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 startup witt' uncalibrated LPRMs for an exposure range of BOC to 500 MWd/MTU.
S...... Sknem Power Conmomin
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-6 3.3 Power-Dependent MCPR and LHGR Limits Figures 3.12 and 3.13 present the base case operation TSSS ATRIUM-9B and GE9 MCPRp limits for Cycle 9. Figures 3.14 and 3.15 present the ATRIUM-9B and GE9 MCPRP limits for base case operation with NSS insertion times. The limits are based on the eACPR results from the limiting system transient analyses discussed above and a MCPR safety limit of 1.11.
Relative to the TSSS MCPRi 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 LHGRFACý multipliers to the steady-state LHGR limit ensures that the LHGR during ACOs 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-9B LHGRFACp multipliers for Cycle 9 TSSS and NSS insertion times are presented in Figures 3.16 and 3.17, respectively.
3.4 Flow-Dependent MCPR 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/1 05%F.
MCPRf 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 MCPRf 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 Semnms Power Cowraton
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-7 determine the 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 and GE9 fuel. MCPRf limits 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 EODIEOOS 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 MCPR, 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 LHGRFACI analyses with the CASMO-3G/MICROBURN-B core simulator codes. The analysis assumes that the recirculation flow increases slowly along the limiting roC 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 LHGRFACI 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 LHGRFACI 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 Nuclear Instrument 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 desi' of the fuel assembly, the in-channel void fraction, and the fuel exposure. The neutron lifetinlnr>
are similar for the SPC and GE LaSalle fuel with typical values of 39(10 ) to 40(104) seconds
ý2__
K%_.
t%
EMF-2440 LaSalle Unit 2 Cycle 9 Revision 0 Plant Transient Analysis Page 3-8 for the ATRIUM-9B lattices and 41 (104) to 43(1 O1) 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-9B and GE9 fuel, and a full core of GE9 fuel are essentially equivalent.
LaSalle Unit 2 Cycle 9 DInr 1# -'
neihnt Anniu ie Table 3.1 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 Mlbm/hr Core bypass flow*
14.8 Mlbmlhr Core inlet enthalpy 523.9 Btu/ibm Vessel pressures Steam dome 1001 psia Core exit (upper-plenum) 1013 psia Lower-plenum 1038 psia Turbine pressure 948 psia Feedwater I steam flow 15.145 Mlbm/hr Feedwater enthalpy 406.6 Btullbm Recirculating pump flow 15.83 Mlbm/hr (per pump)
-Core average gap 1162 Btu/hr-ft 2-OF coefficient (EOC)
Includes water channel flow.
EMF-2440 Revision 0 Paae 3-9 "I
Ica1 Z =ilil
,li~el
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440 Revision 0 Page 3-10 Table 3.2 Scram Speed Insertion Times Control Rod TSSS NSS Position Time Time (notch)
(sec)
(sec) 48 (full-out) 0.00O 0.000 48*
0.200*
0.200*
45 0.430 0.380
.. 39 0.860 0.680
-25
_1.930 1.680 5
3.490 2.680 0 (fulý-in) 3.880 2.804 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).
LaSalle Unit 2 Cycle 9 0l-q,*t "
e,**;Ont Annl
~ie EMF-2440 Revision 0 Paoe 3-11 A)
Table 3.3 EOC Base Case LRNB Transient Results Peak Peak Power/
ATRIUM-9B ATRIUM-9B GE9 Neutron Flux Heat Flux Flow ACPR LHGRFACp 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 801105 0.29 1.04 0.39 324 100 80157.2 0.29 1.05 0.39 265 96 60/ 105 60/35.1 0.27 1.06 0.36 245
- 4.
- 4.
4 4
4.
0.17 1.13 0.21 96 73 63 40/105 0.23" 1.13 0.27 100" 46" 251105 0O.1r 1.22*
0.19*
44 27*
NSS Insertion Times 100/105 0.28 1.02 0.37 380 124 100/81 0.22 1.03 0.30 358 120 80/105 0.27 1.04 0.36 302 95 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 1.15 0.14 76 60 40 1105 0.20 1.14 0.27 115 47 25/105 0.15-1.22 0.17 42r 27*
The analysis results are from an earlier cycle exposure. The ACPR and LHGRFACp results are conservatively used to establish the thermal limits.
4a 193"l Z=II*i~lI*tlJ*
l
LaSalle Unit 2 Cycle 9 931ft# rs iar,~
nwt Anoklcie
- 5F.US S
EMF-2440 Revision 0 Paoe 3-12 Table 3.4 EOC Base Case FWCF Transient Results Peak Neutron Peak Power/
ATRIUM-9B ATRIUM-9B GE9 Flux Heat Flux Flow ACPR LHGRFACý ACPR
(% rated)
(% rated) 7TSSS Insertion Times 1001105 0.25 1.09 0.31 298 123 1001 100 0.24 1.11 0.31 288 122 100181 0.23 1.09 0.28 285 121 801105 0.28 1.07 0.35 253 101 80/57.2 0.19 1.16° 0.23
.154 91 60 105 0.35' 1.02' 0.41 154' 77' 60135.1 0.11 1.25 0.14 74 63 40 /105 I
0.51' 0.94-0.57 104-51r 25/105 0.80' 0.79-0.88-69*
44" NSS Insertion Times 1001105 0.23 1.10 0.29-263 120 100/81 0.18 1.11 0.22 237 116 80/105, 0.27 1.10 0.33,
235 99 80-057.2 0.15 1.20 07.17
.131 88 60/105 0.33 1.05""
0.40 188 79 60/35.1
-0.11 1.28- -
0.13 65 63 40/105-..-0.48' 0.95' 0.55' 96',,
57' 25/105 0.78' o.79' o.86-,
66' 44-The analysis results are from an earlier cycle exposure. The ACPR and LHGRFACp results are conservatively used to establish the thermal Omits.
Siemens Power Cornoration 9 43",ý L=
LaSalle Unit 2 Cycle 9 IDlestv l'rr-ine;anf Anftl*,
- I*
- l z
- l.
il-EMF-2440 Revtsion 0 Paae 3-13
<0j Table 3.5 Input for MCPR Safety Limit Analysis Fuel-Related Uncertainties Source Statistical Parameter Document Treatment ANFB correlation° ATRIUM-9B Reference 17 Convoluted GE9 Reference 12 Convoluted Radial power 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 peaking and standard deviation of bow data (see Reference 18)
Convoluted1 114 1 Nominal Values and Plant Measurement Uncertainties Uncertainty (%)
Statistical Parameter Value (Reference 8)
Treatment Feedwater flow ratet (Mlbmnhr) 22.4 1.76 Convoluted Feedwater temperature (OF) 426.5 0.76 Convoluted Core pressure (psia) 1031.35 0.50 Convoluted Total core flow (Mlbmlhr) 113.9 2.50 Convoluted Core powerl (MWth) 5167.29 Additive constant uncertainties values are used.
j Feedwater flow rate and core power were increased above design values to attain desired core MCPR for safety limit evaluation consistent wit Reference 5 methodology 0
t
LaSalle Unit 2 Cycle 9 931f~t "r
&,cm;nt Anot ic Table 3.6 Flow-Dependent MCPR Results 105%
Core Maximum Core Flow Flow
(% rated)
GE9 ATRIUM-9g 30 1.52 1.52 40 1.46 1.46 50 1.41 1.42 60 1.37 1.38 70 1.31 1.32 80 1.26 1.27 90 1.20 1.21 100 1.14 1.14 105 1.11 11 SwnmPtnwir Cwmnwnuin 163
- t
ý
- i EMF-2440 Revision 0 Paae 3-14
LaSalle Unit 2 Cycle 9 Plant Transient Analysis w
U.
0 I.
z CJ w
W In TIME, SECONDS Figure 3.1 EOC Load Rejection No Bypass at 1001105 - TSSS Key Parameters EMF-2440 Revision 0 Page 3-15
LaSalle Unit 2 Cycle 9 Plant Transient Analysis 0
N I
0 z
-J a:
LLI w
w I
-_J LJ U)
UI)
.0 1.0
- u.
5.0 4.0 "IME, SECONDS Figure 3.2 EOC Load Rejection No Bypass at 1001105 - TSSS Vessel Watei Level Siemens Power Corpoiuion EMF-2440 Revision 0 Page 3-16
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440 Revision 0 Page 3-17
<9 U) M Q.
Li a:
VIA D
En w Ir 0
a*
TIM.
SECONDS Figure 3.3 EOC Load Rejection No Bypass at 1001105 - TSSS Dome Pressure F;mnwm I~Per evrnwatinl
LaSalle Unit 2 Cycle 9 Plant Transient Analysis CORE POWER HEAT FLUX CORE FLOW STEAM FLOW FEED FLOW Mf
.0 U
ý mftw
!0v TA C15.0 rJEScot*
20.
Figure 3.4 EOC Feedwater Controller Failure at 1001105,, TSSS Key Parameters Rim mum Pvmmr tnaaywvmfntn EMF-2440 Revision 0 Page 3-18 Alu1J 2=1 C
LL I-0 z
w IL 0.
imn Ir
, 1/
3M 2!
LaSalle Unit 2 Cycle 9 Plant Transient Analysis 15.0 TIE. SECONDS Figure 3.5 EOC Feedwater Controller Failure at 1001105 - TSSS Vessel Water Level N'
R.inumw Poawm CmnnmraM EMF-2440 Revision 0 Paqe 3-19 Paqe 3-19
j) 0 m
N U, z_
w 0
z
-j w
-J w
w I,
-J Id
- LaSalle Unit 2 Cycle 9 Plant Transient Analysis 5
In IL 0
ai TIE, SECONDS Figure 3.6 EOC FeedwaterController Failure at I001105-TSSS Dome Pressure Siemens Power Corporaton EMF-2440 Revision 0 Page 3-20
LaSalle Unit 2 Cycle 9 Plant Transient Analysis 200 EMF-2440 Revision 0 Pane 3,-21 I
I I
I II I
i i
I i
I I
I I
T Ff.1 175 [
150 "o 125 C
m 100 L
U)
.0 E 75 z
50 25 0
.0.
.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 3.7 Radial Power Distribution for SLMCPR Determination Siemens Power Comobtion 4 &
LaSalle Unit 2 Cycle 9 1
ol Rod Corner Figure 3.8 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCAS-3911B-14G8.0-1 0DM With Channel Bow bly Exposure of 18,000 M W dIM TU )
o nt r EMF-2440 Revision 0 Paae 3-22 C
a n
t r
0 I
R a
d C
0 r
n e
r 1.052 1.045 1.088 1.088 1.104 1.079 1.068 1.013 1.005 1.045 0.951 1.019 0.996 0.852 0.986 0.998 0.914 0.991 1.088 1.019 1.001 1.059 1.089 1.051
'0.982 0.981 1.027 1.088 0.996 1.059 0.905 0.957 1.050 Internal 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 0.956 r-idilL 10" Vt.8I*I "a
=l l
LaSalle Unit 2 Cycle 9 Plant Transient Analysis o ntr oI Rod Corner Figure 3.9. LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-410B-19G8.0-1DOM With Channel Bow (Assembly Exposure of 17,500 MWdIMTU)
EMF-2440 Revision 0 Pane 3-23 K)
C 0
n t
r 0
R 0
d C
0 r
n e
r 1.058 1.049 1.092 1.091 1.107 1.082 1.072 1.017 1.010 1.049 0.945 1.020 0.996 0.843 0.987 0.998 0.906 0.995 1.092 1.020 1.002 1.061 1.090 1.052 0.981 0.980 1.030 1.091 0.996 1.061 0.894 0.955 1.053 Internal 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 1.067 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
LaSalle Unit 2 Cycle 9 I*,t----&4 r--*
A V% Ii ri hl.;
fliLn rIaIIeIIII r-,a a, i ontrol Rod Corner
'Figure 3.10 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-383B416G8.0-10DM With Channel Bow (Assembly Exposure of 17,500 MWdIMTU)
EMF-2440 Revision 0 Paoe 3-24 C
0 In t
r 0 1 R
0 d
C 0
r n
e r
1.017 1.017 1.068 1.083 1.107 1.074 1.048 0.985 0.970 1.017 0.986 1.024 1.000
-0.885 0.992 1.004
-0.956 0.965 1.068 1.024
'0.890 1.063
'1.091 1.055 0.990 0.989 1.009 "1.083 1.000 1.063 0.944 0.966 1.055 Internal 1.107 0.885 1.091 Water 1.074 0.846 1.040 Channel 1.074 0.992 1.055 1.032 0.951 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
LaSalle Unit 2 Cycle 9 Plant Transient Analvsis ontr 0
EMF-2440 Revision 0 Plant~~~D.
TrnietAn)si g
I Rod Corner C
0 n
t r
0 R
0 d
C 0
r n
e r
Figure 3.11 LaSalle Unit 2 Cycle 9 Safety Limit Local Peaking Factors SPCA9-396B-12GZ-1OOM With Channel Bow (Assembly Exposure of 15,000 MWdMTU)
~*.qnww Pnvmr Ctvnnnmtin 1.025 1.058 1.062 1.117 1.100 1.108 1.043 1.026 0.979 1.058 0.934 1.018 0.852 1.003 0.845 0.999 0.903 1.005 1.062 1.018 1.003 1.067 1.092 1.058 0.984 0.983 1.006 1.117 0.852 1.067 1.046 0.823 1.056 Internal 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 f, I k1i
LaSalle Unit 2 Cycle 9 C)Iftn "Tr-neiiart Analv it S
.~.fl
- oe#3-26 I
EMF-2440 Revision 0 Paoe 3-26 0
10 20 30 40 s0 0
70 so 9D 100 110 PrMofRMd)
Power MCPRP
(%)
Umit 100 1.41 60 1.46 25 1.91 25 2.20 0
2.70 Figure 3.12 EOC Base Case Power-Dependent MCPR Umits for
.ATRUM-9B Fuel - TSSS Insertion Times Siemen Power Copoation
EMF-2440 Revision 0 Page 3-27 LaSalle Unit 2 Cycle 9 Plant Transient Analysis CL 0
10 20 30 40 5o O0 PmW r (%OF P8Md)
"0*,
70 s0 s0 100 110 Power
- MCPR,
(%)
Umit 100 1.51 60 1.52 25 1.99 25 2.20 0
2.70 Figure 3.13 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel - TSSS Insertion Times Siemens Power Corporation
LaSalle Unit 2 Cycle 9 DOlnft runeiant Analvlir EMF-2440 Revision 0 Page 3-28 0
10 20 0
40 50 60 70 so 100 110 P~fM o~f abd)
Power MCPRp
- (%)
LUmit 100 1.39 60 1.44 -
1.69 25 2.20
_0 -
2.70 Figure 3.114 EOC Base Case Power-Dependent MCPR Limits for "ATRUM-gB Fu6l - NSS Insertion Times
- a.
- a. K dE=
LaSalle Unit 2 Cycle 9 Plant Transient Analysis 25M 245 235 2M5 215 IA 1.7S 1.35, 1.75 1.15 EMF-2440 Revision 0
- IA -~
0 10 2D 30 40 SD so 70 O,
- o 100 110 POMM (% of d61 Power MCPI:
(%)
Limit 100 1.48 60 1.51 25 1.97 25 2.20 0
2.70 Figure 3.15 EOC Base Case Power-Dependent MCPR Limits for GE9 Fuel - NSS Insertion Times N
Siemens Power Coorabtion Kj
ý I
LaSalle Unit 2 Cycle 9 Plant Transient Analysis EMF-2440 Revision 0 Paoe 3-30 0
10 2
30 40 50 G0 70 so s
100 110 Fo.m-r..)
Power' LHGRFACV
-(%)
Multiplier 100 1.00
-60 1.00 25--
0.78
.25 0.78 0
0.78 Figure 3.16 EOC Base Case P ower-Dependent LHGR Multipliers for U M -9B Fuel - TSSS Insertion Tim es I
II i
I
LaSalle Unit 2 Cycle 9 Plant Transient Analvsis C.
U 1.30 1.25 1.20 1.15 1.10 1.05 IO~l 0.05 0.90 0.85, 0.100 0.75T 0.70 0.65.
0.60 0
10 20 30 40 50 60 Pomr (MM) 70 80 90 100 110 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
4 Siemens Power Comoration EMF-2440 Revision 0 Pane 3-31 a
SoLRW FTC7 UaRFC U a I
a U