ML20196D890
| ML20196D890 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/30/1998 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML20138L907 | List: |
| References | |
| SCE-9801-NP, NUDOCS 9812020349 | |
| Download: ML20196D890 (258) | |
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l2 SCE-9801-NP
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- Reload Analysis Methodology for the San Onofre Nuclear Generating Station f
Units 2 and 3 y ,
November 1998' i
Southern California Edison An Edison International" Companyc
.u Nuclear FuelManagement Division a
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m 9812O20349 981130 PDR ADOCK 05000361 P PDR
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Reload Analysis Methodology for the San Onofre Nuclear Generating Station 1 Units 2 and 3 I
November 1998 I
I Southern California Edison An Edison International
- Company Nuclear Fuel Management Division l
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DISCLAIMER I This document was prepared by Southern Califomia Edison Company for its own use. The use of information contained in this document by anyone other than Southern Califomia Edison Company is not authorized, and in regard to unauthorized use neither Southem Califomia Edison Company or any of its officers, directors, agents, or employees assumes any obligation, responsibility or liability, or makes any wananty or representation, with respect to the contents of this document, or its accuracy or completeness.
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- Southern California Edison i November 1998
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ABSTRACT I This report presents a summary of the Southem Califomia Edison (SCE) program to acquire the Asea Brown Boveri/ Combustion Engineering (ABB CE) technology to perform reload licensing designs for the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3. This repon describes the ABB CE/SCE Reload Technology Program and SCE's reload analysis process, including the major analysis areas, calculation results, implementing procedures, and major computer codes. Implementation of the reload analysis process is accomplished within the SCE Reactor Core Design and Monitoring Program which is a site wide program that meets or l exceeds all recommendations in INPO's Significant Operating Experience Repon (SOER) 96-02,
" Design and Operation Considerations for Reactor Cores." An outline of SCE's Reactor Core Design and Monitoring Program is presented. Comparisons of the principal results of j independent reload analyses performed by SCE and ABB CE are presented. This benchmark I demonstrates the ability of SCE to independently perform reload analyses required for the design, l licensing, stanup, operation, and monitoring of a SONGS Units 2 and 3 reload cycle. I l
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I l Southern California Edison ii November 1998 I
l Table of Contents List o f Tables . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . viii List of Figu res . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x List of Defini tion s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . x i
1.0 INTRODUCTION
AND
SUMMARY
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 2.0 RELOAD TECIINOLOGY TRANSFER PROGRAM OVERVIEW . . . . . . . . . . . 2 2.1 PIIASE 1- CLASSROOM TRAINING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . c 2.2 PHASE 2-ON TIIE JOB TRAINING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.3 PIIASE 3- INDEPENDENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.3.1 Unit 3 Cycle 9 Independent Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.3.2 Unit 2 Cycle 10 Independent Analysis . . . . . . . . . . . . . . . . . . . . . . . . . 5 2.3.3 Benchma rk A nalyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.0 OVERVIEW OF THE RELOAD ANALYSIS PROCESS . . . . . . . . . . . . . . . . . . . . 6 3.1 PH YSICS ANA LYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1.1 Models and Depletions Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.1.2 Physics Data for Fuel Assembly Mechanical Design . . . . . . . . . . . . . I 1 3.I.3 Generic Physics Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 3.1.4 Physics Data for Fuel Performance Analyses . . . . . . . . . . . . . . . . . . 12 3.1.5 Physics Data for LOCA Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 3.1.6 Physics Data for Core Thermal-Hydraulics Analyses . . . . . . . . . . . 13 3.1.7 Physics Data for Non-LOCA Transient Analyses . . . . . . . . . . . . . . . 14 3.1.8 Physics Data for COLSS/CPCS Analyses . . . . . . . . . . . . . . . . . . . . . 15 3.1.9 As B uilt Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 3.1.10 Physics Data for Spent Fuel Pool Storage and Boraflex Requirements
........................................................18 3.2 CORE THERM AL-HYDRAULICS ANALYSIS . . . . . . . . . . . . . . . . . . . . . 19 3.2.1 DNB R A nalysis Mod els . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 3.2.1.1 Major Inputs to the Core Thermal-Hydraulics Design . 21 3.2.1.2 Major Outputs of the Core Thermal. Hydraulics Design
...............................................22 3.2.2 Statistical Combination of Uncertainties . . . . . . . . . . . . . . . . . . . . . . 23 3.2.2.1 Major Inputs for the SCU Analysis . . . . . . . . . . . . . . . . . 24 3.2.2.2 Major Outputs of the Statistical Combination of Uncertainties Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3.3 FUEL PERFORM ANCE ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 3.3.1 Major Inputs for Fuel Performance . . . . . . . . . . . . . . . . . . . . . . . . . . 30 3.3.2 Major Outputs for Fuel Performance . . . . . . . . . . . . . . . . . . . . . . . . 31 3.4 NON-LOCA TRANSIENT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 3.4.1 Events Not Normally Analyzed in the Reload Analysis . . . . . . . . . . 34 l
Southem Califomia Edison iii November 1998
3.4.1.1 Decrease in Feedwater Temperature . . . . . . . . . . . . . . . . 35 3.4.1.2 Increase in Feedwater Flow . . . . . . . . . . . . . . . . . . . . . . . 35 3.4.13 Loss of Load, Turbine Trip, or Loss of Condenser Vacuum
...............................................35 3.4.1.4 Loss of Normal AC Power To Station Auxiliaries . . . . . 36 3.4.1.5 Loss of Normal Feedwater Flow . . . . . . . . . . . . . . . . . . . . 36 3.4.1.6 Feedwater System Pipe Breaks . . . . . . . . . . . . . . . . . . . . . 36 3.4.1.7 Startup of an Inactive Reactor Coolant Pump . . . . . . . . 36 3.4.1.8 Chemical and Volume Control System Malfunction-Pressurizer Level Control System Malfunction . . . . . . . 37 3.4.1.9 Inadvertent Operation of the ECCS During Power 1 Opera tio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 7 3.4.1.10 Pressurizer Pressure Decrease Events . . . . . . . . . . . . . . . 37 3.4.1.11 Small Primary Line Pipe Break Outside Containment . 37 1 3.4.1.12 Steam Generator Tube Rupture . . . . . . . . . . . . . . . . . . . . 38 3.4.1.13 Inadvertent Opening of a IOSGADV . . . . . . . . . . . . . . . . 38 3.4.2 Events Normally Analyzed in the Reload Analysis . . . . . . . . . . . . . . 38 1 3.4.2.1 Fu el Fail u re E ven ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 8 3.4.2.1.1 Increased Main Steam Flow With Single Failure
...................................... 39 3.4.2.1.2 Steam System Piping Failures . . . . . . . . . . . . . 41 3.4.2.13 Single Reactor Coolant Pump Sheared Shaft (SS) l 3.4.2.1.4
/ Seized Rotor (SR) ..................... 44 CE A Ejection . . . . . . . . . . . . . . . . . . . . . . . . . . 45 3.4.2.2 Margin Setting Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 3.4.2.2.1 Total Loss of Forced Reactor Coolant Flow . 49 3.4.2.2.2 Asymmetric Steam Generator Events . . . . . . 50 3.4.2.23 Uncontrolled CEA Withdrawal . . . . . . . . . . . 52 1 3.4.2.2.4 Single Full Length CEA Drop Event . . . . . . . 54 3.4.2.2.5 Part Length CEA Drop . . . . . . . . . . . . . . . . . . 55 3.4.2.2.6 CEA Subgroup Drop Events . . . . . . . . . . . . . . 57 1 3.4.2.2.7 CEA Withdrawal within Deadband . . . . . . . 58 3.4.2.2.8 Inadvertent Boron Dilution . . . . . . . . . . . . . . . 60 3.4.3 Degraded Performance of CPCS & COLSS Category . . . . . . . .. 61 I 3.4.3.1 COLSS Out-of Service and at Least One CEAC Operable
.......................................61 3.4.3.2 COLSS In-Service and Both CEACs Inoperable . . . . . . 62 3.43.3 COLSS Out-of-Service and Both CEACs Inoperable . . 62 3.4.4 Verification of Transient Related CPCS Constants . . . . . . . . . . . . . 63 l 3.4.4.1 3.4.4.2 Dynamic Compensation . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 Transient Offset Power Penalties . . . . . . . . . . . . . . . . . . . 64 3.4.4.3 Radial Penalty Factor Delay . . . . . . . . . . . . . . . . . . . . . . 64 3.4.4.4 Reactor Coolant Pump Speed Trip Setpoint . . . . . . . . . . 64 3.4.4.5 Variable Overpower Trip Setpoint ................ 64 )
3.4.4.6 Asymmetric Steam Generator Trip Setpoint ......... 65 Southern California Edison iv November 1998
3.5 CORE PROTECTION CALCULATOR SYSTEM ANALYSIS . . . . . . . . 66 3.5.1 Physics Analysis Input to the CPCS Analysis . . . . . . . . . . . . . . . . . . 68 3.5.2 Core Thermal-IIydraulics Analysis Input to the CPCS Analysis . . 69 3.53 Safety Design Input to the CPCS Design . . . . . . . . . . . . . . . . . . . . . . 69 3.5.4 Fuel Performance Analysis input to the CPCS Analysis . . . . . . . . . 69 3.5.5 CPCS Constants Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 3.5.5.1 CPCS Data Base Constants Verification . . . . . . . . . . . . . 70 3.5.5.2 Reload Data Block Constants Analyses . . . . . . . . . . . . . . 70 3.5.5.3 CPCS Addressable Constants Analyses . . . . . . . . . . . . . . 71 3.6 CORE OPERATING LIMIT SUPERVISORY SYSTEM ANALYSIS . . . 74 3.6.1 Physics Analysis Input to the COLSS Analysis . . . . . . . . . . . . . . . . . 75 3.6.2 Core Thermal.Ilydraulics Analysis Input to the COLSS Analysis . 75
- 3.6.3 Safety Analysis Input to the COLSS Analysis . . . . . . . . . . . . . . . . . 75 I 3.6.4 Other Inputs to the COLSS Analysis . . . . . . . . . . . . . . . . . . . . . . . . . 76 3.6.5 COLSS Constants Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76 3.6.5.1 COLSS Data Base Constants Verification . . . . . . . . . . . . 76 3.6.5.2 COLSS Addressable Constants Analyses . . . . . . . . . . . . 76 4.0 REACTOR CORE DESIGN AND MONITORING PROGRAM . . . . . . . . . . . . . 78 I 4.1 4.2 FUEL MANAGEMENT GUIDELINES FOR SONGS UNITS 2 AND 3. . 80 REACTOR CORE DESIGN REVIEW TEAM . . . . . . . . . . . . . . . . . . . . . . 80 4.3 RE LOAD G RO UND RULES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 80 4.4 ABB CE FUEL DESIGN CHANGE INTERFACE . . . . . . . . . . . . . . . . . . . 81 4.5 ABB CE RELOAD ANALYSIS COMPUTER CODES AND M ETH O DO LOG Y INTER FA CE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 4.5.1 Computer Ilardware Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 4.5.2 Computer Software Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 81 4.5.3 Analysis Methodology Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 82 I 4.6 4.5.4 Q u ality Progra m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 83 SCE ENGINEER TRAINING QUALIFICATION GUIDE . . . . . . . . . . . . 84 4.7 I 4.8 CORE RELOAD ANALYSES AND ACTIVITIES CHECKLIST . . . . . . 85 SOURCE VERIFICATION /ABB CE FUEL FABRICATION INTERFACE
.............................................................85 4.9 I ABB CE ENGINEERING INTERFACE . . . . . . . . . . . . . . . . . . . . . . . . . . . 86 4.9.1 L O C A A n a ly ses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... 86 4.9.2 Fuel Mechanical Design Analyses . .......................... 86 I 4.10 4.9.3 A s- b u il t Da ta . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 7 SITE PR OG R AM IMPACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 87 4.11 LICENSING AND DESIGN BASIS UPDATES . . . . . . . . . . . . . . . . . . . . . . 87 4.12 DESIGN PROCESS FLOW AND CONTROLS . . . . . . . . . . . . . . . . . . . . 87 4.13 CO LSS/CPC PR OD UCTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 88 4.14 LOW POWER AND POWER ASCENSION TESTING ............. 89 4.15 CORE AND SFP REQ UIREMENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 4.16 CORE AND FUEL MONITORING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 89 I
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I 5.0 COMPARISONS OF SCE INDEPENDENT ANALYSIS TO ABB CE ANALYSIS
....................................................................91 5.1 UNIT 2 CYCLE 10 FUEL MANAGEMENT GUIDELINES COMPARISON
...............................................................92 5.2 COMPARISON OF PRINCIPAL PIIYSICS RESULTS . . . . . . . . . . . . . . 100 5.2.1 Comparison of General Physics Data . . . . . . . . . . . . . . . . . . . . . . . . 100 5.2.2 Comparison of Physics Data for Fuel Assembly Mechanical Design ;
........................................................I11 l 5.23 Comparison of Physics Data for Fuel Performance, Core Thermal-Ilydraulics, and LOCA Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 13 l 5.2.4 Comparison of Physics Data for Safety Analyses . . . . . . . . . . . . . . 118 5.2.4.1 Comparison ID HERMITE Model Input . . . . . . . . . . . . I18 5.2.4.2 Comparison of Fr Versus Inlet Temperature for Pin Cen su s Even ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 124 ;
5.2.43 Comparison of Physics Data for CEA Ejection Analysis l
..............................................125 i I 5.2.4.4 5.2.4.5 Comparison of CEA (Full Length and Part Length) Single, CEA 2 & 3, and Subgroup Drop Physics Data . . . . . . . 129 Comparison of Physics Data of CEA Withdrawal . . . . 133 I 5.2.4.6 5.2.4.7 Comparison of Physics Data for CEA Deviation Within Dea d b a n d . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 5 Comparison of Physics Data for Post-trip Steam Line l
I 5.2.4.8 B rea k A nalysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 141 Comparison of Physics Input to ASGT Safety Analysis
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I 53 COMPARISON OF CORE THERMAL-HYDRAULICS ANALYSIS R ES U LTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 48 5.3.1 Comparison of Basics Thermal-Hydraulics Data . . . . . . . . . . . . . . 148 5.3.2 Comparison of Core Wide Parameters . . . . . . . . . . . . . . . . . . . . . . 149 53.3 Comparison of CETOP-D Benchmarking . . . . . . . . . . . . . . . . . . . . 149 5.3.4 Comparison of MSCU - Response Surface Verification . . . . . . . . 150 ;
5.3.5 Comparison of Steam Line Break Initial Void Fraction ....... 151 l 5.4 COMPARISON OF FUEL PERFORMANCE ANALYSIS RESULTS . . 152
,g 5.4.1 Comparison of Hot Rod Internal Pressure . . . . . . . . . . . . . . . . . . 154 l 3 5.4.2 Comparison of Hot Rod Gap Conductance . . . . . . . . . . . . . . . . . . . 158 5.4.3 Comparison of Hot Rod Average Temperatures at the Peak Power i I 5.5 5.5.1 Nod e . . . . . . . . . . ........................................162 COMPARISON OF SAFETY ANALYSIS RESULTS . . . . . . . . . . . . . . 166 Comparison of CESEC Basedeck Results . . . . . . . . . . . . ....... 166 5.5.2 Comparison of Boron Dilution Analysis . . . . . . . . . . . .... .... 167
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5.5.3 Comparison of Feedwater Line Break Analysis Results . . . . . . . . 168 5.5.4 Comparison of Pre-Trip Steam Line Break Analysis Results . . . . 171 5.5.5 Comparison of Post-Trip Steam Line Break Analysis Results . . . 172 5.5.6 Comparison of Loss of Normal AC Power (LOAC) Analysis Results
........ ... ...... .. ..............................177 5.5.7 Comparison of CEA Ejection Analysis Results . . . . . . . . . . . . . . . . 179 Southern California Edison vi November 1998
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1 5.5.8 Comparison of Seized Rotor / Sheared Shaft Analysis Results . . . . 180 1 5.5.9 Comparison of Single Part Length Rod Drop Event Results . . . 181 5.5.10 Comparison of CEA Withdrawal within Deadband Event Results l
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5.5.11 Increased Main Steam Flow with LOAC Analysis . . . .. ...... 185 l 5.5.12 Comparison of CEA Misalignment Analysis Results . . . . . . . . . . 186 l 5.5.13 Comparison of CEA Withdrawal Analysis Results . . . . . . . . . . . . 187 l 5.5.14 Comparison of Total Loss of Flow Analysis Results . . . . . . . .. 189 l l 5.6 COMPARISON OF COLSS AND CPC ANALYSIS RESULTS . . . . . . . 190 5.6.1 Comparison of COLSS and CPC Flow Uncertainties Analysis Results
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.......................................................192 i 5.6.2 Comparison of COLSS Power Uncertainties Analysis Results . . . 193 5.6.2.1 Comparison of COLSS Primary Delta-T Power !
Un certain ties . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 193 I 5.6.2.2 Comparison of COLSS Secondary Calorimetric U ncertain ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 194 I c.6.3 Comparison of CPC Calibration Allowances Analysis Results . . . 195 5.6.4 Comparison of COLSS and CPC Digital Setpoints Core Simulation Analysis Results . ....................................... 196 I 5.6.4.1 5.6.4.2 Comparison of Simulated Rod Shadowing Factors [
] Analysis Results . . . . . . . . ..........
Comparison of Measured Fxy Simulations Analysis Results
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............................................19'7 Comparison of PLR Follower Analysis Results . . . .. 198 5.6.5 Comparison of CPC Overall Uncertainty Analysis / Addressable Constants Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 199 5.6.6 Comparison of COLSS Overall Uncertainty Analysis Results . . . 200 5.7 Comparison of As-Hullt Physics Startup Test Predictions . . . . . . . . . . . . 201 I
6.0 REFERENCES
. . . . . . . .... . ..... .. ... ..... .......... ... 203 APPENDIX A: SONGS RELOAD ANALYSIS RELATED PROCEDURES . . . . . . 207 A.1 Nuclear Fuel Management Division ..... ...... .. ... . .. ... 208 A.2 Technical Division .. ........ .. ..
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.... . .... .... ... . 214 A.3 Site Technical Services Division . . .. . . .. . ..... . ... 222 A.4 Nuclear Engineering and Design Organization Division . . . . . . . . . . 224 I I A.5 A.6 A.7 Chemistry Division Licensing Division Nuclear Training Division . . .
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. . 231 226 I A.8 A.9 Nuclear Oversight Division Maintenance Division A.10 Operations Division
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Southern California Elison vii November 1998 l
I I List of Tables l Table 2.1-1 Table 3.5-1 Classroom Training . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 CPCS Auxiliary Trips . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 Table 4.6-1 Reload Analyst Qualification Position . . . . . . . . . . . . . . . . . . . . . . . . . . 85 Table 5.1-1 Unit 2 Cycle 10 Physics Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . 93 Table 5.2.1-1 Physics Design Parameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 102 Table 5.2.1-2 Verification of LCS MTC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 103 Table 5.2.1-3 BOC Maximum CBCs and Minimum IBWs . . . . . . . . . . . . . . . . . . . . 103 Table 5.2.1-5 ARO F,y Ru ndo wn s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 105 I Table 5.2.1-6 Table 5.2.1-7 Table 5.2.1-8 Rodded F,y D ata . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 06 F, (Tilted) Versus Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 109 CEA B ank Worths . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 10 Table 5.2.2-1 Physics Data For Fuel Assembly Mechanical Design . . . . . . . . . . . . . . I 12 Table 5.2.3-1 Physics Data for Core Thermal-Hydraulics, Fuel Performance, and LOCA Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 1 4 I Table 5.2.3-2 Table 5.2.3-3 Table 5.2.3-4 Physics Data for Core Thermal-Hydraulic . . . . . . . . . . . . . . . . . . . . . . . I14 Physics Data for Fuel Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . I 14 Physics Data [ ] for LOCA . . . . . . . . . . I 17 I Table 5.2.4.1-1 Table 5.2.4.1-2 Table 5.2.4.1-3 I D HERMITE Model BOCS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 119 1D HERMITE Model EOCL ...............................120 I D HERMITE GEOMETRY Deck . . . . . . . . . . . . . . . . . . . . . . . . . . . 121 Table 5.2.4.1-4 I D HERMITE Thermal-Hydraulics Deck . . . . . . . . . . . . . . . . . . . . . 123 Table 5.2.4.2-1 Fr Versus Inlet Temperature for Census Events . . . . . . . . . . . . . . . . . 124 Table 5.2.4.3-1 Physics Data for Post-Trip Ejected CEA ................... 126 Table 5.2.4.3-2 Scram Worth for CEA Ejection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 127 Table 5.2.4.3-3 Pin Census -- 100% and 0% Power . . . . . . . . . . . . . . . . . . . . . . . . . . . 128 Table 5.2.4.4-1 FLCEA Total Distortion Factor (TDF) . . . . . . . . . . . . . . . . . . . . . . . . 129 Table 5.2.4.4-2 CEA 2&3 Drop Total Distortion Factor (TDF) . . . . . . . . . . . . . . . . . . 130 Table 5.2.4.4-3 Subgroup Drop Total Distortion Factor (TDF) . . . . . . . . . . . . . . . . . . 130 Table 5.2.4.4-4 PLCEA Drop Physics Results (Positive Reactivity) . . . . . . ....... 131 Table 5.2.4.4-5 PLCEA Drop Physics Results (Negative Reactivity) . . . . . . . ..... 132 Table 5.2.4.4-6 Hot Channel F,, Penalty Radial Distortion Factors (RDF) . . . . . . . . . 132 Table 5.2.4.5-1 Physics Data for CEA Withdrawal . . . . . . . . . . . . . . . . . . . . . . . . . . 134 I Table 5.2.4.6-1 Table 5.2.4.6-2 Deviated Worths for CEA Deviation Within Deadband . . . . . . . . . . . 137 Radial Power Distonion Factors . . ......................... 138 HZP Post Trip SLB Reactivity Balance .............. ........ 142 I Table 5.2.4.7-1 Table 5.2.4.7-2 Table 5.2.4.7-3 HFP Post Trip SLB Reactivity Balance . . . . . . . . . . . . . . . . . . . . . . . . 142 Post Trip SLB Low Flow Reactivity Credits ..... ... ......... 145 I Table 5.2.4.7-4 Table 5.2.4.7-5 Table 5.2.4.7-6 Post Trip SLB low Flow Fq Credits . . . . ............ .... ... 145 Post Trip SLB High Flow Reactivity Credits . ................. 146 '
Post Trip SLB High Flow Fq Credits . . . . . . . . . . . . . . . . . . . . . . . . . . 146 Table 5.2.4.8.1 Physics Data for ASGT Safety Analysis . . . . . . . . . . ............ 147 lI Table 5.3.1-1 Basic Thermal-Hydraulic Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 148 Table 5.3.2-1 Core Parameters . . .............................. ...... . 149 Table 5.3.3-1 CETOP-D Results ......................................149 Southern California Edison viii November 1998
List of Tables Table f.3.4-1 M SCU Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 150 Table 5.3.5-1 SLB Void Reactivity Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 151 Table 5.4-1 Hot Rod Average Power . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 152 Table 5.5.1-1 CESEC Basedeck Calculations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 166 Table 5.5.2-1 Boron Dilution Results - Time to Criticality . . . . . . . . . . . . . . . . . . . . 167 Table 5.5.2-2 Boron Dilution Results - SRM ratio . . . . . . . . . . . . . . . . . . . . . . . . . . 167 Table 5.5.3-1 Feedwater Line Break Analysis Input . . . . . . . . . . . . . . . . . . . . . . . . . 169 Table 5.5.3-2 Feedwater Line Break Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . 170 Table 5.5.4-1 Pre-Trip Steam Line Break Analysis Results . . . . . . . . . . . . . . . . . . . . 171 1 Table 5.5.5-1 Key Parameters for HFP Post-Trip Steam Line Break . . . . . . . . . . . . 174 Table 5.5.5-2 Key Parameters for HZP Post-Trip Steam Line Break . . . . . . . . . . . . 174 Table 5.5.5-3 Post-Trip HZP-LOAC Steam Line Break Results . . . . . . . . . . . . . . . . 175 1 Table 5.5.5-4 Post-Trip HZP-AC Steam Line Break Analysis Results . . . . . . . . . . . 175 Table 5.5.5-5 Post-Trip HFP-LOAC Steam Line Break Analysis Results . . . . . . . . . 176 l Table 5.5.5-6 Post-Trip HFP-AC Steam Line Break Analysis Results . . . . . . . . . . . 176 B Table 5.5.6-1 Key Parameters For less of Normal AC Power Analysis . . . . . . . . . . 178 Table 5.5.6-2 Loss of Normal AC Power Analysis Results . . . . . . . . . . . . . . . . . . . . 178 Table 5.5.7-1 Post Trip CEA Ejection Analysis Results . . . . . . . . . . . . . . . . . . . . . . 179 I Table 5.5.7-2 Pre-Trip CEA Ejection Analysis Results . . . . . . . . . . . . . . . . . . . . . . . 179 Table 5.5.8-1 Seized Rotor / Sheared Shaft Analysis . . . . . . . . . . . . . . . . . . . . . . . . . 180 Table 5.5.9-1 i Table 5.5.10-1 PLR Drop Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 182 Key Parameters for CEA Withdrawal within Deadband Analysis . . . 183 Table 5.5.10-2 CEA Withdrawal within Deadband Analysis: Power > 25% . . . . . . . 184 Table 5.5.10-3 CEA Withdrawal within Deadband Analysis: Power .< 25% . . . . . . . 184 Table 5.5.11-1 Increased Main Steam Flow with LOAC Analysis Results . . . . . . . . . 185 Table 5.5.12-1 CEA Misalignment Power Reduction Analysis Results . . . . . . . . . . . 186 Table 5.5.12-2 Lowest Power Levels with No Power Reduction Required . . . . . . . . . 187 l Table 5.5.13-1 CEAW from Suberitical Power Analysis Results . . . . . . . . . . . . . . . . 188 Table 5.5.13-2 CEAW from HZP Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . 188 Table 5.5.13-3 CEAW from 50% Power Analysis Results . . . . . . . . . . . . . . . . . . . . . 188 Table 5.5.13-4 CEAW from HFP Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . 189 Table 5.5.14-1 Loss of Flow Analysis Results . . . . . . . . . ..................... 189 i Table 5.6.1-1 Table 5.6.2.2-1 COLSS and CPC Flow Uncenainties Analysis Results . . . . . . . . . . . . 192 COLSS Secondary Calorimetric Uncertainty Analysis Results . . . . . . 194 Table 5.6.3.1-1 CPC Power Calibration Allowances Analysis Results . . . . . . . . . . . . 195 I Table 5.6.4.1-1 Simulated [ ] RSFs at 100% Power Analysis Results . . 196 Table 5.6.4.2-1 Simulated Measured Fu Analysis Results . . . . . . . . . . . . . . . . . . . . . 197 Table 5.6.4.3-1 PLR Follower Analysis Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 198 I Table 5.6.5-1 Addressable Constants Analysis Results . . . . . . . . . . . . . c. . . . . . . . . 199 Table 5.6.5-2 [ ] Criteria Analysis Results . . . . . . . . . . . . . 199 Table 5.6.6-1 COLSS Penalty Factor Analysis Results . . . . . . . . . . . . . . . . . . . . . . . 200 Table 5.7-1 Summary of Physics Test Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . 202 L
Southern California Elison ix November 1998
List of Figures Figure 3.0-1 Simplified Diagram of Reload Analysis Process . . . . . . . . . . . . . . . . . . . 8 Figure 3.0-2 Simplified Reload Analysis Network . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 Figure 3.2-1 Channel Geometry of CETOP-D Model . . . . . . . . . . . . . . . . . . . . . . . . 26 l Figure 3.2-2 Figure 3.2-3 Sample First Stage TORC Model Layout . . . . . . . . . . . . . . . . . . . . . . . 27 Sample Second Stage TORC Model Layout . . . . . . . . . . . . . . . . . . . . . 28 Figure 3.2-4 Sample Third Stage TORC Model Layout . . . . . . . . . . . . . . . . . . . . . . . 29 Figure 3.5-1 Simplified CPC Functional Diagram . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 Figure 3.5-2 Simplified CEAC/CPC Block Diagram . . . . . . . . . . . . . . . . . . . . . . . . . 73 Figure 3.6-1 Simplified COLSS Functional Diagram . . . . . . . . . . . . . . . . . . . . . . . . 77 I Figure 4.0-1 SCE Reactor Core Design and Monitoring Program . . . . . . . . . . . . . . . 79 Figure 5.1-1 Enrichment Zone and Erbia Patterns . . . . . . . . . . . . . . . . . . . . . . . . . . . 94 Figure 5.1-2 I Figure 5.1-3 Figure 5.1-4 F,, Historical Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 95 Peak Linear Heat Rate Historical Data . . . . . . . . . . . . . . . . . . . . . . . . . 95 Peak Rod Burnup Historical Data . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 96 I Figure 5.1-5 Figure 5.1-6 Figure 5.1-7 Maximum HFP Boron Concentration Historical Data . . . . . . . . . - . . . . 96 Most Positive HZP MTC Historical Data . . . . . . . . . . . . . . . . . . . . . . . 97 Most Positive MTC 70% Power Historical Data . . . . . . . . . . . . . . . . . . 97 I Figure 5.1-8 Figure 5.1-9 Figure 5.1-10 Most Positive HFP MTC Historical Data . . . . . . . . . . . . . . . . . . . . . . . 98 Minimum HFP Scram Worth Historical Data . . . . . . . . . . . . . . . . . . . . 98 Minimum HZP Scram Wonh Historical Data . . . . . . . . . . . . . . . . . . . . 99 Figure 5.2.1-1 Boron Run Down Curves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 107 I Figure 5.2.1-2 SEP F,y R u nd own . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 108 Figure 5.2.1-3 LEP F,yR u nd o wn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 8 Figure 5.2.3-1 Physics Data [ ] for LOCA . . . . . . . . . . . . . . . . . . . . . I 15 Figure 5.2.3-2 Physics Data [ ] for LOCA . . . . . . . . . . . . . . . . . . . 1 16 Figure 5.2.4.6-1 Deviated Wonh for CEA Deviation Within Deadband . . . . . . . . . . . . 139 Figure 5.2.4.6-2 Distortion Factor for CEA Deviation Within Deadband . . . . . . . . . . . 140 Figure 5.2.4.7-1 HZP Post Trip SLB Cooldown Curve . . . . . . . . . . . . . . . . . . . . . . . . . 143 Figure 5.2.4.7 HFP Post Trip SLB Cooldown Curve . . . . . . . . . . . . . . . . . . . . . . . . . 144 Figure 5.4-1 [
l Figure 5.4.1-1
] .................... 153 L&M Erbia Hot Rod Internal Pressure . . . . . . . . . . . . . . . . . . . . . . . . 155 Figure 5.4.1-2 L&M UO2 Hot Rod Intemal Pressure . . . . . . . . . . . . . . . . . . . . . . . . . 156 Figure 5.4.1-3 K UO2 Hot Rod Intemal Pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . 157 Figure 5.4.2-1 L&M Erbia Hot Rod Gap Conductance . . . . . . . . . . . . . . . . . . . . . . . . 159 Figure 5.4.2-2 I Figure 5.4.2-3 Figure 5.4.3-1 L&M UO2 Hot Rod Gap Conductance . . . . . . . . . . . . . . . . . . . . . . . . 160 K UO2 Hot Rod Gap Conductance . . . . . . . . . . . . . . . . . . . . . . . . . . . 161 L&M Erbia Hot Rod Average Temperature . . . ................ 163 Figure 5.4.3-2 L&M UO2 Hot Rod Average Temperature . . . . . . . . . . . . . . . . . . . . . 164 Figure 5.4.3-3 K UO2 Hot Rod Average Temperature . . . . . . . . . . . . . . . . . . . . . . . . 165 Figure 5.6.2.1-1 COLSS [ ] Analysis Results . . . . . . . . . . . . . . . . . 193 I
Southern California Iklison x Novemter 1998
i List of Definitions ACRONYMS. DEFINITIONS ID One Dimensional Approximation of the Problem 3D Three Dimensional ABB CE Asea Brown Boveri/ Combustion Engineering ADV Atmospheric Dump Valve 3 AFW Auxiliary Feed Water AOO Anticipated Operating Occurrences AOPM Available Over Power Margin AOR Analysis of Record ARI All Rods (CEAs)In ARO All Rods (CEAs) Out ASGT Asymmetric Steam Generator Transient ASI Axial Shape Index BOC Beginning of Cycle BOCL BOC Based on Previous Cycle Running leng BOCS BOC Based on Previous Cycle Running Short BPPCC Boundary Point Power Correlation Coefficient CBC Critical Boron Concentration CBCS COLSS Backup Computer System CEA Control Element Assembly CEAC Control Element Assembly Calculator CEACOOS CEAC Out Of Service CECOR Computer program for incore power distribution monitoring I CEDM Control Element Drive Mechanism CEDMCS Control Element Drive Mechanism Control System CFR Code of Federal Regulations I
L Southem California Edison xi November 1998
List of Definitions ACRONYMS DEFINITIONS CHF Critical Heat Flux CIS COLSS In Service CISAM Cycle Independent Shape Annealing Matrix COLR Core Operating Limits Requirements COLSS Core Operating Limit Supervisory System COOS COLSS Out Of Service CPCs Core Protection Calculators CPCS Core Protection Calculator System DBD Design Basis Document (Topical or System)
DCP Design Change Package (SCE)
DD Doppler Defect DNBR Depanure from Nucleate Boiling Ratio EAB Exclusion Area Boundary ECCS Emergency Core Cooling System EFPD Effective Full Power Days EFPH Effective Full Power Hours (of reactor operation)
EOC End of Cycle EOCL EOC Based on Previous Cycle Running Long EOCS EOC Based on Previous Cycle Running Shon ESFAS Engineered Safety Features Actuation System FCE Facility Change Evaluation (SCE)
FLCEA Full Length Control Element Assembly FPA Fast Power Ascension ,
Fq (Fy ) Maximum Value for the Core of the Ratio of Nodal Peak Pin Power in an Assembly to the Average Pin Power in the Core (3D power peak)
{ Fr (F,) Integrated Radial Peaking Factor Southem California Elison xii November 1998
I List of Definitions l ACRONYMS FTC Fuel Temperature Coefficient DEFINITIONS Fxy (F,y) Planar Radial Peak.ing Factor Fz (F,) Relative Axial Power Peaking Factor HFP Hot Full Power HPSI High Pressure SafetyInjection HZP Hot Zero Power IBW Inverse Boron Worth INPO Institute of Nuclear Power Operation IOSGADV Inadvenent Opening of Steam Generator Atmospheric Dump Valve ITC Isothermal Temperature Coefficient kW/ft Kilowatts per Foot 1* Prompt Neutron Lifetime LCO Limiting Conditions for Operation LCS Licensee Controlled Specifications LEP long End Point (BOCL or EOCL)
LHGR Linear Heat Generation Rate LHR Linear Heat Rate LOAC Loss of Altemating Current Power LOCA Loss of Coolant Accident LPD Local Power Density LTOP low Temperature Overpressure Protection MC Mesh Centered i
MD Moderator Defect .
L MDNBR Minimum value for the Departure from Nucleate Boiling Ratio MOC Middle of Cycle MOCL MOC Based on Previous Cycle Running long Southern Califomia Edison xiii November 1998
I List of Definitions l ACRONYMS MSCU DEFINITIONS Modified Statistical Combination of Uncertainties MSIV Main Steam Isolation Valve MSSV Main Steam Safety Valve MTC Moderator Temperature Coefficient MW Megawatt MWD /MTU Megawatt Days per Metric Ton Uranium N-1 ARI Except for the Worst Rod Stuck Out N-2 ARI Except for the Two (2) Worst Rods Stuck Out (CEA Ejection)
NFM Nuclear Fuel Management Division (SCE)
NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System nyt Neutron Fluence OJT On the Job Training PD Power Defect PDF Probability Distribution Function PDIL Power Dependent insenion Limit
.PLCEA Samc as PLR PLHGR Peak Linear Heat Generation Rate PLR Pan Length Rod or CEA PMS Plant Monitoring System POL Power Operating Limit PPM Parts Per Million PSV Pressurizer Safety Valve ,
PTQAF Peak Power to Quarter Assembly Power Factor RC Reactor Coolant RCP Reactor Coolant Pump Southem California Edison xiv November 1998
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List of Definitions ACRONYMS DEFINITIONS RCS Reactor Coolant System RCTS Regulatory Commitment Tracking System (SCE)
RDB Reload Data Block RIR Reactivity Insertion Rate RMS Root Mean Squared ROCS Reactor Operation and Control Simulator ROPM Requimd Over Power Margin RPS Reactor Protection System RSF Rod Shadowing Factor RTD Resistant Temperature Device RTP Rated Thermal Power RTF Reload Technology Transfer (Program)
SAEOC EOC Extended to the Safety Analysis Limit for the Current Cycle SAFDL Specified Acceptable Fuel Design Limits SAM Shape Annealing Matrix SCE Southern California Edison SCU Statistical Combination of Uncertainties ;
SEP Short End Point of the Previous Cycle (BOCS or EOCS)
SLB Steam Line Break I SNM Special Nuclear Material l SOER INPO's Significant Operating Experience Repon SONGS San Onofre Nuclear Generating Station SR Seized Rotor SS Sheared Shaft T/S Technical Specification T/li Therrnal-Hydraulic Southem California Edison xv November 1998 l l
List of Definitions ACRONYMS DEFINITIONS TDF Total Distonion Factor (instantaneous plus xenon power distortion)
TQAM Topical Quality Assurance Manual (SCE)
UFF Under Flow Fraction UFSAR Updated Final Safety Analysis Repon V&V Verification and Validation (of computer software)
VOPT Variable Over Power Trip l
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Southern California Edison xvi November 1998
l E Section 1.0 Introduction and Summary j
1.0 INTRODUCTION
AND
SUMMARY
l I This repon summarizes the program undertaken by Southem California Edison (SCE) to develop the capability to perform reload licensing analyses for the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3. The purpose of this report is to demonstrate the capability of SCE to i independently perform the reload analyses required for the design, licensing, startup, operation, and monitoring of a reload fuel cycle.
The foundation for the approach and methodology described in this report was for SCE to obtain models and methods previously approved by the Nuclear Regulatory Commission (NRC) rather j than to develop new methodology in-house. Specifically, in 1991 SCE contracted with Asea Brown Boveri/ Combustion Engineering (ABB CE) to provide a program which would enable .
SCE to obtain ABB CE technology to perform reload analysis and licensing activities. ABB CE i designed the ABB CE/SCE Reload Technology Transfer (RTT) Program to provide this training ,
by building on the experience from previous technology transfer programs ABB CE had perform l of other utilities. '
Section 2.0 contains an overview of the RTF Program. This program was implemented in three phases: (1) classroom lecture, (2) on the job training, and (3) independent analysis. Section 3.0 I provides an overview of the reload analysh process for a typical San Onofre Nuclear Generating Station (SONGS) reload analysis and is intended to enhance the understanding of comparisons between the results from the SCE and ABB CE analyses presented in section 5.0. This section l
I also provides references to the previously licensed codes and methodology manuals submitted to the NRC by ABB CE. Section 4.0 describes the Reactor Core Design and Monitoring Program !
which encompasses the reload analysis process obtained in the RTT Program. Section 4.0 shows l I how the reload analyses performed by SCE are integrated into the existing SCE processes activities and procedures for implementing a reactor core design. SCE's Reactor Core Design and Monitoring Program meets or exceeds all recommendations in INPO's SOER 96-02 "
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I Design and Operation Considerations for Reactor Cores." Section 5.0 provides a comparison of the principal results of the independent analysis performed relative to earlier analyses perfomied by SCE and ABB CE or to plant measurements.
I The independent analysis was the final phase of the Reload Technology Transfer Program, l whereby SCE performed reload analyses independent of ABB CE. Based on the comparisons of I the principal design results as presented in Section 5.0, has demonstrated SCE's competency regarding quality assurance practice and technical competency with respect to SCE's ability to set up computer code input decks, execute the codes, and properly interpret the results of ABB CE's reload analysis methodology used for SONGS Units 2 and 3.
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i Southern California Edison 1 November 1998 l l
Section 2.0 Reload Technology Transfer Program Overview 2.0 RELOAD TECHNOLOGY TRANSFER PROGRAM OVERVIEW In 1991 SCE contracted ABB CE to provide a training program to transfer the ABB CE licensed reload engineering methodology to the SCE engineering staff. The scope of the RTT Program ,
included all reload engineering technology except:
loss of Coolant Accident (LOCA)
Fuel assembly mechanical design Currently, SCE does not intend to perform any LOCA analyses, since the LOCA analyses presented in the Updated Final Safety Analysis Report (UFSAR) are typically conservative with respect to reload core design and are performed infrequently. In addition, the fuel assembly mechanical design is a logical area to remain the responsibility of the fuel vendor.
The training program was developed to occur in three phases:
Phase 1 - Classroom training Phase 2 - On the job tmining Phase 3 - Independent analysis 2.1 PHASE 1- CLASSROOM TRAINING The objective of the classroom training phase was to provide basic training in the fundamentals of ABB CE reload design methods and software. This training was provided by ABB CE engineers familiar with the design and licensing analyses of SONGS Units 2 and 3. The topics covemd in this phase included descriptions of the various phenomena being modeled, the basis, limitations, and uncertainties of the models, descriptions of the processes used to generate input for the models, and the use of the models to generate inputs to subsequent analyses. Table 2.1-1 provides a list of the 20 classroom modules presented. During this phase ABB CE provided all documentation, manuals, and calculations that represent the licensing and calculation bases for the reload methodology, computer codes, and analyses of record (AOR). This phase of the program staned in June 1991 and was completed in 1995.
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Southern California Edison 2 November 1998
Section 2.0 Reload Technology Transfer Program Overview Table 2.1-1 Classroom Training Course ID Course Title LD100 Overview of Reload Design Process LD001 Quality Assurance Procedures LD002 Technical Specification Changes LD003 Reload Analysis Report NA001 Overview of Combustion Physics Design Methods NA002 Cross Section Generation NA003 Enrichment Setting and Fuel Management NA004 Determination of Physics Biases and Uncenamties NA005 Radial / Axial Boundary Conditions NA006 CECOR Coefficient Generation and Testing i NA007 Physics Input to Safety Analysis Pan A Physics input to Safety Analysis Part B I SA001 Thermal-Hydraulics Design Analysis FuelPerformance Analysis Part i i SA002 Fuel Performance Analysis Pan 2- DNB Propagation Non-LOCA Transient Analysis Part 1 Non-IDCA Transient Analysis Part 2 Non-LOCA Transient Analysis Pan 3: CEA Eyrtion'HRISE Code SA003 Non-LDCA Transient Analysis Pan 4 The llERMTTE Code Non-LDCA Transient Analysis Pan 5 Summary of Transients i Non-LDCA Transient Analysis Part 6 CLSEC/STRIKIN Non.tDCA Trarment Analysis Pan 7. CENTS Code SA103 COLSS/CPC Related Transient Analysis I CC001 Overview of COMS!CIOCEAC Analysis CC002 Neutronic and Thermal Hydraulic Analysis for COLSS and CPCS CC003 CPC/CEAC Algonthms and Database Venfication CC004 COLSS Algonthms, COLSS Database Update and Testmg CC005 COLSS and CPC Addressable Constants. Startup Support, Margin Assessment i
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- Southern California Edison 3 November 1998
Section 2.0 Reload Technology Transfer Program Overview l
2.2 PIIASE 2-ON THE JOB TRAINING In the On the Job Training (OJT) phase of the Reload Technology Transfer Program, SCE l engineers prepared design analyses for a reactor core design with ABB CE engineers. In this phase SCE engineers performed reload analyses both at ABB CE offices and later at SCE offices as co-authors with ABB CE engineers. SCE engineers participated in significant parts of the
ABB CE engineers independently reviewed the calculations. The Unit 2 Cycle 9 reload analyses l were performed by SCE engineers at SCE offices with ABB CE engineers as co-authors and
. independent reviewers. ABB CE management approved all of these analyses. The OJT phase i included tasks such as: generation ofinput data, execution of related computer programs, interpretation of computer code output, and documentation and review of related calculations. ]
The objective of this phase was to reinforce the classroom training provided in Phase I by I actually performing reload analysis.. l The Unit 2 Cycle 9 reactor core design incorporated the first significant fuel management changes since Cycle 4. Specifically, the cycle length was extended from approximately 525 l
- g feed batch size was decreased from 108 assemblics (% core) to 100 assemblies. The maximum ;
one pin bumup was extended from 52,000 MWD /MTU to 60,000 MWD /MTU. This was the '
I first full batch of urania-erbia fuel which was introduced to replace the B4C burnable neutron absorber pins. The Cycle 9 cores were the second cycle to have the debris resistant bottom grids, (Guardian Grids) and LASER welded grids. The impact of these changes on historical trends in
)
I core wide parameters is presented in Figures 5.1-2 through 5.1-10. Analyzing for these changes expanded the number of calculations and extent of analyses for Cycle 9 beyond what is usually done for a equilibrium fuel cycle. Expansion of the reload analyses' breath and depth enhanced the learning experience for SCE engineers.
Phase 2 staned in 1994 and was completed in early 1997 at the conclusion of the Unit 2 Cycle 9 reload analysis campaign.
2.3 PIIASE 3-INDEPENDENT ANALYSIS I The final phase of the RTT Program was the Independent Analysis Phase. In this phase, SCE l engineers perfomied reload analyses independent of ABB CE. The independent analyses were j done over two reload analysis campaigns: Unit 3 Cycle 9 and Unit 2 Cycle 10. l 2.3.1 Unit 3 Cycle 9 Independent Analysis SCE engineers performed reload analyses independent of ABB CE based on the Unit 3 I Cycle 9 reactor core design. The Unit 3 Cycle 9 reactor core design was very similar to the Unit 2 Cycle 9 reactor core design. All reload analyses were performed and reviewed by SCE engineers. ABB CE engineers performed a second independent review followed by ABB CE managemenapproval. Principal results from these analyses were compared
_ - _ , - 4 m, _
Section 2.0 Reload Technology Transfer Program Overview I to results from the Unit 2 Cycle 9 analyses, discussed in Section 2.2. Observed
- differences and similarities were reconciled. The objective of this phase was to demonstrate that ABB CE's reload technology had been effectively transferred to the SCE engineering staff. This phase started in 1996 and was completed in 1997, officially ending the RTT program.
I Based on SCE's satisfactory performance of all phases the Reload Technology Transfer program, ABB CE certified, in Reference 39, SCE's capability to independently perform reload analyses. These analyses were started in 1996 and completed in July 1997 with the startup of Unit 3 Cycle 9.
2.3.2 Unit 2 Cycle 10 Independent Analysis A second set of independent calculations are being performed by SCE. The Unit 2 Cycle 10 reload design analyses are being performed and independently reviewed by SCE engineers and approved by SCE management using SCE's Quality Assurance program (Reference 36). Principal results from these analyses are being compared to results from the Units 2 and 3 Cycle 9 analyses discussed in Sections 2.2 and 2.3.1. Observed I differences and similarities are being reconciled. As shown in Section 5.1, the reactor core design for Unit 2 Cycle 10 is similar to the Cycle 9 cores. The comparisons between I Unit 2 Cycle 10 and Units 2 and 3 Cycle 9 analysis results do include the effect of the Tcold Reduction Program in addition to reactor core design changes discussed in Section 5.1. The Tcold Reduction Program will lower the reactor inlet coolant temperature as a means to prolong steam generator lifetime. These analyses were initiated in late 1997 and will be completed by the startup of Unit 2 Cycle 10, scheduled for March 1999.
2.3.3 Benchmark Analyses Whenever possible analysis comparisons, presented in Section 5, are based on the I analyses performed by SCE for Unit 2 Cycle 10 relative to Units 2 and/or 3 Cycle 9.
However, comparisons are also presented based on Unit 3 Cycle 9 analyses, relative to Unit 2 Cycle 9, to provide a more complete set of comparisons. In addition, the as-built physics test predictions are for Unit 3 Cycle 9 and are compared to actual plant startup measurements rather than a second set of calculated values. These comparisons of principal results represent the benchmarking to other analyses, required by GL 83-11 (Reference 51) and show SCE's competency regarding quality assurance practice and technical competency with respect to SCE's ability to set up computer code input decks, execute the codes, and properly interpret the results.
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Section 3.0 Overview of the Reload Analysis Process 3.0 OVERVIEW OF THE RELOAD ANALYSIS PROCESS The purpose of this section is to provide a general overview of the reload design process used for the independent analysis phase (Section 2.3). This section is included to enhance the understanding of comparisons between the results from the SCE and ABB CE analyses presented in Section 5.0. This section also provides references to the ABB CE's NRC licensed and approved computer code and methodology manuals.
A simplified diagram of the reload analysis process is shown in Figure 3.0-1. The process inputs to the reload analysis include plant and cycle perfonnance objectives, Technical Specifications, the bases for previously licensed analyses, and the Reload Ground Rules. The Reload Ground Rules document plant data and conditions in a form that can be understood and used by a reload analyst (Section 4.3). The process outputs of the reload analyses include the Facility Change Evaluation (FCE), Technical Specification changes, Licensee Controlled Specifications (LCS) i and Core Operating Limits Requirements (COLR) changes, safety and monitoring setpoints, and plant startup and operations data.
This section discusses typical interfaces of functional areas for preparing the reload products, and is organized to present the reload analysis process in tenns of the major disciplines required to support the reload design analysis process. Figure 3.0-2 shows a simplified network diagram I which illustrates the relationships between the disciplines required to implement the reload design. These disciplines are defined as:
- a. Physics Analyses - responsible for all neutronics related analyses.
I b. LOCA Analyses - responsible for all LOCA analyses and Emergency Core Cooling System (ECCS) performance related analyses. (These analyses will continue to be performed by the fuel vendor.)
l c. Core Thermal-Hydraulics Analyses - responsible for the design and development of models used for Departure from Nucleate Boiling Ratio (DNBR) calculations.
I d. Non-LOCA Transient Analyses - responsible for all transient analyses excluding ,
LOCA.
I e.
Fuel Performance Analyses - responsible for all fuel rod thermal design.
- f. Core Operating Limit Supervisory System (COLSS)/ Core Protection Calculator
{ System (CPCS) Analyses - responsible for the generation of COLSS and CPCS setpoints, data base constants, and core operating margin assessment.
[ g. Fuel Assembly Mechanical Design - responsible for all analyses related to the mechanical design of the fuel assemblies and Control Element Assemblies (CEAs),
L Southern California Edison 6 November 1998
Section 3.0 Overview of the Reload Analysis Process including direct interface with the fuel fabrication facility. (These analyses will continue to be performed by the fuel vendor.)
Currently, SCE relies on the fuel vendor to provide engineering support in the LOCA Analysis and Fuel Assembly Mechanical Design areas since the amount of engineering support required in these areas for a typical reload design is small and the specialized skills and knowledge is high.
The remainder of Section 3.0 provides an overview of the reload design process in each of the disciplines within the SCE scope.
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c m m Section 3.0 Onniew of the Reload Anahsis Process Figure 3.0-1 Simplified Diagram of Reload Analysis Process INPUTS O( TPUTS r .. . . . . .. o y.. ..... "***^='ra"'*-'
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Section 3.1 Physics Analysis 3.1 PHYSICS ANALYSIS This section describes the physics analyses that were performed by SCE for Units 2 and 3 Cycle
- 9. Included in this discussion are brief descriptions of the main inputs and outputs of the physics analyses, and the analyses themselves.
3.1.1 Models and Depletions Analysis The Reactor Operation and Control Simulator (ROCS) and Mesh-Centered (MC)
(References I and 41) physics models' are constmeted for the reload fuel cycle. ROCS incorporates higher order expansion methods for the flux solution based on two-group diffusion theory and can model all aspects of reactor operations from startup to refueling.
I MC is a fine mesh, two energy group diffusion theory neutronics code which calculates fine-mesh (pin-wise) flux, power, and burnup distributions through the application of the nodal imbedded method to individual assemblies using inter-assembly currents calculated I by the coarse mesh ROCS code.
I Three separate depletions are performed with the physics models. Two design depletions are performed at Hot Full Power (HFP) based on two different points within the previous cycle shutdown window, short (best estimate cycle Effective Full Power Days (EFPD) 2 2 I minus delta ) and long (best estimate cycle EFPD plus delta ). These depletions supply core isotopic distributions and nominal HFP power distributions, which bound any possible nominal distributions of the reload fuel cycle as long as the previous cycle ends I between the short and long end points. The third depletion (as-built analysis)is performed using the as-built assembly information and the actual previous cycle bumup, as discussed in Section 3.1.9. .
I The two design depletions utilize inputs from the Reload Ground Rules, fuel management scoping studies, and the previous cycle (s) as-built analysis. The Reload Ground Rules document is a compilation of plant data and operating conditions to be used in the reload engineering effort. The fuel management scoping studies provide a fuel shuffle pattern and new fuel batch design (enrichments and burnable absorber distributions) which have been verified for energy content and acceptable physics parameters using the Reactor Core Design Guidelines (Reference 2). The as-built analysis from the previous cycle furnishes a model of the isotopic characteristics of the fuel that will be retained or I reinserted from prior cycles. The above physics models, isotopic distributions, and nominal HFP power distributions are used as input to the analyses described below.
I L ,
ROCS /MC may be replaced in the future by slMULATE which has been approved by the NRC in Reference 26 for use by SCE in licensing applications, including PWR reload physics analysis, f,eneration of safety analysis
[ inputs, stanup predictions, core physics databooks, and reactor protection rystem ad monitoring system setpoint updates.
Delta equates to approximately % of the bumup window Southern California Edison 10 November 1998
Section 3.1 Physics Analysis 3.1.2 Physics Data for Fuel Assembly Mechanical Design The physics information required for the fuel assembly mechanical design includes both core average and spatially dependent data. This data is extracted from the ROCS /MC results and adjusted by appropriate uncertainties to form the physics data input to the fuel assembly integrity analysis. A typical list of physics data provided for the fuel assembly integrity analysis is shown below:
I i ,
I This analysis has inputs from the Reload Ground Rules and the models and depletions analysis.
3.1.3 Generic Physics Data The following basic core wide physics data is used by many analyses.
- a. Fuel temperature coefficient
- b. Kinetics parameters
- c. Minimum net scram worths with worst rod stuck out and Power Dependent Insertion Limit (PDIL) worth subtracted,
- d. Maximum integrated radial peaking factor in the cycle as a function of power I c. Radial pin census (statistical population of pins based on radial power level) at nominal full power all rods out and selected rodded configurations and burnups.
I f. Axial power distributions
- h. Scram reactivity insertion as a function of scram bank position Southern California Edison 1I Novemter 1998
I Section 3.1 Physics Analysis 1
3.1.4 Physics Data for Fuel Performance Analyses The physics data required for the fuel performance design includes both core average and I spatially dependent data. This data is extracted from ROCS /MC results and convented into the format required for the fuel perfonnance design. A typical list of physics data provided for the fuel performance design is shown below:
[
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I This analysis has inputs from the Reload Ground Rules and the models and depletions analysis.
3.1.5 Physics Data for LOCA Analyses J The physics data required for the LOCA analyses include both core average and spatially l dependent data. The spatially dependent (pin-by-pin) power distribution data is extracted from the ROCS /MC results and converted into the radiant heat transfer factor data input to the LOCA analyses. The remaining core average data is calculated from ROCS /MC, HERMITE and QUIX code outputs (References 1,4, and 5). As noted below, some of this data is calculated in other analyses. A typical list of physics data provided for the LOCA analyses is shown below:
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Section 3.1 Physics Analysis
]
This analysis has inputs from the models and depletions analysis.
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3.1.6 Physics Data for Core Thermal-Hydraulics Analyses The physics information required for the core thermal-hydraulics analyses is fine mesh (pin-by-pin) power distribution data, the maximum burnup for determining rod bow penalties on DNBR, and the number of non-fuel rods (bumable absorber rods) in the core.
The power distribution data is calculated in the models and depletions analysis with the ROCS /MC code. The power distribution data is extracted from the ROCS /MC results I and converted to the pin-by-pin power distribution data input to the core thermal-hydraulics analyses. The maximum bumup for determining rod bow penalty on DNBR is I determined by finding the bumup at which the decrease in relative rod power with bumup offsets the rod bow penalty on DNBR, which is also bumup dependent. A typical list of physics data provided for the core thermal-hydraulics analyses is shou n below:
[
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This analysis has inputs from the models and depletions analysis, the fuel.perfonnance
{ design, and the ABB CE Fuel and Poison Rod Bowing topical report (Reference 6). [
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Southern California Edison 13 November 1998
'I llW Section 3.1 Physics Analysis 3.1.7 Physics Data for Non-LOCA Transient Analyses The typical physics data required for the Updated Final Safety Analysis Report (UFSAR)
Chapter 15 analyses is listed below;
- a. ID HERMITE Model
- b. F, Trend as a Function of RCS Temperature
- c. Boron Dilution Analysis
- 2. Minimum refueling boron concentration
- d. CEA Ejection Analysis Data
- 1. Maximum ejected rod worth with and without doppler feedback
- 2. Pre-ejected and post-ejected peaking factors
- 3. Pin census
- 4. Minimum scram wonh with two (N-2) worst CEAs stuck out g e. Full Length CEA , CEA 2&3, and Subgroup Drop Data g 1. Total Distonion Factor (TDF)(Maximum product of static radial distortion j and xenon redistribution penalty factor for 15 minutes, one (1) hour and two (2) hours
- f. Pan I2ngth CEA Drop Data
- 1. Maximum reactivity insertion
- 2. Maximum product of static radial distortion and xenon redistribution penalty factor (TDF) for 15 minutes, one (1) hour and two (2) hours
- 3. Axial power distributions. before and after drop
- g. CEA Withdrawal Data
- 1. Maximum reactivity insertion rate (RIR)
- 2. Maximum Fq
- 3. Spontaneous fission and Alpha-n neutron source term for subcritical multiplication.
- 4. Axial power distribution
- 5. Maximum radial distortion factor I h. CEA Deviation Within Deadband
- 1. CEA wonhs
- 2. Radial power distortion factors l
Southern California Edison 14 November 1998 i
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I Section 3.1 Physics Analysis
- i. Steam Line Break Event Data
- 1. Moderator and fuel temperature reactivity cooldown curves for all rods in with I worst rod stuck out
- 2. Inverse boron worths
- 4. Local flow and power feedback reactivity credits
- j. Asymmetric Steam Generator Transient Data !
- 1. Radial distortion factor versus core inlet temperature tilt This list includes both core average and spatially dependent data. This data is either !
obtained directly from the ROCS /MC, HERMITE (References 1,3, and 4) and QUIX (References 3 and 5) codes or calculated from quantities found within the outputs of these codes. Additionally, a space-time physics model is required. A One-Dimensional (1D)
HERMITE model is constmeted for this purpose.
This analysis utilizes inputs from the Reload Ground Rules and the models and depletions I
analysis.
3.1.8 Physics Data for COLSS/CPCS Analyses The physics data required for the COLSS/CPCS analyses includes both core average and spatially dependent data. The data is determined from ROCS /MC output. Nominal HFP power distributions calculated in the models and depletions analysis with the ROCS /MC code are surveyed to form the maximum F, data input to COLSS/CPCS analyses. A typical list of physics data provided for the COLSS/CPCS analyses is shown below:
(
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Section 3.1 Physics Analysis l
I i This analysis has inputs from the models and depletions analysis, physics data for fuel I performance design, and physics data for safety design.
3.1.9 As-Built Analysis An as-built ROCS /MC model of the reload is constructed from: (a) the design models from the models and depletions analysis, (b) the actual as-built fresh fuel isotopics calculated from the fuel manufacturing assays, and (c) the humed fuel isotopics calculated at the actual shutdown bumup with the as-built model from the previous cycle.
The physics data calculated with the above described as-built model includes both core average and spatially dependent data. The data is calculated with the ROCS /MC code.
The as-built model is depleted and various restart cases are performed to obtain the predicted as-built power distributions. Some of the spatially dependent data is extracted from the ROCS /MC results and converted to the coefficients used in the CECOR and l COLSS incore monitoring systems. Another product of the as-ouilt analysis is the as-built loading pattem for the next reload fuel cycle. The physics data produced in the as-built analysis is shown below:
- a. As-built loading pattem
- 1. A core loading map is constructed based on the as-built fuel assembly assays.
I This map gives the location, orientation, and serial number of each assembly to be loaded into the core. The new fuel assemblies are judiciously placed l such that the as-built loading variations effect on radial peak and azimuthal tilt I results in the lowest possible radial peak and tilt.
- b. Low power physics test predictions
- 1. CBC
- 2. Isothermal Temperature Coef6cient
- 3. Fuel Temperature Coef6cient
- 4. Change in isothermal temperature coef6cient with respect to change in boron
- 5. Inverse boron worth
- 6. CEA bank wonhs
~
- 7. Delayed neutron parameters Southern California Edison 16 November 1998
Section 3.1 Physics Analysis
- c. Power ascension physics test predictions
- 1. Radialpowerdistributions
- 2. Axialpowerdistributions
- 3. Peaking factors (Fxy, Fr, Fz, Fq)
- 4. Isothermal temperature coefficient, power coefficient, fuel temperature coefficient
- 6. Change in isothermal temperature coefficient with respect to change in boron
- 7. CBC
- d. CECOR Data
- 1. A CECOR library consisting of coupling coefficients, W-primes,1-pin / boxes, axial boundary conditions, azimuthal tilt G-factors.
- 2. A CECOR geometry deck.
l
- 3. A shuffled exposure file.
I e. COLSS data described in item b of Section 3.1.8
- f. Curves to support shutdown margin determination (Plant Physics Data Book)
- 1. Inverse boron worth versus core average temperature, versus boron ppm, versus bumup I 2. Isothermal temperature coefficient versus core average temperature, versus boron ppm, versus bumup I 3. CEA worth data versus core average temperature versus boron ppm, versus bumup
- c. Worth of worst two stuck CEAs I g. Verification of safety analyses
- 1. The physics parameters transmitted for input to the safety analyses, which are sensitive to as-built variations, are reevaluated with respect to data generated based on the design model.
This analysis has inputs from the models and depletions analysis, fuel manufacturing assays, as-built analysis from the previous cycle, and the shutdown burnup from the previous cycle.
E Southern California Edison 17 November 1998
Section 3.1 Physics Analysis 3.1.10 Physics Data for Spent Fuel Pool Storage and Borallex Requirements The physics data includes the bumup and enrichment for fuel assemblies to be offloaded to the spent fuel pool. The fuel assembly bumup is determined from CECOR computer code output, Reference 45. The fuel enrichment is obtained from the fuel batch description. The analysis compares the bumup/ enrichment of each fuel assembly against the minimum requirements of Technical Specification 3.7.18. Unrestricted storage is ,
acceptable if the fuel assembly burnup is less than the minimum rcquired bumup for the given fuel enrichment.
In addition, based on the spent fuel pool storage history and bumup of fuel assemblies to be permanently discharged, analysis is performed to reduce the cumulative exposure to the boraflex panels in the racks. Analysis also confirms that the lifetime gamma dose will be within 1ElI rads.
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Section 3.2 Core Thermal-llydraulics Analysis 3.2 CORE THERMAL-HYDRAULICS ANALYSIS This section describes the core thermal-hydraulics analyses that were performed by SCE using previously approved ABB CE methodology. Included in this discussion is a brief description of the main input / outputs of the core thermal-hydraulics analyses and the analyses themselves. The reload core thermal-hydraulics analyses can be broken into two major categories: (a) DNBR analyses using TORC /CETOP-D to produce a core thermal-hydraulics design model, and (b) calculation of the design DNBR limit by statistically combining system parameter uncertainties.
One of the limits imposed on the power level of a reactor is the Minimum allowable value for the Departure from Nucleate Boiling Ratio (MDNBR), the DNBR SAFDL MDNBR expresses the adequacy of cooling in the most limiting flow channel in the reactor core and is therefore a measure of the core thermal margin. Section 3.2.1 describes the ABB CE open-core thermal-hydraulics code (TORC) used to determine core thermal hydraulic performance and outlines the ABB CE methodology used for the development of a core thermal-hydraulics model.
The principal core thermal-hydraulics design basis was to avoid thermally induced fuel damage I during normal steady state operation and during Anticipated Operational Occurrences (AOOs).
Specified Acceptable Fuel Design Limits (SAFDLs) exist on peak fuel temperature and DNBR to meet this design basis. The DNBR SAFDL was determined statistically such that there was at I least a 95% probability at the 95% confidence (95/95 probability / confidence) level that the limiting fuel rod in the core would not experience DNB. Section 3.2.2 outlines the Statistical Combination of Uncertainties (SCU) methodology used to determine the design DNBR limit.
3.2.1 DNHR Analysis Models I The purpose of the DNBR analysis was to provide a design thermal-hydraulic model(i.e.,
CETOP-D model) for reload safety and COLSS/CPCS design. The steady state DNBR analysis was perfomied using the ABB CE open-core thermal-hydraulics code (TORC) as described in Reference 7. TORC has received generic NRC approval for use in licensing analyses (Reference 9). TORC solves the conservation equations for a 3-dimensional model of the open-lattice core to determine the local coolant conditions at all points within the core. Lateral transfer of mass, momentum, and energy between neighboring flow channels were accounted for in the calculation of the local coolant conditions.
These coolant conditions were then used with the CE-l Critical Heat Flux (CHF) correlation (References 10 and i 1) to determine the MDNBR for the reactor core.
The core thermal-hydraulics modeling in TORC was divided into three stages. (TORC analysis can be done in two (2) stages or even one (!) if the computing capability is available. Problem size generally requires at least two (2) stages.) If the first stage is used (Figure 3.2-2), coolant conditions throughout the core on the coarse-mesh basis were calculated. A core quadrant was modeled, in which the smallest unit represented by a flow channel was a single fuel assembly. The geometry and heat generation of the fuel assemblies were input. The axial distributions of flow and enthalpy in each fuel assembly Southern California Edison 19 Novernber 1998
Section 3.2 Core Thermal-1-lydraulics Analysis were then calculated along with the transport quantities (mass, momentum, and energy) that cross the lateral boundaries of each flow channel. Stage 1 is not typically needed to evaluate reload changes In the second stage, (Figure 3.2-3) a hot assembly was divided into four assembly quadrants. One of these quadrants contains the limiting subchannel. The lateral transports of mass, momentum, and energy from the stage 1 analysis were imposed as boundary conditions on the peripheral boundary enclosing the hot assembly and its neighboring assemblies. The inlet Dow, inlet temperature, and fuel heat generation were also input into the code. The lateral transport of mass, momentum, and energy between the flow channels within the stage 2 mesh were calculated.
The third stage (Figure 3.2-4) involved fine-mesh modeling of the assembly quadrant containing the limiting subchannel. In this stage, the limiting subchannel, all of the other sub-channels, and all of the fuel rods were modeled individually. All of the flow channels used in this stage were hydraulically open to their neighbors. The lateral l transport of mass, momentum, and energy from the stage 2 calculations were imposed on the lateral boundaries of the assembly quadrant containing the limiting subchannel. The local coolant conditions were calculated for each flow channel. These coolant conditions were then input into the DNB correlation, and the MDNBR was determined for the most limiting subchannel in the reactor core.
The CETOP-D code, a variant of the TORC code, is used as a design code for the core I thermal-hydraulics analyses. CETOP-D was developed to reduce the computer time needed for core thermal-hydraulics analyses, while retaining the capabilities of the TORC design model. CETOP-D used transpan coefficients for improved prediction of diversion I cross flow and turbulent mixing between adjoining channels. Furthermore, a prediction-correction method is used to solve the conservation equations, replacing the iterative method used in the TORC code. A complete description of CETOP-D is contained in I Reference 12, and the generic applicability of the CETOP-D to SONGS is detailed in References 13 and 14.
I The CETOP-D code provided an additional simplification to the conservation equations due to the specific geometry of the model. It had a total of four core thermal-hydraulics channels to model the open-core fluid phenomena. Figure 3.2-1 shows a typical layout of these channels. Channel 2 was a quadrant of the hottest assembly in the core and Channel I was an assembly which represented the average coolant conditions for the remaining ponion of the core. The boundary between Channel 1 and 2 was open for crossflow, but there was no turbulent mixing across the boundary. The outer boundaries.of the total geometry were assumed to be impermeable and adiabatic. The lumped Channel 2 includes Channels 3 and 4. Channel 3 comprised the subchannels adjacent to the MDNBR hot Channel 4. f
]. The location of the MDNBR channel was determined from the TORC analysis of the core.
i Southern California Edison 20 November 1998
Section 3.2 CoreThermal-Hydraulics Analysis To produce a core thermal-hydraulics design model for a specific reload core, [
l i
] This DNBR limit was derived using the statistical method I discussed in Section 3.2.2.
3.2.1.1 Major Inputs to the Core Thermal-IIydraulics Design Pin-by-pin power distributions and the corresponding core wide power distribution for the potentially limiting assemblies are required as input to the TORC modeling process. Potentially limiting assemblies are determined based on the inlet flow distribution, assembly power, and the " flatness" of the pin-by-pin power distribution within the assembly.
l The detailed fuel assembly dimensions (clad outer diameter, pitch, spacer grid locations, etc.) and the dimensions of peninent reactor internals (e.g., core shroud) are based on a set of engineering drawings of the fuel and reactor internals. Other inputs include the core inlet flow distribution and core exit pressure distributions obtained I from flow model tests.
l The operating conditions that require definition in a TORC or CETOP-D case include l system pressure, core inlet temperature, core average heat flux and core average mass velocity. These operating conditions are based on the Reload Ground Rules, and are consistent with the safety and COLSS/CPCS design. Core physics data is required for I each reload while most other inputs are cycle independent, barring any design changes to the fuel, other fuel related components, or reactor internals.
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Section 3.2 Core Thermal-Hydraulics Analysis 3.2.1.2 Major Outputs of the Core Thermal-Hydraulics Design A 4-pump CETOP-D model (i.e., a model based on-full RCP flow) was provided for the reload safety design and for use in the COLSS and CPCS design. Typically, a cycle independent CETOP-D model is provided with a penalty applied to the core average heat flux. The CETOP-D model is valid for steady state normal operation allowed by the PDILs and for quasi-steady state analysis of transients, such as:
- a. Increased main steam flow
- c. Loss of normal AC power
- d. Uncontrolled CEA withdrawal from a subcritical or low power condition
- e. Uncontrolled CEA withdrawal at power
- f. CEA misoperation
- g. CEA ejection h.. Total loss of forced reactor coolant flow
- i. Steam system piping failure J. Feedwater system pipe breaks
- k. Reactor coolant pump shaft breaks
- 1. Steam generator tube rupture I Typically, the CETOP-D model is developed for two ranges of applicability, the COLSS (nanow) range and CPCS (wide) range since the COLSS and CPCS have different ranges of operability. Penalty factors applied to the core average heat flux for each of the two ranges are detennined in the benchmark process.
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Section 3.2 Core Thennal-liydraulics Analysis 3.2.2 Statistical Combination of Uncertainties The Statistical Combination of Uncertainties (SCU) methodology described in References 16 and 17 is used to calculate the minimum DNBR limit value (1.31 since Cycle 2), and I the DNBR Probability Density Function (PDF) used with the 95/95 probability / confidence DNBR tolerance limit in the COLSS and CPC uncertainty analyses.
The data required for a detailed thermal-hydraulic analysis are divided into two main groups: system parameters, which describe the physical system and are not monitored during reactor operation, and state parameters, which describe the opemtional state of the reactor and are monitored during operation.
I There is a degree of uncertainty in the value used for each of these parameters. The SCU methodology is used to statistically combine uncertainties of the system parameters and incorporate their effects on DNBR to derive the minimum DNBR limit. The individual uncertainties that are combined in the system parameter SCU analysis are the following:
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1 These uncenainties are statistically combined to yield the DNBR PDF. This DNBR PDF is then deterministically combined with the fuel rod bow and the High Impact Design (HID-1 and IIID-2) grid penalties to determine the minimum DNBR limit, Reference 21, I 50, and 52. This DNBR limit, while using the nominal values of system parameters in .
design analysis, will ensure with at least 95 percent probability and 95 percent confidence level that departure form nucleate boiling will not occur anywhere in the core. This limit is also used in the on-line COLSS DNBR power operating limit calculation and as the CPCS DNBR trip setpoint. The DNBR PDFis also used in the COLSS and CPC overall uncenainty analyses. .
In the Modified SCU methodology (Reference 15) the system parameter uncenainties are combined in the same way to determine the DNBR PDF. However, [
Southern California Edison 23 November 1998
Section 3.2 Core Thermal-Hydraulics Analysis l-The use of a response surface to represent a complicated, multi variate function is an established statistical method. A response surface relating MDNBR to system parameters is created. Conservatism is achieved by selecting the "most adverse set" of state parameters ihat maximizes the sensitivity of MDNBR to system parameter variations.
TORC analyses are performed to determirie the sensitivity of the system parameters at several sets of operating conditions (state parameters). Data to estimate the coefficients of the response surface is generated in an orthogonal composite design using the TORC code with the CE-1 CHF correlation.
The MDNBR PDF is estimated using the response surface in a Monte Carlo simulation.
The estimated MDNBR PDF is approximately normal and a 95/95 probability / confidence limit is assigned using normal theory. The SIGMA code applies Monte Carlo and stratified sampling technique to combine arbitrary PDFs numerically. (The Monte Carlo simulation and the SIGMA code have been reviewed and found acceptable (Reference 17)). This code is used with the response surface to combine system parameter PDFs I with the CHF uncertainty and the TORC code uncertainty into a resultant MDNBR PDF.
3.2.2.1 Major Inputs for the SCU Analysis The TORC model used in the DNBR analysis as discussed in section 3.2.2 is also g used in the development of the MDNBR limit. Physics design information on the I maximum bumup for determining the rod bow penalty on DNBR is an input. Other inputs include the detailed fuel assembly dimensions and tolerance values used to determine many of the system parameter PDFs.
I I
reload analysis.
l Other penalties imposed by the NRC in the course of their review of the Unit 2 Cycle 2 SCU antlysis (Reference 16) are included in the overall uncertainty penalty factors derived in the MSCU Analysis. [
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Southern California Edison 24 November 1998
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Section 3.2 Core hermal-flydraulics Analysis I
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3.2.2.2 Major Outputs of the Statistical Combination of Uncertainties Analysis
]
The DNBR PDF and its tolerance limit were provided for use in CPC and COLSS DNBR overall uncertainty analyses (Sections 3.5.2 and 3.6.2) using the MSCU l method. This PDF included NRC imposed statistical penalties (e.g., a 5% increase on l the CHF standard deviation) but excluded penalties that are applied deterministically l to determine the DNBR limit (e.g., rod bow and grid design penalty).
The DNBR limit with deterministic adjustments was used in calculating the CPC DNBR trip serpoint and the COLSS DNB-Power Operating Limit (POL) alarm j setpoint. Additionally, the DNBR limit was used in the Non-LOCA transient I analyses, and is cited in the Technical Specifications.
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Southern California Edison 25 November 1998
Section 3.2 Core Therrnal-flydraulics Analysis j Figure 3.2-1 l 5 Channel Geometry of CETOP-D Model l CHANNEL NUMBER
- 2 HOT ASSEMBLY =1/4 --*-
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Section 3.2 Core Thermal Ilydraulics Analysis Figure 3.2-2 Sample First Stage TORC Model Layout 1 2 l 3 4 5 6 7 8 9 10 11 12 13 1
14 15 16 17 18 19 20 l 21 22 23 24 25 26 27 28 I 29 30 31 32 33 34 35 36 1
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Section 3.3 FuelPerformance Analysis 3.3 FUEL PERFORMANCE ANALYSIS The primary licensing and design objective of the fuel performance design is to evaluate the steady state fuel thermal and mechanical behavior ofindividual nuclear fuel rods as a function of time or burnup. Generally, this requires generation of representa:ive values for fuel rod temperatures, rod internal gas pressure, and fuel rod deformation. This section describes the fuel performance licensing methodology for the SONGS reload analyses.
The ABB CE fuel evaluation model computer code, FATES, was developed to predict the steady state fuel rod temperature distribution, gap conductance, fuel and clad dimensions, plenum pressure, and stored energy for ABB CE designed fuel. FATES modeled fuel rods consisting of pelletized solid or annular UO2fuel encapsulated in a zircaloy cladding tube. The fuel pellet is l assumed to be a right circular cylinder. The fuel pellet model accommodates dimensional changes due to fuel relocation, densification, thermal expansion, and fission-induced swelling.
The fuel is modeled as a collection of discrete axial segments, where an independent radial l thermal equilibrium calculation is performed at each segment. [
l.
FATES 3B is used for the reload licensing analyses of both SONGS units. Models contained in FATES 3B describe the principal fuel rod behavioral phenomena, including thermal expansion, I relocation, densification, creep, swelling, fission gas generation and release, and elastic deformation. Detailed descriptions of the models are contained in References 18,19, and 20.
When compared to previous version of FATES, FATES 3B has an improved predictive I capability, especially with respect to fission gas generation and release, at high bumup.
FATES 3B has been approved for high bumup in Reference 20 and for urania-erbia fuel in Reference 42.
3.3.1 Major Inputs for Fuel Performance The fuel rod geometric parameters, the actual or projected power history and the core thermal-hydraulics conditions are required inputs to the FATES 3B analysis. These values are selected in accordance with the methodology outlined in Reference 8. The fuel rod geometric input parameters include clad length, clad inner and outer diameter, active fuel length, fuel pellet diameter, pellet dish and chamfer dimensions, and fill gas pressure.
I i l 1 The power history data is described by the fuel rod operating history. Typically, the core or rod average power data with axial and radial multipliers is used to obtain the LHR at each axial node at every time step. [
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Southern California Edison 30 November 1998
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Section 3.3 FuelPerformance Analysis 1
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1 3.3.2 Major Outputs for Fuel Performance The fuel performance analysis performed and data transmitted depends on the specific I application. The safety and LOCA analyses typically require initial fuel rod conditions, maximum rod internal pressure, core minimum and maximum gap conductances, rod minimum gap conductances, the minimum power to fuel centerline melt, the axial I densification factor, and the engineering factor on LHR. The data is in various forms, including values calculated directly by FATES 3B, FATES 3B generated output files used as initial conditions for transient codes, and values calculated extemal to FATES 38.
I Additionally, FATES 3B is used to verify Technical Specifications on permissible LHRs, peaking factors, and other limits to preclude fuel damage.
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u Southern California Edison 31 November 1998
Section 3.4 Non-LOCA Transient Analysis g 3.4 NON-LOCA TRANSIENT ANALYSIS This section discusses the methodology that SCE utilizes to analyze transients and events for reloads. A brief overview of the analysis methodology and the codes and highlights of the analyses will be presented. Two types of events are presented in the following section: UFSAR events whose consequences may be adversely affected by changes in reloads and those that are not. The UFSAR Chapter 15 events that are typically insensitive to changes in core parameters are discussed in Section 3.4.1 and are listed below:
- a. Decrease in Feedwater Temperature
- b. Increase in Feedwater Flow I c. Loss of Load, Turbine Trip, or Loss of Condenser Vacuum
- d. Loss of Normal AC Power to Station Auxiliaries
- e. Loss of Normal Feedwater Flow
- f. Feedwater System Pipe Breaks
- g. Startup of an Inactive Reactor Coolant Pump I h. Chemical and Volume Control System Malfunction - Pressurizer Level Control System Malfunction
- i. Inadvertent Operation of the ECCS during Power Operation
- j. Pressurizer Pressure Decrease Events
- k. Small Primary Line Pipe Break Outside Containment
- 1. Steam Generator Tube Rupture
- m. Inadvertent Opening of a Steam Generator Atmospheric Dump Valve I The preceding events are evaluated each cycle through a summary of transient calculation to account for possible plant design changes, licensing changes (Technical Specifications), and core parameter changes, as reflected in the Reload Ground Rules. Analysis of these events are l performed if necessary during the reload. Recent plant driven changes are Tcold Reduction I Program and increased steam generator tube plugging.
Southern California Edison 32 November 1998
Section 3.4 Non-IDCA Transient Analysis UFSAR Chapter 15 events which are always expected to be affected by a reload design are discussed in Section 3.4.2 ad are listed below:
- a. Fuel Failure Events
- 1. Increase in Main Steam Flow with single failure
- 2. Steam System Piping Failures
- 3. Single Reactor Coolant Pump Shaft Seizure / Sheared Rotor (SS/SR)
- 4. CEA Ejection
- b. Margin Setting Events (Anticipated Operational Occurrences)
- 1. Total Loss of Forced Reactor Coolant Flow
- 2. Asymmetric Steam Generator Events
- 3. Uncontrolled CEA Withdrawal
- 4. Single CEA Drop Events i l
i
- 5. Pan Length CEA Drop j
- 6. CEA Subgroup Drop Events l
- 7. CEA Withdrawal within Deadband
- 8. Inadvenent Boron Dilution
- 9. Increase in Main Steam Flow The inputs to the Non-LOCA transient analyses are derived from the Reload Ground Rules, 4 physics analysis, core thermal-hydraulics analysis, fuel performance analysis, and other cycle specinc parameters that affect the reload analyses. The major outputs of the Non-LOCA transient analyses are (a) de" for the COLSS/CPCS setpoints, and (b) update of appropriate Technical Specincations, Technical Specifications Bases, Licensee Controlled Specifications (LCS), Core Operating Limits Requirements (COLR), Design Bases Documents, and the Updated Final Safety Analysis Report (UFSAR).
Southern California Edison 33 November 1998
Section 3.4 Non-LOCA Transient Analysis The major computer codes used in the Non-LOCA Transient analyses are as follows:
- 1. CESEC-III System code for analyzing plant transient response (References 23 and 30).
- 2. TORC /CETOP-D Thermal-hydraulic codes for DNB calculations (References 7,9, I and 12) 1
- 3. HRISE Used in Non-LOCA transient analysis to determine the MacBeth DNBR of fuel at specific transient conditions (Reference 49).
I 4. CENTS System code for more accurate representation of NSSS response (Reference 48). It is the upgrade for CESEC-III.
- 5. HERMITE Used for steady-state or space-time coarse mesh neutronic and thermal hydraulic calculations used in the evaluation of reactor transients (Reference 4).
I 6. STRIKIN II Calculates the heat transfer from, and the peak clad temperature of the rod in CEA ejection analysis (Reference 27,28,29, and 52).
3.4.1 Events Not Normally Analyzed in the Reload Analysis Several of the safety design events presented in the UFSAR are not normally reanalyzed I as part of a reload design. These events are not typically reanalyzed since the results presented in the UFSAR are either insensitive or unaffected by reload changes or are bounded by other events. Although detailed analyses are generally not performed, each event is evaluated to ensure that the current cycle reload parameters are bounded by the assumptions of the UFSAR analyses. The following sections contain a brief discussion of the major events of this type, including the criteria used to determine if reanalysis is required. A more complete discussion of these events, including bases, acceptance criteria, methodology, and codes, is presented in Chapter 15 of the UFSAR (Reference 3) and the Accident Analysis Design Basis Document (Reference 39). In the following I discussion reference are made to UFSAR sections. Reference is also maje to the cycle specific section of Chapter 15, section 15.10. l Chapter 15 of the UFSAR is divided into two main sections, the original Chapter 15 and a cycle specific section 15.10. The original Chapter 15 consisted of sections 15.0 through 15.9 and Appendices 15.A through 15.F. These sections contain information of the l
LOCA and non-LOCA Transient analysis that has been reviewed and approved by the NRC. To capture cycle to cycle changes in these events, a Section 15.10 was added. This section is organized to match the original Chapter 15 organization. (Example: Increased Southern California Edison 34 November 1998
l Section 3.4 Non-LOCA Transient Analysis Main Steam Flow is presented in section 15.1.2.3. The cycle specific section is presented I in Section 15.10.1.2.3.) The cycle specific section provides a summary of the information in the main Chapter 15 event description including tables of key input data and sequence of events. Results are also presented relative to acceptance criteria. No figures are presented in Section 15.10. Section 15.10 is updated each cycle, as appropriate.
3.4.1.1 Decrease in Feedwater Temperature The Decrease in Feedwater Temperature events (UFSAR, Sections 15.1.1.1 and I 15.1.2.1,15.10.1.1.1, and 15.10.1.2.1) are typically less limiting than the Increased Main Steam Flow event because the smaller cooldown that results from this type of event will cause less or a core power increase and, therefore less of a transient DNBR decrease. The current limiting single failure is the failure of the turbine bypass valves resulting in additional flow through the bypass valves. The decrease in Feedwater Temperature plus single event is less limiting than the increased main steam flow plus single failure event, since AC power remains available, providing greater core flow and a larger margin to the DNBR limit.
3.4.1.2 Increase in Feedwater Flow The Increase in Feedwater Flow event (UFS AR Section 15.1.1.2,15.1.2.2, 15.10.1.1.2 and 15.10.1.2.2) is less limiting than the Increased Main Steam Flow event since the cooldown that results from an increased opening of the feedwater valve or an increase in feedwater pump speed will cause relatively less power increase and, therefore,less of a transient DNBR decrease. The curTent limiting single failure is the failure of the turbine bypass valves resulting in additional flow through the bypass valves.
3.4.1.3 Loss of Load, Turbine Trip, or Loss of Condenser Vacuum The Loss of Load, Turbine Trip, or Loss of Condenser Vacuum analyses presented in the UFSAR, Section 15.2.1.1,15.2.1.2,15.2.1.3, 15.2.2.1, 15.2.2.2, 15.2.2.3,15.10.2.1.1,15.10.2.1.2,15.10.2.1.3,15.10.2.2.1,15.10.2.2.2, and 15.10.2.2.3 were based on generic core physics parameters chosen to bound all future l
i cycles. The Loss of Condenser Vacuum event is the limiting Chapter 15 transient for moderate frequency over-pressurization events, for which primary and secondary transient pressures must not exceed 110% of the system design pressures. In such events, the Pressurizer Safety Valve (PSV) and Main Steam Safety Valve (MSSV) lift setpoints limit the primary and the secondary system maximum pressures. Since these are pressurization events, transient DNBR improves. These events are typically not reanalyzed as part of a reload design unless changes occur in the performance characteristics of the valves, the initial power level, the allowable RCS inlet temperature range or other key plant parameters.
l Southern California Edison 35 November 1998
Section 3.4 Non-LOCATransient Analysis f
3.4.1.4 Loss of Normal AC Power To Station Auxiliaries
( The analysis of this event as presented in UFSAR Sections 15.2.1.4, 15.2.2.4, 15.10.2.1.4, and 15.10.2.2.4 is performed to demonstrate an adequate secondary heat sink to ensure that the criteria regarding fuel damage and RCS and secondary system
( pressure are met. The overpressure consequences of this event are bounded by the Imss of Condenser Vacuum event.
{ 3.4.1.5 Loss of Normal Feedwater Flow The Loss of Normal Feedwater Flow event (UFSAR Sections 15.2.2.5,15.2.3.2,
{ 15.10.2.2.5, and 15.10.2.3.2) is analyzed for peak RCS and secondary pressure and adequate secondary heat sink. This event typically is not re-analyzed if the physics parameters bound the reload fuel cycle values. The cunent limiting single failure for this event is all four steam dump and bypass system valves stuck open.
3.4.1.6 Feedwater System Pipe Breaks The Feedwater Line Break event (UFSAR Section 15.2.3.1 and 15.10.2.3.1) is the limiting transient for very low probability overpressurization events, for which primary and secondary transient pressures must not exceed 120% of the system design pressures, and for over filling of the pressurizer and passing water through the PSVs.
Radiological consequences are bounded by the Post Trip Main Steam Line Break event. The current analysis was based on a limiting 0.2 square foot break size, which will remain bounding unless the generic kinetics parameters or applicable system
{ parameters change. Examples of when reanalysis is needed are if the PSV or MSSV characteristics change, if the deliverable flows from the main or auxiliary feedwater pumps are reduced or if the number of plugged tubes is increased.
3.4.1.7 Startup of an Inactive Reactor Coolant Pump
[ The UFSAR, Section 15.4.1.5 and 15.10.4.1.5 analysis of the startup of an inactive reactor coolant pump event considered both heat up and cooldown cases with sufficiently conservative isothennal temperature coefficients to bound future reload
{ designs. This event is analyzed for Modes 3,4, and 5 since in Modes 1 and 2 all four RCPs must be in operation precluding this event. Mode 6 is not performed because p RCPs are not allowed to operate and the reactor vessel head is detensioned. The i conclusion of this analysis was that this event would do not result in the loss of the minimum required shutdawn margin. Since the shutdown margin is n_ot lost during the event, there is no increase in heat flux and therefore no decrease in minimum DNBR. Also, when the RCS is above the conditions requiring low Temperature Ovegressure Protection (LTOP), the peak RCS pressure will not exceed 110% of design pressure in response to the startup of an inactive reactor coolant pump event.
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Southern California Edison 36 November 1998
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Section 3.4 Non-LOCATransient Analysis While the RCS is in the LTOP mode, the Shutdown Cooling System relief valves will prevent violation of RCS integrity limits.
3.4.1.8 Chemical and Volume Control System Malfunction-Pressurizer Level Control System Malfunction The pressurizer level control system malfunction causes the startup of all three charging pumps and minimizes let down flow, (UFSAR Sections 15.5.1.1,15.5.2.1, 15.10.5.1.1 and 15.10.5.2.1). The pressure transient is due to an increase in primary coolant inventory and not due to thermal expansion, as in the Section 3.4.1.3 events.
The peak pressure reached for the UFSAR case was less than 110% of the design pressure (i.e.,2750 psia) and no liquid flow through the PSV's result. Operator response is required within 15 minutes to prevent pressurizer overfill.. Additionally, since the RCS pressure increases, the transient DNBR improves. These events are typically not reanalyzed, unless changes occur in the performance characteristics of the PSVs, the initial power level, or the RCS inlet temperature range.
3.4.1.9 Inadvertent Operation of the ECCS During Power Operation The inadvertent operation of the ECCS event (UFSAR Sections 15.5.1.2, 15.5.2.2, 15.10.5.1.2 and 15.10.5.2.2) is caused by an inadvenent safety injection actuation signal that stans all charging pumps, isolates let down, stans boric acid makeup pumps, and high pressure safety injection pumps. The consequences of this event are bounded by the Chemical and Volume Control System Malftmetion-Pressurizer Level Control System Malfunction event. The current limiting single factor is the failure of the BAMU pump to start thus causing the charging pumps to switch to the RWST.
3.4.1.10 Pressurizer Pressure Decrease Events The inadvenent opening of a PSV event (UFSAR Section 15.6.3.4 and 15.10.6.3.4)is evaluated by the fuel vendor as pan of the ECCS analyses.
3.4.1.11 Small Primary Line Pipe Break Outside Containment l
In the UFSAR analysis of a small primary line pipe break category (UFSAR Section 15.6.3.1 and 15.10.6.3.1), a double-ended break of the letdown line outside the containment and upstream of the letdown line penetration control valve was selected.
This event was selected since this break location results in the largest release of reactor coolant outside the containment. The transient DNBR remains above the DNBR limit. The radiological release rate was based on an assumed maximum equilibrium value that would occur with and without iodine spiking. Since this limit is not dependent on reload design parameters, this event is typically not reanalyzed.
Southern Califomia Edison 37 November 1998
Section 3.4 Non-LOCA Transient Analysis 3.4.1.12 Steam Generator Tube Rupture UFSAR Sections 15.6.3.2 and 15.10.6.3.2, present the Steam Generator Tube Rupture event with an assumed LOAC. The analysis was performed to determine the radiological doses to the environment and was conservatively based on an assumed initial coolant activity consistent with Technical Specification primary and secondary activity limits, and with and without a coincident (existing) iodine spike. Since the I significant input parameters for this event are not dependent on reload design parameters, this event is typically not reanalyzed.
3.4.1.13 Inadvertent Opening of a IOSGADV The IOSGADV event (UFSAR Sections 15.1.1.4,15.1.2.4,15.10.1.1.4 and I 15.10.1.2.4) is caused by the opening of a single ADV or main steam safety valve.
The current limiting single failure is a loss of AC power. This event is not impacted by changes in fuel management and thus not typically analyzed on a reload basis.
This event is the limiting event for dose releases for non-limiting fault events that do not result in fuel failures.
I 3.4.2 Events Normally Analyzed in the Reload Analysis This section presents the methods used to analyze the events which may change as the result of a reload. Descriptions of the analyses are presented in the following five sub-sections:
- 1. Section 3.4.2.1 - Fuel Failure Events.
- 2. Section 3.4.2.2 - Margin Setting Events (AOOs)
- 3. Section 3.4.2.4 - Degraded Performance of CPCS & COLSS Category
- 4. Section 3.4.2.5 - Verification of Transient Related CPCS Constants.
3.4.2.1 Fuel Failure Events The four events typically evaluated for fuel failure and dose consequences as part of a 1 reload safety design are: (a) Increased Main Steam Flow with Single Failure,(b)
Steam System Piping Failures,(c) Single Reactor Coolant Pump SS/SR, and (d) CEA Ejection. Each event is discussed separately below. ,
e L Southern California Edison 38 November 1998
Section 3.4 Non-LOCA Transient Analysis 3.4.2.1.1 Increased Main Steam Flow With Single Failure
- a. Description of the event The most limiting combination of an increased heat removal event, without a steam line rupture, in terms of DNBR, is the increase in main steam flow with a concurrent single failure, as presented in UFSAR Sections 15.1.2.3 and ,
._ 15.10.1.2.3. A turbine steam bypass valve may be inadvertently opened by the operator due to failure of the associated control system. The opening of any of I these valves increases the rate of heat removal by steam generators, causing a cooldown of the RCS.
I The most limiting single failure is chosen to yield the greatest decrease in DNBR after initiation of a reactor trip signal. In addition, the event is evaluated with the lowest possible pre-trip DNBR. Parametric analysis has determined that the loss of all AC power (LOAC), when a reactor trip condition exists, produces the most adverse consequences following an increased main steam flow. The loss of RCS flow that results from the LOAC causes a greater decrease in DNBR after a reactor trip than any other possible single failure.
This event is currently the limiting event for radiological dose consequences for the infrequent incidents category.
- b. Analysis criteria Since the Increased Main Steam Flow + LOAC is an infrequent event, violation of I the SAFDLs is permissible. This analysis must demonstrate, however, that a coolable core geometry, Reference 35, is maintained and that the radiological consequences are well within the 10CFR100 limits.
- c. Objectives of the analysis I The objective of maintaining coolable geometry is met by showing that DNB propagation does not occur. Each cycle a radiological consequences evaluation / calculation is performed to show that the event doses are well within I the 10CFR100 dose limits or a COLSS required over power margin (ROPM) penalty and/or CPCS power penalty is calculated to reduce the dose consequences to be well within the 10CFR100 dose limits.
~
- d. Basic assumptions and justifications The CESEC-III and TORC (References 30,23 and ~7) codes have been used to simulate the reactor core response during this transient. The CETOP-D code (Reference 12) can be used instead of TORC as CETOP-D is always conservative i Southern California Edison 39 November 1998
I Section 3.4 Non-tDCA Transient Analysis relative to TORC and is easier to u.se. In the analysis of this event,it is .
conservatively assumed that the initial excess load portion of the event reduces the core DNBR tojust above the SAFDL The LOAC then begins at the SAFDL I The High Power Level and CPCS trips (DNBR, local Power Density) provide primary protection during this event. Additional protection is provided by other trip signals including Low Steam Generator Water level, Low Steam Generator pressure and CPCS' Variable Ovegower Trip (VOPT) or RCP Speed Out of Range. Other assumptions that are used in this analysis are listed below.
I 1. The most conservative Moderator Temperature Coefficient (MTC) within the LCS is used (i.e., the MTC which produces the most adverse DNBR results).
I 2. A conservative Fr is used, such that the transient initial condition DNBR is at the DNBR limit.
I 3. A conservative flow fraction versus time function is used to represent the LOAC.
- 4. Minimum fuel-clad gap conductance is used. This is conservative because a smaller gap conductance delays the heat flux decay due to reactor trip.
- e. Analysis method This transient is initiated at SAFDL conditions as determined using the TORC code or conservatively by the CETOP-D code. The CESEC-IH code is then used to produce the time dependent core average heat flux, hot assembly heat flux, and the core inlet coolant mass flux. The TORC or CETOP-D code is used to determine the actual transient DNBR The amount of fuel failure is then I calculated based the minimum DNBR and on the limiting pin census3 . The radiological consequences are calculated in accordance with UFSAR including Appendices 15B and 15.108. i If the dose consequences exceed the well withir.10CFR100 limit for these events, then a COLSS required overpower margin (ROPM) penalty and/or a CPCS power l I penalty will be calculated to reduce the dose consequences to well within 10CFR100 limits for these events.
i 3
See secuan 3 I.i.2 1
Southern California I.dison 40 November 1998
Section 3.4 Non-LOCA Transient Analysis
- f. Current additional conservatisms The analysis includes the following conservatisms:
- 1. No credit is taken for the delay between the turbine trip and the LOAC.
The DNBR due to the fourpump coast down resulting from the LOAC would be less severe if this delay was included in the simulation.
- 3. Fuel failure is not determined through the use of DNBR statistical convolution methodology, Reference 14.
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3.4.2.1.2 Steam System Piping Failures
- a. Description of the event As described in UFSAR Section 15.1.3.1 and 15.10.1.3.1, a pipe break in the main I steam piping increases the steam flow from the affected and unaffected steam generators. This greatly increased steam flow enhances the rate of RCS heat removal by the secondary system. The increased rate of heat removal causes a I decrease in core coolant inlet temperature. In the presence of a conservatively assumed negative MTC, the decrease in core inlet temperature causes an increase in core reactivity and results in an increase in core power. The excursion in core I power is terminated by the action of one of the following RPS trips: CPCS generated trip, Low Steam Generator Pressure, High Containment Pressure, High Linear Power Level, or VON.
Following the reactor trip, the continued decrease in inlet temperature could cause the shutdown margin to be degraded to the point that the core retums to power.
Steam line break cases are chosen to maximize the potential for (a) a post-trip return to power or (b) degradation in fuel cladding performance, which would maximize dose at the site Exclusion Area Boundary (EAB). Of the nine cases presented in the UFSAR Section 15.1.3.1, five cases were chosen to maximize potential for a post-trip return to power and one case (defined as pre-trip power excursion case) was chosen to maximize potential for degradation in fuel per-fonnance and three radiological assumption cases were chosen to maximize dose at the site EAB.
In the reload design, the two limiting cases are analyzed to determine if a post-trip core return to power will occur: 1) hot full power case and 2) hot zero power case.
I Additionally, the most limiting steam line break outside containment during full Southern California Edison 41 November 1998
I Section 3.4 Non-LOCA Transient Analysis I power operation, the pre-trip power excursion case, is evaluated to determine the worst case fuel performance.
- b. Analysis criteria The steam line break event is a limiting fault event. The results of the analysis must show that:
- 1. Fission power remains sufficiently low following a reactor trip to preclude degradation in fuel performance as a result of post-trip return to power.
- 2. Degradations in the fuel performance prior to trip is of sufficiently limited extent.
- 3. The core will remain in place and intact with no loss of core cooling capability.
- 4. Doses are within the 10CFR100 requirements as specified in the Standard Review Plan.
- c. Objectives of the analysis I The objective of the analysis is to demonstrate that the dose consequences for this event do not exceed the 10CFR100 limits for these events. If the 10CFR100 limits are violated, the dose consequences are reduced by either providing a COLSS required overpower margin (ROPM) penalty and/or a CPCS power penalty to reduce the dose consequences or by obtaining more detailed physics and/or pin census data.
- d. Basic assumptions and justification The analysis assumptions are similar to those given in Reference 3. Only those parameters that are reload dependent are discussed here:
- 1. Assumptions used in the cases chosen to maximize potential for a post-trip retum to power:
- a. Reactivity insenion versus fuel temperature based on the most negative End of Cycle (EOC) Fuel Temperature Coefficient (FFC) j is used.
l
- b. The EOC core kinetics parameters are used. The smallest delayed neutron fraction existing at the EOC is chosen to enhance the power transient.
Southern California Edison 42 November 1998 I
y c. A minimum gap conductance is used for the analysis which tends g
to retain heat in the fuel. Moderator temperatures will not increase as much if a relatively higher gap conductance is used. Therefore, I the reactivity effect due to moderator temperature is conservatively modeled.
- d. The reactivity insertion versus moderator temperature with All I Rods In (ARI) with Worst Rod Stuck Out for EOC is used. This is calculated at EOC with the most negative MTC to result in the greatest positive reactivity addition during the event.
I This is the least amount of reactivity that would be inserted after the reactor trip.
- 2. Assumptions used in the cases chosen to maximize potential for deg-radation in fuel per'ormance:
l a. The initial conditions and event initiators are chosen to result in the most adverse power excursion and fuel perfomiance degradation for this event.
I b. The least negative FTC of reactivity is used. This will minimize the amount of negative reactivity feedback generated by the fuel I during the power excursion.
- c. RPS trip setpoints are conservatively modeled to account for I uncertainties which result in conservative response times.
- e. Analysis method
( The Nuclear Steam Supply System (NSSS) response to the steam line break is
_[ simulated using the CESEC-Ill computer code (References 23 and 30). The I
CETOP-D computer code (Reference 12) is used to determine the MDNBR for
! the fuel performance case. The llRISE computer code (Reference 49) is used to determine the MDNBR for the retum to power cases. More details on simulating pre-trip and post-trip cases can be found in Appendices 15E and 15.10E of
!, Reference 3.
- f. Cun ent additional conservatisms l
Conservatism are included in the analysis, such as using only one high and low pressure safety injection pump and including uncertainties in MTC and FTC of reactivity.
u Southern California Edison 43 November 1998
Section 3.4 Non-LOCA Transient Analysis 3.4.2.1.3 Single Reactor Coolant Pump Sheared Shaft (SS) / Seized Rotor (SR)
- a. Description of the event A SR event can be caused by seizure of the upper or lower reactor coolant pump thrust journal bearings. A single reactor coolant pump SS could be caused by a mechanical failure of the pump shaft. Following an event of this type, the core flow decreases asymptotically to the 3-pump flow rate. The reduction in coolant flow rate causes an increase in the core average coolant temperature and could result in some fuel pins experiencing DNB.
In the case of a SR event, a reactor trip on pump speed occurs soon after the pump shaft seizes. For a SS event, a reactor trip is generated when the rapid flow reduction causes the pressure drop across the steam generator (in the affected loop) to decrease below the trip setpoint.
- b. Analysis criteria The SS/SR events are limiting fault events, however, for these events the corresponding off-site doses must be a small fraction of 10CFR100 limits (UFSAR Sections 15.3.3.2 and 15.10.3.3.2) and this analysis must also demonstrate that a coolable core geometry is maintained.
- c. Objective of the analysis The objective of the analysis is to demonstrate that radiological dose consequences remain a small fraction oft he 10CFR100 limits or the dose consequences are reduced by either providing a required overpower margin penalty (ROPM) in COLSS or by obtaining more detailed physics and/or pin cen-sus data.
- d. Basic assumptions and justification The analysis assumptions are similar to those given in UFSAR Sections 15.3.3.2 and 15.10.3.3.2 The initial conditions are chosen to maximize the amount of fuel failure. For example:
1
- 1. Of the two events, the SS event is the more limiting because the SR flow coastdown is used but SS has a longer response time to trip. Therefore, only the SS event is typically analyzed.
Southern California lifison 44 November 1998
Section 3.4 Non-LOCA Transient Analysis
- 2. The least negative Doppler curve is used to minimize the reactivity change during core heat up. This least negative data results in a higher core power at the time of reactor trip and also delays the core power decrease, resulting in a later DNBR tumaround and a lower core flow at the time of MDNBR.
- 3. The Beginning of Cycle (BOC) core kinetics parameters are used. This I delays the core power decrease after reactor trip, resulting in a later DNBR turnaround and a lower core flow at the time of MDNBR.
I 4. The most positive MTC is used to provide a more adverse reactivity insenion due to moderator feedback.
I
- 5. The minimum gap conductance is used because this delays core heat flux decrease after trip, thus resulting in a lower core flow at the time of l minimum DNBR
- 6. The use of the a limiting negative axial shape index produces the most I limiting results.
- e. Analysis method This analysis used the CESEC-III computer code (References 23 and 30) to model the NSSS response and the TORC code (Reference 7) or the conservative I CETOP-D code (Reference 12) is used to determine the transient DNBR response.
Statistical convolution technique was used to describe the number of failed pins (References 13 and 14).
- f. Current additional conservatisms A number of conservatisms are included in the analysis, such as: uncertainties which are conservatively applied to core power, core inlet temperature, RCS pressure, Axial Shape index (ASI) and integrated radial peak, and use of the most positive MTC per the LCS.
3.4.2.1.4 CEA Ejection l
- a. Description of the event A CEA ejection is postulated by the circumferential rupture of a Control Element Drive Mechanism (CEDM) housing or nozzle. The CEA ejection which results in the most rapid positive reactivity addition is evaluated. A CEA ejection would cause a large increase in the overall power and a highly skewed and severely peaked power distribution. Protection against the effects of a CEA ejection is Southern California Iklison 45 November 1998
Section 3.4 Non-LOCA Transient Analysis j provided by the inherent Doppler feedback, log Power Trip, High Pressurizer u Pressure trip and the VOPT.
- b. Analysis criteria The CEA ejection is classified as a Limiting Fault event. The criteria used are:
- 1. All fuel pins which are calculated to experience DNB, based on the synthesis method described in Reference 25, are assumed to fail.
- 2. All fuel pins with greater than 250 cal /gm total centerline enthalpy are assumed to fail.
- 3. All fuel pins with greater than 200 cal /gm total average enthalpy of the hottest fuel pellet are assumed to fail.
1
- 4. No fuel pins shall exceed 280 cal /gm total radial average enthalpy.
- 5. The peak pressures of the primary and secondary systems are below the emergency condition limits (120% of the design values) as defined in Section III of the ASME Boiler and Pressure Vessel Code (3000 psia and 1500 psia, respectively).
- 6. The off-site doses are within 10CFR100 requirement.
- c. Objective of the analysis The objective of the analysis is to demonstrate that the dose consequences for this ,
event are less than the 10CFR100 limits for this event. This ensures that the site l boundary doses are less than those previously reported. If the 10CFR100 limits are violated, the dose consequences are reduced by either providing more Required Overpower Margin (ROPM) penalty in COLSS or by obtaining more detailed physics and/or pin census data.
- d. Basic assumptions and justification The analysis assumptions are similar to those given in UFSAR Sections 15.4.3.2 and 15.10.4.3.2. Only those parameters that are reload dependent are discussed below.
l
- 1. The most conservative ejected rod worth as a function of power level and rod insertion limits is used. The ejected rod worth largely defines the magnitude of the power spike.
l I
, _- c_ _ . - ,,,8
l Section 3.4 Non-LOCA Transient Analysis
- 2. The most conservative Fr is used.
- 3. Conservative, normalized pin censi are used for fuel failure calculations.
- e. Analysis method g The analysis method described in References 13,14, and 25 is used for this B analysis. For cases where an immediate high flux trip is generated, the STRIKIN code (References 27,28,29, and 52) is used to credit the delay in the fuel pin thermal response.
l For cases where the ejected rod worth is insufficient for an immediate high flux trip (referred to as delayed, slow, or no trip CEA ejection case), the CESEC-III code (References 23 and 30) is used to predict system response due to its inclu-sion of a secondary system model. At each power level, a worse case set of elevated power and RCS temperature conditions are used to determine the j' resultant 3D peaking factor (Fq) achievable during that CEA ejection. ROCS /MC code (References 1 and 41) are used to determine the post-ejected Fq assuming bounding fuel and RCS temperature feedback options. The resulting Fq must be less than the limiting Fq allowed by the LHR SAFDL. This subset of CEA ejection cases does not credit any trip functions by assuming steady state I operation at the elevated power and temperatures.
- f. Current additional conservatisms f A number of conservatisms are included in the analysis as discussed above.
3.4.2.2 Margin Setting Events l'
Many of the events presented in this section have a close relationship with the reactor protection and monitoring systems. The Core Protection Calculator System (CPCS) l and Core Operating Limit Supervisory System (COLSS) are key subsystems within j the plant protection and monitoring systems. The following text presents a brief
,I discussion of the relationship of the safety design to these systems and provides some l key definitions.
I The Plant Protection System (PPS) is composed of two subsystems: the Engineered l
l
. Safety Features Actuation System (ESFAS) and the Reactor Protection System. The !
CPCS has a direct relationship to the Plant Protection System and Reactor Protection System, since it provides four independent channels for two of the RPS trip functions-the low DNBR trip and the high Local Power Density (LPD) trip. The CPCS design ,
basis also includes provisions for auxiliary trip functions, which provide protection l for cenain design basis events. The auxiliary trips are used to: [ !
l Southem California Illison 47 November 1998
1 Section 3.4 Non-LOCA Transient Analysis
]
The CPCS has a bidirectional relationship with the safety analyses. In one direction, the analysis of some events determine the values of CPCS setpoints or constants. In the other direction, many transients credit CPCS response to provide protection or mitigate event consequences.
The COLSS program is contained in the PMS and receives input from and provides output to other parts of the Plant Monitoring System (PMS). It also has a direct relationship to the Technical Specifications, since the system is designed to provide information to assist the reactor operators in monitoring the Limiting Conditions for Operation (LCOs) for LHR mr.rgin, azimuthal tilt, DNBR margin and Axial Shape Index (ASI). These LCOs require that if the COLSS is in service, the core power must be less than the COLSS DNBR and LHR Power Operating Limits (POLS), and licensed power limit, and the COLSS calculated azimuthal tilt and ASI must be within the specified limits. By assisting the operator in maintaining these LCOs, COLSS maintains the initial conditions assumed in the safety analyses and for the calculation of certain Plant Protection System and Reactor Protection System setpoints, including CPCS constants. Most safety and setpoint analyses assume initial conditions at a COLSS calculated POL in order to assure the limiting condition of I minimum initial thermal margin. In addition, the analysis of events'such as loss of flow, CEA drop, and asymmetric steam generator transient events, directly determine COLSS thermal margin design requirements to preclude any fuel pin failures, as I described below.
Thermal margin enters into the relationship between the core monitoring and I protection systems (COLSS and CPCS) and reactor operation, since the amount of thermal margin affects the flexibility and power capability of the reactor. In order to understand the importance of themial margin, it is necessary to define the following:
Available Overpower Marcin (AOPM) is defined as the ratio of the power to the I DNBR SAFDL at any given combination of the other thermal hydraulic parameters (mass flowrate, RCS pressure, core inlet temperature, radial peaking factor, and axial power distribution) to the actual core power at the time point ofinterest.
Reauired Overpower Marnin (ROPM) is the ratio of the AOPM at the begirming of a transient to the AOPM at the point of minimum DNBR during the transient, as discussed in UFSAR Appendices 15A and 15.10A.
The required initial margin is inherently reserved in the COLSS calculated POL through the use of two methods: a) the Under Flow Fraction (UFF) and b) the ROPM.
I d
i.e.. events categorized as anticipated operational occunences ( AOOs)
Southern Califomia Edison 4h November 1998
In the ROPM method, the COLSS calculates the core power at which the DNBR SAFDL is reached based on the measured temperature, pressure, Dow, radial peaking factor, and axial power distribution. The power is then divided by the ROPM to obtain the POL. In the UFF method, a POL is calculated based on the measured temperature, pressure, and power distribution, but not the measured flow. Instead, the flow equivalent to the product of the measured flow and the UFF is used. The minimum value of these two POLS is then compared to the measured core power. If the calculated power limit is less than the measured power, a COLSS alarm occurs so that appropriate operator action can be taken. Usually, the UFF is used in COLSS to reserve the required margin for the loss of flow event. This is because the flow I coastdown during loss of flow type events produce the major effect on the DNBR degradation and the required margin can be more accurately calculated. The ROPM is typically used to preserve the margin for all other transients.
A number of events are evaluated during each reload to determine the margin l requirements, namely UFF and ROPM, that are needed to ensure that there is no fuel failure for these events. For events initiated at less than 70% power, ROPM requirements are offset against AOPM when possible to ensure no failed fuel. These events are discussed below.
3.4.2.2.1 Total Loss of Forced Reactor Coolant Flow
- a. Description of the event A loss of flow event may result from a LOAC to one or more of the four Reactor
' Coolant Pumps (RCPs). As the flow through the core decreases, the system temperature and pressure increase. The total loss of flow from four (4) RCPs event is analyzed to determine the minimal initial margin that must be maintained
[ by the Technical Specification LCOs, such that the DNBR SAFDL is not violated during the event. The initial margin is monitored by the COLSS through a required UFF or ROPM as described in section 3.4.2.2.
- b. Analysis Criteria
( The loss of Dow event is an AOO for which the following criteria must be met:
(1) the transient MDNBR must be greater than or equal to the DNBR SAFDL, and (2) the peak LiiR must be less than or equal to the LHR S AFDL.
- c. Objectives of the analysis The objective of this analysis is to determine the required UFF as a function of ASI or ROPM that will ensure that the analysis criteria are met. The resulting required UFF versus ASI function or ROPM is implemented in the COLSS design to maintain the required initial margin for a loss of flow event. Since there is no Southern California IWon 49 November 1998
Section 3.4 Non-IDCA Transient Analysis power excursion during the transient, the LHR SAFDL is not challenged during the event.
- d. Basic assumptions and justification In general, all initial conditions, such as F,, ASI, temperature, pressure, etc., are chosen for maximum adverse sensitivity during a loss of flow event. The key reload dependent assumption for this analysis is:
- 1. The most positive MTC allowed by the LCS is used. This is conservative because the positive reactivity inserted (as a result of the reactor coolant system heat up just prior to the reactor trip) will delay the power decay following the scram and maximize the initial margin required to accommodate DNBR degradation during the transient.
- c. Analysis method The methodology used for this event is described in UFSAR Sections 15.3.2 and I 15.10.3.2 and Appendices 15F and 15.10F, and in Reference 13. The analysis uses a ID HERMITE code (Reference 4) for the short term simulation. The I NSSS response is simulated using the CESEC-III code (References 23 and 30) for the extended simulation. For a typical reload analysis only the shon term simu-lation is performed in conjunction with the CETOP-D code (Reference 12) to determine the transient DNBR response.]
- f. Current additional conservatisms A number of conservatisms are included in the analysis, such as not crediting the pressure increase during the transient and assuming a conservative initial RCS flow rate.
3.4.2.2.2 Asymmetric Steam Generator Events I a. Description of the event There are four postulated events that could cause an Asymmetric Steam Generator Transient (ASGT) event: (1) loss of load to one steam generator, (2) loss of feedwater flow to one steam generator,(3) excess feedwater to one steam generator, and (4) excess load to one steam generator. Of these four events, the loss of load to one steam generator produces the largest core inlet temperature distonion which results in the largest power distonion and is described below.
I l The reactor is assumed to be initially operating at full power when the Main Steam Isolation Valves (MSIVs) on one of the steam generators instantaneously Southern California !!dison 5n November 1998
- Section 3.4 Non-LOCA Transient Analysis close, isolating the steam flow from one steam generator. With the loss ofload to one steam generator caused by a spurious closure of the MSIV, the pressure and temperature of the RCS increases. Also, the water level of the isolated steam generator rapidly drops as the increasing secondary system pressure and temperature collapse the steam bubbles in the liquid inventory. The pressure could continue to increase until the secondary safety valves open. The pressure of the other steam generator remains steady or drops depending on the mode of operation of the turbine generator control system.
l la An asymmetry in the core inlet temperature distribution occurs when the temperature in the primary coolant loop associated with the isolcred steam generator increases due to a reduction in the primary-to secondary heat transfer rate caused by the termination of steam flow from the isolated steam generator.
The core inlet temperature in the primary coolant loop associated with the unaffected steam generator decreases due to the cooling action of the unaffected l generator which " picks up" the load lost by the isolated steam generator.
In the presence of a negative MTC and fuel temperature coefficient, the radial l power increase will occur on the cold side of the core where no mixing of the core inlet flows is assumed. The outermost fuel bundles experience the greater increase in radial peak (resulting in a higher hot pin power). A greater increase in I the hot pin power occurs since the local MTC at the hot spot is more negative than the core average MTC. This radial peaking factor increase leads to an increase in the peak LHR and a decrease in DNBR which is partially mitigated by the I decreasing coolant temperature at the peak location. By contrast, the power on the hot side of the core decreases due to negative reactivity addition arising from moderator temperature effects. Doppler reactivity feedback effects act to re-I stabilize the core and flatten the radial power distribution.
A reactor trip occurs when the asymmetry in cold leg temperatures exceeds a I setpoint value implemented in the CPCS. Sufficient initial ROPM must be preserved so that the SAFDLs will not be violated for the ASGT type of events.
- b. Analysis criteria The ASGT events are classified as AOOs for which the following criteria must be met: (1) the transient MDNBR must be greater than or equal to the DNBR SAFDL and (2) the peak LHR must be less than or equal to the LHR SAFDL.
I c. Objectives of the analysis l The calculated ROPM must be such that the DNBR LCO preserved by the COLSS, in conjunction with the CPCS generated reactor trip, assures that the DNBR and LHR SAFDL are not violated.
s Southern California Iklison 5l November 199X
I Section 3.4 Non-LOCA Transient Analysis
- d. Basic assumptions and justification Many of the assumptions in this analysis, such as temperature stratification in the core, instantaneous MSIV closure, pressurizer pressure, and steam bypass control system status, are the same as in UFSAR Sections 15.9.1.1 and 15.10.9.1.1. Other assumptions that are reload related are listed below.
- 1. The BOC kinetic parameters are used to minimize the heat flux decay after reactor trip.
- 2. The EOC Doppler (most negative) reactivity and uncertainties are used to minimize the heat flux decay after trip.
- 3. The most negative MTC within the LCS is used to maximize the reactivity feedback. This increases peak power and heat flux, and maximizes F,. As a result, power shifts to the cold side of the core, with the power in the outer fuel assemblies usually experiencing the greatest increase, while power on the hot side decreases. This maximizes the calculated ROPM since the distonion of the radial peak is maximized.
- 4. The minimum gap conductance is used to minimize the decay of the heat flux after the trip.
- e. Analysis method The NSSS response to the ASGT event is simulated using the CESEC-III computer program (References 23 and 30). The CETOP-D code (Reference 12) is I used to determine the ROPM required so that the SAFDLs are not violated.
- f. Current additional conservatisms None 3.4.2.2.3 Uncontrolled CEA Withdrawal
- a. Description of the event I h A malfunction of the Control Element Drive Mechanism Control System (CEDMCS) or rod regulating system could cause an uncontrolled withdrawal of the CEAs. The core power would increase due to the positive reactivity addition.
The power excursion is terminated by the VOPT when the reactor is critical, and l by the CPC trip or Log Power trip when the reactor is in a suberitical condition prior to the event.
I
- Southern California Edison 52 November 1998
___a
I I
Section 3.4 Non-LOCA Transient Analysis l
- b. Analysis criteria '
The uncontrolled CEA withdrawal is a moderate frequency event. Therefore, the j g fuel temperature must not exceed fuel centerline melt criteria and DNBR must be ;
E greater than the DNBR SAFDL.
- c. Objectives of the analysis The objective of the analysis is to demonstrate that the fuel temperature remains I less than the temperature for fuel centerline melt and DNBR is greater than the l SAFDL Peak RCS Pressure is not limiting for this event.
- d. Basic assumptions and justification l
! The analysis assumptions are similar to those given in UFS AR Sections 15.4.1.1 and 15.10.4.1.1. Only those parameters that are reload dependent are discussed below.
- 1. The initial conditions are chosen to maximize fuel temperature during the transient. For example, the most conservative core ASI and MTC are used.
!I 2. Conservative CEA reactivity insenion rates (RIR) are used. The rates are determined for suberitical, low, and full power initial conditions.
- 3. The maximum peaking is determined for suberitical, low, and full power initial conditions.
f 4. Peak RCS pressure is calculated for the limiting low power case I
. e. Analysis method I
I The CESEC-III code (References 23 and 30) is run twice for each initial l condition. The initial run determines the time of reactor trip. The second run
( models the insertion of the withdrawn CEAs at the rate determined by the scram position versus time curve after the reactor trip. The resulting core thermal-hydraalics boundary conditions determined by CESEC-Ill are used to verify that the LHR, based on the peak heat flux,is acceptable. Integrated deposited energy may be calculated by a hand calculation instead of LHR to show that no fuel melt-ing occurs. CETOP-D is used to ensure that DNBR SAFDL is not violated.
- f. Current additional conservatisms None Southern California Edison 53 November 1998
1 1
I Section 3.4 Non-LOCA Transient Analysis 3.4.2.2.4 Single Full Length CEA Drop Event i 1
- a. Description of the event The dropping of a single full length CEA initially causes a reduction in core power and a primary to secondary side power to load mismatch. This mismatch I results in a cooldown of the RCS due to the excess heat removal by the secondary system. In the presence of a negative MTC, the cooldown will add positive reac-tivity and the core power will tend to return to its pre-drop level. This return to I power is accompanied by a power distribution distorted by the full insertion of the CEA. This combination of effects leads to a loss of thermal margin for the CEA drop transient.
I The SONGS full length CEAs have either 4 or 5 full length fingers, each with B4C extending the full length.
I b. Analysis criteria Single CEA deviation events are AOOs. and violation of the SAFDLs during the transient is not allowed.
- c. Objec6e of the analysis I The objective of the analysis is to detennine the amount of COLSS ROPM necessary to provide protection against violation of the SAFDLs due to the drop of any single 4 or 5 fingered full length CEA. The COLSS ROPM is detennined I such that it is large enough to offset the immediate radial power distortion effects and the additional radial power distortion due to changes in the xenon redistribu-tion in the core. COLSS provides protection for up to 15 minutes following a I single CEA drop. Other Technical Specification requirements provide for long term protection for 120 minutes.
- d. Basic assumptions and justifications None.
I l
- e. Analysis method 1
This event is analyzed consistent with UFSAR Sections 15.4.1.3 and 15.10.4.1.3.
j l l L Southern California lidison 54 November 1998 i r-
1 Section 3.4 Non-LOCA Transient Analysis
}
- f. Current additional conservatisms I 1 1 3.4.2.2.5 Part Length CEA Drop
- a. Description of the event A pan length CEA (PLR) is not axially uniform and absorbs neutrons differently along its length. The lower 50% of the PLR is Inconel, a mild absorber. The top 10% is B C, a strong absorber, and the remainder is open to the RCS allowing water filling.
The Pan Length CEA drop event is typically analyzed from a s50 percent initial power condition as discussed in UFSAR Sections 15.4.1.3 and 15.10.4.1.3.
I Above this power level, the Technical Specifications limit PLR insertion is such that the drop of the PLR can only add negative reactivity and the power distribution distonion is less than that of the drop of a single full length CEA.
I Below 50 percent power, it is possible that dropping a PLR out of an axial region of high power could have the effect of adding positive reactivity to the core. This h positive reactivity insenion, along with a constant secondary system demand, B
could result in the core power generation being higher than that removed by the secondary system. The power to load mismatch will increase RCS temperatures. -
The RCS heat up, in the presence of a positive MTC, could result in a further l increase in core power. The event is postulated to continue until a new quasi-steady state condition is reached due to the action of the secondary safety valves, I or is terminated by either a liigh Pressurizer Pressure Trip, or a CPC Trip, or VOPT.
- b. Analysis criteria Since the PLR drop event is an AOO, the fuel must remain within the DNBR J SAFDL.
[
Southern California Edison 55 November 1998
Section 3.4 Non-LOCA Transient Analysis
- c. Objectives of the analysis The objective of the analysis is to determine the amount of COLSS ROPM necessary to ensure that there is no violation of the SAFDLs in the event of a PLR drop.
- d. Basic assumptions and justifications Only those assumptions that are reload dependent are discussed here.
- 1. The most positive MTC allowed by the LCS that can occur at a given initial power level over the life of the cycle is used. This assumption enhances the heat up of the core.
- 2. The least negative Doppler coefficient is selected to provide the smallest amount of negative reactivity feedback in response to the power increase.
- 3. The maximum F, for each power level is used.
I
- 4. Cycle specific positive reactivity and radial power distortions are used.
- e. Analysis method The CESEC-III computer code (References 23 and 3'0) is used to model the NSSS i response to the PLR drop. The output from CESEC is input into CETOP-D code to find the time of minimum DNBR along with core coolant mass flux, core heat flux, core coolant inlet temperature and core pressure at the time of minimum I DNBR. ROPM is calculated from CETOP-D by comparing the POLR at initiation of the event (pre-drop) to the POLR at the time of minimum DNBR and for times up to two (2) hours after the drop.
- f. Conservatisms in the analysis
- 1. EOC kinetics (minimum Beta) are used to maximize the core power rise for the event.
- 2. The most positive MTC, within the LCS that does not cause a reactor trip is used.
I
- 3. Maximum fuel gap conductance is used because the resulting lower fuel temperature causes less doppler feedback during heat up.
I
- 4. The least negative (BOC) doppler curve is used.
1
- Southern California Edhon 56 Novemter 1998
Section 3.4 Non-LOCA Transient Analysis 3.4.2.2.6 CEA Subgroup Drop Events
- a. Description of the event The dropping of a CEA subgroup is much like the dropping of a single CEA, in that the drop inserts negative reactivity into the core, reducing power. Reactivity feedback effects will respond gradually to bring the core power back to its l pre-drop level if the secondary system is held at its pre drop level. However, the I core power distribution will be distoned by the presence of the CEAs. When the core returns to the initial power, the distorted power distribution may cause the hot channel DNBR to approach the SAFDLs.
A CEA subgroup drop will cause the generation of a CPC penalty factor sufficient to cause a reactor trip before the core approaches violation of the SAFDLs. As discussed in UFSAR Sections 15.4.1.3 and 15.10.4.1.3, typically no non-LOCA Transient analysis is needed since the reactor trips when a CEA subgroup drops to avoid violation of the SAFDLs or the resulting power distortions are less than the FLCEA Drop event .
With both CEACs inoperable, the CPCs conservatively assumes that it is due to I failed RSPT indication. For this case a subgroup drop will not generate a penalty and it is necessary to reserve COLSS margin to protect the DNBR SAFDL
- b. Analysis criteria CEA subgroup drops are AOOs and violation of the SAFDLs during the transient l is not allowed.
- c. Objectives of the analysis The objective of the analysis is to determine the amount of COLSS ROPM i necessary to provide protection against violation of the SAFDLs due to the subgroup drop when both CEACs are inoperable.
- d. Basic assumptions and justifications The margin degradation is due only to changes in the radial power peaking factor, xenon redistribution and CEA position which are cycle specific. The adverse change in system pressure is compensated by a favorable change in RCS temperature. Power, which decreases with the negative reactivity insenion, will retum to its original value because of the negative moderator temperature
- coef6cient.
m P
, Southern California Elison 57 November 1998
Section 3.4 Non-LOCA Transient Analysis
- e. Analysis method A conservative hand calculation or a CETOP-D calculation, if necessary, is used to calculate the ROPM necessary to reserve in COLSS when both CEACs are inoperable.
- f. Current additional conservatisms None 3.4.2.2.7 CEA Withdrawal within Deadband
- a. Description of the event Misalignment within CPC/CEAC CEA position deadbands do not generate CEAC or CPC penalty factors. Sufficient thermal margin must be preserved in COLSS to enable the plant to ride through the consequences of a CEA withdrawal within the deadband without resulting in violation of the DNBR SAFDLs.
The withdrawal of the CEA within the deadband inserts positive reactivity, which causes reactor power to increase. Combined with a constant secondary system power demand, the power to load mismatch causes an increase in RCS temperatures. The addition of positive reactivity is counteracted by the negative moderator temperature and doppler reactivity coefficients adding sufficient negative reactivity to halt the power increase. With a positive moderator temperature coefficient, reactor power and temperature and steam generator temperature increase until the secondary safety valves open preventing a further increase.
The cause of the thermal margin degradation is a power increase due to the addition of positive reactivity with the withdrawal, an upward shift in the axial power distribution as the power follows the upward motion of the CEAs, and xenon redistributes over time. The change in the radial power distribution will be undetected by the CPCs because the CEA movement occurs with the allowable deadband.
- b. Analysis criteria CEA Withdrawal Within Deadband events are AOOs and violation of the SAFDLs during the transient is not allowed.
g Southem California Edison 58 November 1998
l I
Section 3.4 Non-LOCA Transient Analysis
- c. Objective of the analysis The objective of the analysis is to determine the amount of COLSS ROPM necessary to provide protection against violation of the SAFDLs due to the withdrawal within deadband. The COLSS ROPM is determined to prevent violation of the SAFDLs during the transient.
- d. Basic assumptions andjustifications Only those assumptions that are reload dependent are discussed here.
- 1. The use of the most positive MTC allowed by plant ifs can cause the calculation of excessive thermal margin requirements. During the reload analysis, time-in-life dependent physics data is used. The analyses of the events are divided into two pans, a BOC analysis of the events in which the most positive MTC is used, and a post-BOC analysis which uses a negative MTC value within the LCS.
- 2. Cycle specific reactivity insertions at specific power levels are used. Various other physics data generated for the withdrawal event is presented in Section 3.1.1.2.
- e. Analysis method The CESEC-III computer code (References 23 and 30) is used to model the NSSS response to the withdrawal events. The changes in core thermal margin are analyzed using the CETOP-D code (Reference 12) based on boundary conditions I provided by the CESEC code.
The CESEC code is used to model the plant response to the addition of positive reactivity accompanying the single CEA withdrawal. The time period over which the reactivity is added is equal to the time required for the CEDMs to withdraw a CEA within the deadband. The reactivity is held constant after this time. The turbine demand is held constant at its initial value. The increase in reactor power and constant turbine power demand results in a power to load mismatch.
I f. Conservatisms in the analysis All parameters that impact the analyses are assumed at their worst value.
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Southern California Iklison 59 November 1998
Section 3.4 Non-LOCA Transient Analysis 3.4.2.2.8 Inadvertent Boron Dilution
- a. Description of the event Controlled boron dilution normally occurs during startup. An inadvertent boron dilution event may lead to unplanned changes in reactivity if the dilution is not monitored and controlled. Sufficient time must be available for the reactor operators to take corrective action to prevent criticality if an inadvertent boron dilution event occurs.
- b. Analysis criteria An inadvenent boron dilution event is a moderate frequency event. The fuel integrity must be assured by allowing sufficient time for operator action to prevent loss of shutdown margin, The analysis is required to demonstrate acceptable results assuming at least 15 minutes (30 minutes in Mode 6) is required between the time that the operator is made aware of the boron dilution and the loss of shut-down margin. The UFSAR analysis described in Sections 15.4.1.4 and 15.10.4.1.4 represents the limiting worse case.
- c. Objective of the analysis The objective of the inadvenent boron dilution analysis for a reload is to find I when the core shutdown margin is lost for Modes 2 through 6. In addition, the boron dilution alarm multiplication factor setpoints are also determined.
- d. Basic assumptions and justification The analysis assumptions are as given in UFSAR Section 15.4.1.4 and 15.10.4.1.4. Only those parameters that are reload dependent are discussed below.
- 1. The minimum Inverse Boron Worth (IBW) is conservatively used for any initial operating mode.
- 2. Conservatively high critical boron concentration values are utilized in the analysis.
I Southem Califomia Iklison 60 November 1998
I Section 3.4 Non-LOCA Transient Analysis
- e. Analysis method The ordinary differential equation which describes a well stirred mixing tank is used for the calculations:
M dC dt
= -(W C) where, M is RCS water mass C is RCS boron concentration and
- W is charging mass flowrate of unborated water l
I The equation is evaluated to determine the time necessary for !be boron to change from its initial concentration to the critical concentration. All combinations of RCS volume and possible operating modes are examined. The shutdown margin requirements are evaluated and Technical Specification changes are made when the requirements are altered.
- f. Current additional conservatisms l l
The calculations are based on the assumption that the RCS is being diluted with I unborated water. The limiting case is based on the assumption that the IBW and CBC are conservatively biased and that all three charging pumps are operating in g Modes 1,2,3,4 and 5(RCS full). In Modes 6 and 5(Reduced inventory - Middle g of RCS Hot Leg) only one charging pump is operational.
I 3.4.3 Degraded Performance of CPCS & COLSS Category l The events described in Section 3.4.2.2 are based on the assumption that the CPCS and the COL.SS are not in a degraded condition. Since potential hardware failures could
,' affect the operability of these systems, the events must also be evaluated for the following l three conditions: (a) COLSS out-of-service and at least one CEAC operable, (b) COLSS in-service and both CEACs inoperable, and (c) COLSS out-of-service and both CEACs l inoperable. A brief description of the impact of each of these modes on the reload event I analysis is presented in the following sections.
l 3.4.3.1 COLSS Out-of-Service and at Least One CEAC Operable i
When COLSS is out-of-service, the CPCS calculated DNBR is used to monitor the DNBR LCO (or COLSS ROPM) per Technical Specification 3.2.4. This specification requires that the CPCS calculated DNBR must be larger than a pre-determined CPCS DNBR vs. ASI limit line. Analyses are performed to determine the ROPM versus power curve for the condition of COLSS out-of-service and CEACs operable. This curve is used to determine a conservative CPCS DNBR limit line (to Southern California Edison 61 November 1998 1
Section 3.4 Non-LOCA Transient Analysis be used in Technical Specification 3.2.4) which will ensure that the DNBR LCOs are maintained if the core is operated within the Technical Specifications limits. Since the Technical Specifications require a more restrictive PDIL when COLSS is out-of-service, the ROPMs for several events are smaller than those when COLSS is in service because a smaller amount of CEA insertion is allowed. Analysis methods are similar to those given previously.
3.4.3.2 COLSS In-Service and Both CEACs Inoperable When at least one of the CEACs is operable, the CPCS monitors the individual and I subgroup CEA movement directly. When both CEACs are inoperable, information on CEA position and the associated radial peaking factors for the DNBR calculation are not provided for the CPCs. As a result additional thermal margin is needed to accommodate CEA misoperation events when the CEACs are inoperable and [
1 l Typically, the two design basis e ents that are protected by CPCS trips due to CEA misoperations are: (a) a CEA susgroup drop and (b) withdrawal of a single CEA (allowed inserted under the CEACs inoperable Technical Specifications) and are discussed below.
During periods in which CEACs are inoperable, the Technical Specification on RPS I operability requires that the lead bank CEA insertion be restricted to the top 15 percent of the core. As described above the CEACs are unable to provide the CPCs, with penalties to compensate for CEA deviations from this configuration. Inward I deviation is more than sufficient to compensate for the lack of radial peaking infomiation. Withdrawals from these conditions lead to an increase and a change in the core power distribution which are undetected by the protection system. It is I necessary to ensure that sufficient margin is maintained to enable the plant to ride through the transients without violation of SAFDLs. Analysis criteria, objective of the analysis, etc, are similar to the CEA Withdrawal within Deadband presented in I Section 3.4.2.2.7.
As a result, with CEACs inoperable, restrictions are placed upon the allowed insertion of the CEAs (typically limited to top 15 percent). This restriction is found in the RPS operability Technical Specifications. Additional extra margin is maintained at high power levels by forcing the operators to reduce the POL calculated by COLSS per the requirements of Technical Specific itions Figure 3.2-1.
f 3.4.3.3 COLSS Out-of-Service and Hoth CEACs Inoperable When COLSS is out-of-service and both CEACs are inoperable, the CPCS calculated DNBR is used to monitor the COLSS DNBR LCO. However, when both CEACs are inoperable, CEA movement cannot be detected by the CPCs (similar to Section 3.4.3.2). As a result, CEA misoperation events must be analyzed for this condition.
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Section 3.4 Non-LOCA Transient Analysis Again, in the case of COLSS in-service and both CEACs inoperable, a single CEA withdrawal (allowed insened per Technical Specifications) needs to be protected.
These events are used to determine the ROPM versus power curve for the condition of COLSS out-of-service and both CEACs inoperable. This curve is used to determine a conservative CPCS DNBR limit line (to be used in Technical l Specification 3.2.4), which ensures that the DNBR LCOs are maintained if the core is operated within the CPCS limits.
Methodology similar to that described previously for CEA misoperation events is used to determine the ROPM when both CEACs are inoperable. Therefore, the results obtained for COLSS in-service and CEACs inoperable apply for COLSS out.
of-service and CEACs inoperable conditions also. The most restrictive event (usually the subgroup CEA drop when both CEACs are inoperable) results in the m iximum ROPM requirements and is used to determine the CPCS DNBR limit line to be used with Technical Specification 3.2.4.
I 3.4.4 Verification of Transient Related CPCS Constants l The CPCS is a set of four digital computers and associated software which initiates two of the RPS trips (see discussion in Section 3.5). Each CPC continuously calculates the DNBR and LPD. A trip signal from any two CPC computers will initiate a reactor trip i when needed to prevent a violatim of the SAFDLs. The CPCS performs the required calculations through the use of algorithms and associated constants and setpoints. The CPCS dynamically processes information related to the NSSS parameters that affect the I margin to the fuel design limits. The output of the CPCS is the generation, if necessary, of a reactor trip on either low DNBR, high LHR generation, or on several auxiliary trip functions such as cold leg differential temperature. The purpose of the CPCS transient I response analysis is to ensure that the CPCS calculations remain conservative relative to the actual system state during postulated system transients. The analyses in this section describe the transient related CPCS constants analyses typically performed each reload.
3.4.4.1 Dynamic Compensation The CPCS is provided with filter algorithms and associated constants which provide dynamic compensation of instrument response. The temperature sensors, in panicular, have sizable time lags associated with dynamic process changes. Because of this, the cold leg temperatures and power signals are dynamically compensated.
The coefficients used in the filter calculations are established based on the assumed l instrument characteristics and loop transport times. Conservative cold leg temperature instmment time constants and excore detector " shadowing" factors are verified. The rapid temperature changes introduced by overcooling and heat up events are used to verify conservative temperature filter performance. The CEA withdrawal event is typically used for verifying the power filter response.
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j 3.4.4.2 Transient Offset Power Penalties !
The CPCS allows modification of the received thermal and neutron power signals by !
. fixed multipliers or offset terms. The offset terms ensure that the overall response of the CPCS is conservative by increasing the power signals above the values actually ,
received by the plant sensors. The conditions which have required the applications of l power penalties are:
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During the reload safety design, these types of conditions are evaluated to determine
- if CPCS power offset adjustments are required. !
3.4.4.3 Radial Penalty Factor Delay i The CEA positions are monitored by the two CEACs. The penalty factor is imposed immediately if both CEACs simultaneously sense a dropped CEA. If only one CEAC senses a dropped CEA, a delay time is imposed before the penalty is applied. The I delay time allows the CEAC to clear any spurious signal, thereby preventing )
unnecessary reactor trips. The delay in trip time is compensated by setting aside j appropriate amount of ROPM each cycle. This delay time is considered sufficiently i large enough to allow CEAC auto restart as a result of a false signal to prevent an ,
inadvenent trip. This delay time is implemented as an addressable constant in the !
CPCS and is evaluated each reload to assure that the fuel design limits are not I exceeded d' iring the delay.
3.4.4.4 Reactor Coolant Pump Speed Trip Setpoint An addressable constant is included in the CPCS which represents the fraction of rated RCP speed above which a pump is assumed to be operating. When any RCP speed drops below this setpoint, a reactor trip signal is generated. This setpoint may be used in the IOSGADV+LOAC, loss of flow analyses and seized rotor analyses, and is typically verified for each reload.
i 3.4.4.5 Variable Overpower Trip Setpoint i A VOPT function is implemented in the CPCS as an auxiliary trip. The trip is similar in operation to the analog VOPT in the RPS, which SONGS does not have. Since the ,
l VON is credited in several transient analyses, the constants associated with the trip function must be reviewed each reload, but generally do not change.
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i Section 3.4 Non-LOCA Transient Analysis 3.4.4.6 Asymmetric Steam Generator Trip Setpoint The analysis of the ASGT event credits a reactor trip generated by the CPCS. The ASGT trip setpoint is based on a cold leg temperature differential as perceived by the CPCS, and is evaluated each reload.
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Section 3.5 Core Protection Calculator System Analysis 3.5 CORE PROTECTION CALCULATOR SYSTEM ANALYSIS This section includes a brief description of the CPCS and the related a.ialyses perfonned to l
I support a typical reload design. The CPCS is a set of digital computers and associated software which initiates the DNBR and LPD reactor trip signals. The CPCS is comprised of four Core Protection Calculators (CPCs) and two Control Element Assembly Calculators (CEACs). As shown on Figure 3.5-1, each CPC continuously calculates the core DNBR and LPD and will I initiate a reactor trip when needed to prevent a violation of the SAFDLs. Each CEAC continuously measures the position of all individual CEAs to detect deviations and provide appropriate penalties to each of the CPCs as shown on Figure 3.5-2. These penalties are applied as appropriate in the DNBR and LPD calculation in the CPCs. The CPCS design basis also includes provisions for auxiliary trip functions, which provide protection for certain design basis events, Section 3.4. The auxiliary trips are used to: (a) provide protection more conveniently than the DNBR or LPD trips, (b) aid the CPCS in meeting the primary design basis, (c) assure core conditions are within the analyzed operating space and (d) assure CPCS protection in the presence of hardware failures. These auxiliary trip functions are shown on Table 3.5-1.
Table 3.5-1 l CPCS Auxiliary Trips I i i
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Section 3.5 Core Protection Calculator System Analysis The CPCS (including the six computers, associated software, sensors, and trip outputs) is a safety grade system designed to take action as needed during cenain transients. It is primarily designed to provide DNBR and LPD protection for anticipated operational occurrences, but is qualified also to assist the reactor protection system and the ESFAF in limiting the consequences of certain postulated limiting fault events. A list of typical CPCS design basis events are shown below:
- 1. Uncontrolled Axial Xenon Oscillations.
- 2. CEA Misoperation Events from Critical Conditions, including Single CEA Withdrawal, Single CEA Sticking with Remainder of CEAs in that Group Moving, Subgroup Deviation (Insertion, Withdrawal or Drop), Out of Sequence Group Withdrawal or Insertion, Uncontrolled Sequential Group Withdrawal, and Excessive Insertion of Pan Length CEAs.
I 3. Excess Heat Removal (Excess Load) Due to Secondary System Malfunctions, including Excess Feedwater Flow, Excess Steam Flow Caused by Inadvenent Opening of Turbine Bypass Valves, Excess Steam Flow Due to Inadvertent Opening of Turbine Control Valves, and Decrease in Feedwater Enthalpy.
- 4. Loss of Forced Reactor Coolant Flow.
- 5. RCS Depressurization (Spray Malfunction).
- 6. Loss of Feedwater Flow.
- 7. Loss of Extemal Load.
- 8. Complete Loss of AC Power to the Station Auxiliaries.
- 9. Uncontrolled Boron Dilution at Power.
- 10. Asymmetric Steam Generator Transients due to Instantaneous Closure of One MSIV.
- b. Postulated Accidents:
l 1. Reactor Coolant Pump Shaft Seizure.
{
- 2. Steam Generator Tube Rupture.
I 3. Steam Line Break Outside Containment.
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Section 3.5 Core Protection Calculator System Analysis The functional design of the CPCS is defined by the description of the algorithms which were created to meet the design basis of the system. The functional design is given in Reference 32 for a CPC and in Reference 33 for a CEAC. The functional design does not include the constants I which are required by the algorithms. These constants are chosen based on analysis that is either generic (such as physical properties), reload fuel cycle independent (such as core configuration),
or reload fuel cycle specific (such as power distribution related information).
The CPCS constants are divided into three types: dam base, Reload Data Block (RDB), and addressable.
I Data base constants are built into the CPCS software along with the CPCS algorithms.
Data base constants cannot be changed except by changing the software, which involves considerable testing and requires approval of the NRC prior to implementation. For a typical reload, the CPCS software is not changed and, therefore, the data base constants are not changed.
RDB constants are located in protected memory of the CPCS and are separate from the I algorithms and non RDB data base constants. RDB constants are loaded from a separate disk and can be changed without requiring a CPCS software change (currently all RDB disks are created by the fuel vendor).
Addressable constants can be changed by the plant operator under administrative controls.
Addressable constants are changed at various frequencies from once per reload to daily during operation.
Since data base constants are not changed during a reload, the RDB and addressable constants are the only constants which are considered for change during the reload design. The first step in the CPCS reload design analysis is to determine the appropriate values of the RDB constants. Once the RDB constants have been determined, the addressable constants are calculated to ensure that the CPCS design basis is met. Some CPCS constants are defined by other reload design functional areas, and other constants are determined based on cycle specific CPCS analyses. In the following sections, the information provided by other functional design areas that is used to determine CPCS constants will be presented first, followed by a discussion of the CPCS RDB and addressable constants analyses.
3.5.1 Physics Analysis Input to the CPCS Analysis Typical physics design inputs to the CPCS reload design process were discussed in Section 3.1 and include a neutronics model, physics related data base constants, and physics related RDB constants. A neutronics model is used as a reactor core simulator g for CPCS constants analyses. The physics design also provides [
[5
] This data is used to verify CPCS data base constants for the drop of any single CEA. The RDB constants which are based directly on the physics design include Southern Calit'ornia Iklison 68 Nosernber 1998 1
Section 3.5 Core Protection Calculator System Analysis constants such as [
] In most cases, the previous cycle values for these constants are verified as applicable to the upcoming cycle.
3.5.2 Core Thermal-Ilydraulics Analysis Input to the CPCS Analysis Core thermal-hydraulics design input to the CPCS reload design process includes a CETOP-D model, associated DNBR limit, and statistical data. As discussed in Section 3.2, the CETOP-D model is the design core thermal-hydraulics code benchmarked to the TORC model over various operating ranges. 'Ihe model is used initially to develop and tune the on-line core thermal-hydraulics algorithms in the CPCS and subsequently the model is used cycle-by-cycle for CPCS constants analyses. Typically, a cycle independent CETOP-D model is provided with a cycle specific penalty to be applied whenever the model is used. The DNBR limit and statistical data provided by the core -
thermal-hydraulics design includes the mean and standard deviation for the DNBR limit statistics (not including deterministic adjustments). The net DNBR limit with deterministic adjustments is calculated as described in Reference 15. The mean and standard deviation is used with its 95/95 probability / confidence tolerance limit in the CPCS analyses as discussed in Section 3.2. The DNBR limit with deterministic adjustments is used in calculating the CPC DNBR addressable constant trip setpoint.
3.5.3 Safety Design Input to the CPCS Design The safety design verifies relevant RDB constants and provides input to calculations of several addressable constants, as described in Sections 3.4.2.4 and 3.4.2.5. The values of these constants are selected to ensure that the CPCS will respond conservatively relative to the assumptions made in the analysis of postulated transients. In addition, the safety design defines the protective steps which must be taken to compensate for the potential degraded state of the CPCS and COLSS. Specifically, the safety design provides the design margins which must be preserved if the COLSS is out-of-service and/or the CEACs are inoperable. These margins are typically different .han the margins provided for normal operation of these systems. The difference is due to !he fact that the ROPMs are determined based on more restrictive requirements (such as a reduced Technical Specification PDIL) when the COLSS is out-of-service and/or the CEACs are inoperable.
3.5.4 Fuel Performance Analysis input to the CPCS Analysis The fuel performance design provides the fuel axial densification factor and the engi-neering factor on LHR which are used in the CPCS addressable constants analyses.
Additionally, the fuel perfomiance design provides the LHR limit based on fuel centerline melt for use in determining the LPD trip limit addiessable constant value.
Southern California Elison 69 November 1998
Section 3.5 Core Protection Calculator System Ana ysis 3.5.5 CPCS Constants Analyses The purpose of the CPCS constants analyses is to verify reload specific data base con-stants and determine appropriate RDB and addressable constants which ensure that the CPCS design basis is met.
3.5.5.1 CPCS Data Hase Constants Verification For a typical reload the CPCS software is not changed, and, therefore, the data base i constants are not changed. Because of this, the data base constants were previously chosen to be bounding for all future cycles, if practical. In some cases, it was not possible to determine bounding values without adversely impacting operating space.
As a result, some data base constants are evaluated each reload to ensure that they are conservative. These cycle specific values are compared to the penalty factors included in the CPCS data base and the addressable or RDB constants are adjusted as l required. The remainder of the CPCS constants analyses involves calculation and/or verification of RDB and addressable constants.
3.5.5.2 Reload Data Block Constants Analyses in addition to the RDB constants that are determined by the physics, core thermal-I hydraulic, fuel performance, or safety design, several RDB constants are verified based on other cycle specific requirements of the CPCS. The following design considerations are evaluated each cycle and appropriate RDB constants are adjusted I as required.
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I I i RDB constants change based on these design considerations. An adjustment can be made to addressable constants,if required. These analyses result-in a list of RDB constants for use in CPCS design analyses. In addition, this list is used to generate the disk for installation in the CPCS prior to startup.
Southem California filison 70 November 1998
Section 3.5 Core Protection Calculator System Analysis 3.5.5.3 CPCS Addressable Constants Analyses The CPCS addressable constants analysis is the Gnal step in the CPCS reload design process. The CPCS addressable constants consist of constants measured during startup, calibration constants, trip setpoints, uncertainty factors, and penalties. The CPCS addressable constants analyses provide the values for all addressable constants including initial values for the startup related constants. After reload startup testing is completed, final values of these constants are installed. The calibration constants such as flow and power calibration constants are routinely checked and adjusted during operation. The DNBR trip setpoint is determined by the thermal-hydraulic and safety design. The fuel performance design typically determines the trip setpoint for LPD.
The CPCS overall uncertainty analysis provides the final verification of the CPCS functional design and data base. This is done by calculating uncertainty factors which assure the CPCS conservatively calculates the MDNBR and LPD to at least a 95/95 probability / confidence level based on the values of the data base and cycle specific RDB constants. The calculation of the uncertainty factors and penalties is described I in detail in Reference 15.
After the uncertainty factors and penalties are calculated, the final values of the \
addressable constants are assembled based on the results of all constant and uncertainty analyses, along with input from other functional areas. The initial I addressable constants are installed in the CPCS prior to startup.
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l Section 3.6 Core Operating Limit Supervisory System Analysis 3.6 CORE OPERATING LIMIT SUPERVISORY SYSTEM ANALYSIS This section includes a brief description of the COLSS and the related analyses to be performed to support a typical reload design. The COLSS is a digital computer program in the PMS which assists the plant operator in maintaining Technical Specification LCOs. An overview description of the COLSS is provided in Reference 34 and a simplified block diagram of the COLSS is shown on Figure 3.6-1. As shown on this figure, the COLSS program monitors the margin to the power operating limits as required by the Technical Specifications for:
- a. Peak LHR (or LPD)
- b. DNBR I c. Licensed power Additionally, the COLSS monitors the core azimuthal tilt and ASI to assist the operator in maintaining the Technical Specification LCOs for these parameters. The COLSS design basis specifies that COLSS must provide infonnation to the operator during normal operation.
Therefore, there are no design basis events for which COLSS must be applicable. Instead, there are a number of AOOs and postulated events which are analyzed in order to provide initial margin requirements to the COLSS analyses. These events may vary from cycle to cycle due to I changes in fuel management, fuel design, or other requirements. The input to the COLSS analyses is based on the choice of AOOs and postulated limiting fault events. The basis for the choice of events to analyze and the methodology for developing the input to COLSS is provided I in Section 3.4.2.2.
The functional design of the COLSS is defined as the description of the algorithms which were I created to meet the design basis of the system. The functional design does not include the values of the constants which are required by the algorithms. These constants are listed in a COLSS data base document which is updated each reload as required.
There are two types of COLSS constants, data base constants and addressable constants. The I COLSS data base constants are specified in a data base document, and are compiled into the COLSS program. The COLSS data base constants are typically specified once for each reload fuel cycle and implemented prior to startup. Addressable constants can be changed as required during operation. Similar to the CPCS data base, the COLSS data base contains default values of addressable constants. The default values are updated based on the results of the COLSS analyses and the startup testing. Section 3.6.1 describes the analysis required to determine the appropriate reload specific data base and addressable constants.
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Section 3.6 Core Operating Limit Supervisory System Analysis The first step in the COLSS reload design analysis is to determine the appropriate values of the data base constants. Once the data base constants have been determined, the addressable constants are calculated. Some of the COLSS constants are defined by other reload design functional areas and other constants are determined based on cycle specific COLSS analyses.
The information provided by other functional design areas that is used to determine COLSS constants will be presented first, followed by a discussion of the addressable constants analyses.
3.6.1 Physics Analysis Input to the COLSS Analysis Physics Design inputs to the COLSS reload design typically include a neutronics model and physics related data base constants. A neutronics model is used in the COLSS analyses as a reactor core simulator for COLSS constants analyses. The physics design also provides [
] Data base constants which are based directly on the physics design include [
l ] In most cases, the previous cycle or implemented values for these constants are verified as applicable to the upcoming cycle.
3.6.2 Core Thermal-Ilydraulics Analysis Input to the COLSS Analysis Core thermal-hydraulics input to the COLSS reload design typically includes a CETOP-D model, associated DNBR limit, and statistical data. As discussed in Section 3.2, the I CETOP-D model is the design core thermal-hydraulics code benchmarked to TORC over various operating ranges. It is used initially to develop and tune the online core thermal-hydraulics algorithms in COLSS and is used cycle-by-cycle as a base core thermal-I hydraulics model for COLSS analyses. Typically, a cycle independent CETOP-D model
' is provided with a cycle specific penalty to be applied to the core average heat flux whenever the model is used. The DNBR limit and statistical data include the mean and I standard deviation for the DNBR limit statistics (not including deterministic adjustments) along with the net DNBR limit with deterministic adjustments as described in Reference 15.
3.6.3 Safety Analysis Input to the COLSS Analysis The safety design provides data base constants, such as: (a) the LHR limit typically based on the LOCA design and (b) the constants related to the Required Overpower Margin (ROPM). The constants related to ROPM include [
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Section 3.6 Core Operating Limit Supervisory System Analysis 3.6.4 Other Inputs to the COLSS Analysis Other inputs required for the COLSS reload analyses include [
]
3.6.5 COLSS Constants Analyses This section provides an overview of the COLSS constants analyses for a typical reload design. These analyses verify data base constants and determine addressable constants which 2nsure that the COLSS design basis is met.
3.6.5.1 COLSS Data Base Constants Verification Most COLSS data base constants do not change from cycle to cycle. The physics, thermal-hydraulic, and safety design provide input to cycle dependent constants as described in the previous sections. This results in a list of data base constants for use in the COLSS addressable constants analyses. In addition, this list is used to generate the data base file for installation in the COLSS prior to stanup.
3.6.5.2 COLSS Addressable Constants Analyses The COLSS addressable constants analysis is the final step in the COLSS reload design process. The COLSS addressable constants consist of constants measured during startup, calibration constants, uncertainty factors, and penalties. The COLSS addressable constants analyses provides the default values for all addressable constants, including initial values for the startup related constants. After reload startup testing is completed. final values of these constants are installed.
The COLSS overall uncertainty analysis provides the final verification of the COLSS functional design and data base. This is done by calculating uncertainty factors which assure that the COLSS conservatively calculates the DNB-POL and LHR-POL to at least a 95/95 probability / confidence level based on the values of the data base constants. The calculation of the uncertainty factors and penalties is described in detail in Reference 15.
I After the uncertainty factors and penalties are calculated, the final values of the addressable constants are assembled based on the results of all constant and uncertainty analyses, along with input from other functional areas. The initial addressable constants are installed in the COLSS prior to stanup.
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Section 4.0 Reactor Core Design and Monitoring Program 4.0 REACTOR CORE DESIGN AND MONITORING PROGRAM SCE's Reactor Core Design and Monitoring Program is a comprehensive site wide program to ensure the thorough engineering and safety evaluation of all potential consequences of each reactor core design and operation. This program meets or exceeds all recommendations discussed in INPO's SOER 96-02, " Design and Operating Considerations for Reactor Cores."
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Figure 4.0-1 presents a simplified diagram of this program. Many of the components of the Reactor Design and Monitoring Program existed at SCE prior to the ABB CE/SCE Reload Technology Transfer program, as denoted in Figures 3.0-1 and 3.0-2. (As an example, SCE has I always had a procedure to control the installation of CPC addressable constants that were calculated and provided to SCE by ABB CE. The Reload Technology Transfer program has provide SCE with the capability to independently calculate the CPC addressable constants. The procedures and processes used by SCE to control the installation of these constants is unchanged by the Reload Technology Transfer Program.) Each major aspect of the Reactor Core Design and Monitoring Program is summarized in the following sections:
4.1 FUEL MANAGEAIENT GUIDELINES FOR SONGS UNITS 2 AND 3 4.2 REACTOR CORE DESIGN REVCW TEA.\1 4.3 RELOAD GROUND RULES 1 4.4 ABB CE FUEL DESIGN CilANGE INTERFACE 4.5.1 Computer liardware Control 4.5.2 Computer Software Control 4.5.3 Analysis .\lethodology Control 4.5.4 Quality Program 4.6 SCE ENGINEER TRAINING QUALIFICATION GUIDE 4.7 CORE RELOAD ANALYSES AND ACTIVITIES CIIECKLIST I 4.8 4.9 SOURCE VERIFICATION /ABB CE FUEL FABRICATION INTERFACE ABB CE ENGINEERING INTERFACE 4.9.1 LOCA Analyses I 4.9.2 4.9.3 Fuel Slechanical Design Analyses As-huilt Data 4.10 SITE PROGRANI I.\lPACT 4.11 I 4.12 4.13 LICENSING AND DESIGN BASIS UPDATES DESIGN PROCESS FLOW AND CONTROLS COLSS/CPC PRODUCTS 4.14 LOW POWER AND POWER ASCENSION TESTING I 4.15 4.16 CORE AND SFP REQUIRE.\lENTS CORE AND FUEL SIONITORING l Appendix A provides a listing of the SCE procedures, referred to in this section, that are currently used to implement SCE's Reactor Core Design and Monitoring Program.
_ Southern California Edison 78 November 1998
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.r-..-..a M o nitorin g Southern California Edison 79 November 1998
l Section 4.0 Reactor Core Design and Monitoring Prograrn 4.1 FUEL MANAGEMENT GUIDELINES FOR SONGS UNITS 2 AND 3 The Fuel Management Guidelines (Reference 2) are designed to ensure the reactor core design has a high probability of being bounded by the current safety analyses, maintains sufficient operating margins and CEA wonh, and meets all technical and economic requirements. Lessons leamed from fuel and plant industry performance issues and events are incorporated into the fuel management guidelines. The SONGS reactor core designs are initially developed using state-of-the-art optimization codes. The resulting designs are modeled in licensing physics codes such as ABB CE's DIT/ ROCS /MC (References 1 and 41) or CASMO/ SIMULATE (Reference 26). Key physics parameters of the reactor core design calculated per the guidelines are compared against Design Criteria. The Design Criteria must be satisfied by each core design. As presented in Table 5.1-1, these include maximum power peaking, MTC, fuel bumup limits, etc. The Design Criteria are not expected to change frequently as they are derived from basic SAFDLs or licensing acceptance criteria. The reactor core design is also compared to Design Recommendations that, while not mandatory, will produce a core which has a high probability of I meeting the current operating margins, producing an economic core design, and producing an easily manufactured core design. The list of Design Recommendations is expected to be updated occasionally, as new developments and events produce lessons leamed which need to be I incorporated into the reload design process. The Design Recommendations include guidance on fresh fuel placement, minimum excore detector response, placement of reinsened assemblies next to the core shroud, etc. The reactor core design that meets the Design Criteria and best I meets the Design Recommendations in Reference 2 is presented to SONGS executive management for review and approval.
4.2 REACTOR CORE DESIGN REVIEW TEAM After a reactor core design has been approved by SONGS executive management, a review team may be assembled composed of representatives from all potentially affected site groups, such as Operations, Training, etc. The goal of this team to is review the reactor core design and any fuel design changes offered by the fuel vendor to determine potential site impact at the earliest possible time in the design process. The team also reconvenes as necessary during the design process to review progress and resolves any open issues. This is done independent of the source of the reactor core design.
4.3 RELOAD GROUND RULES l The Reload Ground Rules (RGR) is a compilation plant design data, including key fuel design l data, in a form useable by reload analysts. All plant design data in the RGR is referenced to l design disclosure documents calculations or procedures. The RGR was originally established to provide a complete and concise set of plant data to the fuel vendor, who did not have easy access to all plant design documents. With the completion of the Reload Technology Transfer program, the RGR is now used by SCE and updated for each new fuel cycle for each SONGS unit to reflect planned changes to fuel management, fuel design, Technical Specifications, and plant design. Due to the potential scope of the changes that can affect the RGR, refueling interval Southern California Iklison 80 November 1998 i
Section 4.0 Reactor Core Design and Monitoring Program changes are reviewed and approved by many site organizations, including Operations, Engineering, and Licensing. The review, modification, approval, and documentation of the RUR is controlled by a Nuclear Fuel Management (NFM) procedure, Appendix A. The procedure also specifies the requirements for updating the RGR between refueling intervals and the process for individuals to notify management of potential errors or changes to the RGR.
4.4 ABB CE FUEL DESIGN CHANGE INTERFACE ABB CE has introduced several design and manufacturing improvements to the basic 16X16 fuel design used at SONGS. (As an example, the initial 16X16 grid design was assembled using a Tungsten Inert Gas (TIG) welding process. In 1995, ABB CE proposed and SCE reviewed and approved an improved LASER welding process.) Proposed fuel design changes are first reviewed by the NFM Division and then by all affected site groups consistent with SCE review and approval process used for design change document review and approval controls. Approved fuel design changes are factored into the reactor core design and Reload Ground Rules, as appropriate. The Reactor Core Design Review Team may also review the changes to identify potential site impact as early as possible. These changes are also reviewed as part of the design change process discussed in Section 4.12. This interface process is performed independent of whether the reload analyses are performed by ABB CE or SCE.
4.5 ABB CE RELOAD ANALYSIS COMPUTER CODES AND METIIODOLOGY INTERFACE As part of the Reload Technology Transfer program, ABB CE provided SCE with computer codes and NRC approved reload analysis methodology. The computer codes were installed and verified and validated (V&V'd) on SCE computers and the analysis methodology was used in preparing analyses, as discussed in Section 2.0. As SCE proceeds to independently perform reload analyses, these computer codes and methodology will be updated and controlled by SCE in accordance with established SCE procedures, Appendix A.
4.5.1 Computer liardware Control SCE currently uses an IBM AIX-RS/6000 computer network. The system has significant computing and file storage capabilities. System and user files are routinely backed up for offsite storage. Disaster recovery and system re-build is through backup processes.
Configuration and user access is controlled by Nuclear Fuel Management procedures in cooperation with the corporate Information Technology Department.
4.5.2 Computer Software Control The ABB CE reload analysis process involves over 50 computer codes. In accordance with the SCE's software control procedure each of the reload analysis codes is assigned a
} " Code Sponsor" to oversee code installation and maintenance. Each computer code is installed and tested using ABB CE's software verification and validation (V&V) report, Southern California Eklison 81 November 1998
1 Section 4.0 Reactor Core Design and Monitoring Program I
user manual, and test cases and corresponding results. A separate SCE installation report I is prepared for each computer code. Currently, ABB CE provides SCE with executable and source listing for each computer code. Any source code changes made by SCE are l
B implemented by SCE's software control procedure (Appendix A). Should SCE make any source code updates, independent of ABB CE which changes the calculated values of any safety related parameters, SCE will perform a V&V prior to use of the modified source code for licensing related activities. If appropriate, a topical report covering the source I code modifications would be prepared for NRC review and approval. All code upgrades and improvements provided by ABB CE are installed using SCE's software control procedure.
Concems and issues raised by SCE regarding ABB CE's reload analysis computer codes are reported to ABB CE using their quality affecting problem reporting process, currently the Condition Alert Program. These issues are evaluated by ABB CE per their procedures and SCE is notified of ABB CE's findings and error notices, as appropriate.
l ABB CE is required to provide SCE with all computer code error notices for all reload analysis computer codes supplied for SCE's implementation of ABB CE's reload analysis methodology. These notifications provide ABB CE's explanation of the error and their evaluation of the impact of each error on reload analyses. Operabililty, reportability, root cause and corrective actions are controlled by SCE's problem reporting process, currently the Action Request Program (Appendix A), and tracked to completion.
ABB CE is continually refining and updating their NRC approved reload analysis computer codes or developing new reload analysis computer codes for NRC review and I approval. These computer code improvements will be implemented at SCE's discretion.
4.5.3 Analysis Methodology Control The Reload Technology Transfer program provided SCE with an analysis methodology that was captured in class notebooks, reports, documentation, and calculations from past I cycles. In addition, the methodology discussion in the calculations performed for Units 2 and 3 Cycle 9 (Section 2.3) was enhanced to further document the methodology transferred to SCE. Section 3 provides a summary of SCE's reload analysis I methodology. Deviations in this established reload analysis methodology are controlled by NFM's calculation documentation procedure.
I NFM's reload analysis documentation meets all pertinent requirements of SCE's Topical Quality Assurance Manual (TQAM) Reference 36. NFM's procedure for documenting reload fuel cycle analyses functionally duplicates the recorded calculation format used by ABB CE. Any deviations in established reload analysis methodology, computer codes, and other significant changes are documented and approved on a deviation page. For purposes of NFM calculations, a methodology change is defined as any deviation from the methodology used the last time the analysis was performed, usually the last reload cycle. Computer code changes include changes in revision and in the use of any new s Southem Califomia Edison 82 November 199M
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l Section 4.0 Reactor Core Design and Monitoring Program NRC approved codes (within the limitations of the code SER). Other deviations include significant plant changes, significant calculation format deviations, or input data not being obtained from the results section of a calculation. All deviations are reviewed and approved by the analyst and NFM supervision and management.
Significant deviations from established methodology that increases the conservatism in a methodology approved by the NRC, as described in Section 3, will be reviewed by SCE management. Significant deviations from established methodology that reduces known conservatism but are still compliant with the methodology approved by the NRC, as described in Section 3, will be reviewed by SCE management and/or ABB CE, as I appropriate. Methodology changes that are not compliant with the NRC approved methodology, as described in Section 3, will not be implemented without prior NRC approval.
Concerns and issues raised by SCE regarding ABB CE's reload analysis methodology are reported to ABB CE using their quality affecting problem reponing process, currently the Condition Alen Program. These issues are evaluated by ABB CE per their procedures and SCE is notified of ABB CE's findings and error notices, as appropriate. ABB CE is required to notify SCE of all methodology enors. NFM utilizes existing SCE problem reporting procedures, currently the Action Request Program, to document NFM's evaluation of the impact of a methodology enor on reload analysis results. Operabililty, I reponability, root cause and corrective actions are controlled by SCE's problem reponing process, currently the Action Request Program, and tracked to completion.
ABB CE is continually refining and updating their NRC approved reload methodologies or developing new reload methodologies for NRC review and approval. These methodology improvements will be implemented at SCE's discretion.
4.5.4 Quality Program To ensure the highest quality analyses and processes, the Nuclear Fuel Management Division has establish a process of continuous improvement. The quality improvement lu program includes performance expectations, analysis tailboard, the Analysis Review Committee, self assessments and independent assessments. A tailboard is performed by the cognizant functional supervisor with an analyst near the initiation of a calculation to help ensure that the analyst understands the analysis to be performed, the procedures to be used, and the training requirements. The goal of the tailboard is to ensure that the majority of NFM's calculatiors are initiated with clear performance expectations and analysis goals, which should reduce errors. The Analysis Review Committee consists of Nuclear Fuel Management supervision and management. The calculation is reviewed by the Analysis Review Committee with the analyst prior to being independently reviewed.
l The goal of the Analysis Review Committee is to review the majority of calculations to ensure that all basic requirements for the calculation have been met, all deviations are acceptable, and th t all deviations have been recognized and recorded by the analyst and I
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1 Section 4.0 Reactor Core Design and Monitoring Program approved by supervision and management. NFM performs continuous self-assessment of its processes and products in accordance with the SONGS site self-assessment program.
The NFM quality program requires periodic independent assessment of NFM's reload analysis program.
4.6 SCE ENGINEER TRAINING QUALIFICATION GUIDE NFM engineering staff must meet ANSI-N18.1-1971 requirements for Technical Support Personnel. Initial engineering crientation training is consistent with 10CFR50.120," Training and Qualification of Nuclear Power Plant Personnel" and ACAD 98-004," Guidelines for Training and Qualification of Engineering Personnel." Nuclear Fuel Management Qualification Guide (Reference 37) specifics and administratively controls the Nuclear Fuel Management engineer qualification requirements. All NFM engineers obtain classroom /self study training on processes generic to every engineer's job function and on most of the procedures needed by the engineers. Some procedures are only used for specific tasks or positions, and training is provided on an as-needed basis prior to performing the tasks.
I With the completion of Reload Technology Transfer, the Nuclear Fuel Management Qualification Guide was updated to provide the requirements for an NFM engineer to obtain and maintain qualifications in one or more reload analyst positions, listed in Table 4.6-1. This I involves classroom /self study of the same training courses provided by ABB CE during the Reload Technology Transfer process as described in Table 2.1-1. In addition, training is required on any position specific procedures. Finally, the analyst must complete several calculations in I the specific analyst area under the guidance of a qualified reload analyst that has also completed On-The-Job Training (OJT) requirements.
I Each reload analyst position, listed in Table 4.6-1, contains multiple objectives that must be evaluated prior to being signed-off as complete. Each objective requires the NFM engineer to be both trained and evaluated. The engineer is expected to have reviewed all documents that support I the training prior to being trained. During the training process, the engineer and OJT Trainer shall ensure that the engineer has attained the knowledge and skills required by the objectives.
When the engineer and OJT Trainer are confident that the task can be performed without significant assistance, an OJT Evaluator will conduct an evaluation.
At the completion of the OJT evaluation, the analyst will be able to demonstrate an ability to:
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Discuss the design inputs, assumptions, methodology, computer codes used, and results obtained relative to acceptance criteria in the calculations performed for qualification in the position Demonstrate knowledge of the procedures applicable to the position Demonstrate overall knowledge and skills required for competent performance of position tasks I
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Section 4.0 Reactor Core Design and Monitoring Program The NFM engineer needs to demonstrate a level of knowledge sufficient to adequately complete an analysis of high quality in their reload analyst qualification position with minimal assistance.
Table 4.6-1 Reload Analyst Qualification Position Nuclear Design Analyst Plant Transient Analyst
' COLSS/CPCs Analyst Thermal-Hydraulic Analyst Fuel Performance Analyst Radiological Dose Analyst Pressure-Temperature Response Analyst (Not an RTT position)
The NFM Engineer must complete the applicable section to be qualified and work independently I in their respective qualification area. The qualification guide also establishes the requirements for periodic re-qualification for each qualified position.
l 4.7 CORE RELOAD ANALYSES AND ACTIVITIES CHECKLIST This Nuclear Fuel Management procedure (Appendix A) identify the reload analyses and I activities involved in completing a core reload fuel cycle analysis at SONGS. The NFM procedures provide a method for indicating the completion of tasks and any possible impact on the licensing, design basis, and site programs. For each analysis and activity the analyst is I provided a short summary of the principal inputs and significant products and where they are used. The procedure also lists the major computer codes needed to complete each analysis and is signed off at each step of the reload analyses. Reload t,nalyses and activities are not completed I until the startup physics testing has been successfully completed.
4.8 SOURCE VERIFICATION /ABB CE FUEL FABRICATION I INTERFACE l The Nuclear Oversight Division (NOD) has responsibility to ensure vendors of quality affecting equipment and services supplied to SONGS are in compliance with Criterion VII of 10CFR50 Appendix B. SCE's reload process ensures that NOD's QA inspectors use the reactor core design information developed by the Reactor Core Design Fuel Management Guidelines in auditing ABB CE's fabrication of SONGS fuel. The source verification procedure provides instructions for planning, performing, and reporting SCE's source verification activities.
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Section 4.0 Reactor Core Design and Monitoring Program 4.9 AIG CE ENGINEERING INTERFACE As shown in Figure 3.0-2, SCE did not obtain technology transfer for two analysis areas, LOCA and Fuel Mechanical Design. ABB CE will continue to perform these analyses. For each reload analysis for each SONGS unit, ABB CE sends a data request transmittal to SCE requesting analysis data and computer 61es needed to perform these analyses. The requested data and computer files are generated by SCE using the same processes used for the rest of the reload analyses. The data and computer files are then formally transmitted to ABB CE with computer files being sent on computer disks or electronically via the INTERNET. The results of ABB CE's completed LOCA and Fuel Mechanical Design analyses are transmitted to SCE for determination of potential licensing, plant design, and site impacts as well as incorporation into the final documentation used in the Design Process Flow and Controls discussed in Section 4.12.
I The Core Reload Analyses and Activities Checklist (Section 4.7) ensures that these interfaces are completed.
4.9.1 LOCA Analyses ABB CE will continue to perform the Large Break LOCA, Small Break LOCA, and Post l LOCA Long Temi Cooling analyses under contract for SCE. ABB CE will need data and computer files to perform these analyses. In addition to the data presented in the Reload i Ground Rules, Section 4.3, physics and fuel performance data discussed in Sections 3.1.1.4 and 3.3.2, respectively, are calculated by SCE and fonnally transmitted to ABB CE. The interface between SCE and ABB CE with respect to reload specific data and computer files is achieved via formal transmittals. ABB CE provides a " data request" I that specifies the data and computer files needed to perform the LOCA analyses. After completing the analyses described in Sections 3.1.1.4 and 3.3.2, SCE prepares a " data transmittal" to ABB CE detailing the cycle specific data and computer files. The final results of ABB CE's LOCA analysis are formally sent to SCE for incorporation into the FCE.
I 4.9.2 Fuel Mechanical Design Analyses ABB CE will continue to perfonn all fuel assembly mechanical design analysis. As is the case with the LOCA analyses. ABB CE requests physics data needed to verify the fuel assembly mechanical desigr, for the new reactor core design. ABB CE provides a " data request" that specifies the data needed to verify the fuel assembly mechanical design.
After completing the analyses described in Section 3.1.1.6, SCE prepares a " data transmittal" to ABB CE detailing the cycle specific data. The final results of ABB CE's I mechanical design analysis are formally sent to SCE for incorporation into the FCE.
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Section 4.0 Reactor Core Design and Monitoring Program 4.9.3 As.hullt Data l In order for SCE to perform the as-built analyses described in Section 3.1.1.8, ABB CE must provide the actual as-built fresh fuel isotopics calculated from the fuel I manufacturing assays. ABB CE formally transmits this data to SCE.
4.10 SITE PROGRAM IMPACT This is a site wide program to identify and assess the impact to site programs, procedures, or instructions for any change in plant design, configuration, or work methods. The Reload Analysis and Checklist procedure specifically defers to this process to ensure that all potential site impacts due to the reactor core design are identified, tracked, and managed. (An example would be the impact on reactor coolant system (RCS) chemistry due to an increase in soluble boron levels needed to support longer cycles. In this example, the Chemistry Division would be tasked with evaluating this impact, in consultation with other SCE divisions and with ABB CE, l and implementing the necessary changes to procedures and/or practices.) SCE uses this same process to control the site impact resulting from other changes including design changes and would be done independently of whether SCE or ABB CE perform the reload analyses.
4.11 LICENSING AND DESIGN BASIS UPDATES Similar to the site impact program, there are several site wide programs in place to identify and assess the impact to the licensing and design bases documentation. The Reload Analysis and l Checklist procedure specifically defers to these processes to ensure that all potential licensing I and design bases documentation changes due to the reactor core design are identified, tracked, and implemented. It is necessary that the system and topical design basis documents are kept current and accurately reflect any changes in the plant. ir.cluding the Accident Analysis Design Basis Document, Reference 40. Similar processes are used to control and maintain the Licensee Controlled Specifications, the Technical Specification Bases, and the UFSAR. These same processes are used for all other site programs and processes for controlling changes to the SONGS licensing and design bases documents. These procedures would be used by SCE independently of whether SCE or ABB CE performed the reload analyses.
4.12 DESIGN PROCESS FLOW AND CONTROLS f The implementation of a reactor core design is achieved using the same series of procedures used to implement a major design change to the plant, including guidelines for determining when a 10CFR50.59 safety evaluation is required. The level of documentation, review, and approval for a reactor core design is consistent with that used for a major plant design change. - SCE's design change process meets all requirements of 10CFR50 Appendix B and Electric Power Research
] Institute recommendations in Reference 47. A plant design change is implemented via a Design J Change Package (DCP) where as a reactor core design uses a Facility Change Evaluation (FCE).
The FCE process was established by adapting the DCP to fully augment the contractual and Southern California ikikon 87 November 1998
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Section 4.0 Reactor Core Design and Monitoring Program procedural requirements already existing for nuclear fuel. Both the DCP and FCE are controlled by the same design process flow and control procedures, Appendix A.
4.13 COLSS/CPC PRODUCTS The SONGS COLSS and CPC constants are installed in the CPCS and COLSS systems. The CPCS constants are divided into three types: data base, Reload Data Block (RDB) and addressable. These constants are verified by SCE each cycle, as described in Section 3.5.5.1.
However, the RTT program does not include the creating, testing or verifying CPCS software changes.
Data base constants are built into the CPCS software along with the CPCS algorithms. Data base constants cannot be changed except by changing the software, which involves considerable testing. Depending on the type of changes, implementation of these constants is controlled by a Facility Change Evaluation (Section 4.12) and/or with prior approval by the NRC.
RDB constants are located in protected memory of the CPCS and are separate from the algorithms and non RDB data base constants. RDB constants are loaded from a separate disk and can be changed without requiring a CPCS software change (currently all RDB disks are created by the fuel vendor). RDB constants are verified each cycle or changed each cycle, as discussed in Section 3.5.5.2.
Addressable constants can be changed by the plant operator under administrative controls. Type I I Addressable Constants are those that are subject to frequent (sometimes daily) changes. Type II Addressable Constants are those that are not changed frequently (typically once per cycle),
Section 3.5.5.3.
The COLSS constants are divided into two types: data base constants and addressable constants.
COLSS data base constants are specified in a data base document, and are implemented into the I COLSS program in the form of a compiled data base file. The COLSS data base constants are typically specified once for each reload fuel cycle and implemented prior to startup. The COLSS data base constants are calculated by SCE as discussed in Section 3.6.5.1. Addressable constants I can be changed as required during operation. Similar to the CPCS data base, the COLSS data base contains default values of addressable constants. The default values are updated based on the results of the COLSS analyses and the startup testing, Section 3.6.5.2.
I The Core Reload Analyses and Activities Checklist (Section 4.7) ensures that the CPCS constants (Section 3.5.5) and the COLSS constants (Section 3.6.5) are calculated and transmitted to Reactor Engineering for implementation for each unit and cycle. Implementation and control of all COLSS and CPCS constants and software is performed in accordance with Reactor Engineering and Computer Engineering procedures and is performed independent of whether I ABB CE or SCE performs the reload analyses.
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Section 4.0 Reactor Core Design and Monitoring Program 4.14 LOW POWER AND POWER ASCENSION TESTING Using as-built data, discussed in Section 4.9.3, and the actual cycle shutdown bumup data, ROCS /MC models are constmeted to provide the basis for the physics test predictions, incere detector constants, and initial COLSS and CPC addressable constants, as discussed in Section 3.1.1.8. These are transmitted to Reactor Engineering in accordance with the Reload Analysis and Activities Checklist. Reactor Engineering has responsibility for performing the low power and power ascensian testing, which is performed in accordance with Reactor Engineering and Operations procedures as required by the FCE. Deviations from expected core performance, as defined in Reactor Engineering procedures, are reported and tracked to completion by Reactor Engineering using a site wide problem reporting process, currently the Action Request Program.
Determination of operability, reportability, root cause, and corrective actions are controlled by the problem reporting process.
4.15 CORE AND SFP REQUIREMENTS Using as-built data, discussed in Section 4.9.3, and the actual cycle shutdown bumup data, fresh fuel assemblies are selectively placed to provide a balanced full core loading pattern as discussed in Section 3.1.1.8. The full core loading pattern is transmitted to Site Technical Services for implementation in accordance with the Reload Analysis and Activities Checklist.
As discussed in Section 3.1.10, to facilitate fuel assembly handling, temporary fuel placement guidelines are also developed and transmitted to Site Technical Services for core on-load and off-lod. Th . bumup and enrichment for fuel assemblies to be offloaded to the spent fuel pool are comped to the minimum requirements of Technical Specification. 3.7.18 to determine if the proposed storage is acceptable. Evaluation of boraflex coupon performance is performed to determine the best candidate for testing. Placement of spent fuel assemblies is planned to minimize neutron fluence of the boraflex panels.
4.16 CORE AND FUEL MONITORING The final step in the Reactor Core Design and Monitoring program is to monitor the performance of the fuel, core, key COLSS and CPCS parameters and key plant parameters during the cycle.
The fuel performance monitoring program incorporates the recommendations of INPO's SOER 90-02, Nuclear Fuel Defects, including tracking of the Fuel Reliability Index with associated review and action levels. RCS radionuclide inventory assessments, needed for the fuel performance monitoring program, are provided by the Chemistry Division in accordance with their procedures.
Nuclear Fuel Management Division performs the core follow program using plant data collected by Reactor Engineering. Procedural guidelines are provided for monitoring and evaluating reactorphysics, COLSS, and CPCS performance parameters. The measured reactor core performance and COLSS and CPCS performance are compared to predictions to ensure that the I core is performing as designed. The procedure sets forth the actions to be taken in response to larger than expected deviations in measured versus predicted reactor physics core performance Southern California Edison 89 November 1998
Section 4.0 Reactor Core Design and Monitoring Program parameters. Detennination of operability, reportibility, root cause and corrective actions are I tracked to completion using a site wide problem reporting process, currently the Action Request I program.
NFM core monitoring procedures also evaluate and trend estimated critical position performance during startups to ensure that there are no adverse trends. However, this procedure does not l control or perform the plant surveillance activities to monitor compliance with Technical Specification limits, which is the responsibility of the Reactor Engineering Group.
This same reactor ccre performance data and data taken during the low power and power ascension testing, Section 4.14, is used by NFM to evaluate and re-verify physics biases and l uncertainties. Specifically, the observed differences between the reactor physics measurements l l
and the predictions are evaluated to ensure continued validity of the existing physics biases and l Uncertainties.
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Southern California lidison 90 November 1998
Section 5.0 Comparison of SCE independent Analysis to ABB CE Analysis 5.0 COMPARISONS OF SCE INDEPENDENT ANALYSIS TO ABB CE AN (LYSIS The final pl ase of the RTI' Program was the Independent Analysis Phase as discussed in Section 2.3. In this phase, SCE engineers performed reload analyses independent of ABB CE based on the Unit 3 Cycle 9 reactor core design. All reload analyses were performed and reviewed by SCE engineers. In each calculation, comparisons of principal Unit 3 Cycle 9 analysis results were made to those from Unit 2 Cycle 9. Observed differences and similarities were reconciled. ABB CE engineers performed a second independent review followed by ABB CE management approval to independently evaluate SCE's capabilities. The objective of this phase was to demonstrate that the reload technology had been effectively transferred to the SCE engineering staff.
Based on SCE's satisfactory performance of all phases the Reload Technology Transfer program, ABB CE certified, in Reference 39, SCE's capability to independently perform reload analyses.
A second set ofindependent calculations are being performed by SCE. The Unit 2 Cycle 10 reload design analyses are being performed and independently reviewed by SCE engineers and I approved by SCE management using SCE's QA program. Principal results from these analyses were compared te the Units 2 and 3 Cycle 9 analyses. Observed differences and similarities are I being reconciled. The comparison between Unit 2 Cycle 10 and Units 2 and 3 Cycle 9 includes the effect of the Tcold Reduction Program in addition to reactor core designs changes shown in Section 5.1.
Whenever possible analysis comparisons presented in this section are based on the analyses performed by SCE for Unit 2 Cycle 10 relative to Units 2 and 3 Cycle 9. However, comparisons are also presented based on Unit 3 Cycle 9 analyses, relative to Unit 2 Cycle 9, to provide a more complete set of comparisons. In addition, the as-built physics test predictions are for Unit 3 Cycle 9 and are compared to actual plant stanup measurements rather than calculated values.
The input data and resu!!s presented in this section are used solely for comparison to other benchmark analyses. These comparisons of principal results represent the benchmarking to other analyses, as required by GL 83-11 (Reference 51), show SCE's competency regarding quality assurance practice and technical competency with respect to SCE's ability to set up computer code input decks, execute the codes, and properly interpret the results.
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m Southern California filison 91 November 1998
I Section 5.1 Unit 2 Cycle 10 Fuel Management Guidelines Comparison 5.1 UNIT 2 CYCLE 10 FUEL MANAGEMENT GUIDELINES COMPARISON Table 5.1-1 presents a comparison of key physics parameters from Table 3.1 of the Fuel Management Guidelines (Reference 2) showing that the Unit 2 Cycle 10 reactor core design meets all Design Criteria. The development of reactor core designs using the Fuel Management Guidelines is discussed in Section 4.1. The subsequent figures (Figures 5.1-1 to 5.1.10) compare key Unit 2 Cycle 10 physics parameters to historical SONGS values including Units 2 and 3 Cycle 9 as well as the design target values. These comparisons are used to determine the acceptability of the reactor core design and do not necessarily represent values used in the safety analyses and may represent best estimate values.
Southem California Edison 92 November 1998
Section 5.1 Unit 2 Cycle 10 Fuel Management Guidelines Comparison Table 5.1-1 Unit 2 Cycle 10 Physics Parameters Undt 2 Cycle 10 FSIGN PARAMETER CONDITIONS DFRGN.
NO- TARGET 4.45 w/o (586 EFPD)
- I" (ROCS /MC) 3.1 Maximum ARO IIFP F ,(Equilibrium) 7# # *""
1IR O (Figure .1-2) 1.55 s 1.50 3.2 Peak Linear IIcat Rate (kw/ft) I,#' 1 .0 s 10.1 g O (Figu .1-3) 3.3 Peak Rod Burnup (MWD /MTU) LEP, SAEOC 59,943 (Figure 5.1-4) 60,000 s 60,000 3.4 SEP, BOC, IIFP I800 s 1674 (ppm) (h,.gure 5.1.5) 3.5
"""*""' #"#""8 '"
SEP, BOC,200*F Concentration (ppm) 2594 2800 s2600 3.6 +
Most Positive MTC at IIZP (Ap/*F) SEP, BOC, HZP (F ure 5. )
3.7 Most Positive MTC at 70%P (Ap/*F) SEP, BOC,70%P *
+0.0 x 10 4 s -0.31 x 10'
[ 51 )
3 .51 Most Positive MTC at IIFP (Ap/*F) SEP, BOC, IIFP
- 51 )
3.9 Most Negative MTC at liFP (Ap/*F) SAEOC,11FP -3.12 x 10 -3.7 x 10 ' 2 -3.45 x 10 '
3.10 Minimum liFP Scram Worth (%Ap) N-I, IIFP, BOC 7.08 (Figure 5.1-9) 6.00 2 6.00 3.11 Minimum IIZP Scram Worth (%Ap) N-1, ilZP, BOC 5.61 (Figure 5.1-10) 5.15 2 5.15 3.12 Maximum Pellet Enrichment (wtW) Yaximum in Feed 4.45 4.80 s 4.80 Batch 3.13 B,C Loading (gram B' / inch )in BPR "*",',""'"'## N/A s 0.025 s 0.025 3.14 Erbia Loading (w t% Er203 in UO2 ) in Feed "h" g 2.1 s 2.5 s 2.1 5 2
'Ihe U " enrichment in the rods carrying the erbia is reduced by 0.4 wt%
3.15 Fuel Enrichment (wt%) of Erbia Rods 2 l below the peak U " enrichment in the assembly containing the erbia rod hee Ficure 5.1-1.1 m
Southern California Edison 93 November 1998
I l
section 5.1 unii 2 Cycle 10 Fuel Management Guidelines Comparison Figure 5.1-1 Enrichment Zone and Erbia Patterns for Unit 2 Cycle 10 Batch MO, M1, M2, M3, M4, and M5 Fuel Assemblies (586 EFPD, @ EOC9 = 555 EFPD)
I i fI I II l l l l l 1 m I i i m- E III I I I E E II E E l l II E
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M0 Fuel MI Fuel M2 Fuel iii Iil E I E E I E Ii 11 M i iM m: :s
! .esu=re El ,E ii ii E iE E I E II( l Ii M3 Fuel M4 Fuel M5 Fuel FUEL No. of ENRICHMENT ENRICHMENT No. Of Pins Containing 2.1%
TYPE ASSEMBLIES W/O U-235 W/O U-235 Erbia in 4.05 w/o Fuel F E E MO 16 4.45 4.05 0 M1 M2 8 4.45 4.05 48 M3 16 4.45 4.05 60 M4 M5 60 4.45 4.05 80 Total 100 l Southern California Edison 94 Novernber 1998
Section 5.1 Unit 2 Cycle 10 Fuel Management Guidelines Comparison Figure 5.1-2 Fxy Historical Data.
liFP Equibbnum Conditions. ARO 1.55
.m 1 ,
I x
"- \Eb d b. m, i 1.4 g-- v v 3 y3 I -
Cycle 4 Cycle 5 Cycle 6 Cycle 7 Cycle 8 Cycle 9 Cycle 10 m Cycle Number
--- Unit 2 Fxv -*-- Unit 3 Fe Design Recommended Fxy l
Figure 5.1-3 l Peak Linear Heat Rate IIistorical Data 15 IIFP. Equilibrium Conditions. ARO I c 13 2
1 $
5 11 g
I .
M g.m _C "- h _
k A_
~ ,kg -
Cycle 4 Cycle 5 Cvcle 6 Cycle 7 Cycle 8 Cycle 9 Cycle 10
-*- Unit 21.1IR --*- t ' nit 31.1(R LOCA Lunit Southern Califorma Edison 95 Novenber 1998
I Section 5.1 Unit 2 Cycle 10 Fuel Management Guidelites Comparison Figure 5.1-4 Peak Rod Burnup Historical Data 65 6 63 o 62 g et sn43
$ 60 so m; _
,]
b 59 I O 58 y 57 l g Se ~ . ~ _ , - _ , , _ , , _ . . _ - .
55.- ,
54 I CYCLE 4 CYCLE 5 CYCLE 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 CYCLE NUMBER 1
I Figure 5.1-5 l 2(XX)
Maximum HFP Boron Concentration Historical Data I 8'"
Z l800 DESION ilMIT I ;s i- o H
Z 1500 l u.!
l U 1400 Z
~ O 1300 U
Z 1200 m r
O ^ ~
( x 11(X)
O cc 1000 CYCLE 4 CYC1.E $ CYCLE 6 CYCLE 7 CYCLEH CYCIE9 CYCLE 10
[ CYCLE NUMBER Southern Califonna Edison 96 November 1998
Section 5.1 Unit 2 Cycle 10 Fuel Management Gttidelines Companson
. Figure 5.1-6 I 0.60 Most Positive HZP MTC IIistorical Data I 0.50 SONGS 2&3 TECH SPEC LIMIT
.I , 0.40 ua 0 0.30 ID a J
7
.a h0 I v .20 C~
0.10 0 00 CYCLE 4 CYCLE 5 CYCLE 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 CYCLE NUMBER l Figure 5.1-7 Most Positive MTC at 70% Power Historical Data oi,
,? 00 SONGS 2A3 TECil SPEC LIMIT y 4I mm O
I -02 a
43 ,
i
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\ N ID #
45
's
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I CYCL E 4 CYCLE 5 CYCL E 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 CYCLE NUMBER Southern California Edison 97 November 1998
Section 5.1 Unit 2 Cycle 10 Fuel Management Guidelines Comparison Figure 5.1-8 Most Positive HFP MTC Historical Data 01 C.0 SONG 2&.)11X31 SPEC 1AGT b
-- 42 2
0 4.3 3 .o4
=
4.h Q 46 -
g 47 -. _
E o:
49 CYCtr 4 CYCu 3 CYCII6 CYCH 7 CYC118 CYCII9 CYCII 10 I CYCLE NUMBER Figure 5.1-9 Minimum HFP Scram Worth Historical Data Z 7 50 M
I <
m 7M A O
I &*
% 6.50 h
I Due to the ECP Program, the HFP Scrum W<r1h was not Explzitty Calculated for g qd-s 6 00 -- -
7 M
I 5 50
=
h 5 00 m CYCLE 4 CYCLE 5 CYCLEti CYCI.E 7 CYCI.ER CYCLE 9 CYCLE 10 CYCLE NUMBER DESIGN LIMIT + SONGS-2 DATA --*-- SONGS-3 DATA Southem Caltforma Edison 98 November 1998
Section 5.1 Unit 2 Cycle 10 Fuel Management Guidelines Companson Figure 5.1-10 Minimum HZP Scram Worth Historical Data
^
O
=
5.75 f--
5.50
^
Iig 5.25 I$
N/
2 SONGS 2&3 TECil SPEC LIMIT h 5.00 Is c
d 4.75 CYC1.E 4 CYCLES CYCLE 6 CYCLE 7 CYCLE 8 CYCLE 9 CYCLE 10 I 8 CYCLE NUMBER m
--V- SONGS-2 DATA--dr- SONGS-3 DATA I
1
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c Southern Caldonna Edison 99 November 1998
l Section 5.2 Comparison of Principal Physics Results 5.2 COMPARISON OF PRINCIPAL PHYSICS RESULTS After the reactor core design is approved, physics data for all analyses must be generated. To enhance understanding of the data trends, to obtain an insight into the impact on given analyses, and to provide a check on the validity of the calculated values, a comparison of key parameters is performed for each physics calculation. These calculation comparisons are provided in the following sections.
Section Description l 5.2.1 General Core Physics Analysis 5.2.2 Input to Mechanical Design 5.2.3 Input to Fuel Performance, Core Thermal-Hydraulics and LOCA Analysis 5.2.4 Input to Safety Analysis Physics input to COLSS and CPCS analyses, beyond basic core physics data, is presented in Section 5.6.4 as part of the COLSS and CPCS analysis discussion.
5.2.1 Comparison of General Physics Data The comparison between the Unit 2 Cycle 10 and Units 2 and 3 Cycle 9 physics analysis parameters is reported in Tables 5.2.1-1 through 5.2.1-6.
Table 5.2.1-1 shows that the BOC and EOC doppler defects and BOC moderator defect are close. However, the Unit 2 Cycle 10 EOC moderator defect differs from that of Units 2 and 3 Cycle 9 by about 0.43%Ap. The reason for this deviation is the Tcold I reduction program. The Units 2 and 3 Cycle 9 moderator defect was based on a decrease of the moderator temperature (ATu) of 38 'F (from HFP 583 F to HZP 545 "F). For Unit 2 Cycle 10, ATy was only 26"F(from HFP 571"F to HZP 545 F) due to the Tcold Reduction Program. At BOC, this difference in temperature has a very small effect on the reactivity defect (raw MTC
- ATy) since the MTC is very small. However, at EOC, the MTC is more than 10 times larger and therefore the change in ATu leads to a larger difference in the moderator defect.
At BOC and EOC, the HFP Xenon worths are about the same.
Southern California Edison 100 November 1998
l l
Section 5.2 Comparison of Principal Physics Results I The minimum refueling boron concentrations, Mode 6, (assuming a moderator temperature of 200 'F, k,g - 0.95) are 2582 ppm and 2610 ppm for Unit 3 Cycle 9 and for Unit 2 Cycle 10 respectively. The difference is 28 ppm. Based on this Mode 6 difference I the expected HFP difference should be approximately 28 ppm. However, the HFP ARO critical boron concentrations is 1674 ppm for Unit 3 Cycle 9 compared to 1649 ppm for Unit 2 Cycle 10. The difference is -25 ppm or a net change of 53 ppm relative to the expected boron concentration based on the Mode 6 values. This reverse in trend is mainly due to the fact the BOC HZP MTC for Unit 2 Cycle 10 is smaller (less positive) than that of Unit 3 Cycle 9. For example, the raw MTC at HZP BOC for Unit 2 Cycle 10 is d
0.311x10 Ap/"Fcompared to 0.504x10dap/*F for Unit 3 Cycle 9. Therefore, the negative reactivity insertion corresponding to a decrease in temperature of 345 F (from HZP condition of 545 F to 200F) is smaller for Unit 2 Cycle 10. Consequently, the required minimum refueling boron is slightly higher for Unit 2 Cycle 10.
At BOC, the 95/95 MTC values are comparable for Unit 2 Cycle 10 and Unit 3 Cycle 9.
However, as mentioned above, the raw MTC values for cycles 9 and 10 differ by about 0.2 d
x10 Ap/ F. This comparison reDects the new MTC bias applied in Unit 2 Cycle 10.
The changes in the Unit 2 Cycle 10 and Unit 3 Cycle 9 maximum CBCs and the corresponding IBWs for the BOC and EOC conditions are shown in Table 5.2.1-3 and Table 5.2.1-4. The largest difference in the predicted CBC between the two cycles is 36 ppm. The largest difference in the predicted IBWs is less than 4 ppm /%Ap. The HFP critical boron run down comparison is presented in Figure 5.2.1-1 Unit 2 Cycle 9 and Unit 3 Cycle 9 have higher peaking factors Fxy (ARO and rodded) and Fr (tilted) when compared with those for Unit 2 Cycle 10 at each power level. The power I distribution depends on the assembly bumups and locations, and fresh fuel enrichment.
The peaking factors are summarized in Tables 5.2.1-5,5.2.1-6 and 5.3.1-7 and Figures 5.2.1-2 and 5.2.1-3.
Table 5.2.1-8 shows that there are some minor differences in the calculated bank wonhs between Unit 2 Cycle 10 and Unit 3 Cycle 9. However, the order of magnitude is comparable for both cases and the small differences are due to local changes in the power.
These tables show that the changes in the overall neutronic parameters are small and are consistent with the normal variations expected from cycle to cycle.
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. Southern Calif ornia Elison 101 November 1998
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Section 5.2 Comparison of Principal Physics Results Table 5.2.1-1 g Physics Design Parameters Design Parameters Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 (Unit 2 Cycle 10
. Unit 3 Cycle 9)
Boron (ppm)
BOC HFP 1649 1674 1658 -25 BOC HZP (Tmod=200 F) 2610 2582 2544 28 Doppler Defect (Mp)
BOC 0.9872 1.0410 1.030 -0.05 EOC 0.7712 0.8130 0.8081 -0.04 Moderator Defect ( u p)
BOC 0.1093 0.0927 0.0749 0.02 EOC 1.3528 1.7819 1.7531 -0.43 I Xenon Worth ( Mp)
-2.4041
-2.8118
-2.4519
-2.8628
-2.4583
-2.8635 0.05 0.05 MTC (x10E-4 ap/F)
BOC (100, 20% power) -0.33,0.19 -0.36,0.22 -0.33,0.23 0.03,-0.03 I BOC HZP 0.37 0.43 0.46 -0.06 MOC (90% power) -0.99 -0.98 N/A -0.01 EOC HFP -3,02 -3.27 -3.24 0.25 FTC HFP (x10E-3 ap/F)
BOC 1.54 1.54 1.52 0.00 I EOC ITC ARI-Stuck; BOCS 1.68 1.67 1.66 0.01 (x10-4 ap/AT)
Most Neg(500F, 0 ppm) -2.2286 -2.2183 N/A -0.01 Most pos(300F, 2610 ppm) 0.2377 0.0559 N/A 0.18 ITC ARI-stuck; EOCL (x10-4 ao/T)
Most Neg(500 F,0 ppm) -2.8104 -2.7261 N/A -0.08 l
l Most Pos(212F,2000 ppm) 0.2388 0.1632 N/A 0.08 Beta Effective BOC 0.0064 0.0064 0.0064 0.0000 EOC 0.0052 0.0051 0.0051 0.0001 Neutron Lifetime (psec)
BOC 15.7 16.3 16.5 -0.6 EOC 26.0 27.8 27.6 -1.8 l
l b
Section 5.2 Comparison of Principal Physics Results Table 5.2.1-2 Verification of LCS MTC Timepoint Power Xenon MTC Values (95/95)(x 10-' Ap/*F)
Level
(%) Unit 2 Unit 3 Unit 2 Groundrules Cycle 10 Cycle 9 Cycle 9 (RGR Figure I-1)
BOC 100 EQ -0.33 -0.36 -0.33 <-0.30 1 BOC 70 EQ -0.13 -0.12 -0.08 <-0.06 BOC 20 EQ +0.19 +0.22 +0.23 <+0.34 BOC 0 None +0.37 +0.43 +0.46 <+0.50 NOC 90 EQ -0.99 -0.98 N/A N/A MOC 70 EQ -0.9 -0.83 N/A N/A MOC 50 EQ -0.71 -0.64 N/A N/A I MOC 25 EQ -0.59 -0.48 N/A N/A MOC 20 EQ -0.52 -0.42 N/A N/A I EOC 100 EQ -3.02 -3.27 -3.24 >-3.70 I
Table 5.2.1-3 l BOC Maximum CBCs and Minimum IBWs Mode low liigh ARO ARI N-1 Temp Temp ACBC alilW ACilC AIBW ACBC AlllW (PPM) (PPht/7 cap) (PPM) (PPM /7 cap) (PPM) (PPM /%Ap)
I 1 350 620 -36 3.4 19 -3.8 -22 -3.1 2 350 620 -36 3.4 19 -3.8 -22 -3.1 3 350 620 36 3.4 19 -3.8 -22 -3.1 4 2(X) 350 6 -2.9 25 -3.2 -9 -2.9 5 68 200 10 -2.4 27 -3 -3 -3.0 6 6x 'm to .' a '7 1 _1 1o Note: A - Unit 2 Cycle 10 Minus Unit 3 Cycle 9 I
Southern California filison 103 November 1998
section 5.2 Comparison of Principal Physics Results Table 5.2.1-4 EOC Maximum CBCs and Minimum IBWs I
Mode low liigh ARO ARI N-1 Temp Temp ACBC AIBW ACBC AIBW ACBC AIBW (PPM) (PPM /%Ap) (PPM) (PPM /%Ap) (PPM) (PPM /%Ap)
I 1 350 620 -6 -2.1 -7 -2.9 0 -2.4 2 350 620 -6 -2.1 -7 -2.9 0 -2.4 3 350 620 -6 -2.1 -7 -2.9 0 -2.4 4 200 350 11 -1.7 4 -2.6 8 -1.9 I 5 68 200 11 -1.5 8 -2.3 11 -1.8 6 68 200 11 -1.5 8 -2.3 11 -1.8 Note: A - Unit 2 Cycle 10 Minus Unit 3 Cycle 9 l
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l Southern California lidison 104 November 1998
s <- - - um m- m m -- w Section 5.2 Comparison of Principal Physics Results Table 5.2.1-5 ARO Fm Rundowns Unit 2 Unit 3 Unit 2 Cycle Unit 2 Unit 3 Unit 2 Cycle 9 Cycle 9 Cycle 10 IOSEP Cycle 9 Cycle 9 SEP SEP LEP LEP LEP
- Fxy Box Fxy Box Fry Box Fry Bor Fry Box Fry Box 0.00 1.4224 40 1.4R01 23 1.4715 22 f.4369 31 1.5468 23 1.5149 22 0.00 1.4233 40 1.481 23 1.4718 22 1.4374 31 1.5475 23 1.5152 22 0.25 1.4209 40 1.4668 23 1.4574 22 1.4293 31 1.5309 23 f.4997 22 0.50 1.4204 40 1.4623 17 1.44 % 23 1.4277 31 1.5217 23 1.4899 22 1 1.4213 40 1.4609 17 1.4446 23 1.4283 40 1.5137 23 1.4814 23 2 1.421 40 146R5 35 1.4402 48 1.4287 40 1.505 17 1.4688 23 3 14178 40 1.4731 35 1.4423 48 1.4266 26 1.4988 17 f.452I 23 4 14124 40 1.4744 35 1.4416 4R I.4261 26 1.4911 17 1.4482 4R 5 1.4074 26 1.4747 35 1.4192 4R I.4232 26 1.4831 17 f.4466 48 6 I.4025 26 I.4729 35 1.4361 19 1.4194 26 1.4767 17 1.4442 19 7 13 %5 26 1.4696 35 1.4318 19 1.414 26 I.4707 17 I.441 19 8 1.3R95 26 f.4642 35 1.4272 19 1.4077 26 l.4677 35 1.4384 50 9 13835 2R I.4517 35 1.4223 19 1.4005 26 8.4645 35 1.437 50 10 13765 28 1.4493 35 1.4165 19 13927 26 I.459 35 1.4336 50 Il 13683 2R 1.4395 35 1.4108 35 13839 26 1.4527 44 1.4286 50 12 13593 2R I.43 44 1.4038 35 13747 26 1.4467 44 1.4222 50 13 1.35 28 1.4212 44 13966 19 1365 26 1.43 % 44 1.4153 50 14 1.3419 28 f.4143 17 13894 19 1.3555 26 1.4333 17 1.4081 44 15 13321 28 1.4098 17 1381f 19 13459 28 1.4265 17 I.4012 44 16 13237 31 1.4045 17 13751 Il 13375 31 1.4198 17 13914 44 17 13189 31 13995 11 13716 11 13322 31 1.4138 11 13875 11 18 13139 11 13947 11 13679 11 13299 11 1.4084 11 13839 11 19 13148 11 13893 11 13641 11 133 11 I.4027 il 13799 il 20 13142 il 13R36 11 13597 11 13289 11 13966 11 1.3756 Il 21 1.3122 11 13775 11 13556 Il 13265 13902 Il Il 13718 II 22 13095 11 13712 11 13502 )
Il I.3231 11 13837 11 13651 Il
?1 1 1059 11 11649 11 1 1491 11 Southem Califomia Edison 105 November 1998 e _ _ _ _ .
Section 5.2 Comparison of Principal Physics Results Table 5.2.1-6 Rodded F,, Data BOCS EOCS BOCL CEA EOCL Configuration Unit 2 Unit 3 Unit 2 Unit 2 Unit 3 Unit 2 Unit 2 Unit 3 Unit 2 Unit 2 Unit 3 Unit 2 Cycle 10 Cycle 9 Cycle 9 Cycle 10 cvele 9 Cycle 9 Cvele 10 Cycle 9 Cvele 9 Cvele 10 Cycle 9 Cyrle 9 ARO l.4232 1.481 1.4718 1306 13649 1.3452 1.4374 1.5474 1.5153 13233 13836 1365 BK6 1.5099 1.6012 1.5935 1.4054 1.4335 1.4137 1.5402 1.6655 1.6387 1.4253 1.4541 1.4316 BK6+5 1.4617 1.5266 1.5164 1.4789 1.4872 1.4927 1.534 1.5675 1.5636 1.4961 1.5032 1.5089 BK6+5+4 1.5692 1.699! l.6919 1.5306 1.5732 1.5421 1.6198 1.7598 1.739 1.5561 1.5973 1.5669 BK6+5+4+3 13553 1.6972 1.6806 1.61 % 1.891 1.904 1.6113 1.7455 1.7337 1.6335 1.9458 1.9368 B K6+5 +4+ 3+ P 1.5664 1.7182 1.6 % 2 1.6317 1.8925 1.9072 1.624 1.7521 1.7467 1.6467 1.9474 1.938 B K6+5 +4+ P 1.5833 I.7082 I.705 1.5586 1.5722 1.5566 1.632 1.7665 1.7498 1.5828 1.5942 l.5799 B K6+5 + P I.5056 1.537 1.5432 1.5087 1.5168 1.5195 1.5775 1.5722 1.589 1.5251 1.5299 1.5348 BK6+P I.5163 1.6112 1.6078 1.4133 1.4497 1.4359 1.5554 1.6733 1.6515 1.4316 1.467 1.4544 BKP l.4127 1.4018 I.4R62 13418 13717 I.3723 1.4461 1.558 1.5289 13587 138R1 13801 Southern Califomia Edison 106 November 1998
- fl1 l1l!l l s
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Section 5.2 Companson of Principal Physics Results Figure 5.2.1-2 SEP Fxy Rundown .
1.5 Nese: Assumes Cyde 9 Shandson A1 Early End of Aan> juts Window 4,-
2 2 -
8 45 u
b 1.4
'~.,
~. >
1.35 -
N 1.3 ' ' ' ' ' '
~ ' '
O 2 3 4 6 9 1 5 7 8 10 11 12 13 14 15 16 17 13 19 20 21 22 23 Burnup (GWD/T)
U2C10 - * - U3C9 + U2C9 Figure 5.2.1-3 LEP Fxy Rundown i6 1.55, Now Assumm. Osa 9 shadens as im Ead d Anahus Wadow 1.5 h.
I'd --
T~ _. _.
1.4 1.35
[
I.3 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 Burnnp (GWD/T)
U2C10 - -*-- U3C9 + U2C9 Southern Califorma Edtson 108 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.1-7 F,(Tilted) Versus Power Pcwer (%) Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 100 1.601 1.7204 1.6935 90 1.6189 1.7404 1.7151 80 1.633 1.7561 1.7321 70 1.643 1.7652 1.742 60 1.6591 1.7735 1.751 50 1.6795 1.7982 1,7768 40 1.6858 1.8004 1.7792 30 1.6893 1.8004 1.7792 25 1.7549 1.8999 1.8814 20 1.9124 2.1105 2.099 Tilt applied per Technical Specification 3.2.3.
N I
I I
I L
Southern California Edison 109 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.1-8 CEA Bank Worths Design Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 Delta Parameters (%Ap) (%Ap) (%Ap) (%Ap)
Bank 6 BOCS -0.3597 -0.3451 -0.3426 -0.014 MOCL -0.4022 -0.3985 -0.4024 -0.004 EOCL -0.3896 -0.4006 -0.4008 0.011 Bank 5 BOCS -0.3513 -0.3169 -0.3514 -0.034 MOCL -0.3427 -0.3260 -0.3360 -0.017 EOCL -0.3646 -0.3595 -03644 -0.005 Bank 4 BOCS -0.6972 -0.6467 o.6708 -0.051 MOCL -0.7413 -0.6828 -0.7163 -0.059 EOCL -0.7678 -0.7408 -0.7605 -0.027 Ilank 3 BOCS -0.7542 -0.9103 -0.8710 0.156 MOCL -0.9209 -1.0272 0.9818 0.106 EOCL -1.0644 -1.I298 -1.0995 0.065 Bank P BOCS -0.1944 -0.1841 -0.1880 -0.010 MOCL -0.2527 -0.2617 -0.2626 0.009 EOCL -0.2903 -0.3087 -0.3052 0.018 Southern California Edison i10 November 1998
Section 5.2 Comparison of Principal Physics Results 5.2.2 Comparison of Physics Data for Fuel Assembly Mechanical Design As discussed in Sections 3.1.2 and 4.9.2, physics data is requested by ABB CE for the fuel assembly mechanical design analyses which will continue to be performed by the fuel vendor. The following discussion is the specialized physics data typically needed for the fuel mechanical assembly design analysis, in addition to general physics data, presented in Section 5.2.1, which is used by many analyses. A comparison of the Unit 2 Cycle 10 specialized physics data relative to Units 2 and 3 Cycle 9 is provided in Table 5.2.2-1.
1 [ ] is an anticipated consequence of increased cycle length for Cycles 9 and 10, Section 5.0.
I The significant decrease [
] and the fact that the assembly where the maximum occurs is no longer surrounded by fresh fuel assemblies (as it was in Unit 2/3 Cycle 9). Starting with Unit 2 Cycle 9, the [ ], Section 5.0.
The notable reduction in Unit 2 Cycle 10 [
] is due to increased fuel loading in Cycle 10 l versus Cycle 9, again due to the use of erbia. Erbia is integral with the fuel while B4C displaces fuel.
The differences between Unit 3 Cycle 9 [ ] and Unit 2 Cycle 10 [
] are due to the conservative assumptions (simplified approach) used in the conversion from computer codes ROCS 4 to ROCS 5 in Cycle 7 which results in overly I conservative [ ] for a few cycles. (Unit 3 Cycle 9 ROCS concentration files still contain conservatisms (flat bumup option) from the conversion in Cycle 7. The flat burnup option used in converting between code versions, refers to assuming a flat intra-assembly fluence distribution. This assumption saves the re-depletion of multiple past cycles in ROCS 5 with the net effect of over predicting fluence for a few cycles.)
l The Unit 2 Cycle 10 [ ] while comparable to the Unit 2 Cycle 9 [
], is higher when compared to the Unit 2 Cycle 9 [
] This occurs because: (1) the maximum [ ]is in a different assembly than the maximum [ ] values (unlike Unit 2 Cycle 9 where all maximum I ] occurred ir the same assembly), and l (2) the maximum [ ] is in an assembly which is face-adjacent to four fresh fuel assemblies which results in an increase in fast neutrons in this assembly with a conesponding increase in [ ]
Southern Califomia Edison iII November 1998 1
l Section 5.2 Comparison of Principal Physics Results h Table 5.2.2-1
'l Physics Data For Fuel Assembly Mechanical Design ITEM Unit 2 Unit 2 Unit 3 Cycle 10 Cycle 9 Cycle 9 I
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' he bumup value is an analytical hmit which includes bumup uncenainties apphcable to the mechanical design analys operation willnot exceed the hmet of 60.000 MWD'MRJ, Reference 38.
r Southern California filison I12 November 1998
f Section 5.2 Comparison of Principal Physics Results f
5.2.3 Comparison of Physics Data for Fuel Performance, Core Thermal-Hydraulics,and LOCA Analyses As discussed in Section 3.1.3,3.1.4, and 3.1.5 (Fuel Performance, LOCA, and Core Thermal-Hydraulics, respectively), specialized physics data is needed by these three analyses in addition to some of the general physics data presented in Section 5.2.1, which is used by many analyses. In the case of the LOCA analyses (Sections 3.1.4 and 4.9.1),
j the data is requested by ABB CE because these analyses will continue to be performed by I them. The following discussion is for this specialized physics data.
The results of the Unit 2 Cycle 10 analysis, Table 5.2.3-1, are compared to those of Unit 3 Cycle 9 and Unit 2 Cycle 9. The results in the first table, which are used for all three analysis areas, are comparable. Differences can be attributed to two factors: 1) the number 6
of erbia fuel rods more than doubled and 2) the number of B4C shims has been reduced by a factor of ten. The increased core erbia has the effect of flattening the core power distribution. The reduction in shims (displaced with fuel rods) has decreased the core average LHR and slightly increased the minimum fast flux. The increased flux is also due, in part, to an increase in core average burnup.
Table 5.2.3-2 shows that the limiting [
] is lower than previous analyses due to the increased erbia core content.
The fuel performance [
l } Table 5.2.3 3 are comparable to previous analyses because the fuel management and peaking for this cycle is similar to Unit 2 Cycle 9 and Unit 3 Cycle 9.
As shown Figures 5.2.3-1 and 5.2.3-2, the LOCA [
]. More pins are at a higher [
] are becoming more evenly distributed.
The[ ] for LOCA Table 5.2.3-4, is comparable to Unit 3 Cycle 9.
cycle 10 has the second full batch of crtua fuel Southern California lidison i13 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.3-1 Physics Data for Core Thermal-Hydraulics, Fuel Performance, and LOCA Analyses Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 Core Average Linear ficat Generhtion Rate 5.311 KW/IT 5.444 KW/FT 5.444 KW/FT Maximum Integrated Radial Peak Fr:
Best Estimate F, 1.4130 1.4863 1.4536 F, at Upper Tolerance Limit 1.4729 1.5493 1.5152 Minimum fast flux value (4, , %) 0.571 x 10"n/cm 2/sec 0.569 x 10"n/cm 2/sec 0.568 x 10"n/cm 2/sec PIN-TO-BOX Factor (PTBF) 1.035 1.057 1.053 Maximum RCS Boron Concentration 1731 PPM 1755 PPM 1739 PPM l >
i l Table 5.2.3-2 Physics Data for Core Thermal-Hydraulic Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 I i I
I i Table 5.2.3-3 Physics Data for Fuel Performance Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 K - 59.4 GWDT II - 60.3 GWDT* J - 60 GWD/T L - 53.2 GWDTT K - 54.3 GWDT K - 53 GWD/T M - 30.9 GWDrT L - 32.0 GWD T L - 32 GWD/T
- NOTE: The value for Batch 11 in Unit 3 Cycle 9 is an analytical limit and operation will not exceed the limit of 60 GWD'MTU, Reference 38.
(
Southern California Edison i14 November 1998
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Section 5.2 Comparison of Principal Physics Results Table 5.2.3-4 Physics Data [ ] for LOCA.
I 1 I 1 [ ] I 1 [ ]
Unit 2 Cycle 10 Unit 2 Cycle 9 Unit 3 Cycle 9 I
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t- i Southern California E!ison i17 November 1998
Section 5.2 Comparison of Principal Physics Results 5.2.4 Comparison of Physics Data for Safety Analyses In addition to the data presented in Sections 5.2.1, the safety analyses can require specialized data, including a 1 D HERMITE model, the dependence of power peaking on changes in reactor coolant system temperatures, effect of single CEA and bank CEA drops, and withdrawals and reactivity effects of reactor coolant system cooldowns. This specialized data is discussed in Section 3.1.7.
5.2.4.1 Comparison ID HERMITE Model Input A 1D HERMITE model is typically used in safety analyses that needs to model space time effects on the axial power distribution during the scram such as the single reactor coolant pump sheared shaft / seized rotor analysis (Section 3.4.2.1.3).
I The HERMITE code can not simulate a core depletion such as ROCS /MC.
Therefore, the HERMITE code must be tuned to ROCS /MC at two times in life, BOC and EOC. The tuning criteria, how close HERMITE matches ROCS /MC, I are the values ofinterest in a cycle to cycle comparison. As shown in Tables 5.2.4.1-1 through 5.2.4.1-4, the Unit 2 Cycle 10 HERMITE tuning was consistent with past cycle results.
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Southern California Elison 118 November 1998
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Section 5.2 Comparison of Principal Physics Results Table 5.2.4.1-1 1D HERMITE Model BOCS I Unit 3 Unit 2 Unit 2 Cycle 9 #""##
Cycle 9 Cycle 10 Quantity IIERMITE. IIERMITE- IIERMITE-ROCSAIC ROCSAIC ROCS /MC C-A CB A B C IBW (ppm /%Ap) 1.48 1.24 0 -1.5 -1.2 M'IC (Ap/*F) 0.022e-04 0.0208e-04 0.0205e-04 3.0e-07 4.2e-07 1(X)-70% DD (%Ap) -0.012 -0.0099 -0.0139 -0.002 -0.0040 100-20% DD (%Ap) -0.027 -0.0169 -0.0331 -0.006 -0.0162 100-0% DD (%Ap) -0.034 -0.0187 -0.042 -0.008 -0.0233 MD(571-540)(%Ap) -0.059 -0.0519 -0.0118 0.047 0.0400 PD(100-0%)(%Ap) -0.093 -0.0706 -0.0538 0.039 0.0170 ASI at 1117 0.001 0.0 -0.(X)2 -0.001 -0.002 ASI at ilZP 0.041 0.04 0.008 -0.033 -0.032 IIFP Axial Peak (Fz) 0.001 0.001 0.00061 0.000 0.000 1 Xenon Worth (%Ap) -0.001 0.0024 0.003 0.004 0.(X)l0 ALBEDOS
>j _
Top Group I 0.00004 0.00242 0.(XX)199 0.00016 0.(X)222 Top Group 2 -0.00022 -0.00179 0.(XXX)52 0.00027 0.00184 Bottom Group I 0.(XX)l5 0.(XX)42 0.(XX)21 0.00006 -0.(XX)21 flottom Group 2 -0.00019 -0.(XII HH 0.(XY)182 0.(XX)37 0 00206 I
l Southern California Iklison 119 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.1-2 ID IIERMITE Model EOCL Unit 3 Unit 2 Unit 2 C3cle 9 Cycle 9 Cycle 10 Difference Quantitt
~
HERMITE- HERMITE- HERMITE-ROCS /MC ROCS /MC ROCS /MC C-A C-B A B C IBW(ppm /%Ap) 3.I 2.72 2 -1.100 -0.72 MTC (Ap/*F) 0.0370e-04 0.0362e-04 0.0230c-04 -1.40c-06 -1.32e-06 100-70%DD (%Ap) -0.009 -0.0103 -0.0093 0.000 0.0010 100-20%DD (%Ap) -0.024 -00269 -0.0265 -0.003 0.0004 100-0%DD (%Ap) -0.033 -0.0371 -0.0366 -0.004 0.0005 MD(572-540)(%Ap) -0.198 -0.1853 -0.1145 0.084 0.0708 PDt 100-0%) -0.232 -0.2224 -0.1511 0.081 0.0713 ASI at HFP 0.001 0.004 -0.001 -0.002 -0.005 ASI at HZP 0.04 0 04 0.027 0.013 -0.013 HFP Asial Peak (F,) 0.003 0.01 -0 00189 -0.005 -0.012 HFP Saddic indes -0.001 0.0035 -0 00483 -0.004 -0.0083 Xenon Worth (%Ap) -0.002 0.0022 0.0099 0.012 0.0077 ALBEDOS Top Group 1 0.00002 0 00037 0.000125 0.00011 -0.00025 Top Group 2 0.00001 000041 0.000029 0.00002 -0.00038 Bottom Group 1 -0.0006 -0 00075 0.000096 0.00070 0.00085 Bottom Group 2 -0.00063 -0 00058 0.000134 0.00076 0.00071 Southern California Edison 120 Nosember 1998
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Section 5.2 Comparison of Principal Physics Results Table 5.2.4.1-3 ID HERMITE GEOMETRY Deck Unit 3 Unit 2 Unit 2 Cycle 9 Cycle 9 '""#'
Quantity Cycle 10 A B C C-A C.B Power (watts) 3.39e+09 3.39e+09 3.39e+09 0 0 Core area (em**2) 94640.52 94640.52 94640 52 0 0 Axial boundary node (cm) 9.05 9.05 9.05 0 0 Adjusted I yield 0.0266244, 0.0266244, 0.0266244, 0 0 Xe yield 0.00106744 0.00106744 0.00106744 Core height (cm) 381 381 381 0 0 Gp-l velocity (cm/s) 17792000 17747000 17903200 111230 156230 Gp-2 velocity (cm/s) 400530 400260 398805 -1724 -1454 Lambda (BOCS)
Op-l 0.012748 0.012748 0.012749 0 0 Gp-2 0.03165 0.03165 0.031652 0 0 Op-3 0.1196 0.1196 0.11966 0.00006 0.00006 Gp-4 0.32018 0.32016 0.32027 0.00009 0.0001I Gp-5 1.402 1.4019 1.4017 -0.0003 -0.(X)02 Gp-6 3.8766 3.8763 3.8771 0.0005 0.(XX)8 Beta-eff (BOCS)
Gp-l 0.000209 0.(XX)209 0.00207 0 0 Gp-2 0.0012943 0.0012936 0.0012849 0 0 Gp-3 0.0011742 0.(X)! l 735 0.0011665 0 0 Gp-4 0.0025375 0.0025355 0.0025208 0 0 I Gp-5 0.000926 0.(XX)925 0.000921 0 0 Gp-6 0.000224 0.(XX)224 0.000223 0 0 I
Southern California lilison i2i November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.1-3 l ID IIERMITE GEOMETRY Deck Unit 3 Unit 2 Unit 2 Cycle 9 '"" ##
Quantity Cycle 9 Cycle 10 A B C C-A C.B Lambda (EOCL)
Gp-l 0.012795 0.012795 0.012795 -0.012795 0 Gp-2 0.031399 0.031398 0.031405 0 0 Op-3 0.12355 0.12357 0.12354 0 0 Gp-4 0.32754 0.32758 0.32751 0 -0.0001 Gp-5 1.4084 1.4084 1.408 -0.0004 -0.0004 Op-6 3.7884 3.7879 3.7908 0.0024 0.0029 Beta-eff (EOCL)
Gp-l 0.000151 0.000151 0.000152 0 0 Gp-2 0.0010739 0.0010713 0.0010764 0 0 Gp-3 0.000951 0.(XX)949 0.000955 0 0 Gp-4 0.0019877 0.0019823 0.0019958 0 0.00001 Gp-5 0.000775 0.000773 0.000779 0 0 Gp-6 0.000192 0.000191 0.(X)0193 0 0 l
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I Southern California Edison 122 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.1-4 1D HERMITE Thermal-Hydraulics Deck Unit 3 Unit 2 Unit 2 Cycle 9 Cycle 9 Cycle 10 Quantity A B C C-A C.B Fuel Pellet Dia. (in.) 0.3255 0.3255 0.3255 0 0 Gap thickness (in.) 0.00355 0.00355 0.00355 0 0 Clad thickness (in.) 0.02505 0.02505 0.02505 0 0 Stack density (gms/cc) 10.203 10.22 10.264 0.061 0.044 Clad density (gms/ce) 6.56 6.56 6.56 0 0 Pressure (ps:a) 2250 2250 2250 0 0 Nominal Flow (Ibs/hr) 140.6E6 140.6E6 140.6E6 0 0 Number of fuel rods 49820 49820 51068 1248 1248 Ileat deposited in coolant 0.025 0.025 0.025 0 0 Wetted Perim (ft) 5409.92 5409.92 5409.92 0 0 Coolant area (ft**2) 55.26 55.26 55.26 0 0 l
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- Southern California Edison 123 November 1998
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Section 5.2 Comparison of Principal Physics Results 5.2.4.2 Comparison of Fr Versus Inlet Temperature for Pin Census Events A comparison of Unit 3 Cycle 9 and Unit 2 Cycle 9 with the Unit 2 Cycle 10 %A F, / AT, data is presented in Table 5.2.4.2-1 for decreasing temperatures. Two time points are used (BOCS and EOCL) and the maximum (absolute) values are chosen between the time points. Note:[
l.
As shown in the Table 5.2.4.2-1, the Unit 2 Cycle 10 %A F, / ATs for decreasing T is comparable to Unit 3 Cycle 9 values. This is due to the similar fuel management pattems and flatter overall power distributions of the Unit 2 Cycle 10 and Unit 3 Cycle 9 cores.
l Table 5.2.4.2-1 Fr Versus Inlet Temperature for Census Events I
%A F,/ AT,, (T-in Decreasing) 15 F 30 F Unit 2 Unit 3 Unit 2 Unit 2 Unit 3 Unit 2 I Cycle 10 Cycle 9 Cycle 9 Cycle 10 Cycle 9 Cycle 9
-0.103 -0.099 -0.I21 -0.102 -0.099 -0.I21 I
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Southern California Edison 124 November 1998
Section 5.2 Comparison of Principal Physics Results 5.2.4.3 Comparison of Physics Data for CEA Ejection Analysis Comparison of Fq, maximum ejected CEA worth, and minimum scram wonh is in Tables 5.2.4.3-1 through 5.2.4.3-2. In general, Unit 2 Cycle 10 results are comparable to Unit 2 Cycle 9 and Unit 3 Cycle 9 results. Except at 20% (EOC) and 25% (BOC) power, post-ejected Fq is lower for Unit 2 Cycle 10. Except at 0% power, ejected wonh is a higher for Unit 2 Cycle 10. Except at 0% and 20%
power, minimum scram worth is lower for Unit 2 Cycle 10.
Cycles 9 and 10 results are comparable because the fuel management patterns are similar -- 100 fresh erbia assemblies in about the same locations. However, the differences in fuel management between Cycles 9 and 10 are responsible for the differences in Fq, ejected worth, and scram worth described above.
For example, Cycle 10 has fresh assemblies in full core locations 17 and 28, and once burned fuel in full core locations 41,42 and 56. Cycle 9 has fresh assemblies in Full core locations 17 and 28, a once bumed assembly in location 41, and twice bumed fuel in locations 42 and 56. At 0% power this causes the worst ejected rod to change from full core location 17 (Cycle 9) to 41 (Cycle 10) at BOC, but not EOC. Thus, the Cycle 10 BOC minimum scram wonh has ejected I and stuck CEAs in a once bumed and a fresh assembly, and the Cycle 9 BOC minimum scram wonh has ejected and stuck CEAs in two fresh assemblies. This causes the final Cycle 10 scram wonh to be taken from EOC (smaller than BOC),
whereas the Cycle 9 final scram worth is taken from BOC (smaller than EOC).
1 D pin census data at 100% and 0% power are compared in Table 5.2.4.3-3. Cycle 9 and 10 results are comparable with no discernible trend.
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I Southern Califomia Edison 125 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.3-1 Physics Data for Post-Trip Ejected CEA Power Unit 2 Unit 3 Unit 2 Unit 2 Cycle Unit 3 Cycle Unit 2 Cycle
(%) Cycle Cycle 9 Cycle 9 10 9 9 10 Post Post Ejected Ejected Ejected Post Fq Fq CEA CEA CEA Fq Wonh Worth Wonh
(%Ap) (%Ap) (%Ap)
COLSS IN-SERVICE - FASTTRIP - BOC 100 3.6601 3.7097 3.7069 0.186 0.1703 0.1703 90 3.6771 3.7488 3.7359 0.1842 0.1679 0.1681 70 4.6621 4.7151 4.8749 0.2125 0.1889 0.1925 50 4.7353 4.7948 4.9611 0.2088 0.184 0.1888 25 6.4581 6.3327 6.6227 0.2968 0.2436 0.2432 20 9.214 10.I155 9.8905 0.3827 0.3751 0.3552 0 9.7582 11.4216 11.7I96 0.3853 0.4092 0.4104 COLSS IN-SERVICE -- FAST TRIP - EOC 100 3.8926 4.1089 4.0751 0.2097 0.2101 0.2095 90 3.9259 4.1503 4.1217 0.2062 0.2065 0.206 70 4.4345 4.6961 4.5973 0.2342 0.2322 0.2322 50 4.4424 4.7092 4.6095 0.2279 0.225 0.225 25 7.I71 7.2201 7.4251 0.3397 0.2929 0.3074 1 20 11.1646 10.892 10.4055 0.5486 0.5479 0.5215 0 10.7994 10.8352 11.3526 l 0.5239 0.5692 0.5365 I
L Southern California Edison 126 November 1998
I Section 5.2 Comparison of Principal Physics Results Table 5.2.4.3-2 Scram Worth for CEA Ejection Power Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9
(%) Minimum (N-2) Minimum (N-2) Minimum (N-2)
SCRAM Worth SCRAM Wonh SCRAM Wonh
(%Ap) (%Ap) (%Ap) 100 -7.1565 -7.429 -7.5088 90 -7.1024 -7.3789 -7.4582 70 -6.6729 -6.9946 -7.0415 50 -6.5525 -6.889 -6.7745 25 -5.659 -5.9583 -5.9691 20 -4.9496 4.7556 -5.1253 0 -4.3505 -4.2113 -4.3591 I
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u Southern California FAison 127 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.3-3 Pin Census -- 100% and 0% Power 100% Power Interval Normalization Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 90 0.89--0.90 229 185 235 91 0.90 - 0.91 179 206 220 92 0.91 - 0.92 190 210 187 93 0.92 - 0.93 204 213 201 94 0.93 -- 0.94 168 185 143 95 0.94-- 0.95 156 153 104
% 0.95 - 0.96 114 140 101 97 0.96 0.97 74 100 60 98 0.97 -- 0.98 53 64 33 99 0.98 - 0.99 37 43 6 100 0.99 -- 1.00 20 16 8 TOTAL I424 1515 1298 0c/c Power Interval Normalization Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 90 0.89 - 0.90 44 14 52 91 0.90 - 0.9I 42 19 66 92 0.91 -- 0.92 46 26 47 93 0.92 -- 0.93 48 21 63 94 0.93 -- 0.94 26 6 36 95 0.94 -- 0.95 19 Ii 41 96 0.95 - 0.96 45 10 29 97 0.96 - 0.97 27 16 32 98 0.97- 0.98 16 13 17 99 0.98 -- 0.99 I5 11 7 100 0.99 -- 1.00 6 3 12 TOTAL 334 150 402 Southern California Iklison 128 November 1998 l
Section 5.2 Comparison of Principal Physics Results 5.2.4.4 Comparison of CEA (Full Length and Part Length) Single, CEA 2 &
3, and Subgroup Drop Physics Data The Unit 2 Cycle 10 values Table 5.2.4.4-1 through 5.2.4.4-6 are very similar to the Units 2 and 3 Cycle 9 values. This is to be expected, since the reactor core designs are similar. Any small differences are due to variations in local power distribution or the overall flatness of the local core power peaking in Cycle 10 relative to Cycle 9.
Table 5.2.4.4-5 shows that the Hot Channel F,y Penalty Factors for Unit 2 Cycle 10 have improved over Unit 3 Cycle 9. This improvement is due to the lower overall power peaking of the Unit 2 Cycle 10 fuel loading pattern.
Table 5.2.4.4-1 I FLCEA Total Distortion Factor (TDF)
Bank 6 Any Rod Time Unit 2 Unit 2 Unit 3 Unit 2 Unit 2 Unit 3 Cycle 10 Cycle 9 Cycle 9 Cycle 10 Cycle 9 Cycle 9 15 Min 1.1228 1.1230 1.1212 1.1408 1.1451 1.1518 1 lir 1.1661 1.1673 1.1646 1.2019 1.2089 1.2230 2 lir 1.2240 1.2263 1.2223 1.2834 1.2940 1.3179 707c Power 15 Min 1.1321 1.1320 1.1300 1.1663 1.1653 1.1736 1 Ilr 1.1758 1.1767 1.1737 1.2527 1.2490 1.2562 2 lir 1.2341 1.2362 1.2320 1.3680 1.3607 1.3665 1 50% Power 15 Min 1.1491 1.1455 1.1437 1.1850 1.1833 1.1912 1 lir 1.2021 1.1906 1.1879 1.2728 1.2683 1.2751 2 lir 1.2727 1.2509 1.2469 1.3900 1.3817 1.3870 m
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r Southern Califomia Edison 129 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.4-2 CEA 2&3 Drop Total Distortion Factor (TDF)
Time Unit 2 Cycle 10 Unit 2 Cycle 9 Unit 3 Cycle 9 90% Power 15 Min 1.0978 1.0857 1.0847 1 lir 1.1187 1.1121 1.1046 2 lir 1.1464 1.1471 1.1311 70% Power 15 Min 1.1233 1.1106 1.1065 I lir 1.1684 1.1539 1.1485 2 lir 1.2285 1.2115 1.2054 50% Power 15 Min 1.1259 1.1130 1.1095 l
I !!r 1.1710 1.1564 1.1520 2 Ifr 1.2313 1.2142 1.2087 l Table 5.2.4.4-3 Subgroup Drop Total Distortion Factor (TDF)
Time Unit 2 Cycle 10 Unit 2 Cycle 9* Unit 3 Cycle 9 90% Power 15 Min 1.1769 -
1.1613 1 lir 1.2019 -
1.1899 2 fir 1.2351 -
1.2279 70% Power 15 Min 1.2013 I
1.1852 I fir 1.2398 -
1.2215 2 fir 1.2911 -
1.2699 50% Power 15 Min 1.2055 -
1.1879 I lir 1.2441 -
1.2243 2 fir 1.2956 -
1.2729 Note:
- No subgroup drop TDFs were generated during Unit 2 Cycle 9. bounding values were used.
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_ Southern California Edison 130 November 1998
I Section 5.2 Comparison of Principal Physics Results I Table 5.2.4.4-4 PLCEA Drop Physics Results (Positive Reactivity)
Parameter Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 20% Power I
Worst Conditions BOC 0.021 %Ap (BOCL) 0.022 %Ap (BOCL) 0.022 % Ap (BOCL)
Reactivity insertion EOC 0.034 % Ap (EOCL) 0.035 %Ap (EOCL) 0.034 %Ap (EOCL)
BOC Worst Conditions 15 min 1.0614 (BOCS) 1.0614 (BOCL) 1.0738 (BOCL)
I Di ortion
' (
Factors 2 hrs 1.1026 (BOCS) 1.1112 (BOCL) 1.1347 (BOCL)
EOC Worst Conditions 15 min 1.1048 (EOCL) 1.1142 (EOCL) 1.1134 (EOCL)
I hr 1.1713 (EOCL) 1.1764 (EOCL)
Di onion 1.1809 (EOCL)
Factors 2 hrs 1.2597 (EOCL) 1.2594 (EOCL) 1.2707 (EOCL) 50% Power I Worst Conditions Reactivity insertion BOC 0.006 %Ap (BOCS) 0.012 %ap (EOCL) 0.006 % Ap (BOCL) 0.013 %ap (EOCL) 0.007 %Ap (BOCL) 0.013 % Ap (EOCL)
I BOC Worst Conditions D o on 15 min I hr 1.0361 (BOCS) 1.0533 (BOCS) 1.0357 (BOCL) 1.0531 (BOCL) 1.0371 (BOCL) 1.0549 (BOCL)
Factors 2 hrs 1.0763 (BOCS) 1.0763 (BOCL) 1.0786 (BOCL)
Worst Conditions 15 min 1.0594 (EOCL) ! 0573 (EOCL) 1.0575 (EOCL)
Di on on I r 1. 8 ( U l. 8 OCU l.M81 (EOCU Factors 2 hrs 1.1512 (EOCL) 1.1517 (EOCL) 1.1521 (EOCI.)
E l
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Southern Califoinia Edison 13i Nosember 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.4-5 PLCEA Drop Physics Results (Negative Reactivity)
Parameter Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 50% Power BOC Worst Conditions 15 min 1.0303 (BOCS) 1.0320 (BOCL) 1.0322 (BOCL)
Total Distonion I hr 1.0474 (BOCS) 1.0494 (BOCL) 1.0499 (BOCL) 2 hrs 1.0703 (BOCS) 1.0726 (BOCL) 1.0735 (BOCL)
Worst Conditions 15 min 1.0376 (EOCL) 1.0420 (EOCL) 1.0413 (EOCL)
Total Distortion I hr 1.0761 (EOCL) 1.0819 (EOCL) 1.0812 (EOCL) 2 hrs 1.1276 (EOCL) 1.1351 (EOCL) 1.1343 (EOCL) 70% Power BOC Worst Conditions 15 min 1.0301 (BOCS) 1.0316 (BOCL) 1.0312 (BOCL)
Total Distortion Factors I hr 1.0472 (BOCS) 1.0490 (BOCL) 1.0488 (BOCL) 2 hrs 1.0701 (BOCS) 1.0722 (BOCL) 1.0724 (BOCL)
Worst Conditions 15 min 1.0379 (EOCL) 1.0408 (EOCL) 1.0408 (EOCL)
Total Distortion I hr 1.0765 (EOCL) 1.0806 (EOCL) 1.0806 (EOCL) 2 hrs 1.1279 (EOCL) 1.1337 (EOCL) 1.1338 (EOCL) 90% Power BOC Worst Conditions 15 min 1.0300 (BOCS) 1.031 I (BOCL) 1.0319 (BOCL)
Total Distortion I hr 1.047 I (BOCS) 1.0485 (BOCL) 1.0496 (BOCL) 2 hrs 1.0700 (BOCS) 1.0716 (BOCL) 1.0732 (BOCL)
Worst Conditions 15 min 1.0377 (EOCL) 1.0415 (EOCL) 1.0410 (EOCL)
Total Distortion I hr 1.0763 ( EOCL) 1.0814 (EOCI.) 1.0808 (EOCL) 2 hrs 1.1276 (EOClJ 1.1345 (EOCL) 1.1340 (EOCL)
Table 5.2.4.4-6 Hot Channel F,y Penalty Radial Distortion Factors (RDF)
Bank Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 3 1.1225 1.2574 1.2884 4 No Penalty 1.1598 1.1500 6+5 No Penalty No Penalty No Penalty Southern Califomia Edison 132 November 1998
l Section 5.2 Comparison of Principal Physics Results 5.2.4.5 Comparison of Physics Data of CEA Withdrawal Overall both the Reactivity Insertion Rates (RIRs) and the power peaking for Unit 2 Cycle 10 are reduced when compared to Unit 2 Cycle 9 and Unit 3 Cycle 9.
This is an anticipated consequence of reduced power peaking and lower rod bank worths in Unit 2 Cycle 10.
The 100% and 50% power RIRs and peaking are comparable despite the difference in the initial power condition. This is to be expected, since RIR and peaking are a function of control rod bank configurations which are similar at the two power levels. (Both banks 5 and 6 are used at 50% while just bank 6 is used at 100%--bank 6 is the most worthy in both cases.) Since bank 6 is worth more than bank 5, the most limiting 50% power case actually has a lower RIR because bank 6 remains essentially in during the most limiting 50% power case.
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I Southern California filison 133 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.5-1 Physics Data for CEA Withdrawal Parameter Unit 2 Unit 3 Unit 2 Cycle 10 Cycle 9 Cycle 9 IIFP Maximum Bank RIR (E-4 Ap/sec) 0.40 0.448 0.448 IIFP Maximum Fq 2.747 - -
50% Power Maximum Bank RIR (E-4 Ap/sec) COLSS In/Out 0.34/0.32 - -
50% Power Maximum Fq COLSS In/Out 3.078/2.580 - -
IfZP Maximum Bank RIR (E-4 Ap/sec) 1.78 1.99 1.88 IIZP Maximum Fq 6.564 6.973 6.772 Subcritical BOC Maximum Shutdown Bank RIR (E-4 Ap/sec) 2.62 2.96 2.89 Subcritical BOC Maximum Shutdown Bank Fq 6.762 8.427 8.256 Subcritical BOC Minimum Initial Relative Power for 8.319 9.166 7.857 Keff - 0.99 (E-8 % Power)
Suberitical Maximum Shutdown Bank RIR (E-4 Ap/sec) 3.16 3.34 3.33 Suberitical Maximum Shutdown Bank Fq 7.178 8.206 7.995 i Suberitical Minimum initial Relative Power for Keff - 0.99 (E- 1.020 1.117 1.112 6 r4 Power)
Subcritical Regulating Bank Maximum RIR (E 4 Ap/sec) 1.78 1.99 1.88 Subcritical Regulating Bank Maximum Fq 6.564 6.973 6.772 Suberitical Minimum Initial Relative Power for 1.020 1.117 1.112 Keff - 0.99 (E-6 Q Power) l I
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Southern California Edison 134 November 1998
f Section 5.2 Comparison of Principal Physics Results 5.2.4.6 Comparison of Physics Data for CEA Deviation Within Deadband
{ BOCS and BOCL deviated worth and distonion factors for Unit 2 Cycle 9, Unit 3 Cycle 9, and Unit 2 Cycle 10 at 90%,70%,50%,25%, and 20% power CIS, COOS, and CEACOOS are compared in Tnbles 5.2.4.6-1 and 5.2.4.6-2 and are plotted in Figures 5.2.4.6-1 and 5.2.4.6-2.
The deviated wonhs and distortion factors are slightly higher in Unit 2 Cycle 10 than Unit 3 Cycle 9 and Unit 2 Cycle 9. This is consistent with the physics input to CEA ejection safety analysis which provides much of the input to this analysis, ne differences in points 23,19, and 20 in Figurc 5.2.4.6-1 and points 11 and 23 in Figure 5.2.4.6-2 are appreciably high and are explained below:
Point # 23 in Figure 5.2.4.61:
The wonh in BOCL CIS and COOS in the 20% power case is slightly lower in Unit 2 Cycle 10 than Unit 2 Cycle 9 and Unit 3 Cycle 9. The ejected rod of maximum worth in this case is the rod in full core box 17, which belongs to rod group 3, and the calculation below sho ws that group 3 wonh in Unit 2 Cycle 10 is lower than that in Unit 2 Cycle 9 and Unit 3 Cycle 9. Therefore, the deviated worth is also less in Unit 2 Cycle 10.
The rod worth for Group 3:
Unit 2 Cycle 9: -1.0995 %Ap Unit 3 Cycle 9: -1.1298 %Ap Unit 2 Cycle 10: -1.0664 %Ap Point #19 in Figure 5.2.4.6-l:
The worth in BOCL COOS at 50% power case is slightly lower in Unit 2 Cycle 10 than Unit 2 Cycle 9.
The total rod wonhs are:
Unit 2 Cycle 9 - 0.1335 %Ap Unit 3 Cycle 9 - 0.1136 %Ap Unit 2 Cycle 10 - 0.1141 %Ap Therefore, the worth for BOCL COOS at 50% power is higher in Unit 2 Cycle 9 than in Unit 2 Cycle 10.
Southern California Edison 135 November 1998
Section 5.2 Comparison of Principal Physics Results Point #20 in Figure 5.2.4.6-1 The worth in BOCL CEACOOS at 50% power case is slightly lower in Unit 2 Cycle 10 than Unit 2 Cycle 9.
The total rod worths are:
Unit 2 Cycle 9 -0.1335 %Ap Unit 3 Cycle 9 - 0.1136 %Ap Unit 2 Cycle 10 -0.1141 %Ap Consistently, this is also applicable to BOCL CIS and COOS at 20% power case Point # 23 The total rod worths are:
Unit 2 Cycle 9 - 0.2868 %Ap Unit 3 Cycle 9 - 0.3015 %Ap Unit 2 Cycle 10 - 0.2714 %Ap In conclusion, the data trends in this analysis are consistent with the data trends in the Cycles 9 and 10 physics data for CEA ejection safety analysis.
Point #11 and #23 in Figure 5.2.4.6-2:
The distortion factor in BOCS and BOCL CIS and COOS in the 20% power case shows a slightly higher difference. Comparison of distortion factors for Unit 2 Cycle 9, Unit 3 Cycle 9, and Unit 2 Cycle 10 from CEA Ejection calculations are shown below. Unit 2 Cycle 10 has a higher value than Unit 2 Cycle 9 and Unit 3 Cycle 9. Therefore, this slightly higher difference in distortion factor is consistent with distortion factors calculated in physics input to CEA ejection safety analysis. +
Maximum distortion factors from BOCS and BOCL CEA Ejection calculation:
Point #11 Point #23 BOCS BOCL Unit 2 Cycle 9: 1.8578 1.9027 Unit 3 Cycle 9: 1.8687 1.9751 Unit 2 Cycle 10: 1.9211 2.2574 In conclusion, the data trends in this analysis are consistent with the data trends in the Cycles 9 and 10 physics input to CEA ejection physics analysis.
Southern California Edison 136 November 1998
1 Section 5.2 Comparison of Principal Physics Results Table 5.2.4.6-1 I Deviated Worths for CEA Deviation Within Deadband I Deviated Worth WAp Time in Power Equipment Life level Operability Unit 2 Unit 3 Unit 2 Unit 2 Cycle 9 - Unit 3 Cycle 9 -
Cycle 9 Cycle 9 Cycle 10 Unit 2 Cycle 10 Unit 2 Cycle 10 I BOCS 90 % CIS/ COOS 0.0178 0.0180 0.0198 -0.0020 -0.0018 CEACOOS 0.0322 0.0326 0.0361 -0.0039 -0.0035 70 % CIS 0.0196 0.0196 0.0221 -0.0025 -0.0025 COOS 0.0173 0.0175 0.0194 -0.0021 -0.0019 CEACOOS 0.0312 0.0314 0.0350 -0.0038 -0.0036 50% CIS 0.0196 0.0196 0.0222 -0.0026 -0.0026 COOS 0.0172 0.0173 0.0193 -0.0021 -0.0020 I CEACOOS 0.0307 0.0310 0.0347 -0.0040 -0.0037 25 % CIS/ COOS 0.0268 0.0236 0.0292 -0.0024 -0.0056 CEACOOS 0.0318 0.0314 0.0364 -0.0046 -0.0050 20 % CIS/ COOS 0.0515 0.0547 0.0523 -0.(XX)8 0.0024 CEACOOS 0.0469 0.0471 0.0533 -0.0064 -0.0062 BOCL 90% CIS/ COOS 0.0296 0.0297 0.03(X) -0.(XX)4 -0.0003 CEACOOS 0.0555 0.0556 0.0564 -0.(XX)9 -0.0008 70% CIS 0.0317 0.0317 0.0323 -0.0006 -0.0006 COOS 0.0281 0.0280 0.0284 -0.0003 -0.0004 I CEACOOS 0.0518 0.0518 0.0527 -0.0(XX) -0.0(Xy) 50% CIS 0.0315 0.0305 0.0311 0.(XX)4 -0.0006 COOS 0.0315 0.0267 0.0271 0.0044 -0.0004 CEACOOS 0.0574 0.0486 0.0495 0.0079 -0.(XX)9 25 % CIS/ COOS 0.0354 0.0346 0.0380 -0.0026 -0.0034 CEACOOS 0.0461 0.0461 0.0471 -0.0010 -0.0010 20% CIS/ COOS 0.0891 0.0938 0.0843 0.0048 0.(XF15 CEACOOS 0.0620 0.0620 0.0628 -0.0008 -0.0008 I
l Southern Califomia Edison 137 Nosember 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.6-2 Radial Power Distortion Factors for CEA Deviation Within Deadband Distortion Factors Time in Power Equipment Life level Operability Unit 2 Unit 3 Unit 2 Unit 2 Cycle 9 - Unit 3 Cycle 9 -
Cycle 9 Cycle 9 Cycle 10 . Unit 2 Cycle 10 Unit 2 Cycle 10 DOCS 90% CIS/ COOS 1.014 1.015 1.017 -0.0030 -0.0020 l
CEACOOS 1.030 1.032 1.037 -0.0070 -0.0050 70 % CIS 1.040 1.037 1.042 -0.0020 -0.0050 COOS 1.015 1.017 1.019 -0.0040 -0.0020 CEACOOS 1.032 1.035 1.040 -0.0080 -0.0050 50% CIS 1.043 1.040 1.046 -0.0030 -0.0060 COOS 1.017 1.019 1.021 -0.0040 I CEACOOS 1.034 1.040 1.045 -0.0110
-0.0020
-0.0050 25 % CIS/ COOS 1.091 1.082 1.096 -0.0050 -0.0140 CEACOOS 1.044 1.049 1.055 -0.0110 -0.0060 20 % CIS/ COOS 1.145 1.145 1.161 -0.0160 -0.0160 CEACOOS 1.056 1.062 1.070 -0.0140 -0.0080 BOCL 90 % CIS/ COOS 1.028 1.029 1.029 -0.0010 0.0000 CEACOOS 1.055 1.057 1.055 0.0000 0.0020 1 70 % CIS 1.031 1.033 1.033 -0.0020 0.0000 COOS 1.031 1.031 1.031 0.(XXX) 0.0000 CEACOOS 1.059 1.060 1.059 0.(XXX) 0.(X)l0 50% CIS 1.034 1.037 1.036 -0.0020 0.0010 COOS 1.034 1.034 1.034 0.0000 0.(XXX)
CEACOOS 1.066 1.066 1.065 0.0010 0.0010 25 9 CIS.' COOS l.123 1.116 1.126 -0.0030 -0.01(X)
CEACOOS 1.077 1.077 1.078 -0.0010 -0.0010 20% CIS/ COOS 1.150 1.161 1.217 -0.0670 -0.0560 1 CEACOOS 1.096 1.097 1.098 -0.0020 -0.(X)l0
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Southern Califomia Edison 138 November 1998
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Southem California Edison 140 November 1998 i
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Section 5.2 Comparison of Principal Physics Results 5.2.4.7 Comparison of Physics Data for Post-trip Steam Line Break Analysis The comparison between Unit 2 Cycle 10 and Unit 2/3 Cycle 9 reactivity balance is presented in Tables 5.2.4.7-1 and 5.2.4.7-2 for HZP and HFP, respectively.
Table 5.2.4.7-1 shows that the difference (Unit 2 Cycle 10-Unit 3 Cycle 9) in the HZP reactivity balance is 0.04 %Ap at 300"F, which is equivalent to about 4 ppm boron. Table 5.2.4.7-2 shows that the difference (Unit 2 Cycle 10-Unit 2 Cycle 9) in the HFP reactivity balance is 0.01 %Ap at 300*F, which is about 1 ppm boron.
A comparison of the physics data from Units 2 and 3 Cycle 9 to data for Unit 2 Cycle 10 is also shown by graphing the reactivity balance as a function of the temperature for the HZP and HFP cases in Figure 5.2.4.7-1 and Figure 5.2.4.7-2, respectively. This shows that the cooldown data for both Cycles 9 and 10 are similar in magnitude. This result is expected since : a) the [
]', b) the FTCs are 4 4 almost identical; namely: 1.68x10 and 1.67x10 Ap/*F for Cycles 10 and 9, respectively and c) the raw N-1 scram worths of both cycles are comparable.
I Therefore, the small differences in the cooldown data between the two cycles are consistent with the normal variations expected from cycle to cycle.
The reactivity credit and the Fq comparisons are reported in Tables 5.2.4.7-3 and 5.2.4.7-4 for the low flow condition and in Tables 5.2.4.7-5 and 5.2.4.7-6 for the high flow condition. The comparison shows that the reactivity credits for the low I flow condition are almost identical except that the Power / Flow fractions are slightly different. Because the scaling factors of both cycles are similar I
I ], the reactivity credits for the low flow condition are similar. For the high flow condition, the reactivity credits were explicitly calculated by running ROCS /MC. Since the T/H feedbacks of both I cycle and the worth of the stuck rod are similar (similar core design), the resulting reactivity credits are expected to be similar.
The Fq comparison shows that Cycle 10 power peaking is in general lower than that of Cycle 9 peaking. This difference is mainly due to the change in the assembly burnups and locations, increased use of erbia and fresh fuel enrichments.
I Per Techncial specification 3.1.4 Southern Califomia Edison 141 Novernber 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.7-1 l HZP Post Trip SLB Reactivity Balance Fuel & Mod Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 Temp (*F) (% Ap) (% Ap) (% Ap) 560 -5.15 -5.15 -5.15 542 -4.345 -4.3488 -4.3449 533 -3.7968 N/A N/A 500 -2.619 -2.6338 -2.6261 I 450 -1.1777 -1.2013 -1.1871 400 -0.0883 -0.119 -0.099 350 0.7857 0.749 0.7738 300 1.498 1.4559 1.4848 250 2.2431 2.189 2.221 200 2.8881 2.8218 2.8564 68 4.0516 3.9554 3.9935 Table 5.2.4.-7-2 l HFP Post Trip SLB Reactivity Balance M d Temp (*F) Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 FM Tg*F) (%Ap) (% Ap) (% Ap) (% Ap)
I 990.2 593 N/A N/A -7.9908 989.4 593 N/A -7.9925 N/A 967.8 593 -7.7932 N/A N/A 585 585 N/A -6.9087 -6.9161 573 573 -6.2914 N/A N/A 560 560 -5.6437 -5.6311 -5.6275 542 542 -4.8393 -4.8299 -4.8235 I 533 533 -4.4756 N/A N/A 500 500 -3.299I -3.296 -3.2847 450 450 -1.8604 -1.8635 -1.8485 400 400 -0.7739 -0.7812 -0.7637 l 350 350 0.0969 0.0868 0.1058 300 300 0.8059 0.7937 0.8138 250 250 1.5486 1.5267 1.5477 200 200 2.1912 2.1596 2.1811 68 68 3.3498 3.2931 3.3149 r
Southem Califomia Edison 142 November 1998
Section 5.2 Comparison of Principal Physics Results Figure 5.2.4.7-1 IIZP Post Trip SLB Cooldown Curve 6
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Southern Califoi sia Elison 143 November 1998
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- Southern California Edison 144 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.7-3 Post Trip SLB Low Flow Reactivity Credits Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10
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I l Table 5.2.4.7-4 Post Trip SLB Low Flow Fq Credits Power Fraction Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10 I
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- Southern Califomia Edison 145 November 1998
Section 5.2 Comparison of Principal Physics Results Table 5.2.4.7-5 Post Trip SLB High Flow Reactivity Credits Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10 Power Fraction (%Ap) (ckAp) (ckAp) 1 I
Table 5.2.4.7-6 Post Trip SLB High Flow Fq Credits Power Fraction Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10 l I I
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Southern California Edison 146 November 1998
Section 5.2 Comparison of Principal Physics Results 5.2.4.8 Comparison of Physics Input to ASGT Safety Analysis A comparison of Unit 2 Cycle 10 with Unit 2 Cycle 9, Unit 3 Cycle 9, and Units 2 and 3 generic Physics input data for ASGT analysis is shown in the following Table 5.2.4.8.1.
l p
All Unit 2 Cycle 10 values are very similar to Cycle 9 values. In addition, Cycle 10 values at 90%,70%, and 50% power levels are very close to the generic values.
A slightly larger value is observed at 20% power.
Generic values were calculated using ROCS /MC at 90% and 20% power levels with the [
]. The intermediate values were linearly interpolated. The slightly higher Unit 2 Cycle 10 value at 90% power is due to the more negative current MTC (-3.7x10" ApfF). In addition, the generic calculation l did not use the [ ] calculated at 20% power which in effect used a slightly less negative MTC than -3.5x104 Ap/ F. Therefore, a slightly larger difference is observed at 20% power.
I Table 5.2.4.8.1 Physics Data for ASGT Safety Analysis Power F, Distortion Factor for [ ] A F, Distonion / 'F I (%)
Unit 2/3 Unit 2 Unit 3 Unit 2 Unit 2 Unit 3 Unit 2 Generic Cycle 9 Cycle 9 Cycle 10 Cycle 9 Cycle 9 Cycle 10 90 1.4060 1.4576 1.4540 1.4540 0.00915 0.00908 0.00908 70 1.5450 1.5592 1.5530 1.5580 0.01118 0.01106 0.01116 50 1.6850 n/a n/a 1.7115 n/a n/a 0.01423 20 1.8940 n/a n/a 2.1200 n/a n/a 0.02240 I
I Southem California Edison 147 November 1998
Section 5.3 Comparison of Core Thermal-Ilydraulics Design Results ;
( 5.3 COMPARISON OF CORE THERMAL-HYDRAULICS ANALYSIS RESULTS The core thermal-hydraulics analyses affect directly and indirectly the non-LOCA Transients and
' the COLSS and CPC events discussed in Sections 3.4,3.5 and 3.6. To estimate the impact of the core thermal-hydraulics analyses and to evaluate the validity of the analysis results, a series of comparisons are performed on key innut parameters and key results.
5.3..t Comparison of Basics Thermal-Hydraulics Data Table 5.3.1-1 Basic Thermal-Hydraulic Data Description Unit 2 Cycle 10 Units Unit 2 Cycle 9 Unit 3 Cycle 9 Core Power: 3390 Mwth 3390 3390 RCS Flow rate (w.cs) 95%: 143.3 lbm'hr 140.6 140.6 Core Bypass Flow: 3% of Design 3% 3%
Core Inlet Temperature Tn): 540 'F $53 553 Core inlet Enthalpy (h,,): 534.9 Bru/lbm 550.9 550.9 Core Flow Area (A,,,,): 54.816 ft 2 54.816 54.816 RCS System Pressure (P,cs): 2250 psia 2250 2250 l Fluid Film Coefficient (Avg) 6000 Bru/hr-ft 'F 3
6000 Film Temperature Difference (Avg) 30.3 'F 30.3 30.3 Core Average lleat Flux 0.1771 MBlu/hr-ft' O.1815 0.1815 Total lleat Transfer Area 6.3712'10* ft' 6.2148*10' 6.2148*10' I 1 Average Linear lleat Rate 5.189 Kw/ft 5.319 5.3187 (densified)
Core Enthalpy Rise 83.22 Btu /lbm 84.8 84.81 Core Average Enthalpy 576.5 Btu /lbm 593.3 593.3 Care Average Reynolds Number 4.6512*10' 4.837*10' 4.7054*10' i
1 Core Pressure Drop 18.15 lbf/in 2 17.6 17.684 Veuel Preuure Dron 42 45 lbflin' 41.4 41.5 l l
The above table shows expected results given the increased number of fuel rods (reduced number of B4C shim rods) the Tcold Reduction Program, and the increase in Guardian Grid type fuel in the Unit 2 Cycle 10 core.
I Southem California Edison 148 November 1998 l
Section 5.3 Comparison of Core Thermal-flydraulics Design Results 5.3.2 Comparison of Core Wide Parameters A comparison of the core parameter calculation results is shown below:
Table 5.3.2-1 Core Parameters Parameter Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 Units Core Averagelleat 1.15672
- 10 l.15672
- 10'" 1.1569*10 Btu /hr Generation 4
Core IIcat Transfer Area 6.3840*10 6.2273
- 10* 6.2272
- 10 4
ft2 l 1
[ ] 6.3712*104 6.2148 *10' 6.2148
- 104 ft' 6
Core Average IIcat Flux 0.1816
- 10 0.1857*106 0.1858*106 Bru/hr-ft' I
I 1
Core Mass Flux 2.6692
- 10 6 2.6189*106 2.61849*106 lbnVhr-ft' Changes in the values are attributed to 1) the reduction in T , and 2) the reduced number of B4C Thim rods in the core for Unit 2 Cycle 10.
5.3.3 Comparison of CETOP-D Benchmarking g Table 5.3.3-1 CETOP-D Results CETOP Penalties Unit 3 Cycle 9 Unit 2 Cycle 9 Unit 2 Cycle 10 I i Cycle 10 shows a increase in penalty due to the consideration of other assemblies as pctentially limiting locations for MDNBR. CETOP penalties were driven by Assembly 1 ), while prior analyses have focused their analysis efforts on Assembly [ ]. Unit 2 Cycle 10 results for assembly [ ] were similar to prior analyses results, however consideration of Assemblies [ ] and [ ] resulted in a more significant CETOP penalty.
(See Figure 3.2-2 for thennal-hydraulic analysis model layout and assembly identification 1 numbers.)
Southern California Edison 149 November 1998
Section 5.3 Comparison of Core Thermal-Ilydraulics Design Results 5.3.4 Comparison of MSCU - Response Surface Verification A comparison of the Unit 2 Cycle 9 MSCU analysis results and the Unit 2 Cycle 10 MSCU analysis results is shown in Table 5.3.4-1..
Table 5.3.4-1 MSCU Results Case Description [ l }
. I 1 l
Unit 2 Cycle 9 [ ]
5 Unit 3 Cycle 9 [ ]
Unit 2 Cycle 10 Limiting assembly 9 nxxiel [ ]
Unit 2 Cycle 10 Limiting assembly 23 model [ ]
Unit 2 Cycle 10 Limiting assembly 46 model [ ]
l
[ ]
Unit 2 Cycle 9 [ ]
Unit 3 Cycle 9 [ ]
Unit 2 Cycle 10 Limiting assembly 9 model [ ]
I Unit 2 Cycle 10 Limiting assembly 23 model [ ]
Unit 2 Cycle 10 Limiting assembly 46 model I }
All values for the Unit 2 Cycle 10 TORC models were less than the reference response surface sensitivities, and the results were comparable to the results for Unit 2 Cycle 9 and I Unit 3 Cycle 9. Even with the changes in Tcold and the additional assemblies considered, all results were bounded by the existing response surface.
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Section 5.3 Comparison of Core Thermal-Ilydraulics Design Results 5.3.5 Comparison of Steam Line Break Initial Void Fraction The SLB initial void fraction analysis was performed for Unit 2 Cycle 10. The void fraction was determined at Unit 2 Cycle 10 SLB conditions using both the limiting assenibly 9 and limiting assembly 46 quarter assembly TORC models. (See Figure 3.2-2 for thermal-hydraulic analysis model layout and assembly identification numbers.)
The Unit 2 Cycle 9 core average void fraction was 0.021%. The Unit 2 Cycle 10 core average void fraction was 0.023% (limiting assembly 9) and 0.020% (limiting assembly 46).
The results show good agreement with the Unit 2 Cycle 9 analysis, as there was no signiGcant change in void fraction.
Table 5.3.5-1 SLB Void Reactivity Results Parameter Unit 2 Cycle 9 Unit 2 Cycle 10 Units Reactor Power 102 % 102 %
I ]
Heat Flux (100%) 0.1862 0.1816 Mbtu/hr-ft2 Core inlet Temp 560 560 F System Pressure 2200 2200 psia RCS Flow Rate 90 % 90 %
Nominal Mass Flux 2.6286 (553 "F) 2.6692 (540 F) Mlbm/hr-ft2 Mass Flux at 560 Fand 1007c 2.6010 l How 2.5952 Mlbm/hr-ft2 Mass Flux at 560 *F and 907c 2.3409 2.3356 Mlbm/hr-ft2 l How CORE AVERAGE VOID FRACTION 0.021 % 0.023 %
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Southern Califomia Edison 151 November 1998
Section 5.4 Comparison of Fuel Performance Analysis Results y 5.4 COMPARISON OF FUEL PERFORMANCE ANALYSIS RESULTS
- E The Unit 2 Cycle 10 fuel performance analysis was computed using the same methodology as was used for Unit 2 Cycle 9 and Unit 3 Cycle 9. The fresh fuel (Batch M-UO2 and erbia) was
! identical in design to the Batch L fuel loaded in Unit 2 Cycle 9 and Unit 3 cycle 9 with the exception ofinitial enrichment. The core average linear heat rate and therefore long-term rod average power was lower for Unit 2 Cycle 10 than Unit 2 Cycle 9 and Unit 3 Cycle 9 (5.311 kW/ft versus 5.444 kW/ft for Cycle 10 versus Cycle 9, respectively.) However, the shor;-term peak linear heat rate, that is the primary input determining peak temperatures and pressures, were identical at beginning oflife. The major sources of the differences in Unit 2 Cycle 10, Unit 2 Cycle 9, and Unit 3 Cycle 9 is the [ ]. The Unit 2 Cycle 10 analysis also uses an inlet temperature of 540 F versus 553 'F, in accordance with the Tcold Reduction Program.
The major Unit 2 Cycle 10 parameters which are compared with Unit 2 and 3 Cycle 9 fuel performance analysis results are hot rod internal pressures, hot rod gap conductance, and hot rod average temperatures at the peak power node as a function of bumup for erbia and UO 2 fuel. In order to compare the data, a graph of[
] for Unit 2 Cycle 10 Unit 3 Cycle 9, and Unit 2 Cycle 9 are provided in Figure 5.4-1. Additionally, the [ ] being compared are provided in Table 5.4-1.
i Table 5.4-1 Hot Rod Average Power Unit / Cycle [ ] Rod Average Power (kW/ft)
Unit 2 Cycle 10 1 ]
Unit 3 Cycle 9 [ ]
Unit 2 Cycle 9 l 1 4
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Section 5.4 Comparison of Fuel Performance Analysis Results 5 5.4.1 Comparison ofIlot Rod Internal Pressure i The hot rod intemal pressure as a function of bumup is plotted in Figure 5.4.1-1 for the Unit 2 Cycle 10 Batch M erbia hot rod, the Unit 3 Cycle 9 Batch L erbia hot rod, and the Unit 2 Cycle 9 Batch L erbia hot rod. Similar plots comparing Batches L and M UO2 hot rod internal pressures, and Batch K UO2 hot rod intemal pressures are provided in Figures I 5.4.1-2 and 5.4.1-3, respectively.
l The graphs of the internal pressures are consistent wi'h [
]. The Unit 2 Cycle 10 hot rod internal pressures are slightly less than the Cycle 9 hot rod internal pressures up to a burnup of 31,000 MWD /MTU due to the lower
[ ] for Cycle 10 compared to Cycle 9, resulting in a larger gap volume at low bumup and release of fewer fission gases. [ ]for Cycle 10 remains higher than for Cycle 9 in the 34,000 to 45,000 MWD /MTU power I range. This is reflected in the higher Cycle 10 hot rod internal pressures compared to Cycle 9. The pressures for the Cycle 10 and Cycle 9 hot rods remain consistent with the respective [ ] for the remaining burnup range. The behavior of the I hot rod internal pressures is consistent with the input data.
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Southern California Edison 154 November 1998
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l Section 5.4 Comparison of Fuel Perfonnance Analysis Results 5.4.2 Comparison ofIlot Rod Gap Conductance The hot rod gap conductance is plotted as a function of bumup for Unit 2 Cycle 10, Unit 3 Cycle 9, and Unit 2 Cycle 9 for Batches L and M erbia fuel, Batches L and M UO2 fuel, and Batch K UO2 fuel in Figures 5.4.2-1,5.4.2-2 and 5.4.2-3, respectively. It should be noted that the gap conductances plotted are at the peak power axial node for a given bumup, and that the peak power node is not at the same location for consecutive burnups or for the different units and cycles being considered. The behavior of the gap conductance, however, is similar for the different units and cycles, and can be compared.
As shown in the figures, the gap conductances behave predictably and consistently. The gap conductances initially drop due to fuel densification, and then rise as a result of gap closure. As fission gases build up, the gap conductances decrease. The gap conductances reach a second peak when the [ ], due to a decrease in [
] The gap conductances decrease as fission gases continue to build up. The gap conductances
[ ] again at burnups of 53,000 GWD/MTU and above consistent with the [
] and resulting changes in fission gas production.
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E Southern Califomia Edison 158 November 1998
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i -r- 1 m n-Section 5.4 Comparison of Fuel Performance Analysis Results Figure 5.4.2-3 K UO2 Ilot Rod Gap Conductance use--
umumummu ammm-Southern California Edison 161 November 1998
Section 5.4 Comparison of Fuel Performance Analysis Results 5.4.3 Comparison of Hot Rod Average Temperatures at the Peak Power Node The hot rod average temperatures at the peak power nodes are plotted in Figures 5.4.3-1, 5.4.3-2, and 5.4.3-3 for Unit 2 Cycle 10, Unit 3 Cycle 9, and Unit 2 Cycle 9 Batches L and M erbia fuel, Batches L and M UO2 fuel, and Batch K UO2 fuel. The points on the temperature curves correspond to the same locations and burnups as the points on the gap conductance curves shown in Figures 5.4.2-1,5.4.2-2, and 5.4.2-3. Maximum temperatures occur at minimum gap conductances, as expected, and are similar for all cases for similar fuel. The het rod average temperature curves inversely mimic the hot rod gap conductance curves. The temperature curves are consistent with expectations.
Southern California Iklison 162 November 1998
_ _ _ . . . _ . . . - _ . _ _ . . _ _ . . _ _ . _ _ _ ~ . . . . . . _ _ . ~ . . . . _ . _
i Section 5.4 Comparison of Fuel Perfonnance Analysis Results ;
Figure 5.4.3-1 L&M Erbia Ilot Rod Average Temperature t
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Section 5.5 Comparison of Safety Analysis Results 5.5 COMPARISON OF SAFETY ANALYSIS RESULTS Based on the changes to the plant, RGR, licensing basis, and physics data and the fuel performance and core thermal-hydraulics analyses, a subset of the non-LOCA Transient analyses I are performed. A comparison of these event input data and results relative to Units 2 and 3 Cycle 9 is presented in the following sections.
5.5.1 Comparison of CESEC Basedeck Results The results for basic steady state run case showed the same trend as the previous I basedeck calculation. The final conditions of the key parameters were within 1% of the initial conditions. The case converges to the expected steady state values after a short initialization transient. A comparison between the previous and curTent results can be seen in Table 5.5.1-1. Upon examining the results, it can be seen that the current basedeck core temperatures and secondary parameters are lower due to the changes from l the Tcold Reduction Program. The plugged tube cases can not be directly compared to previous results due to different initial conditions and plugged tube arrangements.
l Table 5.5.1-1 CESEC Basedeck Calculations.
~
Parameter Units Unit 2 Cycle 9 Unit 2 Cycle 10 Core Power Fraction of 1.019 1.019 l RTP Core Heat Flux Fraction of 1.020 1.020 RTP I RCS Pressure psia 2260.2 2260.2
' Pressurizer Pressure psia 2251.5 2251.5 Core Flow Fraction 0.9723 0.9894 '
Core Inlet Temperature *F 553.1 540.079 Core Outlet Temperature F 612.1 600.179 I TWFR ft
- F-2 1.7345 1.7345 sec/ Btu Steam Generator Pressure - Right psia 910.2 815.4 Steam Generator Flow - Richt ihm/sec 2145.4 2137.3 Note: 1.
The core flow ratio is based on each time step's core flow rate versus the design total flow rate.
I For this case the design total flow rate is based on 553 *F where as the cunent time step's core flow rate is based on 540.079 *F. This value is equivalent to a total flow rate excluding the 3 %
bypass.
Southern California Edison 166 November 1998
Section 5.5 Comparison of Safety Analysis Results 5.5.2 Comparison of Boron Dilution Analysis l A comparison ofinput data and results was made between Unit 2 Cycle 10 analysis and f that of the previous analyses, Units 2 and 3 Cycle 9. There were no changes in RCS volumes or dilution time constants. There were changes in CBC and IBW which account i
for the differences between this analysis and the previous analysis in the time to reach 7 criticality. For the limiting case, i.e., BOC ARO, the CBC and IBW decreased for Modes 2 and 3, and increased for Modes 4,5, and 6. The time to reach criticality for the limiting
} case decreased for Modes 2 and 3 and increased for Modes 4,5, and 6. The time to reach criticality for all modes of operation, except Mode 5, remains greater than 60 minutes.
The time to reach criticality for Mode 5 is greater than 45 minutes, which allows 30 minutes for detection.
Table 5.5.2-1 Boron Dilution Results - Time to Criticality Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10 Mode 2 108 min 108 min 108 min Mode 3 108 min 108 min 108 min Mode 4 76 min 76 min 78 min Mode 5 (3 charging pumps) 48 min 48 min 49 min Mode 5 (Mid Loop and I charging 143 min 142 min 145 min pumpi Mode 6 (3 charging pumps) 48 min 61 min 59 min Mode 6 (I charging pump) 145 min 183 min 179 min Table 5.5.2-2 Boron Dilution Results - SRM ratio Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10 Mode 2 & 3 8.4 8.4 8.8 Mode 4 6.5 6.5 6.8 Mode 5 3.9 3.9 4.0 Mode 6" 6.4 6.4 6.2 2600 ppm initial refueling boron concentration minus 50 ppm. with one charging pump in operation.
Southern Cahfornia filison 167 November 1998
l Section 5.5 Comparison of Safety Analysis Results 5.5.3 Comparison of Feedwater Line Break Analysis Results The comparisons ofinitial conditions between Unit 2 Cycle 10 and Units 2 and 3 Cycle 9 are shown in Table 5.5.3-1. The comparison of results are shown in Table 5.5.3-2. For approximately the first 60 seconds, the sequence of events and resultant values are similar to the AOR results. The peak primary and secondary pressures are different but are expected due to the multiple differences shown in Table 5.5.3-1. The numerous differences between initial conditions, number of plugged steam generator U-tubes, MTC, etc., have increased the peak primary pressure to 2852.5 psia.
In general, the event trends along the same lines as the previous analysis. The differences between these two events are due to the changes in the AFW model and Decay Heat model. It was found that the steam driven AFW pump flow contributed more negative results to this event since it is aligned to both steam generators at the same time, thus providing half the available AISV flow to each steam generator. This contributed to the l end results by diverting valuable water to suppon the intact steam generator heat sink.
The decay heat model impacts the end result by an increase in decay heat energy added to the event. Limiting the flow to the intact steam generator and increasing decay heat I results in less energy being removed from the primary system which results in a heat up of the primary system and finally lifting the PSV to decrease primary system pressure. By a lifting the PSVs, steam is removed from pressurizer and RCS liquid surges into the pressurizer. Eventually the pressurizer water volume nears the PSV entrance. If the liquid level was at the PSV entrance and the PSVs lifted, liquid would be passed through the PSVs ar..! possible damage the PSVs.
In general the differences between the current analysis and the AOR can be contributed to increase in plugged tubes, initial conditions, decreased AFW flow, and an increase energy I added by the improved decay heat model.
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Southern California Iklison IM November 1998
Section 5.5 Comparison of Safety Analysis Results Table 5.5.3-1 Feedwater Line Break Analysis Input Parameter Units Units 2/3 Cycle 9 Unit 2 Cycle 10 initial Core Power MWt 3477.8 3477.8
- Core inlet Temperature *F 560 560 g Pressurizer Pressure psia 2150 2200 Pressurizer Water Volume ft' 913.9 913.9 RCS Total Flow gpm 356,400 356,400 Number of SG U-Tubes Plugged -
1000 2000 Decay lleat Model -
ANS 1971 Curve ANS 1979 Curve I Decay lleat MoJet Uncertainty -
No uncertainties 20 + Actinides Steam Power Demand Fraction 1.02 1.02 Steam Generator Pressure psia % ).6 885.16 Intact SG Inventory lbm I80,000 180,000 Feedwater Line Break Area ft2 0.2 0.175 I MTC 10 ' op/*F 0.0 -0.3 Doppler Reactivity Curve -
Least Negative least Negative Doppler Uncertainty Fraction 0.75 0.75 SCRAM Reactivity Curve for CEA Insertion asiu, Curve ( ASI, Delay Time (sec)) +0.3,1.01 +0.3* 1 *01 seconds Kinetics -
Max Generic Max Generic ligap BTU /sec ft2 ,.F 0.230 0.230 SCR AM Worth 9 Ap -6.0 -6.0 Pressurizer Pressure Control System
- Automatic Automatic Pressuriier 1rvel Control System Mode of Operation Automatic Automatic Steam Bypass System *
- g inoperative Inoperative MSSV Tolerance 9 +3 +2 PSV Tolerance G +3 +3 PSV Blowdown Percent 9 8.5 8.0 PSV Area per valve Ft' O 02668 0.0223 AFW Flow t % Flow to each AFW Modelmg -
SG until EFAS I '"**#I AFW Isolation AIT Flow Opm 500 Based on SG Pressure Southern California lilison 169 November 1998
Section 5.5 Comparison of Safety Analysis Results Table 5.5.3-2 Feedwater Line Break Analysis ResultS Units 2/3 Cycle 9 Unit 2 Cycle 10 Event Time (sec) Value Time (sec) Value Iligh Pressurizer Pressure Trip 36.95 2434 psia 40.7 2434 psia Condiuons Trip Breakers Open 37.85 -
41.6 Normal On-site Power lost -
Emergency Feedwater Actuation Signal 38.1 26.27 ft 41.75 26.57 ft Affect Steam Generator Empties 38.5 <5000 lbm 42.5 <5000 lbm CEAs Begin to Drop 38.85 -
42.6 Pressurizer Safety Valves Stan to Open 39.6 2575 psia 42.8 2575 psia Maximum RCS Pressure Occurs 41.144 2832.2 psia 44.946 2852.5 psia Peak Secondary Pressure Occurs 45.144 1156.80 psia 52.7 1150.09 psia Pressurizer Safety Valves Close 48.244 2356.12 psia 53.55 2300 psia Auxiliary Feedwater Enters Intact Steam Generator 92 -
86 -
Steam Generator low Pressgre Trip 212.2 675 psia 243.5 675 psia Condition and MSIS Imuated Main Steam isolation Valves Begn t Close 213.1 -
244.4 -
Complete Closure of Main Steam Isolation Valves Terminating Blowdown 223.1 -
252.4 -
from the intact Steam Generator EFAS isolation Signal based on AP between intact and Affected SG 273.6 250 psid Minimum Liquid Mass in the Steam 250.2 6989.94 lbm 289.55 6114.60 lbm Generator to intact Feedline Affected SG AFW Flow Isolated - -
316.3
~
Main Steam Safety Valves Open on the 414.4 1133 psia 437.4 1122 psia intact Steam Generator Maximum Pressurizer Liquid Volume Occurs 47.544 1269.4 ft' 1373.85 1371.1 ft'
Section 5.5 Cornparison of Safety Analysis Results 5.5.4 Comparison of Pre-Trip Steam Line Break Analysis Results
( The results were similar to the results of Units 2 and 3 Cycle 9. The results are presented in Table 5.5.4-1. The major difference between the current and previous calculation is the flatter power distribution. 'Ihe flatter power distribution increases the number of fuel f' rods that would fail when subjected to similar DNB conditions. Figures 5.2.3-1 and 5.2.3-2 shows normalized Fr as compared to previous cycles for both LEP and SEP, respectively. To compensate for this increase in fuel failure, [
). This change maintained the fuel failure below what was previously used in the associated dose calculation.
Due to the [ ], all other parameters ofinterest, LHGR and DNB Propagation were calculated lower than previous cycles. In general, [ ] was used in Unit 2 Cycle 10 than in previous cycles to maintain the end results similar or f lower than previous cycles.
Table 5.5.4-1 Pre-Trip Steam Line Break Analysis Results Parameter Units Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10
}
1 Break Size IC/OC ft' 5.4/5.4 5.4/5.5 5.6/5.5 IC SLH Fuel Failures % <9 <9 9.35 OC-SLB Fuel Failures % 4.97 6.13 4.16 (C-SLB LilGR (1) kw/ft 17.49 17.65 16.26 OC-SLH LilGR(1) kw/ft 17.29 17.31 15.95 DNil Propagation No No No Note: 1.
It should be noted that LilGR for previous cycles was calculated based on a 3.0 second delay in LOAC.
I Southern California Edison 17i November 1998 9
Section 5.5 Comparison of Safety Analysis Results 5.5.5 Comparison of Post-Trip Steam Line Breck Analysis Results Comparison of key input parameters for Unit 2 Cycle 10 to those for Units 2 and 3 Cycle 9 post-trip steam line break analyses are given in Tables 5.5.5-1 and 5.5.5-2 for HFP and HZP, respectively. The comparison of results to those of previous analyses are given in Tables 5.5.5-3 and 5.5.5-4 for HZP, and Tables 5.5.5-5 and 5.5.5-6 for HFP. The relative changes from Unit 3 Cycle 9 to Unit 2 Cycle 10 are discussed. The values for Unit 2 I
Cycle 9 have, also, been included for interest, but are not discussed further herein.
The following comparison of reactivity credits is consistent with the comparison presented in Section 5.2.4.7.
HERMITE Ao Credits - The high flow reactivity credits have decreased from Unit 3 Cycle 9; and the low flow reactivity credits, temperature tilt - 200 F, power / flow ratios have increased by roughly 1.5% from Unit 3 Cycle 9.
High Flow Fa - The Unit 2 Cycle 10 values have decreased by about 2.7% from the Unit 3 Cycle 9 values.
Low Flow Fq - The Unit 2 Cycle 10 values at power fractions of.02 and 0.0 decreased by 2.8% and 4.3%, respectively, from the Unit 3 Cycle 9 values.
Inverse Boron Worth - The minimum (most negative) IBW has decreased for Unit 2 Cycle 10 for the HZP cases and increased for the HFP cases. Thus, for the HZP cases when boron is added via safety injection, the negative reactivity added is more than for Unit 3 Cycle 9. For the HFP cases when boron is added via safety injection, the negative i reactivity added is less than for Unit 3 Cycle 9. Changes in this parameter have a significant impact on the results of this analysis.
HPSI Flow Curve - The safety injection flow curve for Unit 2 Cycle 10 has decreased relative to Unit 3 Cycle 9 (To provide more margin for performance test). Thus. when boron is added via safety injection, the negative reactivity added in a given amount of I time is less for Unit 2 Cycle 10 than for Unit 3 Cycle 9 for a given IBW. Changes in this parameter have a significant impact on the results of this analysis.
Safety Iniection Delay Time - The safety injection delay time increased for Unit 2 Cycle
- 10. This served to worsen the MDNBR slightly for this analysis. However, the impact was small.
Kinetics - The delayed neutron fraction for Unit 2 Cycle 10 is somewhat greater than for Unit 3 Cycle 9. However, the impact of this is insignificant.
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_ Southem California Edison 172 November 1998
Section 5.5 Comparison of Safety Analysis Results SCRAM Worth - The scram worth applicable to the HFP cases for Unit 2 Cycle 10 was slightly lower than for Unit 3 Cycle 9. However, the same space-time reactivity relative insertion curve, (i.e. for scram worth s 10 %), was used for both Unit 2 Cycle 10 and Unit 3 Cycle 9 HFP cases.
HFP - The results for both the HFP cases are similar to those reported in Unit 2 Cycle 9.
Both sets of cases showed no return-to-power as a result of the transient. However, the decrease in maximum post-trip negative reactivity was expected due to the changes in IBW and HPSI flow delivery relative to Unit 3 Cycle 9. Both of these changes served to reduce the negative reactivity addition.
H7P - The results for the HZP-AC case have a mixed trend relative to that for Unit 3 Cycle 9, although none of the differences are very large. The considerable reduction in the IBW for Unit 2 Cycle 10 served to offset, to some degree, the adverse impact of the reduction in the HPSI flow delivery.
The results for the HZP-LOAC case (i.e., the low-flow HZP case), are more strongly impacted by the decrease in HPSI flow than would be for the high-flow (HZP-AC case).
All results for the Unit 2 Cycle 10 HFP-LOAC case are more adverse than those for Unit 3 Cycle 9.
The Unit 2 Cycle 10 HZP-LOAC maximum post-trip fission power is higher than that of Unit 3 Cycle 9 because the total positive reactivity is higher than Unit 3 Cycle 9 at time of interest.
The Unit 2 Cycle 10 HZP-LOAC maximum post-trip reactivity is slightly higher than that of Unit 3 Cycle 9 due to the net effect of the change in the IBW and the HPSI How I delivery. The considerable reduction in the ]BW for Unit 2 Cycle 10 served to offset, to some degree, the adverse impact of the reduction in the HPSI flow delivery.
The Unit 2 Cycle 10 HZP-LOAC minimum post-trip MacBeth DNBR is lower than that for Unit 3 Cycle 9 due to the reduction in the HPSI flow delivery relative to that for Unit 3 Cycle 9.
( The Unit 2 Cycle 10 HZP-LOAC maximum post-trip LHGR is higher than that for Unit 3 Cycle 9 due to the reduction in the HPSI flow delivery relative to that for Unit 3 Cycle 9.
The Unit 2 Cycle 10 HZP-AC maximum post-trip Hssion power is slightly higher than that of Unit 3 Cycle 9 because the total positive reactivity is higher than Unit 2 Cycle 9 at
[ time ofinterest. The Unit 2 Cycle 10 HZP-AC maximum post-trip reactivity is slightly higher than that of Unit 3 Cycle 9 due to the net effect of the change in the IBW and the HPSI Dow delivery. The considerable reduction in the IBW for Unit 2 Cycle 10 served to
{ offset, to some degree, the adverse impact of the reduction in the HPSI flow delivery.
The Unit 2 Cycle 10 HZP-AC minimum post-trip MacBeth DNBR is higher than that for r
L Southem California Edison 173 November 1998
Section 5.5 Comparison of Safety Analysis Results Unit 3 Cycle 9 due to the reduction in the IBW relative to that for Unit 3 Cycle 9. The Unit 2 Cycle 10 HZP-AC maximum post-trip LHGR is lower than that for Unit 3 Cycle 9 due to the reduction in the IBW relative to that for Unit 3 Cycle 9.
Table 5.5.5-1 Key Parameters for IIFP Post-Trip Steam Line Break Parameters Units Unit 2 Unit 3 Unit 2 Cycle 10 Cycle 9 Cycle 9 Initial RCS Flow Rate GPM 376,000 376,000 376,000 Initial Pressurizer Volume ft' 913.9 913.9 914.7 Moderator Temperature 10 '*Ap/*F -3,7 -3,7 -3.7 1 Coefficient (MTC)
CEA Worth at Trip LOAC 9 cap -7.79 -7.993 -7,991 power / no LOAC power I Inverse Boron Worth (IBW) ppm 9Ap 126.7 121.2 120.1 Initial Steam Generator Pressure psia 967.6 960 960 1 Initial Steam Generator Inventory lbm Iligh level fligh Level liigh Level Alarm Alarm Alarm Pressurizer Pressure Control System Automatic Automatic Automatic Kinetics p - 0.00472 p-0.00468 0 - 0.00467 Table 5.5.5-2 Key Parameters for IIZP Post-Trip Steam Line Break l Parameters Units Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 Initial Tin *F 560 560 560 Initial RCS Flow Rate GPM 376,000 376,000 376,000 Initial Pressurizer Volume ft ' 913.9 913.9 914.7 Moderator Temperature 10 -' ApF -3.7 -3.7 -3.7 Coefficient (MTC) inverse Boron Worth (IBW) ppm @Ap 115.0 121.2 120.1 I initial Steam Generator Pressure psia l123 1124 1124 Initial Steam Generator Ibm liigh level liigh level fligh level Inventory Alarm Alarm Alarm Pressurizer Pressure Control Automatic Automatic Automatic System Kinetics p-0.00472 p-0.00468 p - 0.00467 L
Southern California Edison 174 November 1998
Section 5.5 l Comparison of Safety Analysis Results Table 5.5.5-3 Post-Trip IIZP-LOAC Steam Line Break Results Parameter Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 Minimum Post-trip Macbeth 1.42 1.48 1.49 DNBR Maximum Post-trip 13.325 kw/ft 12.7 kw/ft 12.6 kw/ft LliGR Maximum Post-trip Fission 129.69 MW, 117.04 MW, 116.16 MW, Power (at 251.35 seconds) (at 293.4 seconds) (at 295.6 seconds)
Maximum Post-trip Reactivity 1.78 E-3 Ap l (at 75.58 seconds) 1.75E-3 Ap (at 77.1 seconds) 1.9038E 3 Ap (at 75.55 seconds) l Table 5.5.5-4 Post-Trip IIZP-AC Steam Line Break Analysis Results Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 Minimum Post-trip 7.14 6.36 i Macbeth DNBR 6.36 Maximum Post-trip 15.469 kw/ft 16.18 kw/ft 16.6 kw/ft LIIGR I Maximum Post-trip 202.1 MW, 201.34 MW, 197.73 MW, Fission Power (at 190.82 seconds) (at 191.95 seconds) (at 187.4 seconds)
Maximum Post-trip l Reactivity 1.8237 E-3 Ap (at 59.79 seconds) (at 61.95 seconds) 1.82E-3 Ap 1.7977E-3 Ap (at 60.56 seconds) l I
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v i l i r j Section 5.5 Comparison of Safety Analysis Results ,
1
- . Table 5.5.5-5 l 1 Post-Trip HFP-LOAC Steam Line Break Analysis Results j 1 Unit 2 Unit 3 Unit 2 )
- Cycle 10 Cycle 9 Cycle 9 i Minimum Post-trip Macbeth No Return-to-power No Return-to-power No Return-to-power DNBR '
4
- '" p No Return-to-power No Return-to-power No Return-to-power R
Maximum Post-trip Fission l
I
$ Power l M "* * " P st-trip
-1.546 E-3 Ap -1.8E-3 Ap -1.802E-3 Ap Rai y 4
1 h
4 4
- Table 5.5.5-6 Post-Trip HFP-AC Steam Line Break Analysis Results Unit 2 Cycle 10 Unit 3 Cycle 9 Unit 2 Cycle 9 l Minimum Post-trip Macbeth No Return-to-power - No Return-to-power 1 No Return-to-power DNBR
, Masiniuni Post-trip ##"*'
No Return-to-power No Return-to-power ~E *##
{ Maximum Post-trip Fission ll Power i
"*" E R i
-7.247 E-3 Ap -7.3E-3 Ap -7.197E-3 Ap 3-1 4
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1 Southern California Edison l 176 November 1998 l
Section 5.5 Comparison of Safety Analysis Results 5.5.6 Comparison of Loss of Normal AC Power (LOAC) Analysis Results A comparison of the inputs of this analysis with the previous analysis, Units 2 and 3 Cycle 9 is as follows in Table 5.5.6-1. A comparison of the results of this analysis with I the results of the Units 2 and 3 Cycle 9 analyses is shown in Table 5.5.6-2.
The more adverse results in Unit 2 Cycle 10 are because of two major reasons:
- 1. Imwer initial SG inventory: 130,500 lbm in Unit 2 Cycle 10 vs.150,000 lbm in Unit 2 Cycle 9. The lower initial inventory causes a lower final liquid inventory I because the reactor trips on low RCP speed, not Low Steam Generator level.
2.
I Ifigher CESEC decay heat: the Unit 2 Cycle 10 analysis uses the ANS 1971 decay heat option with uncenainty and selects the most conservative time to switch from the reactor kinetics model to the ANS 1971 decay heat option. The Unit 2 Cycle 9 analysis did not apply uncenainty ; ad the time to switch was not the most conservative time.
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Section 5.5 Comparison of Safety Analysis Results Table 5.5.6-1 Key Parameters For Loss of Normal AC Power Analysis Parameter, Unit Units 2 and 3 Cycle 9 Unit 2 Cycle 10 Initial Core Power, Mwt 3478 3478 Initial Inlet Coolant Temperature. *F 560 560 Initial RCS Flow Rate, gpm 376,200 376.200 Initial Pressurizer Pressure, psia 2300 2300 Initial Steam Generator Pressure, psia 970 970 1
Initial Steam Generator Total Mass (liquid + steam), Ibm 150,000 130,500 l MSSV Setpoint tolerance % -3 -3 MSSV Nominal Serpoint (psia), lowest set valve 1100.0 1100.0 MSSV accumulation % 0 0 MSSV blowdown % 15 15 PSV Setpoint tolerance % + 2.9 +3 PSV Nominal Setpoint, psia (both valves) 2500.0 2500.0 PSV necumulation % of setpoint 0 0 4
PSV blowdown % of serpoint 4.5 4.5 Moderator Temperature Coefficient.10' Ap/"F 0.0 0.0 Fuel Temperature Coefficient Multiplier 0.86 0.85 Mmirnum CEA Worth at Trip, % Ap -6.0 -6.0 Steam Bypass Control System inoperative Inoperative Feedwater ReFulating System Manual Manual Pressunzer Pressure Control System Manual Manual j Pressurizer level Control System Manual Manual Auxiliary I cedwater Flow assuming 1 Pump Only. gpm 500 500 Table 5.5.6-2 Loss of Normal AC Power Analysis Results Units 2 and 3 Unit 2 Cycle 9 Cycle 10 1
Minimum Steam Generator 75,280 39,960 Liquid loventory, Ibm Time of Minimum inventory. sec 538 1140 Southern California Edison 178 November 1998
Section 5.5 Comparison of Safety Analysis Results 5.5.7 Comparison of CEA Ejection Analysis Results Table 5.5.7-1 Post Trip CEA Ejection Analysis Results Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10 1IOT Full POWER Radially Averaged Enthalpy, cal /gm 146.5 147.6 142.4 Centerline Enthalpy, cal /gm 230.8 234.1 220.4 Failed Fuel (DNBR < l.31), % 7.71 g 8.19 7.49 B Failed Fuel (Centerline Melt), % 0.00 0.00 0.00 1IOTZERO POWER Radially Averaged Enthalpy 131.4 129.6 116.9 Centerline Enthalpy 185.7 182.9 164.6 I
The results are comparable to the Q ,le 9 results. The slight differences in energy deposition are due to the increase in the number of fuel pins, decrease in radial peaking I factor, decrease in 3D power peaking (for most cases), and increased ROPM requirement for Cycle 10.
The fuel failure value was affected by the increased ROPM requirement relative to Cycle 9 (reduces fuel failure) and the more conservative partial derivative (increases fuel failure).
I l Table 5.5.7-2 Pre-Trip CEA Ejection Analysis Results Pow er (7r ) Units 2/3 Cycle 9 Max Post.Eiection F,, Unit 2 Cycle 10 Max Post Eiection F, 100 3.35 3.I99 90 3.34 3.098 70 3.72 3.127 I 50 3.64 3.400 25 4.32 4.251 20 4.85 4.709 l 0 5.67 5.999 The slow trip results are similar to those of the previous cycle. Generally, Cycle 10 has more margin, due to the slightly lower average linear heat rate and the elimination of the conservatisms added to the Cycle 9 analysis so that one set of calculations would apply to both units.
k Southern California Edison 179 November 1998
Section 5.5 Comparison of Safety Analysis Results 5.5.8 Comparison of Seized Rotor / Sheared Shaft Analysis Results To provide a one to one comparison, the fuel failure results at i15% ROPM was compared to the fuel failure results of Unit 2 Cycle 9 at i15% ROPM. Table 5.5.8-1 shows this comparison.
As can be seen in Table 5.5.8-1, the fuel failure for Cycle 10 is larger than the Cycle 9 fuel failure by about 3.5%. To investigate the reason behind this difference a comparison of the input to the fuel failure calculation is performed in Table 5.5.8-2. Table 5.5.8-2 shows that the Cycle 9 and Cycle 10 input are almost identical. Thus, the only factor that remains is the pin census used for Cycle 9 and the pin census used for Cycle 10. Since Cycle 10 core is flatter than Cycle 9, just based on the pin census, it is expected that Cycle 10 results in a larger fuel failure. The result seen in Table 5.5.8-1 is thus I acceptable.
Table 5.5.8-1 Seized Rotor / Sheared Shaft Analysis ROPM-115%
l Cycle 9 Cycle 10 Fuel Failure Less than 8.05% i1.34 %
Fr v. DNBR 1.7 1.084 1.7 1.0837 1.6 1.253 1.6 1.2573 1.5 1.425 1.5 1.4287 1 1.4 1.624 1.4 1.6299 I I I 1 1
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Section 5.5 Comparison of Safety Analysis Results 5.5.9 Comparison of Single Part Length Rod Drop Event Results Relative to the ROPM determination, the difference in inputs between the SONGS-3 Cycle 9 analysis and the SONGS-2 Cycle 9 analysis were in the following areas:
- 1) Maximum PLCEA drop worths were lower for the Unit 3 50%, BOC cases, but higher for the Unit 3 20% and 25%, EOC cases; i 2) Axial power distributions, post-drop ASI's were less top-peaked for Unit 3 lp 3) for all but the 20%, EOC case; Pre-drop radial peaking factors (Fr) were higher for Unit 3 in all cases;
- 4) total distortion factors (TDF) at 120 minutes were lower for Unit 3 in all cases; I 5) Post-drop radial peaking factors (Fr) were lower (less adverse) for all Unit 3 cases except the 50% BOC and EOC cases;
- 6) The newly reduced MSSV valve area was included in the Unit 3 Cycle 9 analysis.
The impact of this change was insignificant, given that all [
J.
l 7) An additional 10% penalty was included in the Unit 3 Cycle 9 calculation to account for rod worth difference.
Although some of the differences were more adverse, and some were less adverse, all were relatively small. The net effect was that the ROPM decreased for Unit 3 Cycle 9 compared to that for Unit 2 Cycle 9.
The results of the previous analysis are summarized in the table below. The results in both the present and previous analysis show that no margin is required to be set aside in I COLSS for this event. Generally, the AOPM for Unit 3 Cycle 9 increased over that available for Unit 2 Cycle 9. In all cases, the ROPM calculated for Unit 3 Cycle 9 decreased from that for Unit 2 Cycle 9. As a result, in all but one case the difference 1 between AOPM and ROPM increased for Unit 3 Cycle 9 relative to the same cases for Unit 2 Cycle 9. In the one case where the difference between [ ]
decreased (50%,EOC,560 *F), the decrease was by only one percentage point.
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Section 5.5 Comparison of Safety Analysis Results Table 5.5.9-1, PLR Drop Analysis Results Time in Life Power Tinlet Unit 2 Cycle 9 Unit 3 Cycle 9
(%) (*F) ROPM (%) ROPM(%)
BOC 20 520 253 261 BOC 20 560 263 252 BOC 25 520 239 247 BOC 25 560 233 218 BOC 30 520 226 233 BOC 30 533 212 219 BOC 50 533 123 120 BOC 50 560 131 126 h10C 50 533 130 129 h10C 50 560 133 131 EOC 20 520 165 162 EOC 20 560 175 173 EOC 25 520 164 161 EOC 25 560 172 170 EOC 50 533 130 129 EOC 50 560 133 131 I
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Southern California Edison 182 November 1998
Section 5.5 Comparison of Safety Analysis Results 5.5.10 Comparison of CEA Withdrawal within Deadband Event Results The following Tables 5.5.10-1 thru 3 are comparisons of the input data and [ ]
between the Unit 2 Cycle 9 analysis and that of Unit 3 Cycle 9. The input data for the two cyclea is very similar and the results are similar also.
Table 5.5.101 Key Parameters for CEA Withdrawal within Deadband Analysis PARAMETTIR Unit 2 Cycle 9 Unit 3 Cycle 9
]
DEADBAND (inch)
CEAC OP 10.0 10.0 CEACs INOP 25.0 25.0 MTC(10-4 Ap/'F) IlOC MOC BOC MOC 90% RTP -0.22 -0.5 -0.22 -0.5 70% Rl? -0.06 -0.13 -0.06 -0.13 50% RTP +0.1 N/A +0.1 N/A 25% RTP,560 F +.3 .4 +.3 .4
,520 F +. I 4 4
+.14 .4 20% RTP,560 F + .34 .4 +.34 .4
,520 F +.1 .4 +.1 .4 Reactivity Insertion (W ap) f40C EOC BOC (CEAC OP) IE 90% RTP N/A 0.030 N/A 0.030 70% RTP 0.020 N/A 0.020 50% RTP, CIS N/A 0.020 N/A 0.020 N/A
, COOS 0.018 N/A 0.018 N/A l
Reactivity Insertion (% Ap) BOC FOC BOC EOC (CEAC INOP) 90% R1? N/A 0.056 N/A 0.056 70% RTP N/A 0.052 N/A 0.052 50% R1P 0.031 N/A 0.031 N/A 25% RTP 0.032 0.047 0.032 0.047 20% R'IP,560*F 0.047 0.062 0.044 0.%2
, 520T 0.047 0.048 TDF 15 minutes llQC LQC BOC E(X' (CEAC OP) 90% RTP N/A 1.028 N/A 1.029 70% R1P 1.040 N/A 1.037 N/A 50% RTP, CIS l.043 N/A 1.040 N/A
, COOS 1.017 N/A 1.019 N/A TDF 15 rninutes ILOC LQC ]LQC EQC (CEAC INOP) 90% RTP N/A 1.055 N/A 1.057 70% RTP N/A 1.059 N/A 1.060 50% RTP 1.034 N/A 1.040 N/A 25% RTP 1.044 1.077 1.049 1.077 20% R17 1.056 1.096 1.062 1.097 Southern California lifison 183 November 1998
Section 5.5 Comparison of Safety Analysis Results Table 5.5.10-2 CEA Withdrawal within Deadband Analysis: Power > 25%
l Unit 2 Cycle 9 Unit 3 Cycle 9 Reason for significant difference i
I 1 ROPM - 15 minutes lI (CEAC OP) 90% RTP 116 % 116 % NA 70% R1P 122 % 121% NA 50% RTP, CIS I43 % 142 % NA I . COOS 133 % 133 % NA ROPM - 15 minutes (CEAC INOP) 90% R17 124 % 124 % NA 70% RTP 136 % 137 % NA 50% RTP l47% I49 % NA l
Table 5.5.10-3 CEA Withdrawal within Deadband Analysis: Power s 25%
Unit 2 Cycle 9 Unit 3 Cycle 9 Reason for significant difference [
1 ROPM - 15 minutes (CEAC INOP, BOC) 25% RTP,560 F 241% 243 % NA I 520 F 20% RTP,560 F 226 %
334 %
227 %
329 %
NA Ap better in U3C9 520 F 222 % 225 % TDF, Ap worse in U3C9 ROPM 15 minutes I (CEAC INOP, EOC) 25% RTP,560 F 139 % 139 % NA
. 520 F I414 141% NA I 209 R1P.560 F
. 520 F 1549 1579 155 %
1579 NA NA I
Southern California Edison 184 November 1998
Section 5.5 Comparison of Safety Analysis Results 5.5.11 Increased Main Steam Flow with LOAC Analysis As shown in Table 5.5.11-1, the Cycle 10 transient results are similar to those obtained in Cycle 9. Due to the flatter pin census for Cycle 10, the corresponding amount of fuel failure has increased. In order to reduce the amount of fuel failure for Cycle 10,[
1 Ts.ble 5.5.11-1 Increased Main Steam Flow with LOAC Analysis Results Fast Power Ascension Case Slow Power Ascension Case Parameter Unit 2 Unit 3 Unit 2 Unit 2 Unit 3 Unit 2 Cycle 9 Cycle 9 Cycle 10 Cycle 9 Cycle 9 Cycle 10 Limiting case CEA drop time (to 3.0 3.0 3.0 3.4 3.4 3.4 90% insened)
ROPM (%) 116 116 116 N/A N/A N/A Time of minimum DNBR (seconds) 10.80 10.00 10.60 27.60 27.60 27.70 Initial F, 1.70 1.7204 1.601 1.70 1.7204 1.601 Minimum DNBR I.3147 1.3041 1.2961 1.0878 1.0937 1.0927 Fuel failure, l None 0.2% 0.76 % 18.4 % 15.3 % 30.66 %
1 Fuel failure l N/A N/A N/A N/A N/A 18.4 %
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l Southern California Edison 185 November 1998
Section 5.5 Comparison of Safety Analysis Results 5.5.12 Comparison of CEA Misalignment Analysis Results The reactor power reduction requirements from this analysis are compared to the current LCS power reduction curves and the results from the previous Unit 1 cycle 9 analysis.
Table 5.5.12-1 CEA Misalignment Power Reduction Analysis Results TIME AFTER SONGS-3 CURRENT SONGS-2 PROPOSE M MISAIJGNMENT CYCLE 9 LCS CYCLE 9 DifS G
(Minutes) for U3C9 0 0 0 0 0 15 0 0 0 0 Fulllength Bank 6 60 0.99 7 1.4 5 ;
120 6.10 13 6.6 10 )
0 0 0 0 0 i
Full length 15 0 0 0 0 '
Non-Bank 6 60 6.16 10 5.1 10 l 120 13.53 18 11.9 15 0 0 0 0 0 i
Part length initially 15 0 0 0 0 i r 112.5 inches Withdrawn 60 0 0 0 0 i 120 0 5 0 0 1
0 0 0 0 0 l Part length Initially 15 0 0 0 0 1 < 112.5 inches Withdrawn 60 0 0 0 2 120 0 5 0 5 The difference between the previous results and the current results is due primarily to the use of different radial distortion factors (for all sets of conditions) and different initial COLSS ROPM values (at 90% and 70% power).
l The values of the lowest power levels below which no further power reduction is i necessary from this analysis are compared to the LCS and the results from the Unit 2 Cycle 9 analysis.
Southern California Edison 186 November 1998 i
Section 5.5 Comparison of Safety Analysis Results Table 5.5.12-2 B Lowest Power Levels with No Power Reduction Required TYPE OF CEA I MISALIGNMENT SONGS-3 CYCLE 9 CURRENT LCS SONGS-2 CYCLE 9 PROPOSED LCS for U3C9 Full 12ngth Bank 6 67.06 % 62% 66.2 % 50%
Fulllength Non-Bank 6 59.68 % 58% 59.3% $0%
Part length Initially 2 112.5 Not None Not Not Inches Withdrawn Required Given Required Required Part 12ngth Initially < 112.5 50% 50% 50% 50%
1 Inches Withdrawn I The difference between the previous results and the current results is due to the use of different radial distortion factors (for all sets of conditions) and different initial COLSS ROPM values (at 90% and 70% power).
5.5.13 Comparison of CEA Withdrawal Analysis Results Comparison of results between the Unit 2 Cycle 10 analysis results and the Units 2 and 3 l Cycle 9 results are shown in Tables 5.5.13-1 through 5.5.13-4 present the comparisons for DNBR, Linear lieat Generation Rate and/or Fuel Centerline Melt Temperature, and peak RCS Pressure.
The differences between the Unit 2 Cycle 9/ Unit 3 Cycle 9 and Unit 2 Cycle 10 CEA Withdrawal results (Fuel Centerline Melt Temperature) from suberitical power and HZP I are primarily due to Fqs. The Fqs for Unit 2 Cycle 10 are lower than the Fqs for Unit 2 Cycle 9 and Unit 3 Cycle 9 and the enthalpy for Fuel Centerline Melt Temperature is directly proportional to the Fq. Lower Fqs lead to lower Fuel Centerline Melt I Temperature. These Unit 2 Cycle 10 results are in line with the expectations.
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Southern California Edison 187 November 1998
Section 5.5 Comparison of Safety Analysis Results Table 5.5.13-1 CEAW from Subcritical Power Analysis Results Parameter 15.4.1.1 Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 1U Acceptance Criteria Departure from 21.31 2 1.31 2 1.31 2 1.31 Nucleate Boiling Ratio Fuel Centerline s 4706 3510 3550 2920 Melt Temperature
( F)
Peak RCS s 2750 Bounded by CEAW at HZP Pressure (psla) l Table 5.5.13-2 CEAW from IIZP Analysis Results I Pararneter 15.4.1.1 Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10 Acceptance Criteria Departure from 2 1.31 2 1.31 2 1.31 2 1.31 Nucleate Boiling Ratio Fuel Centerline Melt s 4706 3640 3640 3250 Temperature (* F)
Peak RCS Pressure s 2750 2655 2696 2635 (psia) l Table 5.5.13-3 CEAW from 50% Power Analysis Results I Parameter 15.4.1.1 Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10 Acceptance Criteria I Departure from a 1.31 N/A N/A 2 1.31
^
Nucleate Iloiling Ratio Peak Linear lleat s 21 N/A N/A i1.97
[ Rate ( kW/ft)
Peak RCS Pressure s 2750 Bounded by CEAW at IIZP (psia)
Southern California Edison 188 November 1998
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Section 5.5 Comparison of Safety Analysis Results Table 5.5.13-4 CEAW from IIFP Analysis Results Parameter 15.4.1.1 Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10 Acceptance Criteria Departure from 2 1.31 N/A N/A 2 1.31 Nucleate Boiling Ratio Peak Linear Heat s 21 N/A N/A 17.4 t Rate ( kW/ft)
Peak RCS Pressure s 2750 Bounded by CEAW at IIZP (psia) 5.5.14 Comparison of Total Loss of Flow Analysis Results The comparison of SONGS-3 Cycle 9 Loss of Flow analysis results to SONGS-2 Cycle 9 is shown in the Table 5.5.14-1. The Unit 3 Cycle 9 Less of Flow analysis results are essentially unchanged from those in Unit 2 Cycle 9. Fuel management has a small impact on this analysis.
Table 5.5.14-1 Loss of Flow Analysis Results PARAMETERS Unit 2 Cycle 9 Unit 3 Cycle 9 ROPM for POWER 5 70% 118.5% 118.5 %1 l RUIT for POWER 2 70%
l ASI SCRAM MTC TIME (sec)
+ 0.3 3.0 0.0 0.831 0.831 3.2 0.0 0.818 0.819 3.4 0.0 0.808 0.808
-0.3 3.0 0.5 0.856 0.854 3.2 0.5 0.842 0.841 3.4 0.5 0.829 0.828
( ) - CPC low Pump Speed Trip Setpoint 0.95 0.95 Southern California Edison 189 November 1998
Section 5.6 Comparison of COLSS and CPC Analysis Results 5.6 COMPARISON OF COLSS AND CPC ANALYSIS RESULTS Comparisons of COLSS and CPC analyses results are presented as indicated in the following table.
Cycle Section Comparison Title Dependent Cycle Cycle (s)
Calculation Analyzed Compared (Y/N) 5.6.1 COLSS and CPC Flow N Unit 2 Unit 2 Uncenainties Cycle 10 Cycle 9 5.6.2 COLSS Power Uncenainties N Unit 2 Unit 2 Cycle 10 Cycle 9 5.6.3 CPC Calibration Allowances N Unit 2 Unit 2 Cycle 10 Cycle 9 5.6.4 COLSS and CPC Digital Serpoint Y Unit 2 Unit 2 Core Simulation Analysis Cycle 10 Cycle 9 5.6.5 CPC Overall Uncenainty Analysis Y Unit 3 Unit 2 Cycle 9 Cycle 9 5.6.6 COLSS Overall Uncertainty Y Unit 3 Unit 2 Analysis Cycle 9 Cycle 9 Section 5.6.1 presents a comparison of the CPC and COLSS How rate uncertainties. The CPC and COLSS flow rate uncertainties are determined by the [
] are used to determine the uncertainties. The CPC mass How rate uncertainty and the COLSS volumetric flow rate uncenainty are used as input to the CPC and COLSS overall l uncenainty analyses, respectively.
Section 5.6.2 presents a comparison of COLSS primary delta-T power uncertainty l (Section 5.6.2.1) and secondary calorimetric power uncenainty (Section 5.6.2.2). These uncenainties are determined by [
l
[ ].
Section 5.6.3 presents a comparison of CPC calibration allowances. The [
, Southern California Edison 190 November 1998
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Section 5.6 Comparison of COLSS and CPC Analysis Results i i
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i Section 5.6.4 presents a comparison of the CPC and COLSS Digital Setpoints Core l
Simulation Analyses results. The purpose of this analysis is to generate simulations of l core response that adequately cover the operating conditions that may be sensed by CPC l orCOLSS. Approximately[
]
]. Results from neutronics and thermal-hydraulics simulations saved as files are used in the CPC and COLSS overall uncertainty analyses. The comparison section does not compare the resulting files, but does compare other output data which is needed for startup test simulations [ ] or CPC and COLSS overall uncenainty analyses input [ ].
Section 5.6.5 presents a comparison of CPC overall uncertainty analysis results. The purpose of the CPC overall uncertainty analysis is to determine the [
], and the DNBR and LPD penalty factors (BERR0 through BERR4). These uncertainties are determined by the [
] are used to determine the uncenainties. The [
] are used in the CPC addressable constants analysis. The CISAM acceptance criteria is used to verify the validity of the CISAM during the initial power ascension following refueling. The core average ASI uncertainty is used in the COOS DNBR limit line analysis.
Section 5.6.6 presents a comparison of COLSS overall uncenainty analysis results. The purpose of the COLSS overall uncertainty analysis is to determine the core average ASI uncertainty, and the DNB and LHR power operating limit (POL) penalty factor (EPOL2, EPOL4, UNCERT). These uncertainties are determined by the [
] are used to determine the uncertainties.
The DNBR and LHR penalty factors (EPOL2/4 and UNCERT, respectively) are used in the COLSS addressable constants analysis. The core average ASI uncenainty is used to ;
establish the COLSS in-service ASI limits in the Licensee Controlled Specifications l (LCS).
l Southern California Edison 191 November 1998
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i Section 5.6 Comparison of COLSS and CPC Analysis Results 5.6.1 Comparison of COLSS and CPC Flow Uncertainties Analysis Results '
A comparison of the Unit 2 Cycle 10 results with Unit 2 Cycle 9 is shown in Table 5.6.1-
- 1. Table 5.6.1-1 shows the rounded- up final results. It is seen that the final values selected for this cycle are consistent with the Unit 2 Cycle 9. Note that the impact of the
.1% difference in the COLSS volumetric flow uncertainty and CPC mass flow uncertainty ,
from their respective AOR values is very small due to the [ ] of these l values in the COLSS/ CPC OUAs.
The examination of the differences between Unit 2 Cycle 10 and Unit 2 Cycle 9, Table 5.6.1-1, is based on [ ] (computer code) results. Table 5.6.1-1 shows that volumetric flow uncertainty is larger for this analysis due to the larger COLSS instrument errors used in this analysis. However, this increase in volumetric flow I
uncertainty is not observed in the mass flow uncertainties due to the additional l
differences, such as density, rh.ch only impacted the mass flow uncettainties. These i differences are 1) different reference mass flow,2) different reference pressure used in )
Unit 2 Cycle 9.
)
Table 5.6.1-1 COLSS and CPC Flow Uncertainties Analysis Results*
Parameter Unit 2 Cycle 9 Unit 2 Cycle 10 COLSS volumetric flow uncertainty. % uniform 4.6 4.7 COLSS mass flow uncertainty % uniform 4.8 4.7 CPC mass flow uncertainty % uniform 5.4 5.3 COL 5iS one-sided mass flow uncertainty. % uniform 4.5 4.5
- Note: The reference flow uncertainty calculated for Units 2 and 3 Cycle 10 is 3.99%
(t 5.0%) of design flow. Ilowever, in calculating the COLSS and CPCS flow uncenainties a bounding Reference Flow Uncenainty of 4.95% of design flow was used.This was done to preserve ample margin for potential changes in the input and to be consistent with the analysis of record (AOR).
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Southem Califomia Edison 192 November 1998 l
Section 5.6 Comparison of COLSS and CPC Analysis Results 5.6.2 Comparison of COLSS Power Uncertainties Analysis Results 5.6.2.1 Comparison of COLSS Primary Delta-T Power Uncertainties Figure 5.6.2.1-1 is a comparison of the Unit 2 Cycle 10 Primary Delta-T Power Uncertainties, the Unit 2/3 Cycle 9 maximum Delta-T Power Errors, and the Unit 2/3 Bounding Delta-T Power Errors.
The Unit 2 Cycle 10 Primary Delta-T Power Uncertainties were determined using the adjusted COLSS Instrument Channel E Tors, and therefore contain margin I with respect to the actual COLSS Instrument Channel Errors. This is the major component of the increase in the COLSS Delta-T Power Uncertainty. The other small source of change is that the actual [ ] on reactor coolant pump differential pressure did increase.
l Figure 5.6.2.1-1 COLSS [ ] Analysis Results l
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Southern California Iklison 193 November 1998 i.i . . . .m. _ . .i. ,_
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Section 5.6 Comparison of COLSS and CPC Analysis Results l
5.6.2.2 Comparison of COLSS Secondary Calorimetric Uncertainty I t
A direct comparison of the results is not possible for the following reasons:
(1) Differing methodologies regarding the use of the feedwater venturi i i
uncenainties. [
] used a specific value for each loop.
(2) Differing types of uncenainty distributions. [
I ] uncenainties.
i (3) Differing power levels. f I However, [ ]
uncenainties and a comparison at "similar" power levels is possible.
j Table 5.6.2.2-1 presents these comparisons.
1 Table 5.6.2.2-1 i COLSS Secondary Calorimetric Uncertainty Analysis Results l POWER LINCERTAINTY NOMINAL Units 2 Cycle 9 Unit 2 Cycle 10 ( l [ 1 20 N/A 19.926 N/A 7.408 20.502 N/A 7.5319 6.496 25 N/A 25.172 N/A 7.549 50 N/A 50.173 N/A 3.075 51.7 N/A 3.5197 3.036 75 N/A 75.133 N/A 1.938 80 81.18 N/A 2.1824 1.882 The trends of increasing secondary calorimetric power uncertainty as power level decreases is observed for Unit 2/3 Cycle 9 and Unit 2 Cycle 10 data. This is due to the fixed feedwater venturi AP uncertainty becoming a larger fraction of the feedwater venturi AP indication at lower power levels. Comparison of the Unit 2/3 Cycle 9 and Unit 2 Cycle 10 data indicates that the values of uncenainty as a function of power level are similar. The Unit 2 Cycle 10 uncenainty at 20%
I power is less than the uncenainty at 25% power because at 20% power the feedwater venturi Ap uncenainty was reduced to [
].
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Southem California Iklison 194 November 1998
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Section 5.6 Comparison of COLSS and CPC Analysis Results 5.6.3 Comparison of CPC Calibration Allowances Analysis Results In Table 5.6.3-1, the Unit 2 Cycle 9, Unit 3 Cycle 9, and Unit 2 Cycle 10 results are compared. The CPC power calibration allowances are [
] calculated in the CPC Overall Uncenainty Analysis.
These [ ] are required to compensate for the potential non-conservatism in CPC calculated power following calibration of CPC power to COLSS power (primary or secondary calorimetric power) at low power, as required by Technical Specifications. The [
]. The values of the [
]in Table 5.6.3-1 are a function of the values of[
],
respectively. The difference [
].
The difference between the previous results and the current results is due to the increase in COLSS primary calorimetric power uncenainty. The increase in COLSS secondary
, calorimetric power uncertainty does not affect the results because the limiting case occurs for calibration at 20% power where CPC power is calibrated to COLSS primary calorimetric power.
Table 5.6.3.1-1 CPC Power Calibration Allowances Analysis Results Unit 2 Cycle 9 Unit 3 Cycle 9 Unit 2 Cycle 10
[ ] 0.1173 0.1338 0.1088
[ 0.0 0.0 2.0 1
1 ] 0.I185 0.2014 0.1097 l 1 2.05 2.05 2.05 l
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- Southern California Edison 195 November 1998
l Section 5.6 Comparison of COLSS and CPC Analysis Results 5.6.4 Comparison of COLSS and CPC Digital Setpoints Core Simulation Analysis Results 5.6.4.1 Comparison of Simulated Rod Shadowing Factors [ ]
Analysis Results Table 5.6.4.1-1 l Simulated [ ] RSFs at 100% Power Analysis Results CEA Configuration BOCS BOCL lead 1.07000/ 1.07245 1.06924/ 1.07222 12ad + 2nd 1.11479/ 1.12034 1.11323 / 1.1203i PLR 1.03377/ 1.03530 1.03440/ 1.03676 lead + P 1.1 % 78/ 1.11125 1.10653 / 1.11240 Lead + 2nd + P 1.15474/ 1.16261 1.15345/ 1.16392 NOTE: Comparison is displayed as: (Unit 2 Cycle 9 value) / (Unit 2 Cycle 10 value)
The comparison of the simulated RSF for Unit 2 Cycle 9 and Unit 2 Cycle 10 indicates that the RSF are relatively insensitive to small changes in the fuel I loading pattern.
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2 s
Southern California Edison 196 November 1998
Section 5.6 Comparison of COLSS and CPC Analysis Results 5.6.4.2 Comparison of Measured Fxy Simulations Analysis Results Table 5.6.4.2-1 Simulated Measured Fxy Analysis Results CEA Configuration BOCS HOCL MOCL EOCL ARO 1.4853/1.4217 1.5253 / 1.4464 1.4274/ 1.3857 1.3675 / 1.3235 Irad I.6071/1.5141 1.6481/I.5485 1.4814/ I.4617 1.4258/ 1.4289 Irad + 2nd 1.5173/1.4728 1.5657 / 1.5507 1.5044/ 1.4472 1.4960/1.4768 PLR 1.4942/ 1.4020 1.5307/ 1.4303 1.4231/ 1.3799 1.3873/1.3615 lead + P 1.6114/1.5041 1.6476 / 1.5480 1.4757/ 1.4493 1.4488/ 1.428}
lead + 2nd + P 1.5361/ 1.5122 1.5803/1.5871 1.5442/ 1.4785 1.5246/1.5076 NOTE: Cornparison is displayed as: (Unit 2 Cycle 9) / (Unit 2 Cycle 10)
The increased use of erbia and higher average enrichment of the Cycle 9 fuel I returning to Cycle 10 flattens the power distribution and reduces the Fxy values for Unit 2 Cycle 10 compared to Unit 2 Cycle 9. However, the general trends of the simulated Fxy results for Unit 2 Cycle 10 as a function of CEA insertion and I burnup is consistent with the trends for Unit 2 Cycle 9. The trends are similar to those generated in the Unit 2 Cycle 10 physics models and depletions calculation.
An exception to the trends is noted for the " lead + 2nd" CEA configuration. For Unit 2 Cycle 9, Fxy decreased more significantly (SBC/LBC) or increased more significantly (MOC/) in going from " lead" to " lead + 2nd," compared to Unit 2 l Cycle 10. Also, for Unit 2 Cycle 9, for the " lead + 2nd" CEA configuration, Fxy remained constant with burnup from MOC to EOC, whereas for Unit 2 Cycle 10 Fxyincreased slightly in going from MOC to EOC. These exceptions to the general trends are attributed to the use of erbia in all fuel assemblies for Unit 2 Cycle 10 vs. the use of erbia in approximately one-half of the fuel assemblies in Unit 2 Cycle 9 (fresh assemblies only). This changes the radial power distribution characteristics with CEA insenion and burnup.
Southern California Edison 197 November 1998 J
Section 5.6 Comparison of COLSS and CPC Analysis Results 5.6.4.3 Comparison of PLR Follower Analysis Results The Digital Setpoints Core Simulation Analysis calculates Radial Peaking Factors (Fxy) for variety of CEA configurations for use in downstream setpoints calculations. These CEA configurations include configurations in which the Part length Regulating (PLR) Banks have been inserted into the core. When a PLR Bank is present in the core, an Fxy value is calculated for the active portion of the PLR and for the PLR follower region. The PLR consists of a lower solid inconel region, a middle hollow inconel region, and an upper solid B,C region. A I comparison or survey of the Fxy values is made between the Fxy in the PLR follower region and the Fxy assuming the PLR follower in not inserted. If the difference in Fxy is greater than [
] in downstream analyses.
The comparison of the PLR Follower survey results for Unit 2 Cycle 9 and Unit 2 Cycle 10 indicates that all Fxys [ ] except for the PLR Follower configuration at BOCS. This is attributed to the difference in erbia loading in fuel assemblies under PLR locations in Unit 2 Cycle 9 (4.60 I w/o U2" enrichment,2.1% erbia in 72 pins) vs. Unit 2 Cycle 10 (4.05 w/o U2 n enrichment,2.1% erbia in 60 pins).
I Table 5.6.4.3-1 l PLR Follower Analysis Results CEA Configuration BOCS BOCL MOCL EOCL F 0.2I/ 1.77 0.44/-0.93 -0.07 /-0.I4 0.14/ 0.I6 lead + F 0.41/-0.10 0.39/ 0.35 -0.63 /-0.05 <0.009 / 0.23 lead + 2nd + F -0.08/ 0.62 -0.03 / 0.93 0.42 / 0.38 0.65 / 0.07 NOTE: Comparison is displayed as: (Unit 2 Cycle 9) / (Unit 2 Cycle 10)
I I
L
~
Southern California Edison 198 November 1998 s
Section 5.6 Comparison of COLSS and CPC Analysis Results 5.6.5 Comparison of CPC Overall Uncertainty Analysis / Addressable Constants Analysis Results Comparisons with the Unit 2 Cycle 9 analysis are presented in Tables 5.6.5-1 and 5.6.5-
- 2. Table 5.6.5-1 compares the [ ] values and Table 5.6.5-2 compares the [ ] criteria. The differences seen in these tables are comparable to the differences in the key inputs and the differences resulting from the different values of[
] that were used in Unit 2 Cycle 9 [
J.
Table 5.6.5-1 Addressable Constants Analysis Results Power Parameter Unit 3 Cycle 9 Unit 2 Cycig9 Unit 3 Cygle 9 Unit 2 Cygle 9
@) [ ]' I l I 1 [ ]
80 [ ] 0.8694 0.8231 0.8656 0.8231
[ ] 1.0612 1.0628 1.1001 1.1026 I [ ] 0.8694 0.8231 0.8656 0.8231 I J 1.1487 1.1347 1.2438 1.2285
[ ] 4.4242 4.4138 4.5906 4,7003
[ ] -0.0599 -0.0578 -0,0353 -0.0394
[ ] -0.0169 -0.0127 -0.035 -0.0297
[ 0.0585 0.0594 0.0707 I ]
0.0692 20 I 1 1.1933 1.2017 1.2148 1.2164 Table 5.6.5-2 i [ ] Criteria Analysis Results Type of Criteria Unit 3 Cycle 9 Unit 2 Cycle 9 l
Review 6.36 6.70 Action 8.32 8.79 l
" CisAM (Reference 46) and fast power ascendon (Reference 43) are two types of SAM startup tests analyzed to gen Southem Califomia Edison 199 November 1998
Section 5.6 Comparison of COLSS and CPC Analysis Results 5.6.6 Comparison of COLSS Overall Uncertainty Analysis Results A comparison of the SONGS-3 Cycle 9 results and the results from the SONGS-2 Cycle l 9 COLSS Overall Uncertainty Analysis are presented in Table 5.6.6-1. For interest, the I results from the SONGS-3 Cycle 8 COLSS Overall Uncertainty Analysis are also shown.
I Table 5.6.6-1 COLSS Penalty Factor Analysis Results Parameter OCS BOCL MOC EOC C
Unit 3 Cycle 9 1.% 976 1.06669 1.13187 1.10871 N EE Unit 2 Cycle 9 1.07200 1.07214 1.12728 1.10858 Unit 3 Cycle 8 1.06990 1.08207 1.10719 1.11622 Unit 3 Cycle 9 0.01756 0.00788 0.03265 0.03659 EPOL2 Unit 2 Cycle 9 0.02190 0.01971 0.03372 0.03295 Unit 3 Cycle 8 0.00090 -0.00593 0.01009 0.01844 in reviewing the differences between Unit 2 Cycle 9 and Unit 3 Cycle 9 results the following observations are made. The time-in-life at which the maximum EPOL2/4 g penalty occurs is different. For Unit 2 Cycle 9 the maximum value occurred at MOC.
l For Unit 3 Cycle 9 the maximum EPOL value occurred at EOC. The numerical values of EPOL are similar between Unit 2 and Unit 3.
I
{
Southern California Edison 200 November 1998
F Section 5.7 Comparison of Asbuilt Physics Startup Test Predictions 5.7 Comparison of As-Built Physics Startup Test Predictions As discussed in Section 5.0, the startup test predictions are compared to the ultimate data, the measured data. The Unit 3 Cycle 9 startup test prediction was generated by SCE using ROCS /MC code (References 1 and 41) and SIMULATE code (Reference 26) as appropriate.
Southern California Edison 201 November 1998
L I MM Section 5.7 Comparison of Asbuilt Physics Stanup Test Predictions Table 5.7-1 Summary of Physics Test Results Measurement Predicted Value Measured Value Prediction - Measured Acceptance Criteria CBC(Near ARO Config) Bk6 @ 50" Withdrawn (ppm) 2100 2135 -35 150 ITC (Near ARO Config) Bk6 @ 50" Withdrawn (% Ak/k/F) 0.158E-4 0.124E-4 0.034E-4 10.1 CEA Group Worth Regulating Bank 6 (% Ak/k) N/A N/A N/A (By dilution) N/A Regulating Bank 5 (%Ak/k) 0.312 0.295 0.017 0.1 Regulating Bank 4 (% Ak/k) 0.566 0.539 0.027 0.1 Regulating Bank 3 (%Ak/k) 0.872 0.861 0.011 0.1 Regulating Bank 2 (G Akik) 0.409 0.401 0.008 0.1 TOTAL (%Akik) 2.159 2.096 0.063 10%-
Regulating Groups 6-2 Inserted CBC (ppm) 1803 1845 -42 1 50 CEA Exchange Worth 0.546 Bank I (% Ak/k) 0.543 0.003 i 0.1 Shutdown Bank A (%Ak/k) 1.474 1.432 0.042 1 0.221 Shutdown Bank B (% Ak!k) 1.819 1.836 -0.017 i 0.275 CEA Group Worth 0.469 Regulating Bank 6 (% Ak/k) 0.435 0.034 0.1 (By Boration)
Shutdown Margin (%Ak/k) 2 5.15 6.029 N/A 2 5.15 IBW (pprW%Ak/k) 134.8 134.7 0.1 None Southern Califomia Edison 202 November 1998 l
3 Section 6.0 References
[
6.0 REFERENCES
- 1. CENPD-266-P-A,"The ROCS and DIT Computer Codes for Nuclear Design." April, i 1983. (PROPRIETAP.Y) {
- 2. Fuel Management Guidelines for SONGS Units 2 and 3, Rev. 02, July 21,1998.
- 3. SONGS Updated FSAR for Operating License for Unit 2 and Unit 3: NRC Docket No. t 50-361 and 50.362.
- 4. CENPD-188-P-A,"HERMITE, A Multi-Dimensional Space-Time Kinetics Code for PWR Transients." March 1976. (PROPRIETARY)
- 5. CE-CES-79, REV. 0-9, "QUIX User's Manual." May 1987. (PROPRIETARY) ;
i
- 6. CENPD-225-P-A," Fuel and Poison Rod Bowing." June,1983. (PROPRIETARY)
- 7. CENPD-161-P-A," TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core." April,1986. (PROPRIETARY) <
- 8. CEN-193(B)-P, Supplement 2-P," Partial Response to NRC Questions on CEN-161(B)-P, Improvements to Fuel Evaluation Model," March 21,1982 (PROPRIETARY)
- 9. CENPD-206-P-A, TORC Code Verification and Simplified Modeling Method." June
.1981 (PROPRIETARY)
- 10. CENPD-162-P-A," Critical Heat Flux Correlation for CE Fuel Assemblies with Standard Spacer Grids, Part 1. Uniform Axial Distribution." September,1976. (PROPRIETARY)
I 1. CENPD-207-P-A," Critical Heat Flux Correlation for CE Fuel Assemblies with Standard Spacer Grids, Part 2, Non Uniform Axial Power Distribution." December,1984.
(PROPRIETARY)
- 12. CEN-160-(S), Rev.1-P,"CETOP Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3." September,1981. (PROPRIETARY)
- 13. 12tter from George W. Knighton (NRC) to Kenneth P. Baskin (SCE), " Issuance of Amendment No. 32 to Facility Operating License NPF-10 and Amendment No. 21 to Facility Operating License NPF-15 San Onofre Nuclear Generating Station, Units 2 and 3," March 1,1985 Southern California Edison 203 November 1998
Section 6.0 References
- 14. 12tter from George W. Knighton (NRC) to Kenneth P. Baskin (SCE)," Issuance of Amendment No. 47 to Facility Operating License NPF-10 and Amendment No.36 to Facility Operating License NPF-15 San Onofre Nuclear Generating Station, Units 2 and 3," May 16,1986
- 15. CEN-356(V)-P-A, Rev. 01-P-A, " Modified Statistical Combination of Uncertainties."
u May,1988. (PROPRIETARY)
- 16. Enclosure 1-P to LD-82-054," Statistical Combination of System Parameter Uncertainties I in Thermal Margin Analyses for System 80." Submitted by letter from A. E. Scherer (CE) to D. G. Eisenhut (NRC), May 14,1982. (PROPRIETARY)
I 17. CESSAR SER 2 Section 4.4.6," Statistical Combination of Uncenainties."
18.
l CENPD-139-P-A, "CE Fuel Evaluation Model." July,1974. (PROPRIETARY)
- 19. CEN-161(B)-P-A," Improvements to Fuel Evaluation Model." August,1989.
(PROPRIETARY) 20.
CEN 161(B)-P, Supplement 1-P-A," Improvements to Fuel Evaluation Model," January 1992. (PROPRIETARY)
- 21. CEN-283(S)-P, Parts I, II, and III," Statistical Combination of Uncenainties",1984 I (PROPRIETARY)
- 22. CE-NSPD-151-P. "CE Safety Analysis Methods for Calvert Cliffs Units 1 and 2," June 1981. (PROPRIETARY)
- 23. CENPD-107,"CESEC Digital Simulation of a Combustion Engineering Nuclear Steam I Supply System." April 1974. (PROPRIETARY)
- 24. CENPD- 183," Loss of Flow-CE methods for Loss of Flow Analysis." July 1975.
I (PROPRIETARY) 4 25.
l CENPD-190-A,"CE Method for Control Element Assembly Ejection Analysis." July 1976. (PROPRIETARY) 26.
' SCE-9001-A," Southern Califomia Edison Company PWR Reactor Physics Methodology Using CASMO-3/ SIMULATE-3," September 1992 27.
CENPD-135-P,"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program."
April 1974. (PROPRIETARY)
Southern California Edison 204 November 1998
Section 6.0 References
- 28. CENPD-135, Supplement 2-P,"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program." December 1974. (PROPRIETARY)
- 29. CENPD-135, Supplement 4-P,"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," August 1976. (PROPRIETARY)
- 30. LD-82-001 (dated 1/6/82),"CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," Enclosure 1-P to letter from A. E. Scherer to D. G.
Eisenhut, December 1981. (PROPRIETARY)
- 31. NUREGs 0741/0952, ' Technical Specifications San Onofre Nuclear Generating Station Unit No. 2 and No. 3 Docket 50-361/362, February 1982/ November 1982.
- 32. CEN-305-P, Rev. O l-P, " Functional Design Requirement for a Core Protection Calculator." May,1986. (PROPRIETARY)
- 33. CEN-304-P, Rev. Ol-P, " Functional Design Requirement for a Control Elenient Assembly Calculator." May,1986. (PROPRIETARY)
- 34. CEN-312-P, Rev. Ol-P, " Overview Description of the Core Operating Limit Supervisory System (COLSS)." November,1986. (PROPRIETARY)
- 35. CEN-372-P-A, " Fuel Rod Maximum Allowable Gas Pressure," May 1990 (PROPRIETARY)
- 36. SCE-1-A," Quality Assurance Program,"(Southem California Edison, San Onofre Nuclear Generating Station Topical Report)
- 37. TOEN50," Nuclear Fuel Management Qualification Guide *'
- 38. CEN-386-P-A," Verification of the Acceptability of a 1-Pin Bumup Limit of 60 MWD /kgU for Combustion Engineering 16X16 Fuel," August 1992 (PROPRIETARY) 39.
Letter from J. M. Betancoun (ABB CE) to P. D. Myers (SCE), " Competency Criteria For Performing Reload Analysis and ABB CENO Evaluation of Edison's Independent Capability," July 16,1997
- 40. DBD-SO23-TR-AA, Revision 1. " Accident Analysis Topical Repon," December 31, 1997
, 41. CENPD-275-P-A, "C-E Methodology for Core Designs Containing Gadolina-Urania Bumable Absorbers," May,1988 (PROPRIETARY)
Southern California Edison 205 November 1998
Section 6.0 References
- 42. CENPD-382-P-A," Methodology for Core Designs Containing Erbium Bumable Absorbers," August,1993 (PROPRIETARY)
- 43. CE-NPSD-369," Fast Power Ascension Generic Test Guidelines," August 1986 (PROPRIETARY)
- 44. CE-NPSD-359,"CPCS and COLSS Reload Startup Test Requirements," August,1986 (PROPRIETARY)
- 45. CENPD-153-P Rev,1-P-A," INCA /CECOR Power Peaking Uncertainty," May,1980 1 (PROPRIETARY)
I 46. CE-NPSD-984-P," Cycle Independent Shape Annealing Matrix Methodology," January, 1995
- 47. TR-103586," Guidelines for Optimizing the Engineering Change Process for Nuclear Power Plants." March,1994 48.
CENPD-282-P-A, " Technical Manual for the CENTS Code," Volumes 1 - 3, February, 1991
- 49. Letter, Robert A. Clark (NRC) to A.E. Lundvall, Jr., Safety Evaluation by the Office of Nuclear Reactor Regulation, MacBeth Critical Heat Flux Correlation, Baltimore Gas and Electric Company, Calven Cliffs Units I and 2, Docket Nos. 50-317 and 50-318," June, i 1983 50.
CEN-155-(S)-P,"CE-1 Applicability to San Onofre Units 2 and 3 HID-2 Grids, Response 1 to NRC Questions," March,1991 51.
Generic Letter No. 83-11," Licensee Qualification for Performing Safety Analyses in I Suppon of Licensing Actions," February 8,1983 52.
NUREG-0712, Supplement No. 4 " Safety Evaluation Report related to the Operation of 1 San Onofre Nuclear Generating Station, Units 2 and 3 Docket Nos. 50-361 and 50-362,"
January 1982 1 53. CENPD-135, Supplement 5-P,"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program," April 1977. (PROPRIETARY) 1 I
l 1
Southern California Edison 206 November 1998
APPENDIX A *** FOR INFORMATION ONLY
- SONGS Reload Analysis Related Procedures APPENDIX A: SONGS RELOAD ANALYSIS RELATED PROCEDURES The following table presents a compilation summary of the San Onofre Nuclear Generating Station (SONGS) procedures that initiate, control, limit applicability, implement, administer, and monitor reactor core designs for Units 2 and 3 and associated activities. In addition to the procedure number and title, the listing provides a short summary of the procedure purpose and provides references to any connecting procedures, the Reload Ground Rules (Section 4.3), or a data transmittal controlled by the Core Reload Analyses and Activities Checklist (Section 4.7).
As discussed in Section 4, many of these procedures are used by SCE independent of whether SCE or a fuel vendor performs the reload analyses. Not listed are procedures that directly support those procedures listed or are only peripherally impact the reload analyses. Overall, the list is organized by the division responsible for the procedure. Specifically:
Nuclear Fuel Management Station Technical Engineering (includes Reactor Engineering)
Site Technical Services Nuclear Engineering and Design Organization Chemistry Nuclear Training Licensing Nuclear Oversight Maintenance Operations liowever, these procedures are presented for information only. While the general functions and processes presented in these procedures are needed, specific details, division responsibilities, procedure titles, procedure numbers, bettennent improvements, etc., are subject to change without invalidating this topical report.
, Southern California Edison 207 Novernber 1998 l
l
1 r 3 7--- m m y --
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Number Procedure Title Description of Purpose Connecting Paxwiure A.1 Nuclear Fuel Management Division i sol 23-XXXVI.I NUCLEAR FUEL MANAGEMENT (NFM) To describe various elements of the Nuclear Fuel QUALFIY PROGRAM AfTects all other NFM Management (NFM) Quality Program. Procedures.
2 sol 23-XXXVI-I.I ADMINISTRAT10N OF CONTRACT 3 FOR I. This procedure describes the processes for SOI23-XXIV-1.1 NUCLEAR FUEL AND SPENT FUEL the control of documents and the SERVICES establishment of personnel interfaces required for the administration of contracts for nuclear fuel and spent fuel services.
- 2. This procedure is limited to the administrative activities for the support of contracts entered into between Southern Califomia Edison (SCE) and suppliers of nuclear fuel and spent fuel services for the San Onofre Nuclear Generating Station (SONGS).
3 Sol 23-XXXVI.l.2 NUCLEAR FUEL AND SPENT FUEL 1he purpose of this procedure is to provide a SO123-XXXIV-1.1 SERVICES PROPOSAL SOLICITAT10N system for Nuclear Fuel and Spent Fuel Services AND PREPARAT10N AND AMENDMENT proposal solicitation and the preparation, teview, OF CONTRACT DOCUMENTS and approval of appropriate contracts and amendments.
This procedure pertains to those activities undertaken for the procurement of Nuclear Fuel and Spent Fuel Services. Procurement level 11 shall apply to all Nuclear Fuel Management (NFM) pamu .ei.t activities for these services.
Southern California Edison 203 November 1998
T r - ummm - inum m-APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Indes Procedure Nurnber Procedure Title Description of Purpose Connecting Prwniure 4 sol 23-XXXVI-l.3 SPECIAL NUCLEAR 'tATERI AL his procedure establishes and defines the ACCOUNTABILITY sol 23-SN-1 procedures and policies used to maintain proper SOI-X-7 management and safeguards accounting control SO23-X-7 over the Special Nuclear Materials (SNM) used in SO23-X-7.2 conjunction uith the operation of the San Onofre Nuclear Generation Station, Units I,2 and 3. The procedures and policies are designed to meet the requirements of 10CFR70.51 and NUREG BR0006 and -0007.
Implementation of the procedures and policies for nuclear fuel accountability and refueling actisities are shared between Site Technical Services and Nuclear Fuel Management. Station responsibilities are specified in SO123-SN-1, Special Nuclear Material Accountability Program.
SOI-X-7, Nuclear Fuel Movement SO23-X-7, Nuclear Fuel Movement for Refueling Cycles, and SO23-X-7.2, Nuclear Fuel Movement - Spent Fuel Pool.
5 Sol 23-XXXVI-2.5 PREPARATION OF T1IE PLANT PHYSICS This procedure provides the administrative DATABOOK SO!23-XXIV-5.1 direction for preparing the Plant Physics SO123-XXIV-7.15 Databook during each scheduled refueling outage. SO123-XXXVI-2.2 which consists of physics parameters: critical boron concentration, is#.a.nal temperature coefficient, irm coron worth, doppler power coefnint, fuel temperature coefficient, power coefficient, equilibrium shape index, axial and radial power shapes, rod worths, XE and SM parameters, and kinetics data.
6 SOI 23-XXXVI-2.7 CONTROL OF Tile NUCLEAR FUEL The purpose of this procedure is to provide SO123-XXIV-5.1 M ANAGEMENT(NFM) REDUCED instruction for the configuration control of the INSTRUCT 10N SETCOMPUTER (RISC) NFM RISC 6000 computer network.
6000 COMPUTER NEN'ORK Southern California Edison 209 Novernber 1998
APPENDIX A *** FOR INFORMATION ONLY *** SONGS Reload Analysis Rel:ted Procedures Index Procedure Number Procedure Title Description of Purpose Connecting Procedure 7 SO23-XXXVI 1.4 DOCUMENTATION OF RELOAD FUEL 1. This procedure provides the method for sol 23-XXIV-7.15 CYCLE ANALYSES developrnent, review, approval and sol 23-XXIV-5.I processing of documentation of a Reload sol 23-XXVI-4.2 Fuel Cycle Analysis performed for a reload SO123-XV-5.1 fuel cycle or during a fuel cycle. Typical SOI23-XXIV-l.1 applications of this procedure are to document the preparation of Core Physics Analyses, Transient Analyses, and Digital Setpoint Analyvs.
- 2. The intent of this procedure is to meet all of the pertinent requirements of the Topical Quality Assurance Manual (TQAM),while functionally reproducing the Recorded Calculation (RC) format used by ABB-Combustion Engineering.
8 SO23-XXXVI-2.4 INSTALLA110N OF T11E CEFAST INPlfr The purpose of this procedure is to provide an SOI23-XXIV-5.1 DATA BASE FILES approved process for installing the CEFAST SO23-V-2 database and creating / modifying supporting computer files to the datalwse. Dese files are used to perform power ascension testing at refueling or COLSS/CPC axial shape verification.
9 SO23-XXXVI-2.6 EVALUATION OF REACTOR COOLANT I. De purpose of this procedure is to document S 0 123-111-1.1.23 SYSTEM (RCS) ACT1VilY activities for monitoring and evaluating sol 23-V-1.1 Reactor Coolant System (RCS) activity.
- 2. This procedure sets forth actions to be taken in response to various levels of RCS activity.
- 3. This procedure does not address plant emergency conditions or conditions when the RCS is at reduced inventory.
Southern California Edison 210 November 1998
APPENDIX 'A *** FOR INFORMATION ONLY *** SONGS Reload Analysis Related Pmeedures Index Procedure Number Procedure Title Description of Purpose Connectina Pi.x+,:ure 10 SO23-XXXVI-2.9 EVALUATION OF REACTOR PilYSICS 1. The purpose of this procedure is to document SO23-V-1 BIASES AND UNCERTAINT1ES requirements for the evaluation and re- SO23-V-2 verification of biases and uncertainties SO23-XXXVI-2.8 associated with reactor physics SO23-XXXVI-2.10 methodologies. SO123-XX-1 e
- 2. This procedure requires that the observed differences between the Reactor Physics measurements and the predictions be :
evaluated to ensure continued validity of the existing biases and uncertamties.
II SO23-XXXVI-2.8 EVALUAllON OF CORE PERFORM ANCE Dis procedure provides guidelines for monitoring SO23-V-1.12 and evaluating reactor physics core performance sol 23-XX-1 parameters. The measured reactor physics core SOI23-XXXVI-1.4 performance parameters are compared with design SO123-XIV-5.1 1 depletion predictions to ensure that the core sol 23-XXXVI-2.9 performance is as designed. This procedure sets '
forth the actions to be taken in response to larger than expected deviations in measured versus !
predicted reactor physics core performance parameters. It is not the intent of this procedure to
- perfonn the plant surveillance activities to monitor Technical Specification Limits.
This procedure implements R&euwi-Mation 3 cf the INPO Significant Operating Experience Report (SOER 96-2) reganling the design and operating considerations for reactor cores. This ,
procedure also implements the Nuclear Organization Directive D-030," Reactor Core Design" Section C.2.h, to implement a comprehensive core and fuel performance monitoring and trending program, including validation / benchmarking of physics model ,
capability. '
Southern California Edison 211 November 1998
APPENDIX A
Indes Prmrdure Number Procedure Title Description of Purru e Connecting I'rwnture 12 SO23 XXXVI.2.10 CORE RELOAD ANALYSES AND 1. Identify the analyses and activities involved ACTIVITIES CllECKLIST sol 23-XXX-2.3 in completing a Core Reload Fuel Cycle SO123-XXX-2.4 Analysis for San Onorm Units 2 &3 for a SO123-XXX-5.2 given fuel cycle.
SOI 23-XXXVI-2.2 SO123-XXXVI-2.3
- 2. Provide a method for indicating the SOI23-XXXVI-2.4 completion of the tasks necessary to perform SO123-XXXVI-2.5 a Core Reload Cycle Analysis. SO123-XIV-5.1 SOI23-XXXVI-1.4
- 3. Provide a method for indicating if the SO123-XXXVI-4.2 analyses and activities necessary to perform a sol 23-XXXVI-2.8 Core Reload Fuel Cycle Analysis have any SOI23-XXXVI-2.6 impact on the License, Design Basis, or other SO!23-XXIV-10.9 Site Programs. SOI23-X-7 SO123-X-10 SO123-X-7.2 SO123 V-4.29 SO123-V-4.7 SO123-V-1 SO123-V-1.0.1 SO123-V-2 SOI23-V-13 SO123-V-1.12 sol 23-III-1.1.23 SO123-XXXV-1.1 SO123-XXXV-2.1 sol 23-XV-51 l SO123-XII-7.12 Southern Califomia Edison 212 November 1998
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Prewdure Number Procedure Title Description of Purpor.e Connecting Procedure 13 SO23-XXXVI-4.2 RELOAD GROUND RULES (RGR) 1. His procedure provides a rnethod for CONTROL MET 110DOLOGY sol 23-XXIV-1.1 Refueling Interval Modification on the SO123-XXIV-10.16 Reload Ground Rules. The procedure SO123-XXXVI-3.1 controls the review, modification, approval, and documentation of the Refueling Interval RGR. His is done to ensure the RGR reflects the current design, operating, and licensed configurations of the plant at the
( time of each fuel reload.
- 2. This procedure also provides a method for making Interim Changes to the Refueling Interval RGR. His process ensures the RGR will reflect changes to the current design, operating, or licensing configurations of a unit. This is done any time during a reload period.
- 3. His procedure provides a method for '
individuals to notify the Discipline Manager, l
Nuclear Fuel Er.gineering and Analysis (NFE&A) or the RGR Coordinator of potential errors or changes to the RGR.
Southern California Edison 213 November 1998
.. i i. . .. .
i r1 m o 1 m- m m --
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Arialysis Related Procedures Indes Procedure Number Procedure Title Description of Purpose Connecting Procedure A.2 Technical Division i4 SO23-V-1 LOW POWER PilYSICS TES11NG 1. To sequence the activities during I.ow Power Sol 23-XXVI-2.10 Reactor Physics Testing which determine the Transmittal reactivity parameters of the reactor and to SO23-V-1.01 verify the design calculations following a SO23-V-1.03 refueling.
SO23-V-1.05 SO23-V-1.06
- 2. To satisfy the Technical Specification and SO23-12.2.26 Licensee Controlled Specification (LCS) SO23-3-2.I3 Surveillance Requirements listed in Tables SO23-3-3.29 1.2-1,1.2-2, and 1.2-3.
SO23-13-13 SO23-3-1.1 sol 23-XX-1 15 SO23-V- 1.0.1 CRITICALITY FOLLOWING REFUELING To achieve initial criticality following a refueling SO23-XXXVI-2.10 operation.
Transmittal SO23-V-1 16 SO23-V-l .0.3 ISOTIIERM AL TEMPERATLVE 1. To measure the Isothermal Temperature RGR, I.006 COEFFICIENT MEASUREMENT AT llOT Coefficient of reactivity at hot zero power ZERO POWER SO23-V-1 conditions.
SO23-XXXVI-2.10 Transmittal
- 2. To derive the Moderator Temperature Coefficient from the measured Isothermal Temperature Coefficient.
- 3. To satisfy the following Technical Specification Surveillances:
SR 3.1.4.1 Verify MTC within the upper limit.
SR 3.1.4.2 Verify MTC is within the lower limit specific in the COLR.
I Southem Califomia Edison 214 I
November 1998
' < - - ==== - mums amer-APPENDIX A
MEASUREMENT RGR,1.010 SO23-V-4.7
- 2. To determine the reactor coolant system flow SO23-3-2.13 rate using secondary calorimetric power and primary temperatore indications.
- 3. To proside instructions for adjusting the Core Protection Calculator (CPC), Core Operating Limits Supersisory System (COLSS), and COLSS Backup Computer System (CBCS) flow calculations (if necessary).
18 SO23-V-l .0.5 CONDtOL ELEMENT ASSEMBLY To determine the integral and differential wtnth of SO23-V-1 WORT 11 BY BORAllON/ DILUTION Control Element Assembly (CEA) Groups (identified within the picccdure) 19 SO23-V-1.0.6 CONTROL ELEMENT ASSEMBLY 1. To determine the integral wonh of of the SO23-V-1 WORTil BY EXCII ANGE / BORON Control Element Assembly (CEA) Groups 1 ENDPOINT SO23-V-1.05 A, and B using the CEA Exchange method.
- 2. To detennine the critical boron concentration with Regulating Groups 6-2 at the fully inserted position.
20 SO23-V-1.1 REACTIVITY BALANCE 1. To compam measured Boron concentration of CALCULAT10NS SO23-XX-1 the RCS with the predicted Boron SO23-XXXVI-2.10 concentration in order to detect anomalies in Transmittal core reactivity.
- 2. To satisfy Technical Specification Surveillance SR 3.1.3.1 21 SO23-V-1.5 CALCULATION OF CORE AVERAGE To provide the method for determining core BURNUP burnup.
Southern California Edison 215 November 1998
i r summe - muums imuss e m-APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index IhMure Number Procedure Title Description of Puw Connecting PrxMure 22 SO23-V-1.9 ISOTi!ERM AL'I1EMPER ATURE I To sneasure the Isothermal Temperature COEFFICIENT RGR,1.006 Coefficient (lTC) of reactivity at power. SO23-3-2.13 SO23-V-1.1
- 2. To determine the Moderator Temperature SO23-V-1.I l Coeflicient (M1C) from the ITC data. SO23-V-4.7 SO23-V-13
- 3. To satisfy Technical Specification SO23-XXXVI-2.10 Surveillance SR 3.1.4.2. Transmittal S023-5.17 23 l SO23-V-1.3 FIXED INCORE DETECTOR CIIANNEL I. To satisfy the following Licensee Contmiled RGR,11.015 Ci!ECK Specification Surveillance Requirement
- a. SR 3.3.103.1
- 2. To determine whether the incore detector system is operable.
- 3. To prowide a record ofinoperable incore detectors for COLSS Operability Verification.
- 4. To provide methodology for terr,oving or restoring detectors between surveillances.
24 i SO23-V-1.12 POWER DISTRIBU110N MONITORING To satisfy Technical Specification Surveillance RGR, VIII.047 SR 3.2.2.1 RGR, VIII.048 RGR, VIII.050 SO23-XXXVI-2.10 Transmittal SO23-V-13 SO23-V-4.7 SO23-V-1.26 SO23-XX-1 Southern California Edison 216 November 1998
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Pmcedures index Procedure Number Procedure Title Description of Puw Connecting thedure 25 SO23-V-l .12.2 DETERMINATION OF SilAPE NOTE The determination of the Shape ANNEALING M ATRIX SO23-V-2 Annealing Matris during the initial SO23-V-4.7 power ascension following refueling. SO123-IT-1 SO23-3-2.13
- 1. To proside the instructions for the S 0 23-5-1.7 determination and installation of the Shape Annealing Matrix (SAM) and corresponding Boundary Point Power Conelation Coefficients (BPPCCs) anytime during a fuel cycle following the initial power ascension following refueling.
- 2. To provide a means for implementing changes to the Linear Power Subchannel Input Currents for optimization of plant indications.
26 SO23.V.1.14 DETERMINATION OF CPC AZIMUTilAL To determine the bias (cor'ection) between the TILT BI AS SO23-3-3.6 COLSS calculated Azimuthal Tilt and the manually calculated CPC, Azimuthal Tilt.
27 SO23-V- 1.2 I CORE PERFORM ANCE RECORD 1. To document incore power distributions. SO23-V-l .3 SO23-V-1.12
- 2. To verify that the CECOR determined power distribution is in agreement with predicted values.
- 3. To detect possible fuel mistoadings.
- 4. To satisfy Technical Specification Surveillance SR 3.2.3.3 prior to initial operation above 20% power after reload.
28 SO23-V-1.26 COLSS DATA-TPOWER AND'IURBINE 1. To provide instructions for conservatively RGR, Vill.047 POWER CALIBRAT10NS calibrating raw ATpower (BDELT) and raw RGR, Vlil.048 turbine first stage pressure power (B1 TSP) to RGR, Vill.050 allow their use in the event the calculations of 5023-3-2.2.1 secondary calorirnetric power (BSCAL) fails.
- 2. To determine the zero power basis for COLSS thermal power calculation (BDELT).
Southern California Edison 217 November 1998
1 f 1 rm m m m m --
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Number Procedure Title Description of Purpose Connecting P edure 29 SO23-V-2 POWER ASCENSION TESTING l. To sequence the activities during Power SO23-3-3.29 Ascension Testing necessary to perform SO23-XXXVI-2.10 COLSS and CPC calibrations and to verify the Transmittal design calculations.
SO23-V-I SO23-V-l .03
- 2. To satisfy the following Technical SO23-V-1.3 Specification surseillance requirement- SO23-V-I.12 SO23-V-1.14 a) Using the incore detectors, verify the SO23-V-l .19.1 shape annealing matrix elements to be SO23-V-1.20 used by the CPCs. SO23-V-1.21 SO23-V-l.26 b) Verify measured IW,, obtained using the SO23-V-4.7 Incore Detector System is less than or SO23-V-12.2.26 equal to the value ofI#,, used in the SO23-XX-1 COLSS and CPCs. SO23-3-2.13 5023-3-3.2
- 3. To sequence the following Technical SO23-5-I.7 Specification Surveillance procedures.
a) Independently confirm the validity of the COLSS calculated T, by use of the incore detectors.
30 SO23-V-4.1 CPC SOITWARE CONTROL AND 1. To establish the storage requirements of the DOCUMENTATION CPCCEAC software documentation including software design specifications, listings and users manuals.
- 2. To establish a means for ensuring that the latest revisions to documents are available to users and the documents and software stored on magnetic media are the same revision level.
Southern Califomia Edison 218 November 1998
, . - - - - - mesur-APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Nurnber Procedure Title Description of Puw Connecting Prixn!ure 31 SO23-V-3.2.5 REFUELING-INTERNAL TEMPERATURE 1. To satisfy Technical Specifications RGR, VIII.049 SENSOR CAllBRATION surveillances SR 3.3.1.9, 3.3.11.4. and SO23-V-4.7 3.3.1.12.2. SO23-3-2.21
- 2. To determine the Reactor Coolant System RTD and Thermocouple biases at isothennal conditions.
- 3. To adjust the COLLS ATpower bias term for the new cycle.
32 SO2-V-3.12 (Ilistory ) CONTAINMENT INTEGRATED To determine that the potential leakage from SO3-V-3.12 (Ilistory) RGR, X.024 LEAKAGE RATE TEST Containment at the Design Basis Accident (DBA) SO23-XX-1 pressure is within the limits stated in Tecimical Specification 3.6.1.2.
33 SO23-V-4.7 CONTROL OF CORE PROTECTION 1. To provide a mechanism for the orderly flow CALCULA1DR ADDRESSABLE SO23-XXXVI-2.10 of information from the organization of a Core Transmittal CONSTANTS Protection Calculator (CPC) Addressable SO23-3-2.13 Constants Change Request in Reactor SO23-3-3.25 Engineering, including supervisoty review, to SO23-3-3.2 the Control Room for entry into the CPC system.
- 2. To ensure that the CPC Addressable Constants are serified to be at their latest authorized value following any CPC system maintenance.
- 3. To provide methodology for the logging of the currently authorized values for Addressable Constants.
- 4. To identify the type of documentation necessary to justify changes to CPC Addressable Constants.
- 5. To provide guidance for the calculation of CPC Addressable Constants.
Southern California Edison 219 November 1998
7- T M M M M MM APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures
(
Index Procedure Nurnber Procedure Title Description of Purpose Connecting Pre alure 34 SO23-V-4.29 CONIROt.OFCOLSS ADDRESSABLE 1. To provide a mechanism which adequately CONSTANTS SO23-XXVI-2.10 maintains administrative control over Core Transmittal Operating Limit Supersisory System (COLSS) SO123-V-4.7i Addressable Constants.
- 2. To provide a method for changirsg COLSS addressable constants when changes are not initiated or being performed in accordance with another approved procedure.
35 SO23-V-13 CONTROL OF11tE PLAST PilYSICS 1. To describe the method used in con
- rolling of i
DAT A BOOKS, COLSS ADDRESSABLE SO23-XXVI-2.10 '
the Plant Physics Data Books ed OPS for Transmittal CONSTAN13 AND REACTOR Units 2 and 3. SO23-V-4.29 ENGINEERING DATA TRANSMITTALS SO13-XXVI-2.5
- 2. To provide direction for control of infarmation transmittal to Operations using a Reactor Engineering Data Transmittal.
36 SO23-V-4.7 i SOI'TWARE DEVELOPMENT AND Establishes the program for acquiring, developing, MAINTENANCE maintaining and controlling computer software used to support QA activities.
37 SO23-V-12.13.12 COLSS B ACKLT COMPtTIER PULSE 1. To verify that the Pulse Accumulation Counter RGR, VIII.040 ACCUMULATlON COUNTER TEST Card functions within the specified acceptance criteria.
- 2. To check the signal processing accuracy of the digital Pulse Accumulation Counter (PAC) for Reactor Coolant Pump (RCP) speed input to the COLSS Backup Computer System (CBCS).
38 SO23-V-12.13.28 COLSS BACKUP COMPtTTER ANALOG To check calibrations of, and calibrate if RGR, VIII.040 INPlTF CALIBRATION CllECK/ necessary, the analog input (Chassis FI in Cabinet CALIBRATION L529) hereafter referred to as 529Fl.
39 SO23-V-12.15.1 PMS COMPlTIER ANALOG INPUT To check calibrations of, and calibrate if RGR. VIII.040 CALIBRATlON CIIECK! CALIBRATION necessary, the analog input (Chassis R4, RS, and LINK 3 R6 in Cabinet L-98).
40 SO23-V-12.15.2 PMS COMPlTIER ANALOG INPUT To check calibrations of, and calibrate if RGR, VIII.040 CAllBRAT10N CilECK/ CALIBRATION necessary, the analog input (Chassis R I, R2, and LINK 4 R3 in Cabinet L-981.
Southern California Edison 220 November 1998
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Inder Procedure Number Procedure Title Dewription of Purpose Connecting Prncedure 41 SO23.V- 12.15.3 PMS COMPUTER ANALOG INPUT To check calibrations of, and calibrate if RGR, VIII.040 CALIBRAllON CIIECK/CALIBR ATION necessary. the analog input (Chassis R1, R2, and LINKS R3 in Cabinet L-99).
42 SO23-V-12.15.4 PMS COMPUTER ANALOG INPUT To check calibrations of, and calibrate if RGR, VIII.040 CAllBRATION CilECK/ CALIBRATION necessary, the analog input (Chassis R4, RS, and LINK 6 R6 in Cabinet L-99).
43 SO23-V-12.15.12 PLANTCOMPUTER RCP PULSE To verify the Reactor Coolant Pump (RCP) Pulse RGR, VilLO40 ACCUMULAT10N COUNTER TEST Accumulation Counter Card in Chassis 92FS.
slots 16 through 13 functions within the specified criteria.
44 SO23-V-12.2.26 SURVEILLANCE REQUIREMENT I. To provide instructions for CEA drop time RGR, IX.010 CONIROL ELEMENT ASSEMBLY DROP testing and to revicw data for indications of SO23-V-4.7 TIME TEST (24-MONTil INTERVAL) uncoupled CEA extensions shafts.
SO23-V-13
- 2. To satisfy Units 2 and 3 Technical SO!23-XX-1 Specifications 3.1.5, Surveillances SR 3.1.5.4 5023-3-3.25 and SR 3.1.5.5 prior to the first reactor criticality, after each removal of the reactor head.
Southern California Edison 221 November 1998
, ,. , m ,_ _ _ , . . . .
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Number Procedure Title Description of Pu.m Connecting PrcMure A.3 Site Technical Services Division 45 SO123-X-1.7 SPECIAL NUCLEAR MATERIAL Ris ;woculure establishes and defines the ACCOUNTABILITY sol 23-SN-1 procedtres and policies used to maintain proper management and accounting control over the Special Nuclear Materials (SNM) used in conjunction with the operation of San Onofre Nuclear Generating Station Units 1,2, and 3.
46 SO123-SN-1 Special Nuclear Material Accountability 1. To define organization and tesponsibilities for Program Special Nuclear Materials (SNM).
- 2. To estabhsh and define the policies and specific procedures used to maintain proper management and accounting control over the SNM used in conjunction with the operation of SONGS 1. 2, and 3.
47 SO23-X -6.1 RECEIPT. UNPACKING, & INSPECTION
- 1. To direct the receipt, unpacking, and Sol 23-SN-I OF NEW FUEL ASSEMBLIES AND FUEL inspection, of new fuel assemblies (16 x 16)
ASSEMBLY INSERTS SO 123-X-l .7 l and fuel assembly inserts (control element SO123-XX-1 j assemblies and neutron sources). SO23-X-7 SO23-X-7.2
- 2. To direct the transferring of new fuel assemblies and fuel assembly inserts from the new fuel storage racks to the spent fuel pool via the new fuel elevator.
- 3. To direct the packing of new fuel assemblies (16 x 16) into their handling casks for shipment to the fuel supplier.
48 SO23-X-7 NUCLEAR FUEL MOVEMENT FOR To direct fuel and fuel inserts (CEAs) and other SO23-XXXVI-2.10 REFUELING CYCLES component movements in support of the outage Transmittal refueling.
SO23-XX-1 Southern California Edison 222 November 1998
, 1 a rm_ um APPENDIX A
- FOR INFORMATION ONLY coo SONGS Reload Analysis Related Procedures index PirxWure Number Prrx< dure Title Dw Iption of Purm Connecting Prwaure 49 SO23-X-7.2 NUCLEAR ITEL MOVEMENT - SPENT To direct fuel and fuel inserts (CEAs, sources, FUEL POOL SO23-XXXVI-2.10 etc.), and other component movement in the Spent Transmittal Fuel Pool (SFP). SO23-XX-1 SO23-X-7 NOTE I) Activities in the SFP under this SO23-X-8 pmcedure DO NOTinvolve the movement of fuel or fuelinserts into containtnent using the Transfer System.
- 2) The activities of this procedure DO NOTinvolve CORE ALTERATIONS.
50 SO23-X-8 FUEL BUNDLE VISUAL EXAMINAT10N This procedure provides information and SO23-X-7 IN TliE SPENT FUEL POOL instructions for performing visual inspections of SO13-X-7.2 selected fuel assemblies in the spent fuel pool.
51 SO23-X- 10 BOR AFLEX COUPON SURVEILLANCE 1. To specify the physical and chemical testing TEST 1NG SO23-XXXVI-2.10 program for the Units 2 and 311igh Density Transmittal Spent Fuel Storage Rack Boraflex Coupons. SO123-III-I.I.23 NOTE These tests are used to determine the integrity of the Boraflex panels that are installed in the new rscks.
This procedure includes the coupon testing frequency and the acceptance criteria.
i Southern California Edison 223 November 1998
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APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Indes Procedure Number Procedure Title Description of Puryw Connecting Piwnture A.4 Nuclear Engineering and Design Organization Division 52 SO123-XV-51 iDENT1FYING AND ASSESSING IMPACT l. To provide a method for identifying and TO SITE PROGR AMS AND PROCEDURES assessing impacts to site programs, procedures or instructions for any change in plant design, procedures, configuratiert, or work rnethod.
Alternate methods may be described in applicable procedures prmided those methods ensure adequate tracking and close of impact activities.
- 2. To prescribe a method for tracking and management of identified Site Program impact (SPI) items through the use of the Action Request (AR) piwess. * ,
I 53 Sol 23-XXIV-10.9 DESIGN PROCESS IT OW AND Dis procedure provides an overview of the plant SO23-V-4.71 CON 1ROLS SONGS UN1131. 2&3 design change process for SONGS Units I,2, and SO123-XV-Si 3.
SO!23-XXIV-l.I 54 SO123-XXX V-l.1 GUIDELINES FOR SYS111M DBD De intent of the Design Bases Documentation PREPARAT10N SO123-XXIV-1.1 (DBD) and Reconstitution Program is to ensure sol 23-XXIV-5.1 i
design bases documents accurately reflect the {
SOI 23-XXIV-10.9 i source design documentation, design output sol 23-XXXVI-3.1 documents accurately reflect the design bases, and sol 23-XXV-1.2 the piant configuration is according to the design bases and design output documents, ne DBD Program procedures are contained in Series XXXV of the NES&L Quality Procedures Manual. In general, the division is as follows:
SO123-XXXV-l. System DBD procedures SO123-XXXV-2, Topical DBD procedures sol 23-XXXV-5, Administrative DBD procedures Southern Califomia Edison 224 November 1998
i 1 1 m re e mm-APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Indes Procedure Number Procedure Title Description of Pu. rm- Connecting IYocsiure 55 sol 23-XXXV-21 GUIDELINES FOR TOPICAL DBD 'Ihe intent of the Design Bases Documentation PREPARATION SO23-XXV-I.1 (DBD) and Reconstitution Program is to ensure design bases documents accurately reflect the source design documentation, design output documents accurately reflect the design bases, and the plant configuration is according to the design bases and design output documents. The DDD Program procedures are contained in Series XXXV of the NES&L Quality Procedures Manual. In general, the division is as followt So l 23-XXX V-1. System DBD procedures SOI23-XXXV-2, Topical DBD procedures Sol 23-XXXV-5. Administrative DBD procedures 56 Sol 23-XXIV-1.1 DOCUMENTREVIEW AND APPROVAL Prosides document review and verification CONR OL SO123-XV-51 methodology, and established tne level of SO123-XXIV-10.9 authorization for approval, for engineering design documents generated according to Nuclear Organization procedures.
57 SO123-XV-44 GUIDELINES FOR DETERMINING WIIEN To provide guidance for determining the need for SO23-V-4.7 A 10CFR50.59 SAFETY EVALUAT10N IS a safety evaluation,in accordance with the REQUIRED SO23-V-4.7.1 requirements delineated in 10 CFR 50.59, SO23-V-13 whenever a change to the facility, procedures, sol 23-XXII-5.I tests, or experiments are planned. sol 23-XXII-10.9 sol 23-XXX-2.3 SOI23-XXX-2.4 SO123-XXXVI-3.1 58 SO123-XXIV-5.I ENGINEERING,CONSRUCT10N AND This procedure establishes the program for SO23-XXXVI-1.4 FUEL SERVICES sol'1 TARE QUALIT'Y acquiring, developing, qualifying, maintaining, ASSURANCE and controlling Engineering, Construction and Fuel Services (EC&FS) computer software used for Quality Affecting activities.
Southern California Edison 225 November 1998
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APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures index Prr,cedure Number Procedure Title Description of Puryc,se Connecting Pir.cuture A.5 Chemistry Division 59 S 0123-111-1.12 REACTOR COOLANT ACT1VITY To determine the activity of Reactor Coolant (RC) SO23-XXXVI-2.6 to satis'y Technical Specification requirements.
4 60 S 0123-111-1.1.23 UNITS 2/3 CIIEMICAL CONTROL OF To establish a program of parameters, limits, and PRIMARY PLANT AND RELATED SO123-lil-l.12 routine sampling requirements for the Reactor SYSTEMS Coolant System (RCS) and other related systems.
61 50123-111-1.27.23 PRIM ARY CilEMICAL FEED SYSTEMS To provide a method for chemical additions to the OPERATION following Units 2/3 Primary Plant systems:
+
+
Component Cooling Water
+ Spent Fuel Pool 62 SO123-III-2.22.23 UNITS 2/3 STEAM GENER A'IT}R LEAK 1. To describe methods for detecting and RATE DETERMINATION RGR. X.003 measuring primary-to-secondary system leakage during Unit operation.
- 2. To provide guidance for Chemistry laboratory activities including but not limited to analysis trends, instrumentation isolation, guidance to Operations, and radiation surveys.
i Southern California Edison 226 November 1998
t c .. . . . . _ .
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Indes Procedure Number Procedure Title Description of Purpose Connecting Ihslure A.6 Nuclear Training Division 63 Sol 23-XXI-1.8 MANAGING TR AINING COMMIThlENTS 1. To proside a systematic process for incorporating industry operating events into training programs.
- 2. To preside a systemic process for managing training commitments using the MOSAIC-AR resulting from:
a) Extemal Source Documents b) Intemally generated training action items.
- 3. To provide a systemic process for managing Nuclear Training Division (NTD) Regulatory Commitment Tracking System (RCTS) action assignments.
64 SO123-XXI-1.11.4 NON-LICENSED OPER ATOR TR AINING To describe Units 2 and 3 Non-Licensed Operator PROGRAM DESCRIPT10N (NLO) and proside a systematic process for their implementation.
65 sol 23-XXI-1.11.5 REAClOR OPER ATOR/ ASSISTANT To describe Reactor Opentor and Assistant CONTROL OPERATOR 'IRAINING Control Operator (RO/ACO) Training Program PROGRAM DESCRIPTlON and provide a systematic process for its implementation.
66 SO123-XXI-l.I 1.6 1 SENIOR REACIOR OPERATOR / To describe Senior Reactor Operator and Control 1 CONTROL ROOM SUPERVISOR sol 23-XXI-1.11.4 !
Room Supervisor (SRO/CRS) Training Program SOI23-XXI-I.11.5 l TR AINING PROGRAM DESCRIPTION and provide a systematic process for its implementation.
67 SOI23-XXI-l.11.7 LICENSED OPERATOR 1. To provide instructions and guidelines for the REQUALIFICATlON TRAINING implementation of the Licensed Operator PROGRAM DESCRIPTION Requalification Training Program.
- 2. To describe the Licensed Operator Requalification Training Program.
68 SOI23-XXI-1.11.10 Cl{EMISTRY TRAINING PROGRAM To describe the Chemistry Training Program and DESCRIPTION to provide a systematic process for its implementation.
Southern California Edison 227 l November 1998
APPENDIX A *** FOR INFORM ATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Number Procedure Title Description of Purpose Connecting Procedure 69 SOI23-XXI-1.11.1 I ENGINEERING SUPPORT PERSONNEL To define the requirements for the Engineering SO123-TN-1 (ESP) TRAINING PROGRAh! Support Personnel (ESP) Training Program.
DESCRIPTION To provide assurance that ESP personnel possess the knowledge and skills necessary to petform assigned job functions in a competent safe, and efficient manner.
70 sol 23-TN-I NUCLEAR ORGANIZATION TRAINING To define the Nuclear Organization responsibilities relative to the development and maintenance of training programs which ensure that personnel involved in quality-affecting work are fully qualified and trained according to federal, state, and Company requirements.
71 SO23-XXI-3.2.3 SIN 1CLA1DR REVIEW OF PLANT To outline the steps to be used in reviewing and N10DIFICATIONS tracking plant modifications and generating the work orders necessary for implementing those changes into the simulator.
72 SO23-X XI- 1.! 1. I I ENGINEERING SUPPORT PERSONNEL To define the requirements for the Engineer ing (ESP) TR AINING PROGR AM Support Personnel (ESP) Training Program DESCRIPT10N consistent with I-)CFR50.120 to provide assurance that ESP personnel possess the knowiedge and skills necessary to perform assigned job functions in a competent, safe, and efficient manner.
73 SO123-XU-33 PERSONNEL QUALIFICATION To outline actions required for ensuring PROGRAM FOR T1IE SAN ONOFRE candidates selected for positions within the OPERATIONS ORGANIZAllON Nuclear Organization meet ANSI and Regulatory Guide standards for Nuclear Plant personnel.
Southern California Edison 228 November 1998
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Pr&wiure Number l'rocedure Title Description of Purpose Connectina Procedure 74 SO23-XXI-3.3.3 SIMULATOR CORE PIIYSICS TESTING 1. To verify that key parameters associated with SOI23-XXVI-2.10 reactivity in an A!! Rods Out ( ARO) condition Transmittal on the simulator closely correspond to the values predicted in the Plant Physics Data
' Book for the core cyle presently being modeled.
- 2. To verify that CEA reactivity effects on the simulator closely wh predictions in the Pisnt Physics Data Book for the core cycle presently being modeled.
- 3. To verify that any changes in the moderator temperature produce the same reactivity efTects on the simulator as those predicted in the Plant Physics Data Book for the core cycle presently being medeled. l
- 4. To verify that the simulator core model responds to a reactivity insertion to produce a start up rate that closely corresponds to the predictions found in the Plant Physics Data Book for the core cycle presently being modeled.
- 5. To verify that the steady state and transient Xenon reactivity on the simulator closely match the Xenon reactivities predicted in the Plant Physics Data Book for the core cycle presently being modeled.
Southern California Edison 229 November 1998
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Indes Prc,cMure Number Procedure Title Description of Puw Connecting Procedure 75 SO23-XXI-3.16 SIMULATOR TRANSIENTTEST1NG AND 1. Evaluate the simulator response against EVALUAT10S available transient data to assure the simulator response to transients is acceptable for the training and testing of operators.
- 2. Demonstrate that the simulator response to transients meets the acceptance criteria of ANS/ ANSI 3.5
- 3. Verify annually that the simulator response to transients has not degraded and continues to be acceptable response.
l Southern California Edison 230 November 1998
- APPENDIX A *** FOR INFORMATION ONLY 6$0 SONGS Reload An: lysis Rited Procedures Index Procedure Number Procedure Title Description of Purpose Connecting Procedure A.7 Licensing Division 76 SO l 23-XXX-2.3 Pl!EPARATION OF AMENDMENT l 1. The purpose of this procedure is to describe SOI23-XXIV-1.1 APPLICA110NS FOR T11E SONGS I the preparation process to be used by Nuclear SOI23-XXX-5.2 POSSESSION ONLY LICENSE AND TIIE Regulatory Affairs (NRA)for Amendment Unit 2 AND 3 FACILITY OPERATING Applications to the SONGS 1 Possession Only LICENSE License and the Unit 2&3 Facidty OpeWng Licenses.
- 2. As used in this procedure, the term Ol/ POL includes the Facility Operating Licenses and Technical Specifications for Unit 2&3 and the Possession Only License and Technical Specifications for SONGS 1.
77 SO123-XXX-2.4 PREPARATION, REVIEW, AND This procedure defines the process for control and SO123-XXIV-1.1 APPROVAL OF CllANGES TO T11E maintenance of the Units 2/3 IIS, and at a later SO123-XXIV-2.3 SONGS LICENSEE CON 1 ROLLED time a Unit I LCS. This procedure also defines SPECIFICATlONS AND1ECilNICAL the process for control and maintenance of the SPECIFICATION BASES Unit 2&3 Technical Specification Bases and COLR changes.
78 sol 23-XXX-5.2 CONTROL OF LICENSING DOCUMENT The purpose of this procedure is to document the SO123-XXIV-l.1 CllANGES process of: SO123-XXIV-10.9 sol 23-XV-51
- 1. Preparation of changes to the SONGS Units 1 sol 23-XXXVI-3.1 2 & 3 Licensing Documents, specifically the SO123-XXXVI-2.4 UFSAR/UFIIA;
- 2. Preparation of the 10CFR50.59 Facility Change Report for SONGS 1,2 & 3.
Southem Califomia Edison 231 November 1998
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APPENDIX A
- FOR INFORMATION ONLY *00 SONGS Reload Analysis Rel;ted Proceduxs index Procedure Number Ihn:ure Title Description of Purg Connecting P.wa ure A.8 Nuclear Oversight Division 79 50123-Xil-7.12 SOURCE VERIFICATION
- 1. To provide instructions for planning, SO23-XXXIV-2.10 performing, and reporting source verification Transmittal activities by the Nuclear Oversight Division SO123-XX-1 (NOD) in compliance with Criterior. VII of 10CFR50, Appendix B.
- 2. To implement Chapter 3-D of the Topical Quality Assurance Manual (TQAM).
I Southern California Edison 232 November 1998
APPENDIX A
- FOR INFORMATION ONLY CGC SONGS Reload Analysis Related Procedures Index Procedure Number Procedure Titic Description of Purpcse Connecting Prc,cuiure A.9 Maintenance Division 80 SO23-II-8 2I CEDMCS CAllBRATION To test and verify the proper operations of the RGR, II.006 Control Element Drive Mecloism Control System (CEDMCS).
81 SOI23-XX-I ACIlON REQUESTS MAINTENANCE 1. To provide a single systea for reporting of sol 23-XV-51 ORDER INITIATION AND PROCESSING conditions adverse to qualitg events, and RGR, INTRO proposed improvements (equipn; nt and non-equipment related) and for resultant actions.
- 2. To delineate responsibilities for the management and oversight of the Action Request (AR) process.
- 3. To define the process for ensuring that timely j corrective actions are taken, pursuant to TQAM l-F.
- 4. To provide the necessary guidance to facilitate the initiation of a new AR.
82 SOI23-X U-50 EVENT REPORT PROGRAM To define the process for improving plant safety and availability through identification and correction of unwanted plant conditions involving human performance. programmatic, and organizational problems. ,
83 S 0123-1I-8.21 CEDMCS CALIBRATION To test and verify the proper operations of the RGR,11.006 j
Control Element Drive Mechanism Control System (CEDMCS).
Southern California Edison 233 November 1998 A . __-__
~- . __
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Number Procedure Title Description of Purpose Connecting Paedure A.10 Operations Division 84 SO23-I-2.44 CREACUS. CONTROL ROOM 1. His procedure provides the details to RGR, X.015 EMERGENCY AIR CLEAN UP SYST11M remove a representative carbon sample from OPERATION AND OPERABILITY T11ST the Control Room Emergency Air Cleanup SURVEILLANCE System (CREACUS) and have it analyzed for its decontamination efficiency for elemental iodine and organic iodides.
- 2. This procedure provides the details to verify that a Control Room Emergency Air Cleanup System IIEPA filter bank in-place leak test is performed.
- 3. This procedure performs a control room emergency air cleanup system adsorber stage in-place leak test.
- 4. This procedure demonstrates the operability of the Control Room Emergency Air Cleanup System after 720-hours of charcoal adsorber operation.
- 5. This procedure provides the details to verify that, with the ventilation and air conditioning units operating at designed l flow, the pressure drop across the IIEPA ;
Filters, and charcoal adsorber banks are i within acceptable limits. i Southern Califomia Edison 234 November 1998
._ . . - ~ - - -
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Amlysis Related Procedures Index Procedure Number Prc.cnture Title Description of Pu - Connecting Pro (+ Jure 85 SO23-I-2.44 CREACUS, CONTROL ROOM 6. His pnxedure provides the details to verify RGR, X.015 EMERGENCY AIR CLEAN UP SYSTEM that the system maintains the Control Room OPERATION AND OPERABILITY TEST at a positive pressure 21/8 in. W.G. relative SURVEILLANCE (Continued) to the outside atmosphere during sys, tem operation in the emergency Mode.
- 7. His procedure satisfies the requirements as set forth in Unit 2 and Unit 3 Technical Specifications Sections SR 3.7.11.2, SR3.7.11.4, and 5.5.2.12 (Ventilarion Filter Testing Program - VFTP) 86 SO23 I-2.5 111 STING OF MAIN STEAM SAttI Y 1. This procedure tests the relief set pressure RGR, V.015 VALVES SURVEILLANCE of the Main Steam Safety Valves 2/3 SO23-3-3.2.5 ;
PSV-8401 through 2/3 PSV-8418 using a test device which simmers the valves and is not to be confused with a full lift test.
- 2. His procedure satisfies the Surveillance Requirements as set forth in the Units 2 and 3 Technical Specifications. Section 3.7.1, SR 3.7.1.1 and 5.5.2.10.
87 SO23-3-3.5 . CEA/ REACTOR 1 RIP BREAKER Dis surveillance test demonstrates the ability to OPER ATIBILITY 111 STING, RGR,11.009 operate Safe Shutdown components from their SO23-3 2.13 ATTACilMENT 2. OPTIMAL CEA respective second points of control per LCS SR POS1110N VS EFFPD SO23-3-2.19 3.7.113.1.11. SO23-5 1.7 i
Southern California Edison 235 November 1998
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Number Procedure Title Description of Purpose Connecting Procedure 88 S 023-3-3.6 COLSS Otrr OF SERVICE This surveillance test verifies the following power RGR, I.004 St'RVEILLANCE distribution limits ahile operating in Mode I greater than 20% of rated thermal power with the Core Operating Limit Supervisory System (COLSS) and COLSS Backup Computer System out of service.
- 1. Provides guidance and documentation to ensure compliance with Technical Specifications / Licensee Controlled Specifications Actions which permit operating with DNBR (Tech. Spec.
SR 3.2.4.1) and LIIR (Tech. Spec.
SR 3.2.1.2) outside specified limits for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> with COLSS and COLSS Backup Computer OOS. (LCO 3.2.4, LCO 3.2.1) 89 5023-5-1.7 POWER OPER AllONS This procedure is intended to provide procedural RGR,11.011 guidelines and strategies for conducting overall SO23-V-2 plant operations above 15% power. SO!23-V-13 90 l SO23 1.10 PRESSL'RIZER PRESSL'RE AND 1.EVEL This instruction provides the steps required to l
RGR,111.015 '
CONTROL control Pressurizer pressure and level during normal and off-normal conditions.
91 SO23-3-1.8 DRAINING TllE REACTOR COOLANT To provide instructions to drain the Reactor RGR, I.024 SYSTEM, ATTACllMENT6 Coolant System to the proper level for equipment maintenance or refueling preparations.
92 SO23-5-I.8 Sl{UTDOWN OPERATIONS (MODE 5 Provide guidance for plant operation / activities RGR, I.024 AND 6) with RCS Tavg less than 200'F. SOI23-V-13 SO23-X-7 93 SO23-9-5 CONDENSATE STURAGE AND This instruction describes the normal operation of RGR, V.015 TRANSFER SYSTEM the Condensate Storage and Transfer System. S023-3-3.25 outlines prerequisite conditions prior to system startup and provides a detailed procedure for system alignment during various phases of plant operation.
Southern California Edison 236 November 1998
w m v APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Number l*rocedure Title Description of Purpose Connecting 1%cedure 94 5023-1-4 CONTAINMENT NORMAL llEAT To outline the proper status and operation of the REMOVAL RGR, IX.001 following Containment Normal Heat Removal Systems:
- 1. Containmt. . Normal Cooling System
- 2. Control Element Drive Mechanism Cooling System
- 3. Reactor Cavity Cooling System
- 4. Iower 12 vel Circulating Fans 95 SO23-3-3.25 ONCE A DAY SURVEILLANCE (MODE To delineate those readings, channel checks, and 1-4) SO23-V-2 other surveillances required to be performed once RGR. IV.002 a shift on a routine basis when in Modes 1 SO23-3-2.12 through 4, and document compliance with the SO23-3-2.13 Technical Specifications listed in Section 6.0 of SO23-5-1.3
- this procedure.
% SO23-3-3.26 ONCE A DAY SURVEILLANCE (MODE Ris test procedure describes those readings, RGR, IV.010 t-4) channel checks, and other surveillances (except RGR, IX.001 radiation monitoring) required to be performed SO23-I-4 daily while in Modes I 4, and documents 5023-5-I.3 compliance with the Technical Specifications.
97 SO23-3-1.1 REACTOR START-UP To provide instructions necessary to bring the SO23-V-1.01 Reactor from Ilot Standby (MODE 3) conditions SO23-V-13 to Startup (MODE 2) operation with the Reactor S0123-0r-1 stabilized at 2 x 10'% power. S0123-3-l.2 SO23-3-3.29 SO23-5-1.3 i SO23-V-1 I SO23-3-2.13 98 SO23-3-3.27 ONCE A DAY SURVEILLANCE (MODE nis instruction verifies and documents 1-4) SO23-XXXIV-2.10 surveillances required to be performed once a day Transmittal on a routine basis while in Modes 1-4 in S 023-3-3.29 compliance with the Technical Specifications identified in Section 6.0.
99 5023-13-13 MISALIGNED CEA Specify actions to mitigate the effects of SO23-3-I.I misaligned or immovable CEA(s). 5023-3-1.2 S 023-3-2.13 SO23-3-3.6 Southern California Edson 237 November 1998
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Number Procedure Title Description of Pura Connecting ihdure 100 5023-3 2.13 CORE PROTECT 10N CONTROL This instruction outlines the operation of the Core SO23-V-4.7 ELEMENT ASSEMBLY CALCULA1DR Protection Calculator (CPC) and the Control OPERAllON SO23-3-3.25
- Element Assembly Calculator (CEAC)in the SO23-5-I.7 following modes:
- 1. CPC Operation
- 2. Display Function
- 3. Change Value Function
- 4. Sensor Failures
- 5. CEAC Operation
- 7. CEAC Inoperatable
- 8. Reed Suitch Position Transmitter
- 9. CEAC Retum to Service
- 10. CPC/CEAC Snapshots
- 11. Placing CPC Channels into Tripped Condition.
101 S023-3 3.29 DETERMINATION OF REAC1DR I. To provide methods for determining the SO23-V-I SillTIVOWN M ARGIN actual or available Reactor Shutdown SO23-V-l .5 Margin (SDM) for various plant modes and SO23-V-13 CEA conditions. SO23-3-3.25 l
SO23-XXXIV-2.10
- 2. To provide methods for detennining the Transmittal minimum required boron concentration to maintain SDM during plant heat-up and cooldown.
102 SO23-3-2.21 COREOPERATING LIMITS To provide instructions to operate the COLSS SO23-3-2.13 SUPERVISORY SYSTEM (COLSS) Computer System and the COLSS Backup Computer System during normal and abnormal conditions.
Southern California Edison 238 November 1998
APPENDIX A
- FOR INFORMATION ONLY *** SONGS Reload Analysis Related Procedures Index Procedure Number Procedure Title Description of Purpose Connecting Prcadure 103 SO23 3 3.2 EXCORE NUCLEAR I. To adjust the following indications to SO23-V-4.7 INSTRUMENTA110N CAllBR A110N within -l% and +5% of the calorimetric SO23-3-2.12 -
calibration when above 20% rated thermal SO23-3-2.13 power (RTP): (Tech. Spec. SR 3.3.1.4) SO23-3-3.25 SO23-3-3.25.1
- Indicatedlinearpowerlevel
- CPC AT Power
- CPC Nuclear Power
- 2. To integrate the I & C Department and Operations Division responsibilities during Excore calibration.
104 SO23-3-2.11 SPENT FUEL POOL OPERATION 1. This instruction provides the procedures for SO23-IT-1 operating the Spent Fuel Pool Cooling and SO23-V-12.2.26 Purification System under the following conditions:
- a. Normal Spent Fuel Pool Cooling
- h. Fuel Pool Purification
- c. Refueling Water Storage Tank Purification
- d. Attemate methods of SFP Cooling
- 2. This instruction provides the procedures for operations of the Spent Fuel Pool Isolation Gates, and Seal Pressurization System.
- 3. This instruction provides the procedures to adjt st Spent Fuel Pool level and/or boron concentration.
Southern California Edison 239 November 1998
APFENDIX A
- FOR INFORMATION ONLY *S* SONGS Reload Analysis Related Procedures Index Prawture Number Procedure Title Dec.cription of Purisc= Connecting PicMure 105 SOI23-IT-1 INFREQUENILY PERFORMED TESTS 1. To delineate Nuclear Organization AND EVALUATIONS CONTROLS PROGRAM management controls that must be spplied to ensure that Infrequently Performed Tests and Evelutions are properly planned, reviewed, and executed with appropriate Management Oversight to ensure the highest regard for Reactor safety (see Definitions, Attachment A).
- 2. To maintain a definitive list of tests and evolutions which require the extra-ordinary controls discussed herein.
i Southern California Edison 240 November 1998
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f ENCLOSURE 3 PROPRIETARY INFORMATION AFFIDAVIT SIGNED BY ABB CE r
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AFFIDAVIT PURSUANT 1
TO 10 CFR 2.790 l, Ian C. Rickard, depose and say that I am the Director, Nuclear Licensing, of Combustion .
1 Engineering, Inc., duly authorized to make this affidavit, and have reviewed or caused to have reviewed the information which is identified as proprietary and referenced in the paragraph ,
immediately below. I am submitting this affidavit in conformance with the provisions of 10 CFR :
2.790 of the Commission's regulations and in conjunction with the application of the Southern -
California Edison for withholding this information.
The information for which proprietary treatment is sought is contained in the following document:
SCE-9801-P, Rev. O,
- Reload Analysis Methodology for the San Onofre Nuclear !
Generating Station Units 2 and 3", November 1998 This document has been appropriately designated as proprietary.
I have personal knowledge of the criteria and procedures utilized by Combustion Engineering in designating information as a trade secret, privileged or as confidential commercial or financial information. j j
Pursuant to the provisions of paragraph (b) (4) of Section 2.790 of the Commission's regulations, l
the following is furnished for consideration by the Commission in determining whether the i information sought to be withheld from public disclosure, included in the above referenced document, should be withheld. l
- 1. The information sought to be withheld from public disclosure, is owned and has l been held in confidence by Combustion Engineering. It consists of the reload engineering analysis methodology, including the computer programs utilized and l
the scope of analyses performed for the San Onofre Nuclear Generating Station. l
- 2. The information consists of test data or other similar data concerning a process,
)
method or component, the application of which results in substantial competitive !
advantage to Combustion Engineering.
I I
l 1
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- 3. The information is of a type customarily held in confidence by Combustion '
Engineering and not customarily disclosed to the public. Combustion Engineering has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The details of the aforementioned system were provided to the Nuclear Regulatory Commission via letter DP-537 from F. M. Stern to Frank Schroeder dated December 2,1974. This system was applied in determining that the subject document herein is proprietary.
- 4. The information is being transmitted to the Commission in confidence under the 4
provisions of 10 CFR 2.790 with the understanding that it is to be received in I confidence by the Commission.
- 5. The information, to the best of my knowledge and belief, is not available in public I sources, and any disclosure to third parties has been made pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence.
- 6. Public disclosure of the information is likely to cause substantial harm to the competitive position of Combustion Engineering because:
- a. A similar product is manufactured and sold by major pressurized water reactor competitors of Combustion Engineering.
b Development of this information by Combustion Engineering required of millions of dollars and tens of thousands of manhours of effort. A competitor would have to undergo similar expense in I generating equivalent information.
- c. In order to acquire such information, a competitor would also require considerable time and inconvenience to develop the reload engineering analysis methodology, including the computer l programs utilized and the scope of analyses performed for the San l Onofre Nuclear Generating Station. !
I
- d. The information consists of the reload engineering analysis methodology, including the computer programs utilized and tho ,
i scope of analyses performed for the San Onofre Nuclear !
Generating Station, the application of which provides a competitive i
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economic advantage. The availability of such information to competitors would enable them to modify their product to better i compete with Combustion Engineering, take marketing or other a
actions to improve their product's position or impair the position of i.
Combustion Engineering's product, and avoid developing similar data and analyses in support of their processes, methods or !
apparatus.
- e. In pricing Combustion Engineering's products and services, a
significant research, development, engineering, analytical, manufacturing, licensing, quality assurance and other costs and I
expenses must be included. The ability of Combustion
- Engineering's competitors to utilize such information without similar expenditure of resources may enable them to sell at prices ;
l reflecting significantly lower costs.
- f. Use of the information by competitors in the international l marketplace would increase their ability to market nuclear steam >
supply systems by reducing the costs associated with their technology development, in addition, disclosure would have an ;
1 >
adverse economic impact on Combustion Engineering's potential for obtaining or maintaining foreign licensees. i
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Further the deponent sayeth not. !
la Director, Nuclear Licensing Sworn to before me this[6 day of / /F 1998 L -
iY
! ausus -
h) ,
V\ Notary Public //
V j My c'ommission expires:
(/3/97