RS-08-045, LaSalle, Units 1 and 2, Updated Final Safety Analysis Report (Ufsar), Revision 17, Chapter 6.0 - Engineered Safety Features

LaSalle, Units 1 and 2, Updated Final Safety Analysis Report (Ufsar), Revision 17, Chapter 6.0 - Engineered Safety Features

ML081330058
Person / Time
Site:	LaSalle
Issue date:	04/14/2008
From:	Exelon Generation Co, Exelon Nuclear
To:	Office of Nuclear Reactor Regulation
References
RS-08-045
Download: ML081330058 (656)

Text

LSCS-UFSAR 6.0-i REV. 15, APRIL 2004 CHAPTER 6.0 - ENGINEERED SAFETY FEATURES TABLE OF CONTENTS PAGE 6.0 ENGINEERED SAFETY FEATURES 6.0-1 6.1 ENGINEERED SAFETY FEATURE MATERIALS 6.1-1 6.1.1 Metallic Materials 6.1-1 6.1.1.1 Materials Selection and Fabrication 6.1-1 6.1.1.2 Composition, Compatibility and Stability of Containment and Core Spray Coolants 6.1-4 6.1.2 Organic Materials 6.1-4 6.1.3 Postaccident Chemistry 6.1-4 6.2 CONTAINMENT SYSTEMS 6.2-1 6.2.1 Containment Functional Design 6.2-1 6.2.1.1 Containment Structure 6.2-1 6.2.1.1.1 Design Bases 6.2-1 6.2.1.1.2 Design Features 6.2-3 6.2.1.1.3 Design Evaluation 6.2-7 6.2.1.1.3.1 Accident Response Analysis 6.2-8 6.2.1.1.3.1.1 Recirculation Line Rupture 6.2-9 6.2.1.1.3.1.2 Main Steamline Break 6.2-18 6.2.1.1.3.1.3 Intermediate Breaks 6.2-19 6.2.1.1.3.1.4 Small Size Breaks 6.2-20 6.2.1.1.3.2 Accident Analysis Models 6.2-22 6.2.1.1.4 Negative Pressure Design Evaluation 6.2-28 6.2.1.1.5 Suppression Pool Bypass Effects 6.2-28 6.2.1.1.6 Suppression Pool Dynamic Loads 6.2-30 6.2.1.1.7 Asymmetric Loading Conditions 6.2-31 6.2.1.1.8 Containment Ventilation System 6.2-31 6.2.1.1.9 Postaccident Monitoring 6.2-31 6.2.1.1.10 Drywell-to-Wetwell Vacuum Breaker Valves Evaluation for LOCA Loads 6.2-31 6.2.1.1.11 Impact of Increased Initial Suppression Pool Temperature 6.2-32 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 6.0-ii REV. 15, APRIL 2004 6.2.1.2 Containment Subcompartments 6.2-32 6.2.1.2.1 Design Bases 6.2-32 6.2.1.2.2 Design Features 6.2-34 6.2.1.2.3 Design Evaluation 6.2-36 6.2.1.2.4 Impact of Increased Initial Suppression Pool Temp erature

6.2-45 6.2.1.3 Mass and Energy Releas e Analyses for Postulated Loss-of-Coolant Accidents 6.2-45 6.2.1.3.1 Mass and Energy Release Data 6.2-45 6.2.1.3.2 Energy Sources 6.2-46 6.2.1.3.3 Effects of Metal-Water Reaction 6.2-46 6.2.1.3.4 Impact of Increased Initial Suppression Pool Te mperature

6.2-46 6.2.1.4 Mass and Energy Releas e Analysis for Postulated Secondary Systems Pipe Ruptures Inside Containment (PWR) 6.2-46 6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies on Emergency Core Cooling System (PWR) 6.2-46 6.2.1.6 Testing and Inspection 6.2-47 6.2.1.7 Instrumentation Requirements 6.2-47 6.2.1.8 Evaluation of 105

°F Suppression Pool Initial Temperature 6.2-47 6.2.2 Containment Heat Removal System 6.2-48 6.2.2.1 Design Bases 6.2-48 6.2.2.2 System Design 6.2-49 6.2.2.3 Design Evaluation 6.2-49 6.2.2.3.1 RHR Contai nment Cooling Mode 6.2-49 6.2.2.3.2 Summary of Containment Cooling Analysis 6.2-50 6.2.2.3.3 Impact of Increased Initial Suppression Pool Temperature 6.2-50 6.2.2.3.5 Impact of Power Uprate 6.2-51 6.2.2.3.6 Sensitivity of Initiation Time of RHR Containment Cooling Mode 6.2-51 6.2.2.4 Test and Inspections 6.2-51 6.2.2.5 Instrumentation Requirements 6.2-51 6.2.3 Secondary Containment Functional Design 6.2-51 6.2.3.1 Design Bases 6.2-51 6.2.3.2 System Design 6.2-51 6.2.3.3 Design Evaluation 6.2-53 6.2.3.4 Test and Inspections 6.2-53 6.2.3.5 Instrumentation Requirements 6.2-53 6.2.4 Containment Isolation System 6.2-53 6.2.4.1 Design Bases 6.2-54 6.2.4.2 System Design 6.2-55 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 6.0-iii REV. 15, APRIL 2004 6.2.4.2.1 Evaluation Against General Design Criterion 55 6.2-55 6.2.4.2.2 Evaluation Against General Design Criterion 56 6.2-59 6.2.4.2.3 Evaluation Against General Design Criterion 57 6.2-61 6.2.4.2.4 Miscellaneous 6.2-64 6.2.4.3 Design Evaluation 6.2-64 6.2.4.4 Tests and Inspections 6.2-65 6.2.5 Combustible Gas Control in Containment 6.2-65 6.2.5.1 Design Bases 6.2-66 6.2.5.2 System Design 6.2-67 6.2.5.3 Design Evaluation 6.2-70 6.2.5.3.1 General 6.2-70 6.2.5.3.2 Sources of Hydrogen 6.2-71 6.2.5.3.3 Accident Description 6.2-72 6.2.5.3.4 Analysis 6.2-72 6.2.5.4 Testing and Inspections 6.2-73 6.2.5.5 Instrumentation Requirements 6.2-73 6.2.6 Containment Leakage Testing 6.2-73 6.2.6.1 Containment Integrated Leakage Rate Test 6.2-74 6.2.6.2 Containment Penetration Leakage Rate Test 6.2-77 6.2.6.3 Containment Isolation Valve Leakage Rate Test 6.2-80 6.2.6.4 Scheduling and Reporting of Periodic Tests 6.2-80 6.2.6.5 Special Testing Requirements 6.2-80 6.2.7 References 6.2-80 6.3 EMERGENCY CORE COOLING SYSTEMS 6.3-1 6.3.1 Design Bases 6.3-1 6.3.1.1 Summary Description of the Emergency Core Cooling System 6.3-1 6.3.1.1.1 Range of Coolant Ruptures and Leaks 6.3-2 6.3.1.1.2 Fission Product Decay Heat 6.3-2 6.3.1.1.3 Reactivity Required for Cold Shutdown 6.3-2 6.3.1.2 Functional Requirement Design Bases 6.3-2 6.3.1.3 Reliability Requirements Design Bases 6.3-3 6.3.2 System Design 6.3-3 6.3.2.1 Schematic Piping and Instrumentation Diagrams 6.3-4 6.3.2.2 Equipment and Component Descriptions 6.3-4 6.3.2.2.1 High-Pressure Core Spray (HPCS) System 6.3-4 6.3.2.2.2 Automatic Depressurization System (ADS) 6.3-6 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 6.0-iv REV. 15, APRIL 2004 6.3.2.2.3 Low-Pressure Co re Spray (LPCS) System 6.3-6 6.3.2.2.4 Low-Pressure Coolant Injection (LPCI)

Subsystem 6.3-8 6.3.2.2.5 ECCS Discharge Line Fill System 6.3-9 6.3.2.2.6 ECCS Pumps NPSH 6.3-9 6.3.2.2.7 Design Pressu res and Temperatures 6.3-11 6.3.2.2.8 Coolant Quantity 6.3-11 6.3.2.2.9 Pump Characteristics 6.3-11 6.3.2.2.10 Heat Exchanger Characteristics 6.3-12 6.3.2.2.11 ECCS Flow Diagrams 6.3-12 6.3.2.2.12 Relief Valves and Vents 6.3-12 6.3.2.2.13 Motor-Operated Valves and Controls (General) 6.3-13 6.3.2.2.14 Process Instrumentation 6.3-14 6.3.2.2.15 Scram Discharge System Pipe Break 6.3-14a 6.3.2.3 Applicable Codes and Classification 6.3-15 6.3.2.4 Materials Specifications and Compatibility 6.3-15 6.3.2.5 System Reliability 6.3-15 6.3.2.6 Protection Provisions 6.3-16 6.3.2.7 Provisions for Performance Testing 6.3-16 6.3.2.8 Manual Actions 6.3-18 6.3.3 ECCS Performance Evaluation 6.3-18 6.3.3.1 ECCS Bases for Technical Specifications 6.3-19 6.3.3.2 Acceptance Criteria for ECCS Performance 6.3-19 6.3.3.3 Single-Failure Considerations 6.3-20 6.3.3.4 System Performance During the Accident 6.3-21 6.3.3.5 Use of Dual Function Components for ECCS 6.3-22 6.3.3.6 Limits on ECCS Parameters 6.3-22 6.3.3.7 ECCS Analysis for LOCA 6.3-22 6.3.3.7.1.1 GE LOCA Anal ysis Procedures and Input Variables 6.3-22 6.3.3.7.1.2 SPC LOCA An alysis Procedures and Input Variables 6.3-24 6.3.3.7.2 Accident Description 6.3-25 6.3.3.7.3 Break Spectrum Calculations 6.3-26 6.3.3.7.4 Large Recirculation Line Break Calculations 6.3-27 6.3.3.7.4 Deleted 6.3.3.7.4.1 GE LOCA Analysis Large Recirculation Line Break Calculations 6.3-27 6.3.3.7.4.2 SPC LOCA Anal ysis Large Recirculation Line Break Calculations 6.3-27 6.3.3.7.6.1 GE LOCA Analysis Small Recirculation Line Break Calculations 6.3-29 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 6.0-v REV. 17, APRIL 2008 6.3.3.7.6.2 SPC LOCA Anal ysis Small Recirculation Line Break Calculations 6.3-29 6.3.3.7.7.1 GE LOCA Anal ysis Calculations for Other Break Locations 6.3-31 6.3.3.7.7.2 SPC LOCA Analysis Calculations for Other Break Locations 6.3-31 6.3.3.7.8.1 GE Steamline Break Outside Containment Analysis 6.3-32 6.3.3.7.8.2 SPC Steamlin e Break Outside Containment Analysis 6.3-32 6.3.3.8.1 Errors and Changes Affecting LOCA Analysis 6.3-32 6.3.3.9.1 GE LOCA An alysis Conclusions 6.3-34 6.3.3.9.2 AREVA LOCA Analysis Conclusions 6.3-34 6.3.3.10 MSIV Closure Change from Reactor Water Level 2 to Level 1 6.3-34 6.3.4 Tests and Inspections 6.3-35 6.3.5 Instrumentation Requirements 6.3-37 6.3.5.1 HPCS Actuation Instrumentation 6.3-37 6.3.5.2 ADS Actuation Instrumentation 6.3-37 6.3.5.3 LPCS Actuation Instrumentation 6.3-37 6.3.5.4 LPCI Actuation Instrumentation 6.3-37 6.3.6 References 6.3-38

6.4 HABITABILITY SYSTEMS 6.4-1 6.4.1 Design Bases 6.4-1 6.4.2 System Design 6.4-3 6.4.2.1 Definition of Control Room Envelope 6.4-3 6.4.2.2 Ventilation System Design 6.4-3 6.4.2.3 Leaktightness 6.4-3 6.4.2.4 Interaction with Other Zones and Pressure- Containing Equipment 6.4-4 6.4.2.5 Shielding Design 6.4-4 6.4.3 System Operational Procedures 6.4-5 6.4.4 Design Evaluation 6.4-6 6.4.5 Testing and Inspection 6.4-7 6.4.6 Instrumentation Requirements 6.4-8

6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5-1 6.5.1 Engineered Safety Feature (ESF) Filter Systems 6.5-1 6.5.1.1 Design Bases 6.5-1 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 6.0-vi REV. 15, APRIL 2004 6.5.1.1.1 Standby Gas Treatment System 6.5-1 6.5.1.1.2 Emergency Makeup Air Filter Units 6.5-4 6.5.1.2 System Design 6.5-6 6.5.1.2.1 Standby Gas Treatment System 6.5-6 6.5.1.2.2 Emergency Makeup Air Filter Units 6.5-8 6.5.1.2.3 Supply Air Filter Unit Recirculation Filter 6.5-11 6.5.1.3 Design Evaluation 6.5-11 6.5.1.3.1 Standby Gas Treatment System 6.5-11 6.5.1.3.2 Emergency Makeup Air Filter Units 6.5-12 6.5.1.4 Tests and Inspections 6.5-12 6.5.1.4.1 Standby Gas Treatment System 6.5-12 6.5.1.4.2 Emergency Makeup Air Filter Units 6.5-13 6.5.1.5 Instrumentation Requirements 6.5-15 6.5.1.6 Materials 6.5-16 6.5.2 Containment Spray Systems 6.5-17 6.5.3 Fission Product Control System 6.5-17 6.5.4 Ice Condenser as a Fission Product Cleanup System 6.5-17 6.6 INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 COMPONENTS 6.6-1 6.6.1 Components Subject to Examination 6.6-1 6.6.2 Accessibility 6.6-1 6.6.3 Examination Techniques and Procedures 6.6-1 6.6.4 Inspection Intervals 6.6-1 6.6.5 Examination Categories and Requirements 6.6-2 6.6.6 Evaluation of Examination Results 6.6-2 6.6.7 System Pressure Tests 6.6-2 6.6.8 Augmented Inservice Inspection to Protect Against Postulated Piping Failures 6.6-2

6.7 MAIN STEAM ISOLATIO N VALVE LEAKAGE CONTROL SYSTEM (MSIV-LCS)

Unit 2 Deleted, Unit 1 Abandoned In Place 6.7-1 6.8 MAIN STEAM ISOLATION VALVE - ISOLATED CONDENSER LEAKAGE TREATMENT METHOD - UNIT 1

6.8.1 Design Bases 6.8-1 6.8.1.1 Safety Criteria 6.8-1 6.8.1.2 Regulatory Acceptance Criteria 6.8-1 6.8.1.3 Leakage Rate Requirements 6.8-1 6.8.2 System Description 6.8-2 LSCS-UFSAR TABLE OF CONTENTS (Cont'd)

PAGE 6.0-vii REV. 15, APRIL 2004 6.8.2.1 General Description 6.8-2 6.8.2.2 System Operation 6.8-2 6.8.2.3 Equipment Required 6.8-3 6.8.3 System Evaluation 6.8-3 6.8.4 Instrumentation Requirements 6.8-3 6.8.5 Inspection and Testing 6.8-3 ATTACHMENT 6.A ANNULUS PRESSURIZATION 6.A-i ATTACHMENT 6.B RECIRCULATION SYSTEM SINGLE-LOOP OPERATION 6.B-i LSCS-UFSAR 6.0-viii REV. 15, APRIL 2004 CHAPTER 6.0 - ENGINEERED SAFETY FEATURES LIST OF TABLES NUMBER TITLE 6.1-1 Principal Pressure-Retaining Material for ESF Components 6.1-2 Organic Materials Within the Primary Containment 6.2-1 Containment Design Parameters 6.2-2 Engineered Safety Systems Information for Containment Response Analysis (at 3434 MWt) 6.2-3 Initial Conditions Employed in Containment Response Analyses (at 3434 MWt) 6.2-3a Initial Conditions Employed in Containment Response Analyses (at 3559 MWt) 6.2-4 Mass and Energy Release Data for Analysis of Water Pool Pressure-Suppression Containment Accidents Analyses (at 3434 MWt) 6.2-5 LOCA Long Term Primary Containm ent Response Summary Analyses (at 3434 MWt) 6.2-5a LOCA Long Term Primary Containment Response Summary (at 3559 MWt) 6.2-6 Energy Balance for Design-Basis Recirculation Line Break Accident (at 3434 MWt) 6.2-7 Accident Chronology Design-Basis Recirculation Line Break Accident (at 3434 MWt) 6.2-8 Summary of Accident Results for Containment Response to Recirculation Line and Steamline Breaks (at 3434 MWt) 6.2-8a Summary of Accident Results for Short-Term Containment Response to Recirculation Line Breaks (at 3559 MWt) 6.2-9 Subcompartment Nodal Descriptio n Recirculation Outlet Line Break With Shielding Doors 6.2-10 Subcompartment Nodal Descript ion Feedwater Line Break With Shielding Doors 6.2-11 Subcompartment Nodal

Description:

Head Spray Line Break 6.2-12 Subcompartment Nodal Descript ion: Recirculation Line Break 6.2-13 Subcompartment Vent Path Description-Head Spray Line Break 6.2-14 Subcompartment Vent Path

Description:

Recirculation Line Break 6.2-15 Simultaneous Break of the Head Spray Line and RPV Head Vent Line in the Head Cavity Input Data 6.2-16 Recirculation Line Break Input Data 6.2-17 Main Steamline Break Input Data 6.2-18 Reactor Blowdown Data for Re circulation Line Break (at 3434 MWt) 6.2-18a Reactor Blowdown Data for Re circulation Line Break (at 3559 MWt) 6.2-19 Reactor Blowdown Data for Main Steamline Break (at 3434 MWt) 6.2-20 Core Decay Heat Following LOCA for Containment Analyses (at 3434 MWt)

LSCS-UFSAR 6.0-viiia REV. 15, APRIL 2004 6.2-20a Core Decay Heat Following LOCA for Containment Analyses (at 3559 MWt) 6.2-21 Summary of Lines Penetrating the Primary Containment 6.2-22 Parameters Used to Dete rmine Hydrogen Concentration 6.2-23 Containment Leakage Testing 6.2-24 Subcompartment Vent Path Description Recirculation Outlet Line Break with Shielding Doors 6.2-25 Subcompartment Vent Path Description Feedwater Line Break with Shielding Doors

LSCS-UFSAR LIST OF TABLES (Cont'd)

NUMBER TITLE 6.0-ix REV. 15, APRIL 2004 6.2-26 Mass and Energy Release Rate Data Recirculation Outlet Line Break 6.2-27 Mass and Energy Release Rate Data Feedwater Line Break 6.2-28 Primary Containm ent Isolation Valves 6.3-1 DELETED 6.3-2 Significant Input Variables Used in the GE Loss-of-Coolant Accident Analysis 6.3-2a Significant Input Variables Used in FANP Loss-of-Coolant Accident Analysis 6.3-3 Operational Sequence of Emer gency Core Cooling Systems for GE Design-Basis Accident Analysis 6.3-4 Key to Figures and Tables in Section 6.3 6.3-5 ECCS Single Valve Failure Analysis 6.3-6 Single Failures Considered for ECCS Analysis 6.3-6a ATRIUM-9B MAPLHGR Analysis Results 6.3-6b DELETED 6.3-6c DELETED 6.3-6d DELETED 6.3-6e DELETED 6.3-6f DELETED

6.3-6g DELETED 6.3-6h DELETED 6.3.6i ATRIUM-10 MAPLHGR Analysis Results 6.3-7a Event Times for FANP Limiting La rge Break LOCA, 1.0 DEG Pump Suction SF-LPCS/DG for ATRIUM-9B Fuel 6.3.7b Event Times for FANP LOCA, 1.1 ft 2 Pump Discharge SF-HPCS/DG for ATRIUM-9B Fuel 6.3-8 Summary of Results of GE (SAFER/GESTR) LOCA Analysis 6.3-8a Summary of Results of FANP (HU XY) LOCA Analysis (Recirculation Line Breaks) 6.3-8b Summary of Results of FANP (HUXY) LOCA Analysis (Non-Recirculation Line Breaks) 6.3-9 List of Motor-Operated Valves Ha ving Their Thermal Overload Protection Bypassed During Accident Conditions 6.4-1 Dose Rates in the Control Room and Auxiliary Electric Equipment (AEE)

Rooms During Normal Operation 6.4-2 Dose Experienced by Control Room Personnel Following Loss-of-Coolant Accident 6.5-1 Standby Gas Treatment System Components 6.5-2 Standby Gas Treatment System Equipment Failure Analysis 6.7-1 DELETED 6.7-2.1 DELETED 6.8-1 Dose Consequences of MSIV Leakage

LSCS-UFSAR 6.0-x REV. 15, APRIL 2004 CHAPTER 6.0 - ENGINEERED SAFETY FEATURES LIST OF FIGURES AND DRAWINGS FIGURES NUMBER TITLE 6.2-1 Diagram of the Recirculation Line Break Location 6.2-2 Recirculation Line Break Pressure Response (at 3434 MWt) 6.2-2a Short-term Pressure Response Following a Recirculation Line Break (at 3559 MWt) 6.2-3 Temperature Response for Reci rculation Line Break (at 3434 MWt) 6.2-3a Short-term Temperature Response Fo llowing a Recirculation Line Break (at 3559 MWt) 6.2-4 Containment Vent System Flow Rate vs. Time for Recirculation Line Break (at 3434 MWt) 6.2-5 Containment Pressure Response (at 3434 MWt) 6.2-5a Long-Term Containment Pressure Response Following a Recirculation Line Break (at 3559 MWt) - Case C (2 pumps 1 Heat Exchanger Without

Continuous Spray) 6.2-6 Drywell Temperatur e Response (at 3434 MWt) 6.2-6a Long-Term Drywell Temperature Response Following a Recirculation Line Break (at 3559 MWt) - Case C (2 pumps 1 Heat Exchanger Without Continuous Spray) 6.2-7 Pool Temperature Response - Isolation/Scram, 1 RHR Available (at 3434 MWt) 6.2-7a Long-Term Suppression Pool Temperat ure Response Following a Recirculation Line Break (at 3559 MWt) - Case C (2 pumps 1 Heat Exchanger Without

Continuous Spray) 6.2-8 Pressure Response for a Main Steamline Break (at 3434 MWt) 6.2-9 Temperature Response Following a Main Steamline Break (at 3434 MWt) 6.2-10 Pressure Response for 0.1 ft 2 Liquid Line Break (at 3434 MWt) 6.2-11 Temperature Response for 0.1 ft 2 Liquid Line Break (at 3434 MWt) 6.2-12 Schematic of ECCS Loop 6.2-13 Allowable Steam Bypass Leakage Capacity 6.2-14 Containment Response to Large Primary System Breaks 6.2-15 Containment Response to Small Primary System Breaks 6.2-16 Nodalization Schematic For Recirculation Line Break 6.2-17 Nodalization Schematic For Feedwater Line Break 6.2-18 -P vs. Log t About Break - Recirculation Line Break 6.2-19 Head Spray Line Break Nodalization 6.2-20 Recirculation Line Break Nodalization 6.2-21 Pressure Response for Recirculation Line Break 6.2-22 P vs. Log t About Break - Feedwater Line Break LSCS-UFSAR 6.0-xa REV. 15, APRIL 2004 6.2-23 Pressure Response for Feedwater Line Break 6.2-24 Pressure Histories of Nodes for Worst Cases 6.2-25 Pressure Differential for Nodes of the Worst Break Cases 6.2-26 Vessel Liquid Blowdown Rate (at 3434 MWt) 6.2-27 Vessel Steam Blowdown Rate (at 3434 MWt) 6.2-28 Main Steamline Break Response Pa rameters Blowdown Flow (at 3434 MWt) 6.2-29 Temperature Response of Reactor Vessel (at 3434 MWt) 6.2-30 Sensible Energy Transient in the Reactor Vessel and Internal Metals (at 3434 MWt) 6.2-31 Containment Valve Arrangements 6.2-32 Energy Release Rates as a Function of Time 6.2-33 Integrated Energy Release as a Function of Time 6.2-34 Integrated Hydrogen Production as a Function of Time 6.2-35 Uncontrolled Hydrogen and Oxygen Generation 6.2-36 Hydrogen Concentration with 125 SCFM

LSCS-UFSAR FIGURES (Cont'd)

NUMBER TITLE 6.0-xi REV. 15, APRIL 2004 6.2-37 Nodalization Overlay For Recirculation Line Break 6.2-38 Nodalization Overlay For Feedwater Line Break 6.2-39 Nodalization For Original Recirculation Line Break Analysis 6.2-40 "Equivalent" Nodalization (Case A) 6.2-41 Azimuthal Pressure Distribution (At C Recirculation Outlet Nozzle) Original Data and Case A 6.2-42 Axial Pressure Distribution Original Data and Case A 6.2-43 Simplified Nodalization (Case B) 6.2-44 Azimuthal Pressure Distribution (At C Recirculation Outlet Nozzle) Case A and Case B 6.2-45 Axial Pressure Distribution Case A and Case B 6.2-46 Complex Nodalization (Case C) 6.2-47 Azimuthal Pressure Distribution (At C Recirculation Outlet Nozzle) Case A And Case C 6.2-48 Axial Pressure Distribution (Case A and Case C) 6.2-49 Axial Pressure Distribution at t = 0.500 Seconds 6.2-50 Circumferential Pressure Di stribution at t = 0.500 Seconds 6.2-51 Axial Pressure Distribution at t = 0.500 Seconds (Case C) 6.2-52 Circumferential Pressure Distribution at t = 0.500 Seconds (Case C) 6.3-1 HPCS System Process Diagram 6.3-2 Vessel Pressure vs. HPCS Flow Assumed in SPC and GE LOCA Analyses 6.3-3 HPCS Pump Characteristics 6.3-4 LPCS System Process Diagram 6.3-5 Vessel Pressure vs. LPCS Flow Assumed in SPC and GE LOCA Analyses 6.3-6 LPCS Pump Characteristics 6.3-7 Vessel Pressure vs. LPCI Flow Assumed in SPC and GE LOCA Analyses 6.3-8 Residual Heat Removal System (RHR) 6.3-9 LPCI Pump Characteristics 6.3-10 HPCS Minimum Required Pump Head to Meet LOCA Analysis Assumptions 6.3-11 LPCS Minimum Required Pump Head to Meet LOCA Analysis Assumptions 6.3-12 LPCI Minimum Required Pump Head to Meet LOCA Analysis Assumptions 6.3-13 Upper plenum pressure as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-14 Total Break Flow as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-15 Core inlet flow as a function of time during blowdown from RELAX. (1.0 DEG Suction, Sp-LPCS/DG) 6.3-16 Core outlet flow as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG)

LSCS-UFSAR FIGURES (Cont'd)

NUMBER TITLE 6.0-xii REV. 15, APRIL 2004 6.3-17 Lower downcomer mixture level as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-18 Lower plenum liquid mass as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-19 Hot channel high power node quality as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-20 Hot channel high power node heat transfer coefficient as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-21 System pressure as a function of time from FLEX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-22 Lower plenum mixture level as a function of time during refill/reflood from FLEX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-23 Relative entrainment as a function of time during refill/reflood from FLEX. 6.3-24 Core entrained liquid flow as a function of time during refill/reflood from FLEX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-25 ADS flow as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-26 LPCI flow as a functi on of time during blowdo wn from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-27 LPCS flow as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-28 HPCS flow as a function of time during blowdown from RELAX. (1.0 DEG Suction, SF-LPCS/DG) 6.3-29 Peak cladding temperature as a f unction of time from HUXY. (1.0 DEG Suction, SF-LPCS/DG) 6.3-30 Upper plenum pressure as a function of time during blowdown from RELAX.

(1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-31 Total Break Flow as a function of time during blowdown from RELAX.

(1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-32 Core inlet flow as a function of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-33 Core outlet flow as a function of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-24 Lower downcomer mixture level as a function of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-35 Lower plenum liquid mass as a function of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-36 Hot channel high power node quality as a function of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-37 Hot channel high power node heat transfer coefficient as a function of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG)

LSCS-UFSAR FIGURES (Cont'd)

NUMBER TITLE 6.0-xiii REV. 15, APRIL 2004 6.3-38 System pressure as a function of time from FLEX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-39 Lower plenum mixture level as a function of time during refill/reflood from FLEX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-40 Relative entrainment as a function of time during refill/reflood from FLEX.

(1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-41 Core entrained liquid flow as a function of time during refill/reflood from FLEX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-42 ADS flow as a function of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-43 LPCI flow as a functi on of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-44 LPCS flow as a function of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-45 HPCS flow as a function of time during blowdown from RELAX. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-46 Peak cladding temperature as a function of time from HUXY. (1.1 ft 2 Discharge, SF-HPCS/DG) 6.3-47 Schematic of the Therma l Overload Bypass Circuitry 6.3-48 DELETED 6.3-49 DELETED 6.3-50 DELETED

6.3-51 DELETED 6.3-52 DELETED 6.3-53 DELETED 6.3-54 DELETED 6.3-55 DELETED 6.3-56 DELETED 6.3-57 DELETED

6.3-58 DELETED 6.3-59 DELETED 6.3-60 DELETED 6.3-61 DELETED 6.3-62 DELETED 6.3-63 DELETED 6.3-64 DELETED

6.3-65 DELETED 6.3-66 DELETED 6.3-67 DELETED 6.3-68 DELETED 6.3-69 DELETED LSCS-UFSAR FIGURES (Cont'd)

NUMBER TITLE 6.0-xiv REV. 15, APRIL 2004 6.3-70 DELETED 6.3-71 DELETED 6.3-72 DELETED 6.3-73 DELETED 6.3-74 DELETED 6.3-75 DELETED 6.3-76 DELETED 6.3-77 DELETED

6.3-78 DELETED 6.3-79 DELETED 6.3-80 Post-LOCA Time-Pressure in Secondary Containment (Based on One SGTS Equipment Train Operating) 6.4-1 Control and Auxiliary Elec tric Equipment Room Layout 6.4-2 Location of Outside Air Intakes 6.4-3 Control Room Shielding Model 6.7-1 DELETED 6.7-2 DELETED 6.7-3 DELETED DRAWINGS CITED IN THIS CHAPTER*

DRAWING* SUBJECT

M-89 P&ID Standby Gas Treatment System, Units 1 and 2 M-94 P&ID Low Pressure Core Spray (LPCS) System, Unit 1 M-95 P&ID High Pressure Core Spray (HPCS) System, Unit 1 M-100 P&ID Control Rod Drive Hydraulic Piping System, Unit 1 M-130 P&ID Containment Combustible Gas Control System M-140 P&ID Low Pressure Core Spray (LPCS) System, Unit 2 M-141 P&ID High Pressure Core Spray (HPCS) System, Unit 2 M-146 P&ID Control Rod Drive Hydraulic Piping System, Unit 2 M-1443 P&ID Control Room Air Conditioning System M-1468 P&ID Refrigerant Piping Control Room HVAC System M-3443 HVAC C&I Details Control Room Air Conditioning System

• The listed drawings are included as "General References" only; i.e., refer to the drawings to obtain additional detail or to obtain background information. These drawings are not part of the UFSAR. They are controlled by the Controlled Documents Program.

LSCS-UFSAR 6.0-1 REV. 13 CHAPTER 6.0 - ENGINEERED SAFETY FEATURES

The engineered safety features of LaSa lle County Station are those systems whose actions are essential to a safety action required to mitigate the consequences of postulated accidents. The features can be divided into five general groups as follows: containment system s, emergency core cooling systems (ECCS), habitability systems, fission product removal and control systems and other systems. The LSCS engineered safety features, listed by th eir appropriate general grouping, are given below:

GROUP SYSTEM

Containment Systems

Primary Containment Secondary Containment Containment Heat Removal System Combustible Gas Control System Containment Isolation System

Emergency Core Cooling System High-Pressure Core Spray System (HPCS)

Low-Pressure Core Spray System (LPCS)

Low-Pressure Coolant Injection System (LPCI)

Automatic Depressurization System (ADS)

Habitability Systems Control Room HVAC Fission Product Removal and Control Systems

Standby Gas Treatment System Emergency Make-Up Air Filter System LSCS-UFSAR 6.0-2 REV. 13 GROUP SYSTEM Other Systems Main Steamline Isolation Valve Isolated Condenser Leakage Treatment Method

LSCS-UFSAR 6.1-1 REV. 13 6.1 ENGINEERED SAFETY FEATURE MATERIALS The materials utilized in the LSCS engineered safety feature systems have been selected on the basis of an engineering review and evaluation for compatibility with:

a. the normal and accident service conditions of the (engineered safety feature) ESF system, b. the normal and accident environmental conditions associated with the ESF system, c. the maximum expected normal and accident radiation levels to which the ESF will be subjected, and

d. other materials to preclude material interactions that could potentially impair the operation of the ESF systems.

The materials selected for the ESF systems ar e expected to function satisfactorily in their intended service without adverse effects on the service, performance or operation of any ESF.

6.1.1 Metallic Materials

In general, all metallic materials used in ESF systems comply with the material specifications of Section II of the ASME Boiler and Pressure Vessel Code.

Pressure-retaining materials of the ESF systems comply with the stringent quality requirements of their applicable quality group classification and ASME B&PV Code, Section III classification. Adherence to these requirements assures materials of the highest quality for the ESF systems. In those cases where it is not possible to adhere to the ASME material specifications, metallic materials have been selected in compliance with other nationally recognized standards, e.g., ASTM, where practicable, or chosen in compliance with current industry practice.

6.1.1.1 Materials Selection and Fabrication Metallic materials in ESF systems have, in general, been designed for a service life of 40 years, with due consideration of the effects of the service conditions upon the properties of the material, as required by Section III of the ASME B&PV Code, Article NC-2160.

Pressure retaining components of the ECCS have been designed with the following corrosion allowances, in compliance with the general requirement of Section III of the ASME B&PV Code, Article NC-3120:

a. Ferritic Materials LSCS-UFSAR 6.1-2 REV. 13

1. water service 0.08 inches

2. steam service 0.120 inches

b. Austenitic Materials 0.0024 inches For ESF systems other than ECCS, appropriate corrosion allowances, considering the service conditions to which the material will be subjected, have been applied.

The metallic materials of the ESF syst ems have been evaluated for their compatibility with core and containment spray solutions. No radiolytic or pyrolytic decomposition of ESF material will occur during accident conditions, and the integrity of the containment or function of any other ESF will not be effected by the action of core or containment spray solutions.

Material specification for the principal pressure-retaining ferritic, austenitic, and nonferrous metals in each ESF component ar e listed in Table 6.1-1. Materials that would be exposed to the core cooling water and containment sprays in the event of a loss-of-coolant accident are identified in th is table. Sensitization of austenitic stainless steel is prevented by the following actions:

a. Design specifications for austenitic stainless steel components require that the material be cleaned using halide free cleaning solutions and that special care be exercised in the fabrication, shipment, storage, and construc tion to avoid contaminants.

b. Design specifications call for ASME material, which is to be supplied in the solution annealed condition.

c. Design specifications prohibit the use of materials that have been exposed to sensitizing temperatures in the range of 800° F to 1500° F.

Cold-worked austenitic stainless steels wi th yield strengths greater than 90,000 psi are not utilized in ESF systems. Therefore, there are no compatibility problems with core cooling water or the containment sprays.

Metallic reflective thermal insulation is used exclusively inside the primary containment. Premoulded non-hydrophobic Microtherm MPS Insulation with the water resistant Agricoat coating enclosed in a 24 gauge stainless steel jacket is installed on the Unit 2 RVWLIS piping, 2BN86A-3/4" and 2NB88A-3/4", and the main steam high-flow instrument piping, 2MSC6AD-3/4" inside primary containment. Premoulded non-hydrophobic Microtherm MPS insulation enclosed in LSCS-UFSAR 6.1-3 REV. 14, APRIL 2002 24 gauge stainless steel jacket is insta lled on Unit 1 RVWLIS piping 1NB09A-2", 1NB09B-1", 1NB88A-1", 1NB24A-2", and 1NB24B-1", and the main steam high-flow instrument piping, 1MSC6AK-3/4", inside primary containment. The aforementioned Microtherm Insulation is also installed on the Unit 1 main steam high-flow instrument piping, 1MSC6AK-3/4", inside primary containment.

ARMAFLEX insulation is installed on the chilled water system inside primary containment.

Outside containment, calcium silicate or an engineering approved alternative thermal insulation is utilized. Design specifications on the nonmetallic insulation require that it be in accordance with Regulatory Guide 1.36, in order to avoid the possibility of chloride induced stress corrosion cracking in austenitic stainless steel in contact with the insulation.

To avoid hot cracking (fissuring) during weld fabrication and assembly of austenitic stainless steel components of the ESF, the design specifications require the following:

a. Maximum delta ferrite content for wrought and duplex cast components is 5% - 15%.

b. Chemical analyses are performed on undiluted weld deposits, or alternately, on the wire, consumable insert, etc., to verify the delta ferrite content.

c. Delta ferrite content in weld metal is determined using magnetic measurement devices.

d. Maximum interpass temperatur e shall not exceed 350°F during welding. e. Test results as discussed above are included in the qualification test report.

f. Weld materials meet the re quirements of Section III.

g. Production welds are examined to verify that the specified delta-ferrite levels are met.

h. Welds not meeting these leve ls are unacceptable and must be removed.

LSCS-UFSAR 6.1-4 REV. 14, APRIL 2002 6.1.1.2 Composition, Compatibility and Stability of Containment and Core Spray Coolants The core sprays have two possible sources of coolant. The HPCS system is supplied from either the cycled condensate storage tank or the suppression pool. The normal source of water for HPCS is the suppression pool. The capability remains for the HPCS system to draw a suction on the cycl ed condensate tank because the piping to the tank is installed, but isolated by a b lind flange. Establishment of this flowpath is under administrative control. The LPCS and LPCI are supplied from the suppression pool only. Water quality in both of these sources is maintained at a high level of purity with the possible exception of potentially high soluble-iron metallic impurities. Additional discussion of the water qualities are given in Subsections 6.1.3, 9.2.7, and 9.2.11. Limited corrosion inhibitors or other additives (such as zinc and noble metals) are present in either source.

The containment spray utilizes the suppression pool as its source of supply. No radiolytic or pyrolytic decomposition of ESF materials are induced by the

containment sprays. The containment sprays should not be a source of stress-corrosion cracking in austenitic stainless steel during a LOCA.

6.1.2 Organic Materials

Table 6.1-2 lists all the organic compounds that exist within the containment in significant amounts. All these materials in ESF components have been evaluated with regard to the expected service conditions, and have been found to have no adverse effects on service, performance, or operation.

The dry well liner and coated exposed metal surfaces inside containment are prime coated with an inorganic zinc compound that has been fully qualified in accordance with ANSI standards N101.2, N101.4, an d N512 , with the exception of a small quantity (44 gallons) used on pipe hangers and snubber attachments and recirculating pump motors. Uncoated metal surfaces shall be evaluated for acceptability. No radiolytic or pyrolytic decomposition or interaction with other ESF materials will occur.

6.1.3 Postaccident Chemistry

The post-accident chemical environment inside the primary containment will consist of water from the suppression pool and the cycled condensate storage tank, i.e. water sources for the high pressure core spray, low pressure core spray, low pressure core injection, reactor core isolation cooling and containment spray. The

suppression pool may contain trace amounts of corrosion inhibiting chemicals such as hydrogen, zinc and noble metals. Additionally, portions of the Reactor Building Closed Cooling Water (RBCCW) system and the Primary Containment Chilled Water (PCCW) system are inside the containment. Both systems contain limited LSCS-UFSAR 6.1-5 REV. 14, APRIL 2002 amounts of corrosion inhibitors, and have portions of their piping inside containment classified as Seismic Category 2. During a Design Basis Accident (DBA) either or both of these systems can fail and release the corrosion inhibitors to the suppression pool before isolation. Due to the limited quantity (trace amounts) of these chemicals in the secondary systems and the dilution factor as a result of a DBA, the water will be approximately neutral (pH = 7), and there will be no adverse affect to equipment, coatings or other materials during ECCS or RCIC operation.

LSCS-UFSAR TABLE 6.1-1 (SHEET 1 OF 5) TABLE 6.1-1 REV. 13 PRINCIPAL PRESSURE-RETAINING MATERIAL FOR ESF COMPONENTS I. Containment Systems A. Primary Containment

1. Containment Walls 4500 psi Concrete *2. Drywell Liner SA-516, Grade 60 *3. Suppression Chamber Liner SA-240, Type 304

• 4. Drywell Head SA-516, Grade 70 *5 Penetrations

a. Drywell SA-333, Grade 1 or 6 (Seamless) b. Suppression Chamber SA-312, Grade TP 304 (Seamless) *6. Equipment Hatch SA-516, Grade 70 *7. Personnel Access Hatch

a. Drywell SA-516, Grade 70 b. Suppression Chamber SA-240, Type 304 *8. Suppression Vent Downcomers SA-240, Type 304 *9. Vacuum Relief Piping a. Drywell to Suppression Chamber Penetration SA-106, Grade B b. Suppression Chamber Penetration SA-312, Grade TP 304 (Seamless) 10. Vacuum Relief Valves SA-105

• Indicates that material may be subjected to containment spray or core cooling water in the event of a loss-of-coolant accident.

LSCS-UFSAR TABLE 6.1-1 (SHEET 2 OF 5) TABLE 6.1-1 REV. 13

• 11. Pressure Retaining Bolts a. Drywell SA-320, Grade L43 SA-193, Grade B7 SA-194, Grade 7 b. Suppression Chamber SA-193, Class 2, Grade B8C, Type 347 SA-194, Class 2, Grade 83, Type 347 B. Secondary Containment 1. Ducts A-526 2. Dampers A-285, Grade B A-181, Grade 1 C. Containment Heat Removal System 1. RHR Pumps A-516, Grade 70 2. RHR Heat Exchanger

a. Shell Side SA-516, Grade 70 b. Tube Side SA-249, Grade TP 304L *3. Piping SA-106, Grade B *4. Valves SA-216, Grade WCB or SA-105 *5. Pressure-Retaining Bolting SA-193, Grade B7

• 6. Welding Material SFA-5.18E70S-3(F-6, A-1) D. Containment Isolation System

• 1. Piping SA-106, Grade B or SA-312, Grade TP 304 *2. Valves SA-216, Grade WCB or SA-105 or SA-182, Grade 316L or Grade F316 or SA-351, Grade C8FM or SA-351 Grade CF3

• Indicates that material may be subjected to containment spray or core cooling

water in the event of a loss-of-coolant accident.

LSCS-UFSAR TABLE 6.1-1 (SHEET 3 OF 5) TABLE 6.1-1 REV. 13

• 3. Pressure-Retaining Bolting SA-193, Grade B7 *4. Welding Material SFA-5.18E70S-3 (F-6, A-1)

E. Combustible Gas Control System

1. Piping SA-106, Grade B

2. Valves SA-216, Grade WCB

3. Recombiner SA-358, Grade 304

4. Blower 5. Pressure-Retaining Bolting SA-193, Grade B7 6. Welding Material SFA-5.18E70S-3 (F-6, A-1)

II. Emergency Core Cooling System A. High-Pressure Core Spray 1. Pump A-516, Grade 70 2. Piping

• a. Inside Reactor Building SA-106, Grade B

b. Outside Reactor Building SA-409, Grade TP 304

• 3. Valves SA-216, Grade WCB or SA-105 *4. Pressure-Retaining Bolting SA-193, Grade B7

• 5. Welding Materials SFA-5.18E70S-3 (F-6, A-1)

B. Low-Pressure Core Spray

1. Pump A-516, Grade 70 *2. Piping SA-106, Grade B *3. Valves SA-216, Grade WCB or SA-105

• Indicates that material may be subjected to containment spray or core cooling water in the event of a loss-of-coolant accident LSCS-UFSAR TABLE 6.1-1 (SHEET 4 OF 5) TABLE 6.1-1 REV. 13

• 4. Pressure-Retaining Bolting SA-193, Grade B7 *5. Welding Materials SFA-5.18E70S-3 (F-6, A-1)

A. Low-Pressure Coolant Injection 1. RHR Pump A-516, Grade 70

• 2. Piping SA-106, Grade B

• 3. Valves SA-216, Grade WCB or SA-105

• 4. Pressure-Retaining Bolting SA-193, Grade B7

• 5. Welding Materials SFA-5.18E70S-3 (F-6, A-1) B Automatic Depressurization System *1. Piping

a. Inlet SA-155, Grade KCF70

b. Outlet SA-106, Grade B

• 2. Valves

III. Habitability System A. Blowers A-283, A-242 B. Dampers A-285, Grade B A-181, Grade 1 C. Ducts A-526 D. Housing A-36 IV. Fission Product Removal and Control System A. Standby Gas Treatment System

1. a. Piping (Downstream of Filter Unit)

SA-106, Grade B b. Piping (Upstream of Filter Unit)

A-106, Grade B 2. Housing A-36 *Indicates that material may be subjected to containment spray or core cooling water in the event of a loss-of-coolant accident.

LSCS-UFSAR TABLE 6.1-1 (SHEET 5 OF 5) TABLE 6.1-1 REV. 13

3. Valves SA-216, Grade WCB or SA-105, or SA-516, Grade 7 4. Dampers A-285, Grade B A-181, Grade 1 5. Blowers A-283, A-242 6. Pressure-Retaining Bolting a. Pressure-Retaining Bolting (Downstream of Filter Unit)

SA-193, Grade B7 b. Pressure-Retaining Bolting (Upstream of Filter Unit)

A-193, Grade B7 7. Welding Materials SFA-5.18E70S-3 (F-6,A-1) B. Emergency Air Filter System 1. Ducts A-526 2. Dampers A-285, Grade B A-181, Grade 1 3. Housing A-36 4. Blower A-283, A-242 V. Other Systems A. Main Steamline Isolation Valve Leakage Control System (Deleted)

• Indicates that material may be subjected to containment spray or core cooling water in the event of a loss-of-coolant accident LSCS-UFSAR TABLE 6.1-2 (SHEET 1 OF 2) TABLE 6.1-2 REV. 13 ORGANIC MATERIALS WITHIN THE PRIMARY CONTAINMENT MATERIAL USE QUANTITY Acrylomitrile Butadiene/PVC Foam Rubber ARMAFLEX Insulation on the Chilled Water Piping Throughout Drywell Chlorosulfinated Polyethylene (Hypalon)

Low Voltage Electrical

Power Cable Jacketing and Insulation Material Throughout Drywell Etylene Propylene Rubber (EPR) Low Voltage Electrical

Power Cable Jacketing and Insulation Material Throughout Drywell High Temperature Ethylene Propylene Medium Voltage Electrical

Power Cable Jacketing and

Insulation Material Throughout Drywell Hypalon/Hypalon Instrumentation Cable

Insulation/Jacketing Material Throughout Drywell EPR/Hypalon Instrumentation Cable Insulation/Jacketing Material Throughout Drywell Agricoat Water Resistant Coating on the Premoulded non-hydrophobic Microtherm MPS Insulation 25.8 ft 2 - Unit 2 Cross-Linked Polyolefin/Alkaneimide

Polymer Instrumentation Coaxial and Triaxial Insulation/

Jacketing Material Throughout Drywell Modified Phenolic Coating for Exposed

Carbon Steel Surfaces 16 ft 3 Modified Phenolic Surfacer Coating for Exposed Concrete Surfaces 17 ft 3 Modified Phenolic Finish Coating for Exposed Concrete Surfaces 5 ft 3 LSCS-UFSAR TABLE 6.1-2 (SHEET 2 OF 2) TABLE 6.1-2 REV. 17, APRIL 2008 MATERIAL USE QUANTITY Alkyd Primer and Finish Pipe hangers and Snubber Attachments

and GE Recirculating Pump 44 gal. Lube Oil Reactor Recirculation Pump Motor (2 motors/unit) 145 gal. in Unit 1 120 gal. in Unit 2 Silicone Fluid (SF 1147, GE) MSIV Hydraulic Fluid (4

valves within containment) 1 1/2 gal. per valve Non-separating high temperature grease Drywell cooling area

coolers < 1 gal.

Fyrquel EHC Recirculation Control

Valve Hydraulic Fluid (2 valves) 118 gal. per valve Silicone Fluid Lisega Hydraulic Snubbers

< 1 1/2 gal. per snubber Fiberglass Reinforced

Silicone Fabric 1 (2) RF01 and 1 (2) RE02 Sump Cover Mat 400 ft 2 per unit Silicone Sealant 1 (2) RF01 and 1 (2) RE02 Sump Cover Mat < 1 gal. per unit

LSCS-UFSAR 6.2-1 REV. 13 6.2 CONTAINMENT SYSTEMS 6.2.1 Containment Functional Design

This section establishes the design bases for the primary containment structure, describes the major design features of the structure, and presents an evaluation of the capacity of the containment to perform its required safety function during all normal and postulated accident conditions described in this UFSAR.

6.2.1.1 Containment Structure

6.2.1.1.1 Design Bases The primary containment structure has been designed to meet the following safety design bases:

a. Containment Vessel Design

1. The containment structure has the capability to withstand the peak transient pressures and temperatures that could occur due to the postulated design-basis accident (DBA).

2. The containment has the capability to maintain its functional integrity indefinitely after the postulated DBA.

3. The containment structure also withstands the peak environmental transient pressures and temperatures associated

with the postulated small line break inside the drywell.

4. The containment structure has also been designed to withstand the coincident fluid jet forces associated with the flow from the postulated rupture of any pipe within the containment.

5. The containment has also been designed to withstand the hydrodynamic forces associated with a DBA and safety-relief valve discharge, as described in the LaSalle Design Assessment

Report. Design loading combinat ions are also described in the design assessment report: Design pressure and temperature, and the major containment design parameters are listed in Table 6.2-1.

b. Containment Subcompartment Design The internal structures of the containment have been designed to accommodate the peak transient pressures and temperatures LSCS-UFSAR 6.2-2 REV. 13 associated with the postulated design-basis accident (DBA). The effects of subcompartment pressuri zation for the postulated pipe ruptures have been evaluated. Subcompartment pressurization is more fully discussed in Subsection 6.2.1.2.

c. Containment Internals Design The drywell floor has been designed to withstand a downward acting differential pressure of 25 psig in combination with the normal operating loads and safe shutdown earthquake (SSE). The drywell floor has also been designed to accommodate an upward acting deck differential pressure of 5 psig, in order to account for the wetwell pressure increase that could occur after a loss-of-coolant accident (LOCA). d. Containment Design for Mass and Energy Release

1. The maximum postulated rele ase of mass and energy to the containment is based upon the instantaneous circumferential rupture of a 24- inch reactor recirculation line or a 26-inch main

steamline.

2. The effects of metal-water reactions and other chemical reactions following the DBA can be accommodated in the

containment design.

e. Energy Removal Features The RHR system, through the containment cooling mode, is utilized to remove energy from the containment following a LOCA by circulating the suppression pool water throug h a residual heat removal (RHR) heat exchanger for cooling, and returning the water to the pool through the low-pressure core injection (LPCI) in the reactor pressure vessel (RPV) or the suppression chamber spray header. The containment spray mode of the RHR system can also be utilized to condense steam and reduce the temperature in the drywell following a LOCA. A more detailed description is available in Subsection 6.2.2. The RHR containment cooling mode energy removal capability is not affected by a single failure in the system, sinc e a completely redundant loop is available to perform this functi on. Two redundant loops of the containment spray system are also provided.

LSCS-UFSAR 6.2-3 REV. 13 f. Pressure Reduction Features The containment vent system directs the flow from postulated pipe ruptures to the pressure suppression pool, and distributes such flow

uniformly throughout the pool, to condense the steam portion of the flow rapidly, and to limit the pressure differentials between the drywell and wetwell during various postaccident cooling modes.

g. Hydrostatic Loading Design The containment design permits filling the containment system drywell with water to a level 1 foot below the refueling floor to permit removal of fuel assemblies during postaccident recovery.

h. Impact Loading Design The containment system is protected against missiles from internal or external sources and excessive motion of pipes that could directly or indirectly jeopardize containment integrity.

i. Containment Leakage Design The containment limits leakage duri ng and following the postulated DBA to values less than leakage rates that would result in offsite doses greater than 10 CFR 100.

j. Containment Leakage Testability It is possible to conduct periodical leakage tests as may be appropriate to confirm the integrity of the containment at calculated peak pressure resulting from the postulated DBA.

For the purposes of the containment structure design, the design-basis accident (DBA) is defined as a mechanical failure of the reactor primary system equivalent to the circumferential rupture of one of the recirculation lines. During the DBA, the long-term peak suppression pool temperature shall not exceed the design temperature.

6.2.1.1.2 Design Features

The primary containment is a concrete stru cture with the exception of the drywell head and access penetrations, which are fabricated from steel. The major components are shown in Figure 3.8-1. The concrete is designed to resist all loads associated with the design-basis accident.

LSCS-UFSAR 6.2-4 REV. 15, APRIL 2004 The primary containment walls have a steel liner, which acts as a low leakage barrier for release of fission products.

The walls of the primary containment are posttensioned concrete; the base mat is conventional reinforced concrete. The dividing floor between the drywell and suppression chamber is conventional reinforced concrete and is supported on a cylindrical base at its center, on a seri es of concrete co lumns and from the containment wall at the periphery of the slab.

The drywell floor is rigidly connected to the primary containment wall. A full moment and shear connection is provided by dowels and shear lugs welded to the reinforced liner plate as shown in Figure 3.

8-4. The thermal expansion is accounted for in the containment design; the resulting forces and moments are accommodated within the allowable stress limits.

The primary containment walls support the reactor building floor loads and, in addition, also serve as the biological shield. A detailed discussion of the structural design bases is given in Chapter 3.0. The codes, standards, and guides applied in the design of the containment structure and internal structures are identified in Chapter 3.0.

The walls of the primary containment st ructure are posttensioned, using the BBRV system of posttensioning utilizing parallel lay, unbonded type tendons. The tendons are fabricated from 90 one-quarter inch diameter, cold drawn, stress relieved, prestressing grade wire. Each tendon is encased in a conduit. The walls are prestressed both vertically and horizontally for floor elevations below 820 feet. The horizontal tendons are placed in a 240

° system using three buttresses as anchorages with the tendons staggered so that two-thirds of the tendons at each buttress terminate at that buttress. For floor elevations above 820 feet, the horizontal tendons are placed in a 360

° system using two buttresses as anchorages. Access to the tendon anchorages is maintained to allow for periodic inspection. For a typical layout of hoop tendons, see Fi gure 3.8-11. A typical layout of the vertical tendons is illustrated in Figure 3.8-11.

All liner joints have full penetration welds. The field welds have leaktightness testing capability by having a small steel channel section welded over each liner weld. Fittings are provided in the channel for leak testing of the liner welds under pressure. The actual containment leakag e boundary during normal operation and accident conditions consists of the liner an d liner joint butt welds when the leak test channel is vented to the containment atmosphere and the combined containment liner, liner joint butt welds, containment liner leak test channels, channel fillet welds and the leak test connections when the leak test channel test connection plugs are installed. The liner anchorage system considers the effects of temperature, negative pressure, prestressing, and stress transfer around penetrations.

LSCS-UFSAR 6.2-5 REV. 13 Drywell The drywell is a steel-lined posttensioned concrete vessel in the shape of a truncated cone having a base diameter of approximately 83 feet and a top diameter of 32 feet.

The floor of the drywell serves both as a pressure barrier between the drywell and suppression chamber and as the support structure for the reactor pedestal and downcomers. The drywell head is bolted at a steel ring girder attached to the top of the concrete containment wall and is sealed with a double seal. The double seal on the head flange provides a plenum for determining the leaktightness of the bolted connection. The base of the ring serves as the top anchorage for the vertical prestressing tendons and the top of the ring serves as anchorage for the drywell head. The drywell houses the reactor and its asso ciated auxiliary systems. The primary function of the drywell is to contain the effects of a design-basis recirculation line break and direct the steam released from a pipe break into the suppression chamber pool. The drywell is designed to resist the forces of an internal design pressure of 45 psig in combination with thermal, seismic, and other forces as outlined in Chapter 3.0.

The drywell is provided with a 12-foot diameter equipment hatch for removal of equipment for maintenance and an air lock for entry of personnel into the drywell.

Under normal plant operations, the equipment hatch is kept sealed and is opened only when the plant is shut down for refueling and/or maintenance.

The equipment hatch is covered with a st eel dished head bolted to the hatch opening frame which is welded to the steel lin er. A double seal is utilized to ensure leaktightness when the hatch is subjected to either an internal or external pressure. The space between the double seal serves as a plenum for leak testing the hatch seal. The personnel air lock is a cylindrical inta ke welded to the steel liner. The double doors are interlocked to maintain containment integrity during operation.

All welds that make up the vapor barrier have test channels to permit leak testing of the welds: When the leak test channel test connections are plugged, the leak test channel is part of the vapor barrier.

The primary containment ventilation system, as described in Subsection 9.4.9, is provided to maintain drywell te mperatures at approximately 135

° F during normal plant operation.

LSCS-UFSAR 6.2-6 REV. 14, APRIL 2002 The primary containment vent and purge syst em, as described in Subsection 9.4.10, is designed to purge potentially radioactive gases from the drywell and suppression chamber prior to and during personnel access to the containment.

Containment penetration cooling is provided on high temperature penetrations through the primary containment wall by the reactor building closed cooling water system. The penetrations served by this system and the design basis for the cooling loads are described in Subsection 9.2.3.

Pressure Suppression Chamber and Vent System

The primary function of the suppression ch amber is to provide a reservoir of water capable of condensing the steam flow from the drywell and collecting the noncondensable gases in the suppression chamber air space. The suppression chamber is a stainless steel-lined posttensioned concrete vessel in the shape of a cylinder, having an inside diameter of 86 feet 8 inches. The foundation mat serves as the base of the suppression chamber. The suppression chamber is designed for the same internal pressure as the drywell in combination with the thermal, seismic, and other forces. The liner design and te sting are the same as covered previously within this subsection (6.2.1.1.1.2).

The entire suppression chamber is lined with stainless steel. The drywell floor support columns are also provided with a stainless steel liner on the outside surface. Two 36-inch diameter openings are provided for access into the suppression chamber for inspection. Under normal plant operation, these access openings are kept sealed. They are opened only when the plant is shut down for refueling and/or maintenance. The access openings are located in the cylindrical walls of the chamber 14 feet 2 inches above the suppression pool water level. The access openings are closed using a bolted steel hatch cover. The hatch cover is designed with a double seal and test plenum to ensure leaktightness.

The suppression chamber vent system consists of 98 downcomer pipes open to the drywell and submerged 12 feet 4 inches below the low water level of the suppression pool, providing a flow path for uncondensed steam into the water. Each downcomer has a 23.5-inch internal diameter. The downcomers project 6 inches above the drywell floor to prevent flooding from a broken line. Each vent pipe opening is shielded by a 1-inch thick steel deflecto r plate to prevent overloading any single vent pipe by direct flow from a pipe break to that particular vent. The principal parameters for design of the primary containment, suppression pool, reactor

building and the vent downcomers are listed in Table 6.2-1.

LSCS-UFSAR 6.2-7 REV. 14, APRIL 2002 Vacuum Relief System Vacuum relief valves are provided betwee n the drywell and suppression chamber to prevent exceeding the drywell floor negative design pressure and backflooding of the suppression pool water into the drywell.

In the absence of vacuum relief valves, drywell flooding could occur following isolation of a blowdown in the drywell. Condensation of blowdown steam on the drywell walls and structures could result in a negative pressure differential between the drywell and suppression chamber.

The vacuum relief valves are designed to equalize the pressure between the drywell and wetwell air space regions so that the reverse pressure differential across the

diaphragm floor will not exceed the design value of five pounds per square inch.

The vacuum relief valves (four assemblies) are outside the primary containment and form an extension of the primary co ntainment boundary. The vacuum relief valves are mounted in special piping which connects the drywell and suppression chamber, and are evenly distributed around the suppression chamber air volume to prevent any possibility of localized pressure gradients from occurring due to

geometry. In each vacuum breaker asse mbly, two local manual butterfly valves, one on each side of the vacuum breaker, are provided as system isolation valves should failure of the vacuum breaker occur.

The vacuum relief valves are instrumented with redundant position indication and are indicated in the main control room. The valves are provided with the capability for local manual testing. The position indication requirements for the vacuum relief valves are located in the Administrative Technical Requirements. (References 21, 22, and 23)

This design provides adequate assurance of limiting the differential pressure between the drywell and suppression cham ber and assures proper valve operation and testing during normal plant operation.

No vacuum relief valves are provided between the drywell and the reactor building atmosphere. The concrete containment structure has the ability to accommodate subatmospheric pressures of approximately 5 psi absolute.

6.2.1.1.3 Design Evaluation

The key design parameters for the pressure suppression containment being provided for the LaSalle County Statio n (LSCS) are listed in Table 6.2-1.

These design parameters are not determined from a single accident event but from an envelope of accident conditions. As a result, there is no single design-basis accident (DBA) for this containment system.

LSCS-UFSAR 6.2-8 REV. 15, APRIL 2004 The containment system was analyzed originally at 3434 MWt reactor power. Since then, the containment system evaluation was performed for a reactor power of 3559 MWt by analyzing the limiting events at this power level. The results for 3559 MWt power are included in this section, while keeping most of the original analysis

results for 3434 MWt power as a referenc e analysis for historical purposes.

A maximum drywell and suppression chamber pressure of 39.6 psig and 30.6 psig, respectively is predicted near the end of the blowdown phase of a loss-of-coolant accident (LOCA) transient. Approximately the same peak pressure occurs for either the break of a recirculation line or a main steamline. Both accidents are evaluated at 3434 MWt.

For 3559 MWt reactor power, the maximum cont ainment pressure is predicted to be 39.9 psig in the drywell and 27.9 psig in the suppression chamber for the recirculation line break. The main steamline break was not reevaluated for the uprated power level.

The most severe drywell temperature condition is predicted for a small primary system rupture above the reactor water level that results in the blowdown of reactor steam to the drywell. Based upon the thermodynamic conditions this would produce high temperature steam in the drywell.

In order to demonstrate that breaks smaller than the rupture of the largest primary system pipe will not exceed the containment design parameters, the blowdown phase of an intermediate size break is ev aluated. Containment design conditions are not exceeded for this or the other break sizes.

All of the analyses assume that the primary system and containment are at the maximum normal operating conditions. Re ferences are provided that describe relevant experimental verification of the analytical models used to evaluate the containment response.

Table 6.2-1 provides a listing of the key design parameters of the LSCS primary containment system including the design characteristics of the drywell, suppression chamber and the pressure suppression vent system.

Table 6.2-2 provides the performance parame ters of the related engineered safety feature systems which supplement the de sign conditions of Table 6.2-1 for containment cooling purposes during po staccident operation. Performance parameters given include those applicable to full capacity operation and to those reduced capacities employed for containment analyses.

LSCS-UFSAR 6.2-8a REV. 14, APRIL 2002 6.2.1.1.3.1 Accident Response Analysis

The containment functional evaluation performed at 3434 MWt is based upon the consideration of several postulated accident conditions resulting in release of reactor coolant to the containment. These accidents include:

a. an instantaneous guillotine rupture of a recirculation line, b. an instantaneous guillotine rupture of a main steam-line, c. an intermediate size liquid line rupture, and

d. a small size steamline rupture.

Energy release from these accidents is reported in Subsection 6.2.1.3.

LSCS-UFSAR 6.2-9 REV. 14, APRIL 2002 The accident response analysis based on the GE calculations remains applicable to and bounds the SPC ATRIUM-9B fuel. This is determined based on the containment response being dependent on the amount of energy in the system, the containment design, and the failure modes that allow the pressurization to occur rather than the fuel type. The amount of energy in the system is based on initial conditions and the assumed blowdown. As the blowdown assumed for the containment response analysis as shown in Tables 6.2-18 and 6.2-19 bound the blowdown predicted by the SPC LOCA methodology and results, le ss energy would be released to the containment using the SPC blowdown. For this reason SPC ATRIUM-9B fuel and

LOCA results are considered to be bound by the current GE accident response analysis results for the containment.

For 3559 MWt reactor power, the limiting even t, an instantaneous guillotine rupture of a recirculation line, was analyzed to perform the containment functional evaluation. The analysis at 3559 MWt was performed in accordance with the Generic Guidelines for General Electric Boiling Water Reactor Power Uprate, NEDC-31897P-A (Reference 24). This analysis employed essentially the same methodology, while taking a more detailed modeling approach for the reactor vessel blowdown evaluation. The analysis re sults for 3559 MWt reactor powe r are included at the end of this subsection under the heading "Evaluation at 3559 MWt Reactor Power," after a description of the original 3434 MWt analysis which is kept as a reference analysis for historical purposes.

6.2.1.1.3.1.1 Recirculation Line Rupture

The instantaneous guillotine rupture of a main recirculation line results in the maximum flow rate of primary system fluid and energy into the drywell as illustrated in Figure 6.2-1 by the diagram showing th e location of a recirculation line break.

Immediately following the rupture, the flow out of both sides of the break will be limited to the maximum allowed by critical fl ow considerations. Figure 6.2-1 shows a schematic view of the flow paths to the break. Flow in the suction side of the recirculation pump will correspond to critical flow in the 2.565 square foot pipe cross section. Flow in the discharge side of the recirculation pump will correspond to critical flow at the ten jet pump nozzles associated with the broken loop, providing an effective break area of 0.468 ft

2. In addition, there is a 4- inch cleanup line crosstie that will add 0.080 ft 2 to the critical flow area, yielding a total of 3.113 ft

2.

Assumptions for Reactor Blowdown The response of the reactor coolant syst em during the blowdown period of the accident is analyzed using the following assumptions:

LSCS-UFSAR 6.2-9a REV. 14, APRIL 2002

a. At the time the recirculation pipe breaks, the reactor is operating at the most severe condition that maximizes the parameter of interest; that is, primary containment pressure.

b. The recirculation line is considered to be severed instantly. This results in the most rapid coolant loss and depressurization, with coolant being discharged from both ends of the break.

c. The reactor is shut down at the time of accident initiation because of void formation in the core region. Scram also occurs in less than 1 LSCS-UFSAR 6.2-10 REV. 13 second from receipt of the high drywell pressure signal. The difference between shutdown at time zero and 1 second is negligible.

d. The vessel depressurization flow rates are calculated using Moody's critical flow model (Reference 1) assuming "liquid only" outflow, since this assumption maximizes the energy release to the containment: "Liquid only" outflow requires that all vapor formed in the RPV by bulk flashing rises to the surface rather than being entrained in the existing flow. Some of the vapor would be entrained and would significantly reduce the RPV discharge flow rates. Moody's critical flow model, which assumes annular, isentropic flow, thermodynamic flow, thermodynamic phase equilibrium, and maximized slip ratio, accurately predicts vessel outflows through small diameter orific es. However, actual rates through larger flow areas are less than the model indicates because of the effects of a near homogeneous two- phase flow pattern and phase nonequilibrium. This effect is in addition to the reduction caused by vapor entrainment, discussed previously.

e. The core decay heat and the sensible heat released in cooling the fuel to 545° F are included in the reactor pressure vessel depressurization calculation: The rate of ener gy release is calculated using a conservatively high heat transfer coefficient throughout the depressurization period. By maximizing the assumed energy release rate, the RPV is maintained at nearly rated pressure for approximately 20 seconds. The high RPV pressure increases the calculated blowdown flowrates; this is conservative for containment analysis purposes. With the RPV fluid temperature remaining near 545

° F, however, the calculated release of sensible energy stored below 545

° F is negligible during the first 20 seconds. The sens ible energy is released later, but does not affect the peak drywell pres sure. The small effect of sensible energy release on the long-term suppression pool temperature is included.

f. The main steam isolation valves are assumed to start closing at 0.5 seconds after the accident. They are assumed to be fully closed in the shortest possible time of 3 seconds following closure initiation.

Actually, the closure signal for the main steam isolation valves is expected to occur from low water leve l, so these valves may not receive a signal to close for more than 4 seconds, and the closing time could be as long as 5 seconds. By assuming rapid closure of these valves, the RPV is maintained at a high pressure, whic h maximizes the discharge of high energy steam and water into the primary containment: In addition, the rapid closure of the main steam isolation valves cuts off motive power to the steam-driven feedwater pumps.

LSCS-UFSAR 6.2-11 REV. 13 g. Reactor feedwater flow is assumed to stop instantaneously at time zero.

Since cooler feedwater flow tends to depressurize the RPV, thereby reducing the discharge of steam and water into the primary containment, this assumption is considered conservative and consistent with that of assumption f.

With respect to suppression pool temperature, this assumption has been supplemented with an additional evaluation. The purpose being to evaluate the suppression pool long term temperature response. For this evaluation, the feedwater is assumed to have been injected into the suppression pool, by the end of the recirculation piping break blowdown phase (at 600 seconds), in order to assess long term peak pool temperature. See paragraph enti tled "Evaluation of Post-LOCA Feedwater Injection" in this section.

h. A complete loss of offsite power occurs simultaneously with the pipe break. This condition results in th e loss of power conversion system equipment and also requires that all vital systems for long-term cooling be supported by onsite power supplies.

Assumptions for Containment Pressurization

The pressure response of the containment during the blowdown period of the accident is analyzed using the following assumptions:

a. Thermodynamic equilibrium exists in the drywell and suppression chamber. Since nearly complete mixing is achieved, the analysis assumes complete mixing, which is in the conservative direction.

b. The fluid flowing through the drywell-to-suppression chamber vents is formed from a homogeneous mixture of the fluid in the drywell. The use of this assumption results in complete liquid carry-over into the drywell vents. c. The fluid flow in the drywell-to-suppression chamber vents is compressible except for the liquid phase.

d. No heat loss from the gases inside the primary containment is assumed.

This adds extra conservatism to the analysis; that is, the analysis will tend to predict higher containment pressures than would actually result.

Assumptions for Long-Term Cooling

Following the blowdown period, the emergency core cooling systems (ECCS) discussed in Section 6.3 provide water for core flooding and long-term decay heat LSCS-UFSAR 6.2-12 REV. 13 removal. The containment pressure and temperature response during this period are analyzed using the following assumptions:

a. The LPCI pumps are used to flood the core prior to 600 seconds after the accident. The high-pressure core sp ray (HPCS) is assumed available for the entire accident.

b. After 600 seconds, the LPCI pump flow may be diverted from the RPV to the containment spray. This is a manual operation. Actually, the containment spray need not be activated at all to keep the containment pressure below the containment design pressure. Prior to activation of the containment cooling mode (arbit rarily assumed at 600 seconds after the accident), all of the LPCI pump flow will be used only to flood the core. c. The effect of decay energy, stor ed energy, and energy from the metal-water reaction on the suppression pool temperature are considered.

d. During the long-term containment response (after depressurization of the reactor vessel is complete) the suppression pool is assumed to be the only heat sink in the containment system.

e. After approximately 600 seconds, the RHR heat exchangers are activated to remove energy from the containment via recirculation cooling from the suppression pool wi th the RHR service water systems.

f. The performance of the ECCS equipment during the long-term cooling period is evaluated for each of the following three cases of interest:

Case A - Offsite Power Available All ECCS equipment and containment spray operating.

Case B - Loss of Offsite Power Minimum diesel power available for ECCS and containment spray.

Case C - Same as Case B (except no containment spray) Initial Conditions for Accident Analyses Table 6.2-3 provides the initial reactor co olant system and containment conditions used in all the accident response evaluation

s. The tabulation includes parameters for the reactor, the drywell, the suppressi on chamber and the vent system. A supplementary safety evaluation has also been performed, as discussed in LSCS-UFSAR 6.2-13 REV. 13 Section 6.2.1.8, to evaluate an increase in the initial suppression pool temperature value to 105

° F. Table 6.2-4 provides the initial conditio ns and numerical values assumed for the recirculation line break accident as well as the sources of energy considered prior to the postulated pipe rupture. The assumed conditions for the reactor blowdown are also provided.

The mass and energy release sources and rates for the containment response analyses are given in Subsection 6.2.1.3. Short-Term Accident Response The calculated containment pressure and temperature responses for the recirculation line break are shown in Figures 6.2-2 and 6.2-3 respectively. The calculated peak drywell pressure is 39.6 psig, which is 12% below the containment design pressure of 45 psig. The suppression chamber is pressurized by the carryover of noncondensables from the drywell and by heatup of the suppression pool. As the vapor formed in the drywell is condensed in the suppression pool, the temperature of the suppression chamber water approaches 150

° F and the suppression chamber pressure stabilizes at approximately 30 psig. The drywell pressure stabilizes at a slightly higher pressure, the difference being equal to the downcomer submergence. During the RPV depressurization phase, most of the noncondensable gases in the drywell initially are forced into the suppression chamber. However, following the depressurization the noncondensables will redistribute between the drywell and suppression chamber via the vacuum breaker system. This redistribution takes place as pressure is decreased by the steam condensation process occurring in the drywell.

The LPCI and LPCS systems supply sufficient core cooling water to control core heatup and limit metal-water reaction to less than 0.2%. After the RPV is flooded to the height of the jet pump nozzles, the excess flow discharges through the recirculation line break into the drywell. This flow of water (steam flow is negligible) transports the core decay heat out of the RPV, through the broken recirculation line, in the form of hot water which flows into the suppression chamber via the drywell to suppression chamber vent system. This flow, in addition to heat losses to the drywell walls, provides a heat sink for the drywell atmosphere, LSCS-UFSAR 6.2-14 REV. 14, APRIL 2002 causes a depressurization of the containment, and redistributes the noncondensables as the steam in the drywell is condensed.

Table 6.2-8 provides the peak pressure, temperature, and time parameters for the recirculation line break as predicted for the conditions of Table 6.2-1 and in correspondence with Figures 6.2-2 and 6.2-3. The transient peak calculated drywell floor (deck) differential pressure is 24.2 psid, which is 3.2% below the design sustained differential pressure of 25 psid.

During the blowdown period of the LOCA, the pressure suppression vent system conducts the flow of the steam-water gas mixture in the drywell to the suppression pool for condensation of the steam. The pressure differential between the drywell and suppression pool controls this flow vers us time. Figure 6.2-4 provides the mass flow versus time relationship through the vent system for this accident. A supplementary evaluation has been performed for the addition of feedwater to the suppression pool to assess the impact on long term pool temperature. This evaluation estimates that th e peak short term pool temp erature will increase by an additional 15.4

° F. This results in a short term pool temperature (at 600 seconds) of approximately 166

° F . For further discussion, s ee Section 6.2.1.1.3.1.1 in the paragraph titled, "Evaluation of Post-LOCA Feedwater Injection."

Long-Term Accident Responses In order to assess the adequacy of the containment following the initial blowdown transient, an analysis was made of the long-term temperature and pressure response following the accident. The analysis assumptions are those discussed previously for the three cases of interest. The initial pressure response of the containment (the first 600 seconds after th e break) is the same for each case.

Case A - All ECCS Equipment Oper ating (with containment spray)

This case assumes that offsite a-c power is available to operate all cooling systems.

During the first 600 seconds following the pi pe break, the high-pressure core spray (HPCS), low-pressure core spray (LPCS), and all three LPCI pumps are assumed operating. All flow is injected directly into the reactor vessel.

After 600 seconds, both RHR heat exchangers are activated to remove energy from the containment. During this mode of operation the flow from two of the LPCI pumps is routed through the RHR heat exchanger, where it is cooled before being discharged into the containment spray header.

The containment pressure response to this set of conditions is shown as curve A in Figure 6.2-5. The corresponding drywell and suppression pool temperature responses are shown as curve A in Figure s 6.2-6 and 6.2-7. After the initial blowdown and subsequent depressurization due to core spray and LPCI core LSCS-UFSAR 6.2-15 REV. 13 flooding, energy addition due to core decay heat results in a gradual pressure and temperature rise in the containment. When the energy removal rate of the RHR exceeds the energy addition rate from the decay heat, the containment pressure and temperature reach a second peak valu e and decrease gradually. Table 6.2-5 summarizes the cooling equipment operation, the peak containment pressure following the initial blowdown peak, and the peak suppression pool temperature.

Case B - Loss of Offsite Power (with containment spray)

This case assumes no offsite power is available following the accident with only minimum diesel power. The containment sp ray is operating and injecting into the drywell after 600 seconds. During this mode of operation the LPCI flow through one RHR heat exchanger is discharged into the containment spray nozzles.

The containment response to this set of co nditions is shown as curve B in Figure 6.2-5. The corresponding dyrwell and suppression pool temperature responses are

shown as curve B in Figures 6.2-6 and 6.2-7. A summary of this case is given in Table 6.2-5.

Case C - Loss of Offsite Power (no containment spray)

This case assumes that no offsite power is available following the accident, with only minimum diesel power. For the fi rst 600 seconds following the accident, one HPCS and two LPCI pumps are used to cool the core. After 600 seconds the spray may be manually activated to further reduce containment pressure if desired. This analysis assumes that the spray is not activated.

After 600 seconds, one RHR heat exchanger is activated to remove energy from the containment. During this mode of operation, one of the two LPCI pumps is shut down and the service water pumps to the RHR heat exchanger are activated. The LPCI flow is cooled by the RHR heat exchanger before being discharged into the reactor vessel.

The containment pressure response to this set of conditions is shown as curve C in Figure 6.2-5. The corresponding drywell and suppression pool temperature responses are shown as curve C in Figures 6.

2-6 and 6.2-7. A summary of this case is given in Table 6.2-5.

When comparing the "spray" Case B with th e "no spray" Case C, the same duty on the RHR heat exchanger is obtained since the suppression pool temperature response is approximately the same as shown in Figure 6.2-7. Thus, the same amount of energy is removed from the pool whether the exit flow from the RHR heat exchanger is injected into the reactor vessel or into the drywell as spray. However, the peak containment pressure is higher for the "no spray" case, but the pressure is LSCS-UFSAR 6.2-16 REV. 13 still much less than the containment design pressure of 45 psig. (Subsection 6.2.2.3 describes the containment cooling mode of the RHR system.)

A supplemental evaluation has been performed for the purpose of evaluating the suppression pool long term temperature response. For this evaluation, the feedwater is assumed to have been injected into the suppression pool, by the end of the recirculation piping break blowdown phase (at time t = 600 seconds), in order to assess long term peak pool temperature. See paragraph entitled "Evaluation of Post-LOCA Feedwater Injection" in this se ction. Additionally, a slightly reduced RHR pump flow rate of 7200 gpm (versu s 7450 gpm) has been evaluated, as discussed in Section 6.2.2.3.4. Both of these evaluations are evaluated for the DBA-LOCA in Reference 18. The results indica te an increase in the long term peak suppression pool temperature of approximately 8 F due to the feedwater injection and an approximately 1.5

° F increase due to the lower RHR flow rate. The 200

° F peak pool temperature given in Table 6.2-5 is not exceeded. Plant specific safety evaluations have been performed and have concluded that the existing DBA-LOCA analyses referenced above bounds thes e effects on the containment response.

Energy Balance During Accident

In order to establish an energy distribution as a function of time (short term, long term) for this accident, the following energy sources and sinks are required:

a. blowdown energy release rates, b. decay heat rate and fuel relaxation energy, c. sensible heat rate, d. pump heat rate, and

e. heat removal rate from suppression pool.

Items a, b, and c are provided in Subsection 6.2.1.3. The pump heat rate value that has been used in the evaluation of the containment response to a LOCA for Case A is 4881 Btu/sec. A complete energy balance for the recirculation line break accident is given in Table 6.2-6 for the reactor system, the containment, and the containment cooling systems at time zero, at the time of peak drywell pressure, at the end of reactor blowdown, and at the time of the long-term second peak pressure reached in the containment.

The energy and mass balance have been annotated to include the effects of feedwater coastdown/injection on the long te rm peak suppression pool temperature. See paragraph entitled "Evaluation of Post-LOCA Feedwater Injection" in this

section and footnote in Table 6.2-6.

LSCS-UFSAR 6.2-17 REV. 13 Chronology of Accident Events

The complete description of the containment response to the design-basis recirculation line break has been given above. Results for this accident are shown in Figures 6.2-2 through 6.2-7. A chronological sequence of events for this accident from time zero is provided in Table 6.2-7.

The original and 1988 General Electric co ntainment analysis (references 8 & 17), assumed feedwater flow stopped at the initiation of the LOCA. This assumption is conservative for an assessment on the peak cladding temperature (PCT) or containment pressure and temperature response. However, in order to make a more conservative analysis on the suppression pool predicted temperatures, the feedwater energy due to feedwater pump coastdown, or depressurization and resulting feedwater liquid carryover to the pool, should be taken into account in the suppression pool energy balance. A suppl ementary evaluation was performed to assess the impact on peak suppression pool temperature due to the addition of energy from the feedwater system. (Reference 18)

For this evaluation, the feedwater mass downstream of the 2nd Low Pressure Feedwater Heater is injected into the vessel. The feedwater upstream of this feedwater heater is at a temperature less than 212

° F and would not be expected to be injected into the vessel during a DBA-LOCA. The mechanism for FW injection into the vessel during a LOCA with loss of onsite power is flashing of feedwater liquid when the vessel drops below the saturation pressure corresponding to the feedwater liquid temperature. Thus, only feedwater initially at a temperature above 212° F is assumed to flash and be inje cted into the vessel. This is conservative since vessel pressures are expected to remain higher than atmospheric

pressure during the period when the peak pool temperature occurs. The latest revision of plant piping drawings were used as input to determine the feedwater volume.

Additionally, the sensible energy in the feedwater system metal is also added to the feedwater liquid injected into the vessel. It is conservatively assumed that the feedwater flowing into the vessel and coming into contact with hotter feedwater piping metal downstream, will instantaneously achieve thermal equilibrium with the hotter feedwater system metal. This maximizes the metal sensible energy transfer to the feedwater.

For the analysis, all feedwater mass and energy is injected to the vessel and subsequently transferred to the suppre ssion pool by 600 seconds into the LOCA event. This is modeled by adding all the feedwater mass and energy input at time t

= 600 seconds. Based on this previous discussion, this analysis provides a conservative estimate of the amount of en ergy addition to the pool due to feedwater injection.

LSCS-UFSAR 6.2-18 REV. 15, APRIL 2004 The results indicate an increase in the long term peak suppr ession pool temperature of approximately 8

° F (Reference 18). The 200

° F peak pool temperature given in Table 6.2-5 is not exceeded.

Evaluation at 3559 MWt Reactor Power The analysis of an instantaneous guilloti ne rupture of a recirculation line at 3559 MWt reactor power, Reference 25, employed essentially the same methodology as the 3434 MWt analysis, except for the RPV blow down calculation in the short-term containment response analysis. The blowdown calculation was performed using the LAMB break flow model (Reference 26), which models physical phenomena in the pipe and vessel in a more detailed manner. Th e LAMB break flow rate and enthalpy calculated at initial reactor power of 3559 MWt and initial pressure of 1025 psig were used as input to the containment analysis model in the short-term analysis. For the analysis of the long-term containment response, Case C, which was the limiting case among the three cases (Cases A, B, and C) analyzed at 3434 MWt reactor power, was analyzed at 3559 MWt. The analysis of Case C at 3559 MWt had the same assumptions as the original analysis at 3434 MWt with respect to the availability of the ECCS pumps and RHR heat exchanger.

The key input assumptions updated for the analysis at 3559 MWt are: a) the core decay heat is based on the ANSI/ANS 5.1-1979 decay heat model with a two-sigma uncertainty adder (the decay heat calculations also include contributions from miscellaneous actinides and activation products consistent with the recommendation of GE SIL 636.); and b) the water in the feedwater system continues to flow into the RPV until all feedwater above 212ºF is depleted to maximize pool heat-up.

Table 6.2-a shows initial conditions assume d for the analysis of the design basis recirculation line rupture at 3559 MWt.

The analysis results are tabulated and plotted, as follows. Tables 6.2-5a and 6.

2-8a show a summary of the analysis results for the long-term and short-term responses, respectively. The short-term containment pressure and temperature responses are plotted in Figures 6.2-2a and 6.2-3a, respectively. Figure 6.2-5a provides the long-term containment pressure response. The long-term drywell airspace and pool temperature responses are given in Figure

6.2-6a and 6.2-7a respectively.

6.2.1.1.3.1.2 Main Steamline Break The main steamline break, which is not the limiting event with respect to the containment response, was not analyzed at a reactor power of 3559 MWt. The original analysis at 3434 MWt is presented in this subsection.

The sequence of events immediately following the rupture of a main steamline between the reactor vessel and the flow limiter has been determined. The flow on both sides of the break will accelerate to the maximum allowed by critical flow considerations. In the side adjacent to the reactor vessel, the flow will correspond to LSCS-UFSAR 6.2-18a REV. 14, APRIL 2002 critical flow in the 2.98-ft 2 steamline cross section. Blowdown through the other side of the break can occur because the stea mlines are all interconnected at a point upstream of the turbine by the bypass header. This interconnection allows primary system fluid to flow from the three unbroken steamlines, through the header and back into the drywell via the broken line. Flow will be limited by critical flow in the 0.94-ft 2 steamline flow restrictor. The total effective flow area is thus 3.92 ft 2 , which is the sum of the steamline cross-sectional area and the flow restrictor area.

Subsection 6.2.1.3 provides information on the mass and energy release rates.

Immediately following the break, the total steam flow rate leaving the vessel would be approximately 12,000 lb/sec, which exceeds the steam generation rate in the core of 4,500 lb/sec. This steam flow to stea m generation mismatch causes an initial depressurization of the reactor vessel at a rate of 50 psi/sec. The void formation in the reactor vessel water causes a rapid rise in the water level, and it is conservatively assumed that the water level reaches the vessel steam nozzles 1 second after the break occurs. The water level rise time of 1 second is the minimum that could occur under any reactor operating condition. From that time on, a two-phase mixture would be discharged from the break. During the first second of the blowdown, the blowdown flow will consist of saturated reactor steam. This steam will enter the containment in a super-heated condition of approximately 330

° F. Figures 6.2-8 and 6.2-9 show the pressure and temperature response of the drywell and containment during the primary system blowdown phase of the accident.

Figure 6.2-9 shows that the drywell atmosphere temperature approaches 330

° F after 1 second of primary system steam blowdown. At that time, the water level in the vessel will reach the steamline nozzle elevation and the blowdown flow will change to a two-phase mixture. This increased flow causes a more rapid drywell pressure rise. However, the peak differen tial pressure is 24.2 psid, which occurs shortly after the vent clearing transient. As the blowdown proceeds, the primary system pressure and fluid inventory will decrease and this will result in reduced break flow rates.

LSCS-UFSAR 6.2-19 REV. 14, APRIL 2002 As a consequence, the flow rate in the vent system also starts to decrease, and this results in a decreasing differential pressure between the drywell and containment.

Table 6.2-8 presents the peak pressures, peak temperatures, and times of this accident as compared to the recirculation line break.

Approximately 50 seconds after the start of the accident, the primary system pressure will have dropped to the drywell pressure and the blowdown will be over. At this time the drywell will contain pure steam, and the drywell and suppression chamber pressures will stabilize at approximately 30 and 25 psig, respectively; the difference corresponds to the hydrostatic pr essure at the lower end of the submerged vents.

The drywell and containment will remain in this equilibrium condition until the reactor pressure vessel refloods. During this period, the emergency core cooling pumps will be injecting cooling water from the suppression pool into the reactor. This injection of water will eventually flood the reactor vessel to the level of the steamline nozzles, and at this time, the ECCS flow will spill into the drywell. The water spillage will condense the steam in the drywell and thus reduce th e drywell pressure. As soon as the drywell pressure drops below the suppression chamber pressure, the drywell vacuum breakers will open and noncondensable gases from the suppression chamber will flow back into the drywell.

This process w ill continue until the pressures in the two regions equalize and stabilize at approximately 7.5 psig.

6.2.1.1.3.1.3 Intermediate Breaks The intermediate-size break, which is not the limiting event with respect to the containment response, was not analyzed at a reactor power of 3559 MWt. The original analysis at 3434 MWt is presented in this subsection.

The failure of a recirculation line results in the most severe pressure loading on the drywell structure. However, as part of the containment performance evaluation, the consequences of intermediate breaks are also analyzed. This classification covers those breaks for which the blowdown will result in reactor depressurization and operation of the ECCS. This subsection describes the consequences to the containments of a 0.1-ft 2 break below the RPV water level. This break area was chosen as being representative of the intermediate size break area range. These breaks can involve either reactor steam or liquid blowdown.

Following the 0.1-ft 2 break, the drywell pressure increas es at approximately 1 psi/sec. This drywell pressure transient is sufficientl y slow so that the dynamic effect of the water in the vents is negligible and the vents will clear when the drywell-to-wetwell differential pressure is equal to the ve nt submergence pressure. For the LSCS containment design, the maximum distance between the pool surface and the bottom

of the vents is 12 feet 10 inches. Thus, th e water level in the ve nts will reach this point when the drywell-to-containment pr essure differential reaches 5.2 psid.

LSCS-UFSAR 6.2-20 REV. 14, APRIL 2002 Figures 6.2-10 and 6.2-11 show the drywe ll and wetwell pressure and temperature response, respectively. The ECCS respon se is discussed in Section 6.3.

Approximately 5 seconds after the 0.1-ft 2 break occurs, air, steam, and water will start to flow from the drywell to the suppression pool; the steam will be condensed and the air will enter the wetwell free space. After 5 seconds there will be a constant pressure differential of 5.2 psid between the drywell and wetwell. The continual purging of drywell air to the suppression chamber will result in a gradual pressurization of both the wetwell and dryw ell to about 22 and 27 psig, respectively. Some continuing containment pressurization will occur because of the gradual pool heatup. The ECCS will be initiated by the 0.1-ft 2 break and will provide emergency cooling of the core. The operation of these systems is such that the reactor will be depressurized in approximately 600 seconds. This will terminate the blowdown phase of the transient. The drywell w ill be at approximately 27 psig and the suppression chamber at approximately 22 psig.

In addition, the suppression pool temperature will be the same as following the DBA because essentially the same amount of primary system energy would be released during the blowdown. After reactor depressurization, the flow through the break will change to suppression pool water that is being injected into the RPV by the ECCS. This flow will condense the drywell steam and will eventually cause the drywell and containment pressures to equalize in the same manner as following a recirculation line rupture.

The subsequent long-term suppression pool and containment heatup transient that

follows is essentially the same as for the recirculation break.

From this description, it can be concluded that the consequences of an intermediate size break are less severe than those from a recirculation line rupture.

6.2.1.1.3.1.4 Small Size Breaks The small-size break, which is not the limiting event with respect to the containment response, was not analyzed at a reactor power of 3559 MWt. The original analysis at 3434 MWt is presented in this subsection.

Reactor System Blowdown Considerations

This subsection discusses the containment transient associated with small primary system blowdowns. The sizes of primary system ruptures in this category are those blowdowns that will not result in reactor depressurization due either to loss of LSCS-UFSAR 6.2-20a REV. 14, APRIL 2002 reactor coolant or automatic operation of the ECCS equipment. Following the occurrence of a break of this size, it is assumed that the reactor operators will initiate an orderly plant shutdown and depressurization of the reactor system. The thermodynamic process associated with the blowdown of primary system fluid is one of constant enthalpy. If the primary system break is below the water level, the blowdown flow will consist of reactor water. Bl owdown from reactor pressure to the drywell pressure will flash approximately one-third of this water to steam and two-LSCS-UFSAR 6.2-21 REV. 13 thirds will remain as liquid. Both ph ases will be at saturation conditions corresponding to the drywell pressure. Thus, if the drywell is at atmospheric pressure, the steam and liquid associated with a liquid blowdown would be at 212

° F. Similarly, if the containment is assumed to be at its design pressure, the reactor coolant will blow down to approximately 293

° F steam and water.

If the primary system rupture is located so that the blowdown flow consists of reactor steam only, the resultant stea m temperature in the containment is significantly higher than the temperature associated with liquid blowdown. This is because the enthalpy of high-energy saturated steam is nearly twice that of saturated liquid. The higher enthalpy will result in a superheat condition. For example, decompression of 1000-psia steam to atmospheric pressure will result in 298° F superheated steam (86

° F of superheat).

Based upon this thermodynamic process, it is concluded that a small reactor steam leak will impose the most severe temperature conditions on the drywell structures and the safety equipment in the drywell. For larger steamline breaks, the superheat temperature is nearly the same as for small breaks, but the duration of the high-temperature condition is less. This is because the larger breaks will depressurize the reactor more rapidly than the orderly reactor shutdown that is assumed to terminate the small break.

Containment Response For drywell design consideration, the following sequence of events is assumed to occur. With the reactor and containment operating at the maximum normal

conditions, a small break occurs that a llows blowdown of reactor steam to the drywell. The resulting pressure increase in the drywell will lead to a high drywell pressure signal that will scram the reacto r and activate the containment isolation system. The drywell pressure will continue to increase at a rate dependent upon the size of the steam leak. This pressure increase will lower the water level in the vents until the level reaches the bottom of th e vents. At this time, air and steam will start to enter the suppression pool. The steam will be condensed and the air will be carried over to the suppression chamber free space. The air carry-over will result in a gradual pressurization of the containment at a rate dependent upon the size of the steam leak. Once all the drywell air is carried over to the suppression chamber, pressurization of the containment will cease and the system will reach an equilibrium condition with the drywell pressure at 27 psig and the suppression chamber at approximately 22 psig. The drywell will contain only superheated steam, and continued blowdown of reactor steam will condense in the suppression pool.

LSCS-UFSAR 6.2-22 REV. 13 Recovery Operations The reactor operators will be alerted to the incident by the high drywell pressure signal and the reactor scram. For the purposes of evaluating the duration of the superheat condition in the drywell, it is assumed that their response is to cool down the reactor in an orderly manner using any method, but limiting the reactor cooldown rate to 100

° F per hour. The normal method to achieve recovery is by use of the high pressure core spray in conjunction with the automatic depressurization system. This feed and bleed process can be utilized until the reactor is depressurized. Depending upon their availability and the situation, other methods such as the use of turbine bypass valves in conjunction with the main condenser can be utilized to achieve depressurization. This will result in the reactor primary system being depressurized within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Drywell Design Temperature Considerations

For drywell design purposes, it is assume d that there is a blowdown of reactor steam for the 6-hour cooldown period. The corresponding design temperature is determined by finding the combination of primary system pressure and containment pressure that produces the maximum superheat temperature. Thus for design purposes, this results in a temperature condition of 340

° F. 6.2.1.1.3.2 Accident Analysis Models

The short-term pressurization analytical models, assumptions, and methods used by GE to evaluate the containment respon se during the reactor blowdown phase of a LOCA are described in References 2 and 3.

Once the RPV blowdown phase of the LOCA is over, a fairly simple model of the drywell and suppression chamber may be used. During the long-term, post-blowdown containment cooling mode, the ECCS flow path is a closed loop and the suppression pool mass will be constant. Schematically, the cooling model loop is shown in Figure 6.2-12. Since there is no storage other than in the suppression pool (the RPV is reflooded during the blowdown phase of the accident), the mass flowrates shown in the figure are equal, thus:

==eccs S Dmmm O O LSCS-UFSAR 6.2-23 REV. 13 Analytical Assumptions The key assumptions employed in the model are as follows:

a. The drywell and suppression chamber atmosphere are both saturated (100% relative humidity).

b. The drywell atmosphere temperature is equal to the temperature of the coolant spilling from the RPV, or to the spray temperature if the

sprays are activated. c. The suppression chamber atmosphere temperature is equal to the suppression pool temperature or to the spray temperature if the sprays are activated. d. No credit is taken for heat losses from the primary containment or to the containment internal structures. Energy Balance Considerations The rate of change of energy in the suppression pool, E p , is given by: =s h s w M dt d p E dt d ()().s.s s s h dt d M M dt d h w w+= Since _d_d_t (M w s) = 0 (because there is no storage), and for water at the conditions that will exist in the containment:

where: C p = 1.0 for the specific heat of pool water, Btu/ lb-

°F T s = pool temperature, °F.

The pool energy balance yields:

()= h m h m T d t d C M s o sD o D sp s w This equation can be rearranged to yield:

()()s T dt d p C s h dt d=

LSCS-UFSAR 6.2-24 REV. 13 An energy balance on the RHR heat exchanger yields

(6.2-3)

where: h c = enthalphy of ECCS flow entering the reactor, Btu/lb.

Similarly, an energy balance on the RPV will yield:

Combining Equations 6.2-1, 6.2-2, 6.2-3, and 6.2-4 gives This differential equation is integrated by finite difference techniques to yield the suppression pool temperature transient.

Containment Thermodynamic Conditions Once the energy equations are solved, the drywell and suppression chamber atmospheric temperatures can be calculated.

()=s w M sh o sm Dh o D m sT dt d o x s Hsc m qhh= +=eccs m e q D q c h D h ()+=s w M X H q e q D q s T dt d LSCS-UFSAR 6.2-25 REV. 13 For the case in which no containment spray is operating, the suppression chamber temperature, T w, at any time will be equal to the current temperature of the pool, T s, and the drywell temperature, T d, will be equal to the temperature of the fluid leaving the RPV. Thus:

and T w = T s. For the case in which the containment spra y is assumed to be operating, both the drywell and suppression chamber atmosphere will be at the spray temperature, T sp where: eccs m x H q s T sp T= and, T D = T w = T sp. Using the suppression chamber and dr ywell atmosphere temperatures, and assumption (a) (drywell and suppression chamber saturated), it is possible to solve for the containment total pressures, since:

(6.2-6)

(6.2-7)

where: P D = drywell total pressure, psia, P a D = partial pressure of air in drywell, psia, P v D = partial pressure of wate r vapor in drywell, psia, P s = suppression chamber total pressure, psia, P a s = partial pressure of air in the suppression chamber, psia, eccs m x H q e q D q s T D T++=D v P D a P D P+=s v P s a P s P+=

LSCS-UFSAR 6.2-26 REV. 13 P v s = partial pressure of water vapor in the suppression chamber, psia, and, from the Ideal Gas Law:

(6.2-8)

(6.2-9) where: M a D = mass of air in drywell, lb, M a s = mass of air in the suppression chamber, lb, R = gas constant ft-lbf/lb V D = drywell free volume, ft

3. V s = suppression chamber free volume, ft

3. With known values of T D and T w , Equations 6.2-6, 6.2-7, 6.2-8 and 6.2-9 can be solved by transient analysis and iteration.

This iteration procedure is also used to calculate the unknown quantities M a D and M a s. Solution of Equations The transient analysis is based on successive time step integration of the suppression pool temperature. When this integration has been performed and the value of T s at the end of a time step has been calc ulated, a pressure balance is made. Using values of M a D and M a s from the end of the previous time step and the updated values of T D and T s, a check is made to see if P s is greater than or equal to P D using Equations 6.2-6, 6.2-7, 6.2-8, and 6.2-9. If P s is greater than or equal to P D, then the two values are made equal. The vacuum breakers between the drywell and suppression chamber are provided to ensure that P s cannot be greater than P D. 144 D V D RT D a M D a P= 144 s V w RT s a M s a P=

LSCS-UFSAR 6.2-27 REV. 13 Hence, with P D = P s and knowing that:

M a D + M a s = constant; (6.2-10) where the constant is the known total initial mass of air in the suppression chamber and drywell prior to the accident, Equation s 6.2-6, 6.2-7, 6.2-8, and 6.2-9 can be solved for M a s , M a D , and P s/P D.

It is conservatively assumed that the total mass of air remains constant, which ignores any containment leakage that might occur during the transient.

If, as a result of the end-of-time-step pressure check, where: H = submergence of vents, ft, and V w = specific volume of fluid in vent, ft 3/lb then the pressure in the drywell is high er than the pressure in the suppression chamber but not sufficiently so to depress the water to the bottom of the vents and thus permit air to flow from the drywell to the suppression chamber. Under these circumstances, no air transfer is assumed to have occurred during the time step, and Equations 6.2-6, 6.2-7, 6.2-8, and 6.2-9 are solved using the updated temperatures with the same M a s and M a D values from the previous time step.

If the end-of-time step pressure check shows:

then the drywell pressure is set to the value:

(6.2-11)

'w V H s P D P s P+ w V H s P D P+ +=V H s P D P LSCS-UFSAR 6.2-28 REV. 15, APRIL 2004 This requires that the drywell pressure never exceed the suppression chamber pressure by more than the hydrostatic head associated with the submergence of the vents. To maintain this condition, some transfer of drywell air to the suppression chamber will be required. The amount of air transfer is calculated by using Equation 6.2-10 and combining Equations 6.2-6, 6.2-7, 6.2-8, 6.2-9 and 6.2-11 to

give:

ws w s a s v D D D a D v v HV144RTM PV144RTM P++=+

which can be solved for the unknown air masses. The total pressures can then be determined.

6.2.1.1.4 Negative Pressure Design Evaluation

Containment negative pressure has been addressed in Chapter 3.0 and in the Design Assessment Report.

6.2.1.1.5 Suppression Pool Bypass Effects Protection Against Bypass Paths

The pressure boundary between drywell and suppression chamber including the vent pipes, vent header, and downcomers are fabricated, erected, and inspected by nondestructive examination methods in accordance with and to the acceptance standards of the ASME Code Sectio n III, Subsection B, 1971 (Summer 1972 Addenda). This special construction, inspection and quality control ensures the integrity of this boundary. The design pressure and temperature for this boundary was established at 25 psid and 340

° F, which is substantially greater than conditions during a DBA. Actual peak accident differential pressure and temperature across this boundary will be less than their design values during a LOCA. In addition a stainless steel liner has been provided between the drywell and the wetwell as des cribed in Chapter 3.0.

All penetrations of this boundary except the vacuum breaker seats and suppression pool temperature monitoring probe pene trations and testing penetrations are welded. All penetrations are available for periodic visual inspection.

The following paragraphs describe the evaluation of the steam bypass event at 3434MWt. The limiting event was analyzed for a reactor power level of 3559 MWt, and it was concluded that this reactor power has no significant impact on the suppression pool steam bypass.

LSCS-UFSAR LU2000-027I 6.2-28a REV. 14, APRIL 2002 Reactor Blowdown Conditions and Operator Response

In the highly unlikely event of a reactor depressurization to the drywell accompanied by a simultaneous open bypass path between the drywell and suppression chamber, several postulated conditions may occur. For a given primary system break area, the maximum allowabl e leakage capacity can be determined LSCS-UFSAR 6.2-29 REV. 17, APRIL 2008 when the containment pressure reaches the design pressure at the end of reactor blowdown. The most limiting conditions would occur for those primary system break sizes which do not cause rapid reactor depressurization. This corresponds to breaks of less than approximately 0.4 ft 2 which require some operator action to terminate the reactor blowdown.

Immediately after the postulated conditions given above for a small primary system break, there would be a fairly rapid rise in containment pressure as the noncondensable gases in the drywell are carried over to the suppression chamber. During this portion of the transient, it is assumed that the plant operators are unaware that a leakage path exists. Under normal circumstances, the maximum pressure that can occur in the suppression chamber is approximately 25 psig. This is the pressure that would result if all of the noncondensable gases initially in the containment are carried over to the suppression chamber free space. For the

maximum allowable leakage calculations, it was assumed that the plant operators realize a leakage path exists only when the suppression chamber pressure reaches 30 psig. For conservatism, an additional 10-minute delay is assumed before any

corrective action is taken to terminate the transient. The corrective action is also assumed to take 5 minutes to be effective. At that time, the containment pressure would be equal to the design pressure if the allowable leakage had occurred. The specific type of corrective action taken after 10 minutes is not accounted for in the analysis. The operators have several options available to them. If the source of the leakage is undefined, they could depressurize the primary system via either the main condenser or relief valves, or they could activate the containment sprays.

Analytical Assumptions When calculating the allowable leakage capacities for a spectrum of break sizes, the following assumptions are made:

a. Flow through the postulated leakage path is pure steam. For a given leakage path, if the leakage flow consists of a mixture of liquid and vapor, the total leakage mass flowrate is higher, but the steam flowrate is less than for the case of pure steam leakage. Since the steam entering the suppression chamber free space results in the additional containment pressurization, this is a conservative assumption.

b. There is no condensation of the leakage flow on either the suppression pool surface or the containment and vent system structures. Since condensation acts to reduce the suppression chamber pressure, this is a conservative assumption. For an actual containment there will be condensation, especially for the larger primary system breaks where vigorous agitation at the pool surface will occur during blowdown.

Analytical Results

LSCS-UFSAR 6.2-30 REV. 17, APRIL 2008 The LSCS containment has been analyzed to determine the allowable leakage between the drywell and suppression chamber.

Figure 6.2-13 shows the allowable leakage capacity )K/A( as a function of primary system break area. A is the area of the leakage flow path and K is the total geometric loss coefficient associated with the leakage flow path.

The maximum allowable leakage capacity is at )K/A( = .030 ft

2. Since a typical geometric loss factor would be 3 or grea ter, the maximum allowable leakage area would be .052 ft

2. This corresponds to a 3-inch line size.

Figure 6.2-13 is a composite of two curves.

If the break area is greater than approximately 0.4 ft 2, reactor depressurization will terminate the transient and allow higher leakage. However break areas less than 0.4 ft 2 result in continued reactor blowdown which limits the allowable leakage.

Figure 6.2-14 shows the containment response associated with br eaks larger than 0.4 ft 2. The containment pressure would reach design pressure at the end of reactor blowdown. Figure 6.2-15 shows the same response for a typical small break less than 0.4 ft

2. The containment pressure would reach design conditions, in this case, approximately 5 minutes after operator action.

6.2.1.1.6 Suppression Pool Dynamic Loads The manner in which suppression pool dynamic loads resulting from postulated loss-of-coolant accidents, transients, and se ismic events have been integrated into the LSCS design is completely described in the LaSalle Design Assessment Report, which was submitted with the FSAR as a re ference document. The load histories, load combinations, and analyses are all presen ted in detail in this referenced report.

A safety relief valve in-plant test was conducted on unit 1 as committed by Commonwealth Edison per NUREG-0519. A report entitled "Commonwealth Edison Proprietary LaSalle County I In-Plant S/RV Test Initial Evaluation Report" was submitted March 4, 1983 (C. W. Schroed er to A. Schwence r) and resubmitted October 14, 1983 (C.W. Schroeder to H.R.

Denton). The document contains information and data demonstrating the adequacy of existing design basis hydrodynamic loads resulting from safety/relief valve actuation.

Supplementary evaluations have been performe d, as discussed in Section 6.2.1.8, to verify that an increase in the initial suppression pool temperature (from 100

° F to 105° F) would not significantly impact th e dynamic loading scenarios associated with containment response to post ulated LOCAs and SRV operation.

Containment Dynamic Loads were evaluated for power uprate to 3489MWt in Reference 25. The evaluation shows the LOCA and SRV loads remain within the defined limits.

LSCS-UFSAR 6.2-31 REV. 13 6.2.1.1.7 Asymmetric Loading Conditions The manner in which potential asymmetric lo ads were considered for LSCS is fully described in the Design Assessment Report.

A description of the analytical models utilized for these analyses, as well as a description of the containment testing program, is also presented in this report.

6.2.1.1.8 Containment Ventilation System

The primary containment ventilation system is discussed in Section 9.4.

6.2.1.1.9 Postaccident Monitoring A description of the postaccident monitoring system is provided in Section 7.5.

6.2.1.1.10 Drywell-to-Wetwe ll Vacuum Breaker Valves Evaluation for LOCA Loads

During the pool swell phase of a loss-of-c oolant accident, air fl ows from the drywell through the vent pipes and the suppression pool into the suppression chamber air space resulting in a rise of the suppression pool surface and compression of the air space region above it. This transient wetwell air space pressurization may cause the vacuum breaker valves to experience hi gh opening and closing impact velocities. To estimate the valve disc actuation velocities, the Mark II Owner's Group developed a vacuum breaker valve dynami c model described in NEDE-22178-P(1), "Mark II Containment Drywell-to-Wetwell Vacuum Breaker Models," August 1982, which describes the generic methodology us ed to calculate the response of the drywell-to-wetwell vacuum breaker to certain transients in the Mark II containment. The LaSalle plant, however, is unique in that it is the only domestic Mark II plant which has its vacuum breakers located outside containment. Because of this feature, the Mark II Owners Group model was modified to take credit for the pressure losses associated with the exte rnal piping and isolation valves which connect the vacuum breaker between the wetwell and drywell at LaSalle. In a letter dated December 28, 1982, CECo submi tted a report to the NRC, CDI-82-33, "Reanalysis of the LaSalle Wetwell-to-Dry well Vacuum breakers under Pool Swell Loading Condition," December 1982, out lining the valve modeling improvement which have been made to take credit for the pressure losses associated with vacuum breaker piping. This report documents the reduction of the valve impact velocities during pool swell which are attributed to the use of a more realistic hydrodynamic torque on the valve disc. This analysis has been accepted by the NRC. However, because the hydrodynamic loads associated with a loss-of-coolant accident were not considered in the original design of th e vacuum breaker, CECo decided to modify the vacuum breakers to improve performance and reliability, and to further increase the margin of safety. The modifications included material upgrade and/or dimensional changes to strengthen eccentric shaft, hinge arms, hinge plates, fasteners and a load distribution device to reduce the severity of the vacuum LSCS-UFSAR 6.2-32 REV. 14, APRIL 2002 breaker pallet opening impact loading. Th e modified design was tested under an applied mechanical force which produced an opening pallet impact velocity of 20.2 radians/second and a closing impact veloci ty of 25.8 radians/second. The predicted pallet impact velocities for LaSalle are an opening impact velocity of 16.6 radians/second and a closing impact velocity of 24.2 radians/second. After testing, the vacuum breaker leak rate was verified to be within the acceptable limit. The test results verified the operability and fu nctional capability of the vacuum breaker well in excess of the predicted opening and closing impact velocities, and, thus, demonstrated that the modified LaSalle vacuum breakers will function properly under pool swell induced impact loadings with a considerable margin of safety.

6.2.1.1.11 Impact of Increased Initial Suppression Pool Temperature Supplementary safety evaluations have been performed, as discussed in Section 6.2.1.8, to verify that an increase in the initial suppression pool temperature (from 100

° F to 105° F) would not significantly impact the consequences of the various containment line break analyses.

6.2.1.2 Containment Subcompartments

For the most part, the drywell is a large continuous volume interrupted at various locations by piping, grating, ventilation ducting, etc. The only two volumes within the drywell which can be classified as subcompartments are the annular volume between the biological shield and the reactor pressure vessel, and the volume bounded by the drywell head and the reactor vessel head. These regions are referred to as the biological shield a nnulus and head cavity, respectively, and require special design consideration resulting from the postulation of line breaks in

these volumes.

6.2.1.2.1 Design Bases The methodology used to determine the containment subcompartment pressurization loads and the results pertaining to the pressurization loads documented herein are applicable to reactor operation at or below the bounding thermal power level of 3559 MWt (Reference 30).

Biological Shield Annulus

Pressure transients within the biologic al shield annulus are important for two considerations: (1) determination of the design conditions for the shield wall, and (2) determination of the tipping forces on the reactor pressure vessel. It is not a priori clear that one line break will yiel d the most severe conditions for both considerations. Therefore, consequences of two line breaks were studied: (a) a LSCS-UFSAR 6.2-32a REV. 14, APRIL 2002 complete circumferential failure of one of the two recirculation outlet lines at the safe end to pipe weld, and (b) a complete circumferential failure of one of the six feedwater lines at the safe end to pipe weld. While it was assumed that the recirculation line break with its high mass and energy blowdown rates yields most severe shield wall loads, the break of the feedwater line was added to determine the most severe conditions on the vessel. The pressure transien ts following either LSCS-UFSAR 6.2-33 REV. 13 postulated break were used in determin ation of shield wall and pressure vessel design adequacy.

The performed pressurization analyses for the postulated recirculation line break and feedwater line break were based on the nodalization schemes depicted on Figures 6.2-16 and 6.2-17, respectively.

Both nodalization schemes were given careful consideration to assure correct local and overall pressure responses.

Recirculation Line Break

The sudden injection of the subcooled liquid into the shield penetration (Node 35) and adjoining annulus initially causes a sign ificant fraction of the liquid to flash to steam, pressurizing the penetrations and annulus. The responses of the penetration volume and adjoining subcompa rtments are shown on Figure 6.2-18. Within 10 milliseconds after the postulated break both flows out of the penetration have choked. Some 10 milliseconds later, both the penetration pressure and the pressure in the surrounding annulus node peak, reflecting subcooling and inventory effects addressed in the blowdown flow rates. Flow into the annulus initially proceeds in all directions, but soon swings preferentially upward in response to increasing pressures within the dead-ended skirt region. By 0.1 second into the transient, the pressures in and about the penetration have stabilized and shortly

after (by 0.5 seconds), the differential pressures across the shield wall have begun to decrease (Figure 6.2-21). The differential pressure across the shield wall peaks at 115 psid in the region immediately around the penetration. Peak differential pressure across the shield door in the penetration, however, reaches 325 psid.

Feedwater Line Break Pressurization effects of the postulated feedwater line break are much less pronounced than for the recirculation break. Much of the injected fluid finds its way up and out of the annulus and over the top of the shield wall and into the drywell. Nevertheless, the differential pressure across the shield wall surrounding the penetration peaks at 50 psid, while the differential pressure across the shield door in the penetration reaches 205 psid (F igure 6.2-22). By 0.5 second into the transient all the differential pressures ac ross the shield wall have peaked and are decreasing (Figure 6.2-23).

The break area for the recirculation line break was assumed to be time dependent and limited by effects of pipe restraints (see Attachment 6A). The feedwater line break was assumed to provide instantaneous full size break area. Both break models included the effects of subcooled liquid inventory in the determination of mass and energy flux data.

No margins were applied to the calculated differential pressures for this final pressurization analysis.

LSCS-UFSAR 6.2-34 REV. 13 Head Cavity

The head cavity area was analyzed for specific line breaks. They were: 1) a break of the recirculation outlet line within the drywell; and 2) a break of the main steamline within the drywell; and, 3) a simultaneous break of the head spray line and the RPV head vent line within the head cavity. These analyses were carried out to establish the pressure differentials that would exist across the refueling bulkhead plate as a result of these accident conditions. The break of the recirculation outlet line, the drywell DBA, was found to produce the highest pressure differential across the refuelin g bulkhead plate, a value of 9.0 psid upward. The simultaneous break of the he ad spray line and RPV head vent line caused a pressure differential of 7.0 psid downward. The main steamline data are not presented due to the fact that the recirculation line break produced the higher differential pressure value.

The break size, mass flow rate, and energy content for the recirculation line were defined in Subsection 6.2.1.1.3.1 and Tabl e 6.2-18. The supporting assumptions for these data are also supplied in the same su bsection. The break size, mass flow rate, and energy content for the head spray line were determined using Moody's flow through the 3.72-inch diameter head spra y nozzle at reactor conditions with a multiplier of 1.0. Flow from the other side of the head spray line break was neglected. In addition, the simultaneous break of the RPV head vent line was considered because of the lack of whip restraints on the head spray line. The break size, mass flow rate, and energy conten t for the RPV head vent line were determined using Moody's flow at reactor conditions with a multiplier of 1.0. The RPV head vent line was postulated to ruptur e at the four-to-two inch reducer in the line located in the head cavity. The flow occurred at both ends of the break, one having a diameter of 4.0 inch es and the other 2.0 inches.

No margin was applied to the results, since the analysis was done for the final design, and a margin is not required for that situation. However, a margin does exist, and this is indicated in Tables 6.2-11 and 6.2-12.

6.2.1.2.2 Design Features

Biological Shield Annulus

The biological shield annulus is an annul ar space 48.7 feet high and about 2 feet thick formed by the reactor pressure vesse l and its skirt and the biological shield wall. The shield wall is provided with 32 penetrations to allow for routing for the lines connected to the vessel. The shield wall is also pierced to provide 2 HVAC openings and 2 reactor skirt access doors. The 3-1/2 inch thermal insulation divides the shield annulus, except for the lower skir t portion, into 2 almost equal annului. The inner steel shell of the annulus is spanned with vertical and horizontal LSCS-UFSAR 6.2-35 REV. 13 stiffeners which extend 5 inches into the annulus. Egress to the drywell at the top of the shield is partially blocked by the gusset plates supporting the reactor vessel stabilizers (Figures 3.8-23). The penetratio ns in the shield wall are designed with shield doors with a gap of approximately 3 inches between the doors and the thermal insulation on the penetrating lines.

Figure 3.8-39 provides an exterior wall stretchout of the shield wall.

In the annulus pressurization analysis , it was assumed that following the postulated line break the vessel insulation within the annulus was instantaneously displaced to the shield wall. The vessel insulation support structure remains in its original configuration. Venting of the annulus into the drywell was possible through the annulus between the pipe and shield doors in the 32 nozzle penetrations in the shield wall and by mean s of an opening at the top of the shield wall above which the insulation was assume d to blow out instantaneously when the pressure across the insulation above the shield wall reaches 3 psid. Other possible vent paths such as HVAC openings, reac tor skirt access doors, and insulation blowout panels were assumed to remain closed.

Head Cavity Note: The current flow paths have b een changed to include the two manholes between the head cavity and the drywell and the four ducted HVAC vents have

been modified by the addition of discharge nozzles. The impact of this change has been evaluated and it has been determined that the analysis presented here is bounding.

The physical system, shown in Figure 3.8-1, was modeled as three node with two flow paths for this analysis. The head cavity, drywell, and wetwell are all described by single volumes. The model for the simultaneous break of the head spray and RPV head vent lines in the head cavity is shown in Figure 6.2-19, and that for the recirculation line break in the drywell in Figure 6.2-20. The pertinent data

regarding the volumes and flow paths are given in Tables 6.2-11 through 6.2-14.

There are eight HVAC vents in the refuelin g bulkhead plate: four sixteen-inch diameter supply vents, and four eighteen-inch diameter return vents. The return vents have ductwork attached to them. All of the HVAC (supply and return) were modeled for the postulated break in the head cavity since the pressure in the return vents with the ductwork would always be greater than the drywell pressure.

However, only the supply vents were considered to allow flow for the breaks in the drywell. It was assumed that the HVAC re turn ductwork would be crushed by the fast rising drywell pressure. The do wncomer vents between the drywell and wetwell were modeled as one flow path with a valve in the path set to open at 0.824 second for the recirculation line brea

k. The 0.824 second was taken as a conservative estimate of the time normally required to clear the downcomer vents. At this time, the entire vent area becomes available for pressure relief of the drywell and head cavity region. The simultaneous head spray line and RPV head LSCS-UFSAR 6.2-36 REV. 13 vent line break is a much smaller break and results in a relatively slow pressurization of the drywell. A valve was again used in the flow path, but in this instance, the valve opening was dependent upon the drywell pressure exceeding the hydrostatic head at the downcomer exit. The opening differential pressure used was 5.2 psid which is equivalent to a 12-foot downcomer submergence. The flow was carried over directly into the wetwell air volume. No credit was taken for condensation. The flow through both flow paths was taken to be a completely homogeneous mixture.

6.2.1.2.3 Design Evaluation

Biological Shield Annulus The RELAP 4 Mod 3 computer code was used to perform the analyses. The assumptions made in modeling the problem were in accordance with the applicable

USNRC guidelines.

The mass and energy blowdown rates were determined according to the methods described in Attachment 6.A.

Initial conditions in the annulus and drywell are indicated in Tables 6.2-9 and 6.2-10. In subsonic flow conditions, two flow models were used, as defined in RELAP 4 Mode 3: (a) compressible flow, single st ream model was used for the path of major flow direction, and (b) incompressible flow without momentum flux model was used for flow paths other than the paths of the major flow direction. For sonic flow conditions the Moody or sonic choking model were specified with the multiplier 0.6 for the Moody choking model. Homogeneous flow was assumed for the vent mixture.

The biological shield annulus between th e reactor pressure vessel and the shield wall was modeled differently for each of the two postulated line breaks. In either case, advantage was taken of the near sy mmetry of the annular space across the vertical plane passing through the centerline of the failed line.

Nodalization of the biological shield annulus was determined on the basis of natural geometric boundaries and the constraint that the pressure drop within a node be reasonably low as compared to pressure drop across the boundaries of the node. Nodal boundaries were suggested by the pr esence of the reinforcing steel, thermal insulation support structure and nozzles.

Significant pressure drops near the break suggested smaller nodes (by and large limited with two successive obstructions) around the penetration than elsewhere (Fig ures 6.2-37 and 6.2-38). Therefore the assumption was made that since RELAP 4 allo ws input of loss c oefficients only at the junctions between nodes, the junctions should be placed at points where major LSCS-UFSAR 6.2-37 REV. 13 pressure losses occur. Furthermore, it may be concluded that increasing the number of junctions (by making smaller nodes) beyond this point will yield no improvement in the accuracy of the results.

To test this hypothesis, a sensitivity study was performed on the sacrificial shield nodalization. Using the original noda lization (Figure 6.2-39) as a basis, an "equivalent" model was run which maintained the nodalization near the break but drastically reduced the number of nodes fu rther from the break (Figure 6.2-40). This model demonstrated identical pressure response close to the break and only minor differences away from the break (Fig ures 6.2-41 and 6.2-42). This indicated that the nodalization far from the break was sufficiently refined in the original model and that the "equivalent" model could be used to simulate a response close to the break.

Two additional models were run. The first combined the nodes closest to the break into one large node (Figure 6.2-43). The pressure response was not consistent with the original runs (Figures 6.2-44 and 6.

2-45). This indicated that a model which does not locate node boundaries at all flow restrictions close to the break is not acceptable. The last model substituted six nodes for the three original nodes, causing junctions to occur at locations which coincide with no actual flow restriction (Figure 6.2-46). This model showed a net increase of 5% in the force caused by the pressures in the area being investigated. An examination of the axial and circumferential pressure distributions showed only minor differences (Figures 6.2-47 and 6.2-48).

The sensitivity study indicates that the original nodalization provides an adequate description of the pressurization of the sacri ficial shield annulus.

An increase in the complexity of the RELAP 4 model would not result in a significant change in the results. As previously indicated, half of the annulus was nodalized in case of either postulated line break; for the recirculation line break half-annulus consisted of 35 nodes and the half-drywell of 3 nodes (T able 6.2-9), while for the feedwater line break the half-annulus consisted of 29 nodes and the half-drywell of 3 nodes (Table 6.2-10). Volume of each node was calculated as a net volume, that is, the respective volume of the annulus including the volume of penetrations (if any) was corrected for the volume of the insulation and nozzles. The junctions, 85 and 69 for the recirculation line break and feedwater line break respectively, were assigned the smallest flow area anywhere between the centers of two volumes. All partial loss coefficients, k j's, were derived from Reference 6. The total loss coefficient k t was then determined by adding the weighted partial loss coefficients in series:

2 i A t A i K i t k=

LSCS-UFSAR 6.2-38 REV. 13 where A t is the junction area and A i is the area within the junction and pertaining to the partial loss coeffici ent k. When parallel paths, j, were combined, the following relations were utilized:

Only similar junctions were combined in this manner (like 2 or more penetrations connecting drywell with the same volume of the annulus), other junctions were modeled separately.

Inertia coefficients were similarly calculated using simplified conservative approximations to the integrated junction characteristics. Thus, for the junctions with only minor variations, in cross-sect ional flow area along the junction, the inertia, I, was approximated by:

where L i is the distance along the junction wh ere junction's cross-sectional area is A i. In cases where there appear major variations in the cross-sectional flow area (constriction in the conduit) the inertia was estimated by:

where d is a "characteristic" diameter of the constriction of length L o and with area A o (for an orifice the characteristic diameter is taken to be the diameter of the orifice). L 1 , A 1 and L 2 , A 2 are the length and flow area of the conduit partitioned by the constriction. In special cases, where the constriction is not an ordinary orifice, a variation of the above relation was used to evaluate I.

j A j t A=2 i k 1 t A i A i t K=i L i t A 1 I= 2 A d 2 L o Ad2 o L 1 A d 1 L I+++=

LSCS-UFSAR 6.2-39 REV. 13 Parallel paths were characterized by:

To further illustrate methods of determination of the junction characteristics, treatment of selected representative junctions will be shown in detail. The junctions are those for the recirculation lin e break nodalization scheme: 9, 47, 72.

Junction 9 connects the break volume (nod e 35), which consists of the half-annulus in the recirculation line penetration extended from the shield door to the reactor vessel, with the surrounding annular node (34). The minimum junction area was in this case within the break volume, half of the annular area formed by the recirculation line and the penetration wall was calculated to be 7.04 ft

2. In determining the loss coefficient for this junction, Diagram 11-9, Reference 6, was utilized. An upper limit value was set at 0.85 and considered the only loss for this junction.

The inertia coefficient, I, for the junction was calculated as a sum of two contributions: (a) inertia through the half-annulus of the penetration (0.23), and (b) an upper limit estimate of the inertia within the annulus, node 34 (0.07), totaling 0.30 ft-1. Junction 47 is a vertical junction connecting nodes 16 and 21. The junction area is

the related annulus cross-section area redu ced by two constrictions, stiffener and the thermal insulation support structure.

Although the constrictions appear at different elevations (11 inches apart), they were assumed at the same elevation.

This assumption leads to the junction area of 7.72 ft 2 (upstream volume flow area is 11.87 ft 2 and the flow area of the downstream volume is 12.36 ft 2). The loss coefficient was estimated using Diagram 4-9 of Reference 6, at 0.66 for flow area 7.72 ft 2. The total junction loss coefficient is therefore 0.67. The junction area is characterized by the radial width of 1.

45 feet. This width was taken as the characteristic length, d, for the purposes of the inertia coefficient determination.

Then, using a variation of the above described relation for I, it was found that I = 0.45 ft

-1. 1 j I 1 j I= 2 AdL o A d I+=

LSCS-UFSAR 6.2-40 REV. 13 Junction 72 is an example of the vent path through the line penetration and connects annular node 28 with the containment node 37. The actual penetration is located on the boundary between nodes 28 and 29. For this reason, only half of the penetration was treated as the junction 72.

The minimum area of the junction is the cr oss-sectional area of the half of annulus between the shield door and penetration line. It was determined to be 9.71 ft

2. Half-penetration flow area was calculated at 5.33 ft

2. The inertia coefficient for this junction was determined on the basis of the above areas and the characteristic diameter as being the hydraulic diameter at the penetration exit (3.3 ft

-1). The loss coefficient for the junction was, however, determined for the whole penetration and it consisted of a friction loss (0.02 for A = 10.65 ft 2), turning losses at the nozzle and contraction-expansion losses at the shie ld doors. The turning losses were approximated with losses in the branch of a tee section as shown in Diagram 7-21, Reference 6, and estimated at 1.05 bas ed on the penetration area 10.65 ft

2. The loss at the shield door was approximated with a loss due to a discharge from a straight conduit through a thick-walled orifice or grid, Diagram 11-28, Reference 6, and calculated at 1.69 based on the penetration exit area 1.424 ft

2. Then the total loss coefficient based on the area 1.424 ft 2 is 1.71, which is the loss coefficient of the junction.

A complete review of all volume and junction parameters as used in the analyses is given in Tables 6.2-9, 6.2-10, 6.2-24, and 6.2-25. Tables of junction characteristics include an indication whether the junction was choked during the analysis. The junctions closer to the break volume choked very early in the transient; an indication that the pressurization was hardly a function of either assigned loss coefficients or inertia coefficients.

Mass and energy blowdown rates used in th e analysis are given in Tables 6.2-26 and 6.2-27.

Figure 6.2-18 depicts the calculated differ ential pressures across the biological shield wall (doors) for the postulated recirculation line break. Figures 6.2-49 and 6.2-50 show final pressure distribution in axial and circumferential direction, respectively also for the recirculation line break. Figures 6.2-22, 6.2-51, and 6.2-52 give the same information for the postulated feedwater line break.

Head Cavity

Note: The current flow paths have b een changed to include the two manholes between the head cavity and the drywell and the four ducted HVAC vents have been modified by the addition of discharge nozzles. The impact of this change has been evaluated and it has been determined that the analysis presented here is bounding.

LSCS-UFSAR 6.2-41 REV. 13 The computer code utilized for this investigation was RELAP4/Mod 5 (Reference 7) as received from the Argonne Code Center. A listing of the input for each case (Tables 6.2-15 and 6.2-16) is provided to demonstrate the options of the code that were utilized to obtain a solution. The mass and energy inputs were taken from Table 6.2-18 for the recirculation line break, and calculated based on Moody's flow model with a multiplier of 1.0 for the simultaneous head spray line and RPV head vent line break. The details regarding th e data contained in Table 6.2-18 are given in Subsection 6.2.1.1.3.1. The basic assumptions utilized in the analysis are given below. a. Thermodynamic equilibrium exists in each containment subcompartment. The containmen t option of the RELAP4/MOD5 computer code was utilized which a llows for the flow of air, water vapor, and liquid between the nodes.

b. The constituents of the fluid flowing through the subcompartment vents are based on a homogeneous mixture of the fluid in the subcompartment. The consequences of this assumption result in complete liquid carry-over through subcompartment vents.

c. No heat loss from the gases inside the primary containment is assumed. This adds extra conservatism to the analysis, i.e., the analysis will tend to predict higher containment pressures than would actually exist.

d. Incompressible single-stream flow without momentum flux was used for all junctions.

e. The Moody model for critical flow was used when choking occurred in a junction.

f. The stagnation properties which include dynamic velocity effects were used to determine the flow rate in conjunction with the Moody model.

g. A contraction coefficient of 0.6 was implemented with the junction flow areas which reduces the flow and retains higher pressures closer to the break. In addition, a contraction coefficient of 1.0 was utilized for the fill junction which was used to simulate the break.

h. The reactor pressure vessel head insulation remains in place and retains its structural integrity during any postulated accident. This is conservative since the RPV head cavity volume is minimized which will result in higher pressures in the head cavity.

LSCS-UFSAR 6.2-42 REV. 13 i. The manholes between the head cavity and the drywell are assumed to be closed. This reduces the flow area between the volumes increasing the differential pressure across the bulkhead.

j. All of the HVAC vents (supply and return) are modeled for the postulated break in the head cavity since the pressure in the return vents with the ductwork would alwa ys be greater than the drywell pressure. However, only the supply vents are considered to allow flow for the breaks in the drywell. It is assumed that the HVAC return ductwork would be crushed by the rising drywell pressure.

k. To simplify the input to RELAP4/MOD5, the flow area properties of the HVAC vents are combined into one equivalent vent.

l. The downcomers are represented by an equivalent single flow path with a flow area equal to the sum of the actual flow areas.

m. The modeling of downcomer clearing the initiation of flow into the wetwell was modeled in two ways. In the case of the recirculation line break, the downcomer clearing is extremely rapid. To accurately simulate this, the model would have to be rather complex due to the large inertial and frictional effects present in the downcomer. This complexity was avoided by making use of an accident chronology shown in Table 6.2-7 which found th e vent clearing time to be 0.824 second. A valve was placed in the flow path and opened 0.824 second after the line break. The simultan eous head spray line and RPV head vent line break is a much smaller break and results in a relatively slow pressurization of the drywell. A va lve was again used in the flow path, but in this instance, the valve opening was dependent upon the drywell pressure exceeding the hydrostatic head at the downcomer exit. The opening differential pressure used was 5.2 psid which is equivalent to a 12-foot downcomer submergence.

n. No significant depressurization of the reactor pressure vessel occurs during the postulated break.

o. The simultaneous pipe break of the head spray line and the RPV head vent line was considered because of the lack of whip restraints on the head spray line. The resultant whip of the head spray line is assumed to rupture the RPV head vent line. Neither the RCIC nor the RHR system is operating during the time of the head spray line break, i.e., the RHR-RCIC stop valve is assumed to be closed during the time of the accident. The RPV head vent line is connected at the RPV head and at the main steam header. Therefore, a break in this line results in a two direction blowdown, one side feeds directly from the RPV, and LSCS-UFSAR 6.2-43 REV. 14, APRIL 2002 other feeds from the main steamline. The head spray line has a limiting flow area at the head spray nozzle which has a diameter of 3.72 inches. The RPV head vent line is postulated to rupture at the 4-inch to 2-inch reducer in the line located in the head cavity. The steam flow occurs at both ends of the break, one having a diameter of 4.0 inches and the other 2.0 inches. The total flow area was determined to be 0.163 square feet. All of the fl ows are assumed to have the same RPV conditions which are a pressure of 1050.0 psia and an enthalpy of 1190.0 Btu/lbm. Utilizing Moody' s choked flow tables from RELAP4/MOD5, a maximum flow of 2200.0 lbm/sec-ft 2 or 357.9 lbm/sec was calculated. This is used as a constant flow rate for the break in the head cavity.

p. The mass and energy release rates used for the recirculation line break are those given in Table 6.2-18. The break sizes are specified in Subsection 6.2.1.1.3.1.1 and the details regarding line size, break size, orifice size, etc., are given in Table 6.2-4.

q. RELAP4/MOD5 lacks the ability to model steam condensation in the suppression pool. This limitation has no effect on the results obtained prior to vent clearing but will re sult in an overestimation of the pressure rise in the wetwell after vent clearing. Since the maximum differential pressure across the refu eling bulkhead occurs very shortly after downcomer vent clearing in the case of the recirculation line

break, the effect is negligible. However, it is noted that the long-term pressure values are not realistic because of this modeling method. In the case of the break in the head cavity, flow through the downcomers does not begin until long after the peak differential pressure across the refueling bulkhead plate occurs.

r. The initial conditions are taken to be the normal operating conditions as given in Table 6.2-3 except with a relative humidity of 0.1%. In the head cavity and drywell the initial pressure is 15.45 psia, the initial temperature is 135

° F and the relative humidty is 0.1%. In the wetwell the initial pressure is 15.45 psia, the initial temperature is 100° F and the relative humidity is 0.1%.

The node and flow path data specifics are given in Tables 6.2-11 and 6.2-12 for the simultaneous break of the head spray and RPV head vent lines and Tables 6.2-13 and 6.2-14 for the recirculation line break. The nodes and flow paths are graphically depicted in Figure 6.2-19 for the simultaneous break of the head spray line and RPV head vent line, and Figure 6.2-20 for the recirculation line break.

A description of the loss coefficient determination for the flow paths is provided. This problem has only two flow paths to co nsider. The first path connects the head LSCS-UFSAR 6.2-44 REV. 14, APRIL 2002 cavity to the drywell and consists of eight ports through the bulkhead plate. Four of these ports are the HVAC supply ports for the head cavity and do not have any ductwork attached to them. The remain ing four ports are the HVAC return ducts from the head cavity and have ductwork attached to them. All of the HVAC vents (supply and return) were modeled for the po stulated break in the head cavity since the pressure in the return vents with th e ductwork would always be greater than the drywell pressure. The losses considered were the turning losses of the fluid around the RPV head from the break to the HVAC ports in the bulkhead. These losses are very small since the turning radius around the RPV head is so large.

Therefore, this loss was neglected. The ports without the ductwork were considered as thick-edged orifices. This loss coeffi cient was determined using Diagram 4-14 of Reference 6 and was calculated to be 1.52. The ports with the ductwork consist of a 24-inch to 18-inch diameter reducer followed by ductwork which includes a series of elbows and one tee. The flow finally exit s into the drywell through one of the tee branches. Diagrams 3-9, 6-1, and 7-25 of Reference 6 were used to calculate the loss coefficient and it was determined to be 4.

62. Since the flow through the ports with and without ductwork is parallel, the losses were combined for parallel flow and the total loss coefficient was calculated, as described in Subsecti on 6.2.1.2.3, to be 2.62. The flow area for this case is the total of the minimum flow areas through each of the eight HVAC vents. The total flow area was determined to be 11.12 square feet. For the recirculation line break within the drywell, only the supply vents which are without ductwork were considered to allo w for flow. It is assumed that the HVAC return ductwork would crush because the drywell pressure would be greater than the pressure in the ductwork. The loss coefficient for this case is calculated for the ports without the ductwork. The loss coefficient was determined as mentioned earlier and was calculated to be 1.52. The flow area for this case was determined to be 4.92 square feet.

The loss coefficient for the second flow path, through the downcomers, was taken from Table 6.2-1 and is 5.2. No attempt was made to model the inertial effects of the clearing transient. The path was treated as a valve that opened at a prespecified time of 0.824 second for the recirculation line break. For the simultaneous head spray line and RPV head vent line break, the path was treated as a valve that opened when the drywell pressure exceeded the hydrostatic head of 5.2 psid which is equivalent to a 12-foot downcomer submergence. The path model considers no inertial effects; this is a conservative approach, since it has the effect of making the pressure differentials across the bulkhead plate higher.

Figure 6.2-24 depicts the pressure histories of the head cavity and drywell for the break in the head cavity and the recirculati on line break in the head cavity and the recirculation line break in the drywell. The pressure differential histories across the bulkhead plate for the break in the head cavity and the recirculation line break in the drywell are shown in Figure 6.2-25. The peak pressure differential for each break was found to be 9.0 psid upward fo r the recirculation line break and 7.0 psid downward for the simultaneous head spray line and RPV head vent line break. The LSCS-UFSAR 6.2-45 REV. 14, APRIL 2002 differential pressure history as shown for the simultaneous break of the head spray line and RPV head vent line shows two differential pressure peaks. The first differential pressure peak is due to the su dden pressurization of the head cavity and the second peak is due to the sudden opening of the downcomers at a pressure differential between the drywell and wetwell of 5.2 psid. This second peak is erroneous because no inertial effects were modelled in the downcomer flow path and therefore was not considered as the design downward differential pressure. The design pressure differential is 10.6 psid in both directions.

This provides for a margin factor of approximately 1.2 at the final design stage.

6.2.1.2.4 Impact of Increased Initial Suppression Pool Temperature

Supplementary safety evaluations have been performed, as discussed in Section 6.2.1.8, to verify that an increase in the initial suppression pool temperature would not significantly impact the consequences of this accident scenario.

6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-of-Coolant Accidents

This section contains a description of th e transient energy release rates from the reactor primary system to the containment system following a LOCA with minimum ESF performance. In general, a very conservative analytical approach is taken in that all possible sources of energy are accounted for, whereas the suppression pool is assumed to be the only available heat sink. No credit is taken for either the heat that will be stored in the suppression chamber and drywell structures, or the heat that will be transmitted through the containment and dissipated to the environment.

The analysis at 3559 MWt used essentially th e same methodology as the analysis at 3434 MWt, except for the RPV blowdown in the short-term containment response analysis, as noted in Subsection 6.2.1.1.3. The break flow rate and enthalpy used for the short-term containment response analysis at 3559 MWt are given in Table 6.2-18a. For the analysis of the long-t erm containment response, one of the key input assumptions upda ted for the analysis at 3559 MWt is that the core decay heat is based on the ANSI/ANS 5.1-1979 decay he at model with a two sigma uncertainty adder. The core decay heat values used in the 3559 MWt analysis are provided in Table 6.2-20a. The following subsections explain how the transient mass and release rates from the vessel to the containment were determined for the original analysis at 3434 MWt.

6.2.1.3.1 Mass and Energy Release Data Table 6.2-18 provides the mass and enthalpy release data for the containment DBA, recirculation line break. Blowdown steam and liquid flow rates and their respective enthalpies are reported for a 24-hour period following the accident. Figures 6.2-26 LSCS-UFSAR 6.2-45a REV. 14, APRIL 2002 and 6.2-27 show the blowdown flow rates for the recirculation lines break graphically. This data was employed in the DBA containment pressure-temperature transient analyses repo rted in Subsection 6.2.1.1.3.1.

Table 6.2-19 provides the mass and enthal py release data for the main steamline break. Blowdown data is presented for a 24-hour period following the accident.

Figure 6.2-28 shows the vessel blowdown flow rates for the main steamline break as a function of time after the postulated rupture. This information has been employed in the containment response analys es presented in Subsection 6.2.1.1.3.1.

LSCS-UFSAR 6.2-46 REV. 13 6.2.1.3.2 Energy Sources The reactor coolant system conditions prior to the design basis recirculation line break are presented in Tables 6.2-3 and 6.

2-4. Reactor blowdown calculations for containment response analyses are based upon these conditions during a loss-of-coolant accident.

Following each postulated accident event, the stored energy in the reactor system and the energy generated by fission product decay will be released. The rate of release of core decay heat for the evaluation of the containment response to a LOCA is provided in Table 6.2-20 as a function of time after accident initiation. This data is based upon a normalization factor of 3440 MWt and includes the energy of fuel

relaxation.

Following a LOCA, the sensible energy stored in the reactor primary system metal will be transferred to the recirculating ECCS water and will thus contribute to the suppression pool and containment heatup. Figure 6.2-29 shows the temperature transients of the various primary system structures which contribute to this sensible energy transfer. Figure 6.2-30 shows the variation of the sensible heat content of the reactor vessel and internal structures during a recirculation line break accident based upon the temperature transient responses.

6.2.1.3.3 Effects of Metal-Water Reaction

The containment systems shall accommodate the effects of metal-water reactions and other chemical reactions following a postulated DBA. The amount of metal-water reaction is limited to values consistent with the performance objectives of the emergency core cooling systems (ECCS).

6.2.1.3.4 Impact of Increased Initial Suppression Pool Temperature

Supplementary safety evaluations have been performed, as discussed in Section 6.2.1.8, to verify that an increase in the initial suppression pool temperature would not significantly impact the consequences of this accident scenario.

6.2.1.4 Mass and Energy Release Analysis for Postulated Secondary System Pipe Ruptures Inside Containment (PWR)

Not applicable.

6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies on Emergency Core Cooling System (PWR)

Not applicable.

LSCS-UFSAR 6.2-47 REV. 17, APRIL 2008 6.2.1.6 Testing and Inspection Containment testing and inspection programs are fully described in Subsection 6.2.6 and in Chapter 14.0 of the FSAR. The requirements and bases for acceptability are outlined completely in the Technical Specifications.

6.2.1.7 Instrumentation Requirements

A complete description of the instrumentation employed for monitoring the containment conditions and actuating those systems and components having a safety function is presented in Chapter 7.0.

6.2.1.8 Evaluation of 105

° F Suppression Pool Initial Temperature Temperature limits on the suppression pool for Boiling Water Reactors (BWR) with Mark II containment were implemented to minimize the potential for high amplitude loads on the pool during accide nt events. However, some of the limits were implemented with excessive conservatism because the loading phenomena were not completely understood. This suppression pool temperature limit has therefore been historically chosen based on the maximum expected service water temperature. For LaSalle County Statio n Units 1 and 2, the licensing safety evaluations were based upon a 100

° F suppression pool water temperature, which was equivalent to the Ultimate Heat Sink design temperature limit.

Hot weather in Illinois can cause the temperature of the ultimate heat sink to rise to the point where the suppression pool temperature limit of 100

° F may be exceeded. However, the ultimate heat sink design limit will not be exceeded. To prevent an unnecessary plant shutdown during a period of high electrical demand, plant specific safety evaluations have been performed (References 10-20) to demonstrate that plant operation with higher suppression pool temperature is acceptable, i.e., the plant safety limits will still be met with the higher temperatures.

The suppression pool was designed to fu nction as both a heat sink and an emergency water source during transient and accident events as discussed throughout section 6.2. Therefore, performance of the following evaluations were required to support a 5

° F increase in the initial suppression pool temperature as LaSalle County Station Units 1 and 2:

a) Containment loads associated with SRV operation including air clearing loads and steam condensation loads.

b) Containment response associated with LOCA events including the peak pressure and temperature design limits, condensation capability, condensation oscillation load s (CO), and chugging loads.

LSCS-UFSAR 6.2-48 REV. 17, APRIL 2008 c) Equipment performance for design basis events including the impact on the core cooling capability of the ECCS and the parameters which could impact the operability of the ECCS pumps (such as NPSH availability, etc.).

d) Equipment and ECCS performance for other non-LOCA events, e.g., ATWS.

For each of these cases the evaluation showed that the increase of the initial suppression pool temperature would have an insignificant impact on the existing

design margin for the suppression pool and ECC systems. Peak local pool temperature will increase by 3

° F at a 105

° F initial pool bulk temperature for SRV related events.*

The results of this evaluation were subm itted to the NRC (Reference 11), and an approved license amendment to change the maximum suppression pool temperature limit to 105

° F was received (Reference 12). The Ultimate Heat Sink design temperature limit is changed to 104

° F in Reference 32.

6.2.2 Containment Heat Removal System

The containment heat removal system func tion is accomplished by the containment cooling mode of the RHR system. The system is also equipped with spray headers in the drywell and suppression chamber areas. However, no credit was taken for these spray headers for either heat removal or fission product control following a LOCA. 6.2.2.1 Design Bases

The containment heat removal system, consisting of the suppression pool cooling system, is an integral part of the RHR syst em. It meets the following safety design bases: a. The source of water for restoring RPV coolant inventory is located within the containment to establish a closed cooling-water path.

b. A closed loop flow path between the suppression pool and the RHR heat exchangers is established so that the heat removal capability of these heat exchangers can be utilized.

c. This system, in conjunction with the ECC systems, has such diversity and redundancy that no single failure can result in its inability to cool the core adequately (Subsection 6.3.1).

• Peak bulk suppression pool temperature, in the case of LOCA events, is still approximately 10° F below the allowable values.

LSCS-UFSAR 6.2-49 REV. 13 d. To ensure that the RHR containment cooling subsystem operates satisfactorily following a LOCA, each active component shall be testable during operation of the nuclear system.

6.2.2.2 System Design The containment cooling subsystem is an integral part of the RHR system, as described in Subsection 5.4.7. The piping and instrumentation diagram is given in Drawing Nos. M-96 (sheets 1-4) and M-142 (sheets 1-4). Re dundancy is achieved by having two complete containment cooling systems.

Consideration of the fouling of heat exch angers and the selection of temperatures for heat exchanger design are di scussed in Subsection 5.4.7.

6.2.2.3 Design Evaluation

In the event of the postulated LOCA, the short-term energy release from the reactor primary system will be dumped to the suppression pool. This will cause a pool temperature rise of approximately 46

° F. Subsequent to the accident, fission product decay heat will result in a continui ng energy dump to the pool. Unless this energy is removed from the primary containment system, it will eventually result in unacceptable suppression pool temperatur es and containment pressures. The containment cooling mode of the RHR system is used to remove heat from the

suppression pool.

A supplementary evaluation has been performed for the addition of feedwater to the suppression pool to assess the impact on long term pool temperature. This evaluation estimates that th e peak short term pool temp erature will increase by an additional 15.4

° F. This results in a short term pool temperature (at 600 seconds) of approximately 166

° F. Further details are given in Section 6.2.1.1.3.1.1 in the paragraph titled, "Evaluation of Post-LOCA Feedwater Injection".

6.2.2.3.1 RHR Containment Cooling Mode

When the RHR system is in the containment cooling mode, the pumps draw water from the suppression pool, pass it throug h the RHR heat exchangers, and inject it back either to the suppression pool or to the RPV.

In order to evaluate the adequacy of the RHR system, the following limiting case is postulated:

a. Reactor initially at maximum power.

b. Isolation scram occurs.

LSCS-UFSAR 6.2-50 REV. 17, APRIL 2008

c. Manual depressurization discharges heat to suppression pool.

d. Suppression pool cooling is established approximately 10 minutes after the technical specification limit for pool water temperature is reached.

A complete discussion of the suppression pool temperature transients is contained in Chapter 6 of the LSCS-DAR.

The suppression pool temperature transients have been analyzed based on an increased initial suppression pool temperature of 105

° F as discussed in Section 6.2.1.8. The scenarios analyzed are based on those spec ified in NUREG-0783, Reference 15 provides the results of this analysis. For all analyzed cases the long term suppression pool temperature is less than 200

° F. 6.2.2.3.2 Summary of Containment Cooling Analysis

When calculating the long term, post LOCA pool temperature transient, it is assumed that one RHR heat exchanger loop is not available, the suppression pool level initially is at the technical specification minimum, the suppression pool temperature initially is at the technica l specification maximum, and the design RHR heat exchanger fouling factors are used. No credit is taken for heat loss to environs or to the pool structures.

The resultant suppression pool transient maximum temperature for 3434 MWt is 200° F (see References 8, 15, 16, 17, and 18). It is concluded that even with the very conservative assumptions described above, the RHR system in the containment cooling mode can meet its design objectiv e of safely terminating the limiting case temperature transient. See subsection 6.

2.2.3.5 for impact of power uprate to 3489 MWt.

6.2.2.3.3 Impact of Increased Initial Suppression Pool Temperature

Supplementary evaluations have been performe d, as discussed in Section 6.2.1.8, to verify that an increase in the initial suppression pool temperature would not impact the ability of the RHR containment cooling system to meet its design objective.

6.2.2.3.4 Impact of Reduced RHR Suppression Pool Cooling Flow Rate

The original and 1988 General Electric co ntainment analyses (references 8 & 17), has been supplemented with an evaluation which considers an RHR pump flow rate during the suppression pool cooling of 7200 gpm. As noted in Table 6.2-2, the previous analysis used a flow rate of 7450 gpm. Although the RHR pump is capable of such performance, the minimum requ ired Technical Specification flow per specification SR 3.6.2.3.2 is only 7200 gpm.

Since suppression pool cooling is only initiated after 600 seconds into the DBA-LOCA, the affect of this lower flow rate LSCS-UFSAR 6.2-51 REV. 15, APRIL 2004 will be seen as slightly lower efficiency for the RHR heat exchanger and a higher long term suppression pool temperature. The results of th e Reference 18 General Electric analysis indicate an increase in the long term pool temperature of 1.5

° F for the DBA-LOCA case.

For cases which involve SRV blowdown to the suppression pool the lower RHR

pump flow rate was a ssessed in S&L Calculation 3C7-0181-003, Rev. 3 (Reference 15) and the effect on the peak suppression pool temperature was an increase of less than or equal to 1

° F in the peak suppression pool temperature. For all cases examined, the highest peak pool temperature calculated is 195

° F which is still less than 200

° F peak temperature for all cases analyzed. Thus, complete steam condensation is assured with these elevated pool temperatures.

6.2.2.3.5 Impact of Power Uprate

The resultant post-LOCA maximum suppre ssion pool temperature at 102% of uprated reactor thermal power, 3559 MWt, is 196.1º F, as shown in Table 6.2-5a. The maximum suppression pool temper ature at 3559 MWt for NUREG-0783 events is 190.7º F as evaluated in Reference 31.

The suppression pool limit for events with SRV discharge is evaluated in References 25 and 27. In the NRC's Safety Evaluation of Reference 28 for the elimination of local suppression pool temperature limits for plants with T-Quenchers, an additional concern was raised on the pote ntial transfer of non-condensed SRV steam plumes to ECCS suction strainers. An an alysis was performed in Reference 29 that modeled the steam plume formation, de termined the extent of steam plume projection, and verified that the plume can not enter ECCS suction strainers.

However, the analysis determined the existence of a potential steam ingestion concern for the "K" SRV and the Reactor Core Isolation Cooling (RCIC) suction strainer, if the temperature of the suppression pool is above 200º F. Administrative controls have been implemented to caution the operators on th e use of "K" SRV and RCIC simultaneously when the suppression pool temperature is above 200º F.

6.2.2.3.6 Sensitivity of Initiation Time of RHR Containment Cooling Mode A one-time sensitivity analysis was performed to determine the impact on the peak suppression pool temperature, if the star t of the RHR Containment Cooling Mode is delayed for longer than 10 minutes, fo llowing a DBA-LOCA. Manual operator action from the main control room is needed, in order for Suppression pool cooling to be initiated. These actions could require up to a few minutes to accomplish (accounting for valve stroke times, etc.). The impact on peak suppression pool temperature was studied if the start of suppression pool cooling is delayed from 10 minutes to 30 minutes.

LSCS-UFSAR 6.2-51a REV. 15, APRIL 2004 The study utilized power uprate decay heat loads. The results of this study indicate there is a very small impact on peak suppression pool temperature. The 30 minute case results in an increase of 1.24 deg-F, added to the current analysis peak of 193 deg-F, results in a postulated peak temper ature of less than 195 deg-F. This peak temperature does not challenge the suppression pool design limits. The operator actions to re-align RHR are anticipated to require much less time than the additional 20 minutes of this analysis. The increase in peak suppression pool temperature is concluded to be negligible (i.e. less than 1 deg-F) for these anticipated starting times which are only a few minutes longer than 10 minutes.

6.2.2.4 Test and Inspections The operational testing and the periodic inspection of components of the containment heat removal system are described in Subsection 5.4.7.4.

6.2.2.5 Instrumentation Requirements Suppression pool cooling by the RHR system is manually initiated from the control room where sufficient instrumentatio n is provided for that purpose.

6.2.3 Secondary Containment Functional Design The Secondary Containment consists of th e Reactor Building, the equipment access structure, and a portion of the main steam tunnel and has a minimum free volume of 2,875,000 cubic feet.

The reactor building completely encloses the reactor and its primary containment.

The structure provides secondary contai nment when the primary containment is closed and in service, and primary cont ainment when the primary containment is open, as it is during the refueling period. The reactor building houses the refueling and reactor servicing equipmen t, the new and spent fuel st orage facilities, and other reactor auxiliary or service equipment, including the reactor core isolation cooling system, reactor water cleanup demineralizer system, standby liquid control system, control rod drive system equipment, the emergency core cooling system, and electrical equipment components.

6.2.3.1 Design Bases The functional capability of the ventilation system to maintain negative pressure in the secondary containment with respect to ou tdoors is discussed in Subsection 9.4.2.

6.2.3.2 System Design

The reactor building is designed and constructed in accordance with the design criteria outlined in Chapter 3.0. The reactor buildin g exterior walls and superstructure up to the refueling floor are constructed of reinforced concrete.

LSCS-UFSAR 6.2-52 REV. 15, APRIL 2004 Above the level of the refueling floor, the building structure is fabricated of structural steel members, insulated siding and a metal roof. Joints in the superstructure paneling are detailed to assure leaktightness.

Penetrations of the reactor building are designed with leakage characteristics consistent with leakage requirements of the entire building. The reactor building is de signed to limit the inleakage to 100% of the reactor building free volume per day at a negative interior pressure of 0.25 inch H 2 0 gauge, while operating the standby gas tr eatment system. The building structure above the refueling floor is also designed to contain a negative interior pressure of 0.25 inch H 2 0 gauge. Personnel entrance to the reactor building is through an interlocking double door airlock. Rail car access openings in the reactor building at elevation 710 feet 6 inches provided with double doors to assure that building access will not interfere with maintaining integrity of the secondary containment.

Ventilation for the reactor building is provided by means of a once-through ventilation system. Outdoor air is filter ed then evaporatively or chilled glycol cooled to *reduce the supply air dry bulb temperature to increa se the sensible cooling capacity of this air. This air is then preheated as required to satisfy the plant operating conditions.

The equipment is arranged as follows: outside air inlet, filter, chilled glycol/heating coil evaporative *cooler (abandoned-in-place), resistive heating coils, and supply fans. Three 50% vane axial fans are provided, two of which normally operate and one which serves as a standby.

Supply air is distributed to the reactor building by means of a duct system to provide

equipment cooling in various areas as requir ed. Air is routed from clean areas to areas with progressively greater contamination potential. Pressure differential control dampers are used as required to maintain negative pressures in potentially contaminated cubicles. All exhaust air is ro uted through a return duct system to the exhaust fans.

All supply air delivered to the refueling floor level is exhausted from the periphery of

the spent fuel and equipment storage pools and the reactor well. This air is routed directly to the main system exhaust duct.

Three vane axial exhaust fans are provided, two of which normally operate and one of which serves as a standby. The discharge from the exhaust fans is routed to the plant vent where the air is discharged to the atmosphere. All exhaust air is monitored for radiation.

Normal ventilation systems are not required to operate during accident conditions and

are automatically shut down whenever the standby gas treatment system starts. The equipment for this system is not powered from essential buses. To

• Note: The evaporative coolers are abandoned-in-place.

LSCS-UFSAR 6.2-53 REV. 13 maintain the integrity of the secondary containment, two isolation dampers are provided in the supply air duct between the supply fan discharge and the penetration through the secondary containment wall.

The secondary containment structure protects the equipment in the building from externally generated missiles. Piping syst ems within the secondary containment have been analyzed for high energy pipe breaks outside primary containment and pipe whip restraints are provided as required. The effects of jet impingment have also been analyzed and included in the design of the structure and pipe whip restraints. For more information on high energy pipe breaks outside primary containment see Appendix C.

The isolation features and isolation signals for secondary containment are discussed in Section 6.5, Chapter 7.0 and Subsection 9.4.2.

6.2.3.3 Design Evaluation

The design evaluation of secondary containment ventilation system and atmospheric cleanup system is given in Sect ion 6.5 and Subsection 9.4.2.

6.2.3.4 Test and Inspections

The program for initial performance testing is outlined in the Technical Specifications. Periodic functional testing of the second ary containment and secondary containment isolation system is described in the Technical Specifications.

6.2.3.5 Instrumentation Requirements The instrumentation to be employed for the monitoring and actuation of the standby gas treatment system is fully described in Chapter 7.0.

The instrumentation used for the monitori ng and actuation of the ventilation and cleanup system is discussed in Subsections 7.3.8 and 7.6.1.2.

6.2.4 Containment Isolation System

The primary objective of the containment is olation system is to provide protection against the release of radioactive materials to the environment through the fluid system lines penetrating the containment.

This objective is accomplished by ensuring that isolation barriers are provided in all fluid lines that penetrate primary containment, and that automatic closure of the appropriate isolation valves occurs.

LSCS-UFSAR 6.2-54 REV. 13 6.2.4.1 Design Bases The design requirements for containment isolation barriers are:

a. The capability of closure or isolation of pipes or ducts that penetrate the containment is provided to ensure a containment barrier sufficient to

maintain leakage within permissible limits.

b. The arrangements of isolation valving and the criteria used to establish the isolation provisions conform to the requirements of General Design Criteria 54 through 57, as discussed in Section 3.1.

c. The design of all containment isolat ion valves and associated piping and penetrations is Seismic Category I.

d. Containment isolation valves and associated piping and penetrations meet the requirements of the ASME Boiler and Pressure Vessel Code, Section III, for Class 1 or 2 components, as applicable.

e. Isolation valves, actuators, and co ntrols are protected against loss of safety function from missiles and accident environments.

f. Containment isolation valves provide the necessary isolation of the containment in the event of accidents or other conditions to limit the

untreated release of radioactive materi als from the containment in excess of the design limits.

g. Appropriate isolation valves are auto matically closed by the signals listed in Table 6.2-21. The criteria for assigning isolation signals to their associated isolation valves is descri bed in Subsection 7.3.2. Once the isolation function is initiated, it goes to completion.

h. Redundancy and physical separation are required in the electrical and mechanical design to ensure that no single failure in the system prevents the system from performi ng its safety function.

The governing conditions under which cont ainment isolation becomes mandatory are high drywell pressure or low water level in the reactor vessel. One or both of these signals initiate closure of isolation valves not required for emergency shutdown of the plant. These same signals also initiate the ECCS. The valves associated with an ECCS may be closed remote manually from th e control room or close automatically, as appropriate.

Excess flow check valves are used as a means of automatic isolation on all static instrument sensing lines that penetrate the drywell containment and connect to LSCS-UFSAR 6.2-55 REV. 16, APRIL 2006 either the reactor pressure boundary or the drywell atmosphere. The valve is located downstream of the root valve and as close as practical to the outside surface of the containment. This valve is automatically closed to restrict flow in case of a sensing line break outside containment.

Backfill Injection lines have been added to the reference legs originating from Condensing Chambers 1(2) B21-D004A/B/C/D to comply with NRC Bulletin 93-03.

These lines use two simple check valves in series to accomplish the outboard containment isolation function. It is acceptable to use the two simple check valves instead of one excess flow check valve for the backfill injection lines because these

lines would not need the built-in bleed flow path in an excess flow check valve to reopen when appropriate. The 4 lbs./hr. CRD flow would reopen the check valves when it is available. If it is not availabl e, it is not appropriate to reopen the check valves. This meets the Regulatory Guide 1.11 "... the valve should reopen automatically or be capable of being reopened readily under the conditions that prevail when reopening is appropriate. It should not be necessary to break a line to reopen a closed valve." In addition, there is no instrument reading that will be significantly effected by the closure of these check valves.

Dead-end instrument sensing lines that are in communication with the reactor pressure boundary and penetrate the primary containment are equipped with 1/4

inch orifice as close to the process as possible inside the drywell.

6.2.4.2 System Design Table 6.2-21 presents the design information regarding the containment isolation provisions for fluid system lines and instru ment lines penetrating the containment. Containment isolation signals are identified in Table 6.2-21 and valve arrangements are represented in Figure 6.2-31.

The plant protection system signals that initiate closure of the containment isolation valves are listed in Table 7.3-2.

The isolation provisions follow the requirem ents of General Design Criteria 54, 55, 56, and 57. General Design Criteria 54 ap plies to all of the containment isolation valves. Compliance with General Design Crit eria 55, 56, and 57 is described below. The justification for this design is also presented.

6.2.4.2.1 Evaluation Agains t General Design Criterion 55 Feedwater Line

Each feedwater line forming a part of th e reactor coolant pressure boundary is provided with a swing type check valve on Unit 1 and a swing type check valve on Unit 2 inside the containment, and a nonslam type, air operated testable check valve outside the containment, as close as LSCS-UFSAR 6.2-56 REV. 14, APRIL 2002 practicable to the containment wall. In addition, a motor-operated gate valve is installed upstream of the outside isolat ion valve to provide long-term isolation capability.

During a postulated LOCA, it is desirable to maintain reactor coolant makeup from all available sources. Therefore, it would not improve safety to install a feedwater isolation valve that closed automatically on signals indicating a LOCA, and, thereby, eliminate a source of reactor makeup. The provision of the check valves, however, ensure the prevention of a significant lo ss of reactor coolant inventory and offer immediate isolation should a break occur in the feedwater line. For this reason, the outermost valve does not automatically is olate upon signal from the protection system. The valve is remote manually closed from the main control room to provide long-term leakage protection upon operat or determination that continued makeup from the feedwater system is unavailable or unnecessary.

In addition, the outboard check valve is provided with a special actuator that performs the following functions:

a. The actuator is capable of partially moving the valve disc into the flow stream during normal plant operation in order to ensure that the valve is not bound in the open position. The actuator is not capable of fully closing the valve against flow, however, and there is no significant disruption of feedwater flow.

b. The actuator is capable of applying a seating force to the valve at low differential pressures and abnormal conditions. This improves the leaktightness capability of the valves. The actuator will be utilized during leak testing.

ECCS Lines to the RPV

The subject penetration(s) meet the alternate primary containment isolation criteria of NUREG 0800 "Standard Review Plan for the review of Safety Analysis Reports for Nuclear Power Plants" (SRP) instead of the explicit requirements of GDC 55.

The HPCS, LPCS, and LPCI lines penetrate the drywell and inject coolant directly into the reactor pressure vessel. Isolation is provided on each of these lines by a normally closed check valve inside the containment and a normally closed motor-operated gate valve located outside the cont ainment, as close as practicable to the exterior wall of the containment. If a loss-of-coolant accident occurred, each of these valves would be required to open to supply coolant to the RPV. The motor-operated gate valves are automatically opened by their appropriate signals, and the check valves are opened by the coolant flow in th e line. The opening capability of the check valve can be tested by monitoring flow through the valve into the reactor vessel.

LSCS-UFSAR 6.2-57 REV. 16, APRIL 2006 Control Rod Drive Lines The control rod drive system, has two type s of lines to the RPV; the insert and withdraw lines that penetrate the drywell and connect to the control rod drive.

The control rod drive insert and withdraw lines can be isolated by the solenoid valves outside the primary containment.

These lines that extend outside the primary containment are small, and termin ate in a system that is designed to prevent out-leakage. Solenoid valves normally are closed, but open on rod movement and during reactor scram. In addition, a ball check valve located in the control rod drive flange housing automatically seals the insert line in the event of a

break. RHR and RCIC Head Spray Lines The subject penetration(s) meet the alternative primary containment isolation

criteria of NUREG 0800 "Standard Review Plan for the review of Safety Analysis Reports for Nuclear Power Plants" (SRP) in stead of the explicit requirements of GDC 55.

The RHR and RCIC head spray lines meet outside the containment to form a common line which penetrates the drywell and discharges directly into the reactor pressure vessel. The testable check valve in side the drywell is normally closed. The testable check valve is located as close as practicable to the reactor pressure vessel.

Three types of valves, a testable check valve, a normally closed motor-operated remote manual gate valve, and a normally closed motor-operated automatic globe valve, are located outside the containmen

t. The check valve assures immediate isolation of the containment in the event of a line break. The globe valve on the RHR line receives an automatic isolation signal while the gate valve on the RCIC line is remote manually actuated to provide long-term leakage control.

Standby Liquid Control System Lines The standby liquid control system line penetrates the drywell and connects to the reactor pressure vessel. In addition to a simple check valve inside the drywell, a check valve together with an explosive actuated valve are located outside the drywell. Since the standby liquid control li ne is a normally closed, nonflowing line, rupture of this line is extremely remote. The explosive actuated valve, though, functions as a third isolation valve. This valve provides an absolute seal for long-term leakage control as well as preventing leakage of sodium pentaborate into the reactor pressure vessel during normal reactor operation.

LSCS-UFSAR 6.2-57a REV. 14, APRIL 2002 Reactor Water Cleanup System

The reactor water cleanup (RWCU) pumps, heat exchangers, and filter demineralizers are located outside the primary containment. The return line from the filter demineralizers connects to the feedwater line outside the containment between the outside containment feedwate r check valve and the outboard motor-operated gate valve. Isolation of this line is provided by the feedwater system check LSCS-UFSAR 6.2-58 REV. 14, APRIL 2002 valve inside the containment, the feedwater check valve outside the containment, and a motor-operated gate valve which provid es a long term isolation capability.

During the postulated loss-of-coolant accide nt, it is desirable to maintain reactor coolant makeup. For this reason, valves wh ich automatically isolate upon signal are not included in the design of the system.

Consequently, a third valve is required to provide long-term leakage control. Should a break occur in the reactor water cleanup return line, the check valves would prevent significant loss of inventory and offer immediate isolation, while the outermost isolation valve would provide long-term

leakage control.

Recirculation Pump Seal Water Supply Line The recirculation pump seal water line extends from the recirculation pump through the drywell and connects to the CRD supply line outside the primary containment. The seal water line forms a part of the reactor coolant pressure boundary, therefore the consequences of failing this line have been evaluated. This evaluation shows that the consequences of breaking this line is less severe than that of failing an instrument line. The recirculation pump seal water line is 3/4-inch Class B from the recirculation pump through the second check valve (located outs ide the containment). From this valve to the CRD connection the line is Class D. Sh ould this line be postulated to fail and either one of the check valves is assumed not to close (single active failure), the flow rate through the broken line has been calcul ated to be substantially less than that permitted for a broken instrument line. Therefore, the two check valves in series provide sufficient isolation capability for postulated failure of this line.

RHR Shutdown Cooling Return Line The subject penetration(s) meet the altern ative primary containment isolation criteria of NUREG 0800 "Standard Review Plan for the review of Safety Analysis Reports for Nuclear Power Plants" (SRP) instead of the explicit requirements of GDC 55.

The shutdown cooling return lines are connected to the reactor recirculation pump discharge lines. The isolation valve arrangement on these lines is identical to that on the ECCS lines connected to the RPV. Ho wever, the motor-operated valve outside containment closes automatically upon receipt of an isolation signal.

RHR Shutdown Cooling Suction Line The penetration (M-7) has been protected by a relief valve mounted between the inboard automatic isolation and the containm ent penetration. This relief valve was added in response to NRC Generic Lette r GL 96-06 concerns for isolated line overpressurization during a LOCA.

Because the RHR Shutdown Cooling piping up to and including the outer containment penetration automatic isolation valve is part of the RCPB, the penetration configuration must meet GDC 55.

LSCS-UFSAR 6.2-59 REV. 13 Reactor Recirculation System Sample Line

The Reactor Recirculation sample line is a 3/4" line that is an extension of the RCPB to the outboard isolation valve. The containment penetration (M-36) has an automatic isolation inside containment and an automatic isolation outside

containment. A 3/4" bypass line with a check valve has been added around the inboard isolation valve in response to Generic Letter 96-06. The check valve will open to relieve penetration overpressurization following a LOCA. Manual valves between the check valve and the RR 24" process line will be maintained locked open, when required for overpressure protection, to assure a vent path for overpressure protection.

The two automatic valves and the inboard check valve meet the requirements of GDC 55.

6.2.4.2.2 Evaluation Agains t General Design Criterion 56

Primary Containment Chilled Water System The Primary Containment Chilled Water System (PCCW) consists of two independent trains of cooling for the prim ary containment atmosphere. Each train penetrates the containment with a supply and return line. Each line has an inboard and an outboard automatic isolatio n valve. Each penetration (M-25, M-27, M-28, M-26) has been protected by a relief valve mounted between the inboard automatic isolation and the containment penetration. These relief valves were added in response to NRC Generic Lette r GL 96-06 concerns for isolated line overpressurization during a LOCA.

The penetration configuration must meet GDC 56.

RCIC Turbine Exhaust Vacuum Breaker Line Minimum Flow Bypass

The RCIC turbine exhaust line is provided with a vacuum breaker system to prevent condensation of the exhaust steam from inducing a vacuum in the line. The vacuum relief line connects the turbine exhaust line to the suppression chamber atmosphere. Two check valves in-series in the line prevent steam from exhausting to the vapor space above the pool, and two motor-operated globe valves, one on

either side of the aforementioned check valves, provide remote manual isolation capability for the RCIC turbine exhaust vacuum breaker line.

Combustible Gas Control and Post-LOCA Atmosphere Sampling Lines

The post-LOCA sampling system lines which penetrate the containment and connect to the drywell and suppression ch amber air volume are each equipped with LSCS-UFSAR 6.2-60 REV. 13 a single divisional fail-open, solenoid operated isolation valve located outside and as close to the containment as possible. The combustible gas control system lines which penetrate the containment are equipped with two normally closed motor-operated valves in series, located outsid e containment, remote manually actuated from the control room. These valves provide assurance of isolating these lines in the event of a break and also provide long-term leakage control. In addition, the piping is considered an extension of containment boundary since it must be available for long-term usage following a de sign basis loss-of-coolant accident, and, as such, is designed to the same qualit y standards as the primary containment.

Thus, the need for isolation is conditional.

Containment Vent and Purge and Containment Drain Lines The drywell and suppression chamber vent and purge and containment drain lines have test isolation capabilities commensurate with the importance to safety of isolating these lines. Each line has two normally closed, instrument air powered, air cylinder actuated valves located outside the primary containment. The air cylinders are operated by solenoid valv es connected to the control logic. Containment isolation requirements are me t on the basis that the purge and drain lines are normally closed, low-pressure lines constructed to the same quality

standards as the containment and meet th e Branch Technical Position CSB 6-4. These isolation valves are interlocked to preclude opening of the valves while a containment isolation signal exists. Furt hermore, the consequences of a break in these lines result in no significant safety consideration.

Drywell and Suppression Chamber Air Sampling Lines The air sampling lines are used for continuously drawing containment air during normal operation as part of the leak detection system. These lines are equipped with two normally open, solenoid operated, spring to close valves in series, located outside and as close as possible to the containment. This manner of routing the system piping reduces the number of cont ainment penetrations and minimizes the potential pathways for radioactive material release. In addition, the piping upstream of the air sampling isolation valv es is considered an extension of the containment since it must be available fo r long-term usage following a design basis loss-of-coolant accident. The piping is pa rt of the post-LOCA atmosphere sampling system, and as such, is designed and fabr icated to the same quality standards as

the containment. Containment isolation requirements are met on the basis that these lines are low-pressure lines construc ted to the same quality standards as the containment furthermore, the consequences of a break in these lines result in no significant safety consideration.

LSCS-UFSAR 6.2-61 REV. 13 Service Air and Clean Condensate Supply Lines The Service Air and Clean Condensate supply lines, which penetrate the containment, provide air and water service connectors inside the drywell during reactor shutdown and outages. These lines are equipped with two manually operated valves which are locked closed during reactor operations. In addition, each line is equipped with a spool piece which is removed and respective blank flanges installed during reactor operations. The va lves and spool pieces are located outside of and as close as possible to the containment. This manner of routing the system piping reduces the number of containmen t penetrations. Since these lines are isolated during reactor operations, the po tential pathways for radioactive material release is minimized. Furthermore, the consequences of a break in these lines result in no significant safety consideration.

Reactor Building Closed Cooling Water System

The Reactor Building Closed Cooling Water System (RBCCW) inside containment consists of a closed loop providing cooling for the reactor recirculation pump heat loads and penetration heat loads. The system penetrates the containment with a supply and return line. Each line has an inboard and an outboard automatic isolation valve. Each penetration (M-16, M-17) has been protected by a relief valve mounted between the inboard automatic isolation and the containment penetration. These relief valves were added in response to NRC Generic Letter GL 96-06 concerns for isolated line overpressurization during a LOCA.

The penetration configuration must meet GDC 56.

Primary Containment Chilled Water System

The Primary Containment Chilled Water System (PCCW) consists of two independent trains of cooling for the prim ary containment atmosphere. Each train penetrates the containment with a supply and return line. Each line has an inboard and an outboard automatic isolatio n valve. Each penetration (M-25, M-27, M-28, M-26) has been protected by a relief valve mounted between the inboard

automatic isolation and the containment penetration. These relief valves were added in response to NRC Generic Lette r GL 96-06 concerns for isolated line overpressurization during a LOCA.

The penetration configuration must meet GDC 56.

6.2.4.2.3 Evaluation Agains t General Design Criterion 57 Lines penetrating the primary containment for which neither Criterion 55 nor Criterion 56 govern comprise the closed system isolation valve group.

LSCS-UFSAR 6.2-62 REV. 14, APRIL 2002 Influent and effluent lines of this group are isolated by automatic or remote manual isolation valves located as closely as possible to the containment boundary.

ECCS Pump Test Lines and Minimum Flow Bypass Lines

The LPCS, HPCS, and RHR pump test and minimum flow bypass lines have

isolation capabilities. All the pump test lines are equipped with normally closed motor-operated globe valve outside the containment that is opened only during pump testing. The RHR pump test lin es discharge below the surface of the suppression pool. Thus, the lines are not directly open to the containment atmosphere, since the pool acts to seal the discharge from the containment. The LPCS and HPCS lines discharge into the air space above the suppression pool surface. All the test lines are low-pressure lines, constructed to the same quality

standards as the containment. All valves can be remote manually operated from the main control room, and close automatically on a system start signal.

The minimum flow bypass line on the HPCS has a normally closed motor-operated gate valve located outside the containment while the LPCS and RHR are minimum flow bypass lines equipped with a normally open motor-operated gate valve. A high speed valve is utilized to assure that pump minimum flow requirements are met.

The LPCS and RHR valves are closed when adequate flow in the pump discharge lines is established. The minimum flow bypass lines connect into the associated pump test lines outside the containment.

This reduces the number of penetrations through the primary containment, thus minimizing the potential pathways for radioactive material release.

RCIC Turbine Exhaust, Vacuum Pump Discharge and RCIC Pump Minimum Flow Bypass

The RCIC turbine exhaust and vacuum pump discharge lines which penetrate the containment and connect to the suppression chamber are equipped with a normally open, motor-operated, remote manually actuated valve located as close to the containment as possible. The RCIC turbine exhaust line motor-operated isolation valve is a gate valve and the RCIC vacuum pump discharge line moter-operated isolation valve is a globe valve. In addition, there is a simple check valve upstream of the motor-operated valve which provides positive actuation for immediate isolation in the event of a break upstream of this valve. The gate valve in the RCIC turbine exhaust is designed to be locked op en in the control room and interlocked to preclude opening of the inlet steam valve to the turbine while the turbine exhaust valve is not in a full open position. The RCIC vacuum pump discharge line is also normally open but has no requirement for interlocking with the steam inlet valve to the turbine. The RCIC pump minimum flow bypass line is isolated by a normally closed motor-operated globe valve with a check valve installed upstream. This valve is controlled by sensors in the RCIC pump discharge line flow and pressure.

The valve is also remote manually controlled from the main control room.

LSCS-UFSAR 6.2-63 REV. 14, APRIL 2002 The RCIC turbine exhaust line is also provided with a vacuum breaker system to prevent condensation of the exhaust steam from inducing a vacuum in the line. The vacuum relief line connects the turbine exhaust line to the suppression chamber atmosphere.

Two check valves in-series in the line prev ent steam from exhausting to the vapor space above the pool, and two motor-operat ed globe valves provide remote manual isolation capability for the vacuum breaker line.

ECCS and RCIC Safety/Relief Valves

The safety/relief valves which serve the RHR shutdown cooling line located outside primary containment, RHR Pumps A and C suction lines, RHR Pumps A, B, and C discharge lines, RHR Heat Exchanger drain lines to the RCIC System, LPCS and HPCS suction drain lines, RHR Pumps A and B suction drain lines and discharge drain lines, RHR Pump C discharge drain line, LPCS Pump suction and pump discharge lines, and the HPCS Pump suction line and water leg pump discharge line, discharge water into the air space above the suppression pool surface. The safety/relief valve on RHR Pump B su ction line discharges water below the suppression pool surface. The safety/re lief valves on the RHR Heat Exchangers Shell Side and the RCIC steam supply lines to the RHR Heat Exchangers discharge steam below the suppression pool surface. The safety/relief valves are normally closed and provide a containment barrier in the lines. The thermal expansion safety/relief valve on the Unit 1 HPCS pump discharge line discharges water to the reactor building equipment drains and is normally closed. The thermal expansion safety/relief valve on the Unit 2 HPCS pump discharge line discharges water to the Unit 2 HPCS Pump Room and is normally closed. The safety/relief valves on the RCIC Lube Oil Cooler Supply Line, the RCIC System Pump suction line, and the RCIC Barometric Condenser discharge water to the reactor building equipment drains and are normally closed. Block valves cannot be added to the safety/relief valve discharge lines because they would preclude proper operation of the safety/relief valves, and are prohibited by the piping codes.

ECCS and RCIC Pump Suction Lines The RHR, RCIC, LPCS, and HPCS suction lines contain motor-operated, remote manually actuated, gate valves which provide assurance of isolating these lines in the event of a break. These valves also provide long-term leakage control. In addition, the suction piping from the suppression chamber is considered an extension of containment since it must be available for long-term usage following a design basis loss-of-coolant accident, and as such is designed to the same quality LSCS-UFSAR 6.2-63a REV. 14, APRIL 2002 standards as the containment. Thus, the n eed for isolation is conditional since the ECCS pumps take suction from the suppression pool in order to mitigate the consequences of LOCA. Therefore, their proper position for performing their safety fuction is open, not closed.

It should also be noted that the suction line of the ECCS pumps serves as the source of supply to the water leg pumps, which keep the ECCS discharge lines filled to avoid hydrodynamic effects on ECCS pump initiation. Isolating these water leg pumps from their supply source would de grade rather than improve the safe operation of the plant. However, the suction lines are provided with a motor-operated gate valve that can be remote manually closed from the control room, if required by a system line break or other highly unlikely event.

LSCS-UFSAR 6.2-64 REV. 17 APRIL 2008 6.2.4.2.4 Miscellaneous

Compliance with regulatory guides is addressed in Appendix B.

The isolation valves have been designed against loss of function from missiles, jet forces, pipe whip, and earthquake. The containment isolation valves and valve operators have been designed to oper ate under normal plant and postulated accident conditions. The effects of radi ation, humidity, pressure and temperature both inside and outside the containment, as defined in Chapter 3.0, have been

accounted for in the valve design.

Containment isolation valves are provided with adequate mechanical redundancy to preclude common mode failures. The power supplies to the inboard isolation valves are provided from a separate electrical division than those that supply the outboard isolation valves. Therefore, a common mode failure in one electrical division would not prevent containment isolation. The vent and purge valves consist of Air

Operated Valves and Motor Operated Valv es. See Table 6.2-21 for specific valve characteristics.

A complete list of Primary Containmen t Isolation Valves is contained in Table 6.2-28.

A leak detection system has been provided to detect leakage for determining when to isolate the affected systems that require remote manual isolation. This leak

detection system is described in Subsection 5.2.5.

The design provisions for testing the leakage rates of the containment isolation valves are shown in the valve arrangement drawings, Figure 6.2-31 as referenced in Table 6.2-21. The test connections indicated consist of a double-valved test line with provision for a pressure gauge attachment.

The design provision for testing the leakage rates of the containment isolation valves 2FC086 and 2FC115 is shown on va lve arrangement drawing, Figure 6.2-31, Sheet 10C, Detail "AD". The test connection indicated consists of a single valve test line with a provision for a pressure gauge attachment.

6.2.4.3 Design Evaluation

The main objective of the containment isolation system is to provide protection by preventing releases to the environment of radioactive materials. Redundancy is provided in design aspects to satisfy the requirement that an active failure of a single valve or component does not prev ent containment isolation: Mechanical components are redundant, as shown by the isolation valve arrangements.

LSCS-UFSAR 6.2-65 REV. 17 APRIL 2008 Electrical redundancy is provided in isolation valve arrangements to eliminate dependence on a single power source to attain isolation. Electrical cables for isolation valves in the same process line have been routed separately. Cables have been selected based upon the specific environment to which they will be subjected.

Provisions ensure that the position of all nonpowered isolation valves is maintained.

For all powered valves, the position is indicated in the main control room. A discussion of the instrumentation and controls associated with the isolation valves is given in Chapter 7.0.

In single failure analysis of electrical systems, no distinction is made between mechanically active or passive components; all fluid system components such as valves are considered "electrically active" whether or not "mechanical" action is required.

Electrical systems as well as mechanical systems are designed to meet the single failure criterion for both mechanically active and passive fluid system components

regardless of whether that component is required to perform a safety action. Even though a component such as an electrically operated valve is not designed to receive a signal to change state (open or closed) in a safety scheme, it is assumed as a single failure that the system component changes state or fails. Electrically operated valves include valves that are electrically piloted but air operated as well as valves that are directly operated by an electrical device. In addition, all electrically operated valves that are automatically actu ated also can be manually actuated from the main control room. Therefore, a single failure in any electrical system is analyzed regardless of whether the loss of a safety function is caused by component failing to perform a requisite mechanical motion or a component performing an unnecessary mechanical motion.

6.2.4.4 Tests and Inspections

A discussion of the testing and inspection pe rtaining to isolation valves is provided in Subsection 6.2.6, the Technica l Specifications, and Table 6.2-21.

6.2.5 Combustible Gas Control in Containment

In order to assure that the containment integrity is not endangered due to the generation of combustible gases following a postulated LOCA, systems for controlling the relative concentrations of su ch gases are provided within the plant. The system includes subsystems for mixing the containment atmosphere, monitoring hydrogen concentration, reduci ng combustible gas concentrations, and, as a backup, purging. The hydrogen recombining function of the hydrogen recombiners is abandoned in place.

LSCS-UFSAR 6.2-66 REV. 17, APRIL 2008 6.2.5.1 Design Bases The hydrogen recombining function of th e hydrogen recombiners is abandoned in place. The valves that provide RHR cooling water to the hydrogen recombiners are also abandoned in place in the closed posi tion. The blower an d associated piping are not abandoned and remain operational to maintain the drywell mixing function. The design basis information for the hydrogen recombination function remains for historical reference.

The following design bases were used for the combustible gas control system design:

a. A double-ended rupture of a main recirculation line results in the most rapid coolant loss and reactor depressurization, with the coolant being discharged from both ends of th e break. The noncondensable gas initially in the drywell is forced into the suppression chamber during the RPV depressurization phase. This transfer process takes place through downcomers that connect the drywell and suppression chambers. The postulated metal-water reaction begins in the core region and is assumed to produce hydrogen immediately after the recirculation pipe breaks. The reaction would last 2 minutes during

which 0.945% of the active Zircaloy fuel cladding has reacted. The radiolysis of the coolant in the core region, water sump on the drywell floor and suppression pool also is assumed to begin immediately. The hydrogen and oxygen thus generated will evolve to drywell and suppression chamber atmospheres.

b. The combustible gas control system has the capability for monitoring the hydrogen concentration in drywell and suppression chamber and alarming as the hydrogen concentrat ion reaches 4%. It also has the capability of mixing the atmosphere s of both drywell and suppression chamber. It also will control the combustible gas concentrations in the primary containment without reliance on purging and without the release of radioactive material to the environment.

c. The primary systems for combustible gas control, including measuring, meet the design, quality assurance, redundancy, energy source, and instrumentation requirements for an engineered safety feature system according to Appendix A of 10 CFR 50.

d. The combustible gas control system will be activated after a LOCA in time to assure that the hydrogen concentration does not exceed 4 volume percent of hydrogen in either the drywell or wetwell atmospheres. In addition, the LSCS containment is nitrogen inerted to

LSCS-UFSAR 6.2-67 REV. 17, APRIL 2008 an oxygen concentration of 4% by volume. This is below the combustible limit of oxygen in hydrogen but still provides enough oxygen to react with all the hydrogen that would be produced by the metal water reaction.

e. One recombiner system is provid ed for each nuclear unit. Each recombiner is capable of being cross-connected to the other unit to provide 100% redundancy. The recomb iners are located outside of the primary containment in an accessible area and, therefore, routine maintenance, testing and/or inspection can be performed during normal plant operation or shutdown conditions.

f. The components of the combustible gas control system are protected from missiles and pipe whip to assu re proper operation under accident conditions as required for safety-related systems. The system has been designed to perform in the event of failure of any one of its active components.

g. The combustible gas control systems are designed as Seismic Category I devices. As previously mentioned, the units are capable of being cross-connected to provide redundancy and are further capable of withstanding the temperature and pressure transients resulting from a LOCA. All components that can be subjected to containment atmosphere are capable of withstanding the humidity, temperature, pressure, and radiation conditions in the containment following a LOCA. h. The combustible gas control system is designed to remain operable in the postaccident environment in the reactor building. Components subjected to the reactor containment postaccident environment are likewise designed for those conditions.

i. The combustible gas control system recombiner units are located outside of the primary containment in an accessible area. They can be inspected or tested during normal plant operation or during shutdown conditions.

j. The hydrogen recombiner units are fixed units that are permanently installed; therefore, it is not necessary to have the ability to transport

them. k. The recombiner units are remotely started from the control room and the local control panel in the auxiliary electric equipment room. They are designed such that there are no local operating adjustments required on a unit operating in a post-LOCA environment. This fact eliminates the necessity of biological shielding.

LSCS-UFSAR 6.2-67a REV. 17, APRIL 2008 6.2.5.2 System Design The combustible gas control system consists of four subsystems: a mixing system, a hydrogen monitoring system, two hydrogen recombiners, and a purge system. The design features of these four systems are described in the following sections.

The hydrogen recombining function of th e hydrogen recombiners is abandoned in place. The valves that provide RHR cooling water to the hydrogen recombiners are also abandoned in place in the closed posi tion. The blower an d associated piping are not abandoned and remain operational to maintain the drywell mixing function. The design basis information for the hydrogen recombination function remains for historical reference.

LSCS-UFSAR 6.2-68 REV. 14, APRIL 2002 Hydrogen Mixing System The function of the mixing subsystem is to ensure that local concentrations with greater than 4% hydrogen cannot occur within the primary containment following a LOCA. The atmospheres of both drywell proper and suppression chamber area, each of

which is a single compartment, are well mixe

d. The mixing is achieved by natural convection processes. Natural convection occurs as a result of the temperature difference between the bulk gas space in the vessel and the containment wall. The natural convective action is enhanced by the momentum of steam emitted from the point of rupture. Th ere are two interior subcompartments where gases may not achieve thorough mixing with the bulk containment atmosphere. The drywell

head area, which is for reactor vesse l refueling purposes, is one such subcompartment. The other is the control rod drive area immediately below the reactor pressure vessel. The physical arrangements and/or location of the monitoring system and the hydrogen recombiner system are such that concentrations above the 4% limit of combustible gases will not occur.

The atmosphere between the drywell and suppression pools will be mixed during the depressurization phase of the LOCA. The hydrogen recombiner units will also serve to affect mixing between these two compartments. The hydrogen recombiner will take suction on the drywell and discharge to the suppression pool. This will in turn cause the atmosphere from the suppression pool to circulate into the drywell via the vacuum breaker lines.

The monitoring system will alert the operator of the concentration within these subcompartments and the positions of the effluent and suction points of the recombiner will preclude the building of concentrations above the limit in these areas as well as the dryw ell and wetwell proper.

Hydrogen Monitoring System The hydrogen monitoring system form s a part of the primary containment monitoring system which is discussed in Subsection 7.5.2.

Hydrogen Recombiner System The concentration of combustible gases in the primary containment (drywell and suppression pool areas) following a LOCA is controlled by the hydrogen recombiner system. The combustible gas control system contains one hydrogen recombiner per reactor unit. The hydrogen recombiner is located outside of the primary containment. The amount of Hydr ogen in the effluent gas being returned to the wetwell shall not exceed 0.1% by volume. The system will process the primary containment atmosphere at a rate of at least 125 scfm using a blower to supply containment gases to the recombiner. The recombination process LSCS-UFSAR 6.2-69 REV. 14, APRIL 2002 takes place within the recombiner as a resu lt of an exothermic reaction. The steam is then cooled and the resulting water and remaining gases are returned to the primary containment. Suction is taken from the drywell area, and the discharge is returned to the suppression po ol area above water level.

The hydrogen recombiner unit is skid mounted and is an integral package. All pressure containing equipment including pi ping between components is considered as an extension of the containment and, ther efore, is designed as ASME III Class 2. The skid and the equipment mounted on it are designed to meet Seismic Category I

requirements. The hydrogen recombiner system is designed to accommodate conditions present in the containment (temperature and pressure) following a LOCA event. Piping and instrumentation for the system are shown in Drawing No. M-130. The hydrogen recombiner unit, which requires a 1-2 hour warmup period, is initiated manually from the control room and the local control panel in the aux. electric equipment room. It is initiate d prior to primary containment hydrogen concentration reaching 3 volume percent which occurs approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the accident. Based on the original core loading, the time at which containment hydrogen generation reaches 4 volume percent varies with fuel types located in the core. However, this is acceptable based on Design Basis described in Section 6.2.5.1.d. Once placed in operation, the system continues to operate until it is manually shut down when an adequate margin below the hydrogen concentration design limit is reached. The operation of the system can be tested from the control room or the auxiliary equipment room. The test consists of energizing the blower and heaters and observing system operatio n to see if components are performing properly. Flow and pressure measuremen t devices are periodically calibrated.

The hydrogen recombiner system is serviced by electrical power and cooling water systems, which are placed in operation concurrent with a loss-of-coolant accident. Cooling water required for the operation of the system is taken from the residual heat removal system. The cooling water is utilized to cool the water vapor and the residual gases leaving the recombiner prior to returning them to the containment. All hydrogen recombiner unit cooling water is returned to the suppression pool.

Each recombiner unit has the capability of serving either containment; therefore, there is 100% redundancy of all components and controls.

All functions and controls necessary to start the combustible gas control system are also located in the control room and in the auxiliary electric equipment room which is readily accessible from the control room.

LSCS-UFSAR 6.2-70 REV. 17, APRIL 2008 6.2.5.3 Design Evaluation

The hydrogen recombining function of th e hydrogen recombiners is abandoned in place. The valves that provide RHR cooling water to the hydrogen recombiners are also abandoned in place in the closed posi tion. The blower an d associated piping are not abandoned and remain operational to maintain the drywell mixing function. The design basis information for the hydrogen recombination function remains for historical reference.

6.2.5.3.1 General

In evaluating the combustible gas control system design, it was found necessary to

consider:

a. hydrogen generated in the post-LOCA environment, b. resultant drywell and containment concentrations, and

c. the functional requirements of the combustible gas control system.

The following analytical results are provided:

a. The beta, gamma, and beta plus ga mma energy release rates plotted as functions of time (Figure 6.2-32).

b. The integrated beta, gamma and beta plus gamma energy release plotted as functions of time (Figure 6.2-33).

c. The integrated production of combustible gas within the containment (drywell and suppression chamber) plotted as a function of time for each source (i.e., metal-water reac tion and radiolysis) (Figure 6.2-34).

d. The concentration of combustible gas in the drywell and suppression chamber plotted as a function of ti me, if uncontrolled (Figure 6.2-35). This curve establishes the basis for activation of the combustible gas

control system.

e. The combustible gas concentration in the containment (drywell and suppression chamber) plotted as a function of time with (125 scfm) 100% recombiner capacity initiated at 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after LOCA (Figure 6.2-36).

LSCS-UFSAR 6.2-71 REV. 14, APRIL 2002 6.2.5.3.2 Sources of Hydrogen Short-Term Hydrogen Generation

In the period immediately after the LOCA, hydrogen is generated by both radiolysis and metal-water reaction. However, in ev aluating short-term hydrogen generation, the contribution from radiolysis is insignificant when compared to the hydrogen generated by the metal-water reaction. The only metal-water reaction considered to be significant is reaction of water with the zirconium fuel cladding which produces hydrogen by the following reaction:

Zr + 2H 2 O ZrO 2 + 2H 2 Based on loss-of-coolant accident calculat ional procedures and the analyses of emergency core cooling system (ECCS) pe rformance in conformance with 10 CFR 50.46 and Appendix K, the extent of the above chemical reaction is estimated to be

0.1% of the fuel cladding material. However, the metal-water reaction-generated hydrogen based on a core-wide penetrat ion of 0.00023 inches for 764 bundles with each bundle containing 101 pounds of zircon ium in the active fuel cladding, results in a 0.945% metal-water reaction. Therefore, 0.945% of fuel cladding, which is greater than five times the maximum amount calculated in accordance with 10 CFR 50.46, is assumed to react with water to produce hydrogen. The duration of this reaction is assumed to be 120 seconds with a constant re action rate. The resulting hydrogen is assumed to be uniformly distributed in the drywell containment. This assumption is supported by the test data reported in BNWL 1592 of July 1971.

Figure 6.2-34 presents the accumulated hydrogen generation as a result of this chemical reaction.

Long-Term Hydrogen Generation

Hydrogen is also produced by decomposition of water due to absorption of the fission product decay energy immediately after LOCA.

2H 2 O 2H 2 + O 2 Generation of hydrogen and oxygen due to radiolysis of coolant water is an important factor in determining the long-term gas mixture composition within the containment compartments. Conservative assumptions were used to determine the fission product distribution model that applies after the accident and, therefore, the hydrogen generation rates. The incore radiolysis contributes hydrogen to the drywell, and radiolysis of the suppression pool water contributes hydrogen directly to the suppression chamber. Hydrogen is also discharged from the radiolysis of sump water on drywell floor into the drywell atmosphe re. The total decay energy utilized in the analyses was based on American Nuclear Society Standard ANS 5.1-1979 multiplied by a factor of 1.2, conservatively assuming a 1000-day reactor LSCS-UFSAR 6.2-72 REV. 14, APRIL 2002 operating time at constant full power level to determine the fission product buildup.

Halogen and noble gas inventories were determined from TID-14844.

Hydrogen can also be formed by corrosion of metals in the containment. The significant portion of this source is from the co rrosion of zinc and aluminum. Since the spray system uses only demineralized water for the purpose of reducing temperature and pressure inside the drywe ll, the corrosion of aluminum and zinc will contribute a negligible amount of hydrogen to the containment atmosphere.

Hydrogen is, during normal operation of the plant, dissolved in the primary system water. Figure 6.2-35 presents the accumulated hydrogen and oxygen generation from both chemical reaction and radiolysis decomposition of water.

6.2.5.3.3 Accident Description A complete description of the post-LOCA cond itions is found in Subsection 6.2.1 and Section 6.3.

Following the postulated LOCA, the postulated metal-water reaction begins in the core region and is assumed to produce hydrogen immediately after the recirculation pipe breaks. The reaction lasts 2 minut es during which 0.945% of the active zircaloy fuel cladding reacts. The radiolysis of the coolant in the core region, water sump on the drywell floor and suppression pool is assumed to begin immediately.

The hydrogen and oxygen thus generated will evolve to drywell and suppression chamber atmospheres. The hydrogen conc entration in the drywell would, after about 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, approach the flammability limit if uncontrolled. The hydrogen recombiner system is manually activated before the hydrogen concentration reaches 3 volume percent. The recombiner system takes gases from the drywell atmosphere, recombines the hydrogen with oxygen to form water vapor, and returns the resulting cooled water and remaining gases to the suppression chamber. The pressure buildup in the suppression chamber due to the operation of recombiner system taking suction on the drywell and discharging to the suppression pool will cause the opening of the vacuum brea ker valves between the drywell and suppression chamber. As a result, the flow of the gas mixture from the wetwell to the drywell will balance the negative pressure differential between two volumes and

will also result in lower concentrations due to the influx of the wetwell gases.

6.2.5.3.4 Analysis

Based on the above hydrogen sources and the accident description, the hydrogen concentration in the drywell and suppression chamber is calculated as a function of time. In formulating the model of the Mark II containment for these calculations, a conservative assumption is made, name ly the interchange of mass between the drywell and the suppression chamber through downcomers which takes place during blowdown process is neglected, th at is, no hydrogen is removed from the drywell except through the recombiner system. This assumption is conservative, as LSCS-UFSAR 6.2-73 REV. 15, APRIL 2004 it results in a shorter time for the drywell hydrogen concentration to reach the flammability limit. Furthermore, the hydr ogen and oxygen gases can flow back to the drywell from suppression chamber thro ugh vacuum breakers due to pressure increase in the suppression chamber by th e operation of the recombiner system.

Table 6.2-22 gives all of the necessary parameters used to determine the amount of

hydrogen generation in the LSCS analys is. The results of the analyses are presented in Figures 6.2-35 and 6.2-36. It was determined that the uncontrolled hydrogen concentration in the drywell eventually reaches 4% by volume (dry basis) approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after the LOCA. The suppression chamber hydrogen concentration was determined to be 3.0% by volume due to radiolytic hydrogen generation. Prior to the drywell concentrat ion reaching 3% by volume, a recombiner system is activated. A single system is designed to keep the hydrogen concentration below 4% by volume at all times until ra diolytic generation has ceased. The performance of the recombiner system, which is initiated 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after LOCA, is shown in Figure 6.2-36. The hydrogen conc entration is 3.0% by volume at the time of initiation. Thus, the use of a sing le 125 scfm recombiner system provides effective control of hydrogen concentrat ion and, therefore, would prevent the formation of combustible gas mixture in both drywell and suppression chamber.

6.2.5.4 Testing and Inspections

Each active component of the combustible gas control system is testable during normal reactor power operation.

The combustible gas control systems and the containment purge system will be tested periodically to assure that they will operate correctly.

Preoperational tests of the combustible gas control system are conducted during the final stages of plant construction prior to initial startup (Chapter 14.0). These tests assure correct functioning of all controls, instrumentation, recombiners, piping, and valves. System reference characteristics, such as pressure differentials and flow rates, are documented during the preoperational tests and are used as base points for measurements in subsequent operational tests.

6.2.5.5 Instrumentation Requirements

The instrumentation provisions for actuating the combustible gas control system and monitoring the system are described in Subsection 7.3.5.

6.2.6 Containment Leakage Testing This section presents the testing program for the reactor containment, containment penetrations and containment isolation barriers that comply with the requirements of the General Design Criteria and Append ix J to 10 CFR 50. Each of the tests LSCS-UFSAR 6.2-74 REV. 14, APRIL 2002 described in this Subsection was perf ormed as a preopera tional and will be performed as a periodic test.

6.2.6.1 Containment Integrated Leakage Rate Test

Following the completion of the construction, repair, inspection, and testing of welded joints, penetrations, and mechanical closures including the satisfactory completion of the structural integrity test s as described in Subsection 3.8.1.7, a preoperational containment leakage rate test was performed to verify that the

actual containment leak rate does not exceed the design limits.

In order to ensure a successful integrated leak rate test, loca l leakage tests (Type B and C tests) were performed on penetrations and isolation valves, and repairs are made, if necessary, to ensure that leakage through the containment isolation barriers does not exceed

the design limits.

An integrated leakage rate test is then performed on the entire containment in order to determine that the total leakage (exclusive of MSIV leakage) through

containment isolation barriers does not ex ceed the maximum allowable leakage rate of 0.635% per day at the calculated peak co ntainment internal pressure at 39.9 psig. The pertinent test data, including test pressures and acceptance criteria, is presented in Table 6.2-23.

Pretest requirements have been descri bed in the preoperational test abstract included in Chapter 14.0 of the FSAR. As stated therein, power operated isolation valves will be closed by their actuators prior to the start of the integrated leakage rate test.

During the integrated leak rate test the containment systems are configured as follows; a. Reactor building closed cooling water - lined up for normal operation; isolation valves closed and system filled.

b. Primary containment chilled water - lined up for normal operation; isolation valves closed and system filled.

c. Residual heat removal - One loop lined up in shutdown cooling mode. Other loops lined up in low-pressure coolant injection standby mode and isolated, containment and suppression pool spray flow paths isolated, full flow test lines isolated, reactor head cooling flow path isolated, minimum flow isolated, shutdown cooling discharge line isolated on standby system and condensate discharge from RHR heat exchangers shell side flow pa th isolated; system filled.

d. Low-pressure core spray - system filled and isolated.

LSCS-UFSAR 6.2-75 REV. 13 e. High-pressure core spray - system filled and isolated.

f. Reactor core isolation coolin g - isolation valves closed; RCIC condensate filled and isolated. RCI C full flow test return line to suppression pool filled and isolated.

g. Reactor water cleanup - suction line filled and isolated; return line filled and isolated.

h. Standby liquid control - lines filled and isolated.

i. Control rod drive - lined up in scram conditions; pumps off, system filled. j. Reactor recirculation system - pumps off, system filled.

k. RPV and primary containment instrumentation - lines filled and vented to containment instrumentation to the RPV or drywell will be opened. l. Neutron monitoring sytem (TIP) - TIPs will be fully withdrawn and the ball valves closed.

m. Floor and equipment drains - sumps pumped down to low water level, isolation valves closed.

n. Clean condensate - drained and ve nted, isolation valves closed, spool piece removed and blind flange installed or filled and isolated and system leakage added to type A result.

o. Service air - vented, isolation valves closed, spool piece removed and blind flange installed.

p. Feedwater - filled and isolated.

q. Main steam - filled, isolation valves closed.

r. Containment monitoring - post-LOCA monitoring system open to containment, pumps off, valves op en; drywell monitoring and sampling system isolated, pumps off.

s. Post-LOCA hydrogen control - lined up for unit operation, isolation valves open or isolated and system leakage added to type A result.

t. Primary containment instrument air - all accumulators vented, isolation valves closed.

LSCS-UFSAR 6.2-76 REV. 17, APRIL 2008

u. Fuel Pool Cooling - Cycled Condensate to Refueling Bellows filled and isolated, Reactor Well Drain filled and isolated.

v. All accessible liner leak test channel plugs are verified installed.

The Type C leak rates for the following penetrations are added to the Type A test results on a Minimum-Path Basis:

a. reactor building closed cooling water, b. primary containment chilled water, c. RHR shutdown cooling suction, d. reactor core isolation cooling steam supply, e. reactor water cleanup suction, f. reactor water sample, g. floor and equipment drains,

h. inboard MSIV drain,

i. Feedwater Lines, j. RCIC Full Flow Test Return Line to Suppression Pool.

k. Cycled Condensate to Refueling Bellows

l. Reactor Well Drain

Measures will be taken to ensure stabilization of the containment conditions prior to containment leakage rate testing.

The test method utilized is the absolute method, as described in ANSI/ANS 56.8-1994. The test procedure, test equipment and facilities, period of testing, and verification of leak test accuracy also follow the recommendations of ANSI/ANS 56.8-1994.

The acceptance criteria for the preoperational containment integrated leakage rate test are in compliance with the criteria given in Appendix J of 10 CFR 50. except as LSCS-UFSAR 6.2-77 REV. 13 noted below. Structural verification test acceptance criteria are described in Subsection 3.8.1.7.

The acceptance criteria for the periodic containment integrated leakage rate test are in compliance with the criteria given in 10CFR50 Appendix J Option B, NRC Reg Guide 1.163, NEI-94-01, Rev. 0, and ANSI/ANS 56.8-1994. The As-Found Type

A test leakage must be less than the acceptance criterian of 1.0 La (Primary Containment overall leakage rate acceptance criterion). During the first unit startup following testing (prior to enteri ng a mode where containment integrity is required) the As-Left Type A leakag e rate shall not exceed 0.75 La.

6.2.6.2 Containment Penetration Leakage Rate Test Containment penetrations whose design inco rporates resilient seals, gaskets, or sealant compounds; air lock door seals, equipment and access doors with resilient seals or gaskets; and other such penetrat ions received a preoperational and will be periodically leak tested in accordance wi th Appendix J of 10 CFR 50 except as noted in the following paragraph.

The following penetrations were preoperationally and will be periodically tested to Type B criteria:

a. equipment access hatch, b. personnel air lock, by (when co ntainment integrity is required, the personnel airlock should be test ed within 7 days after each containment access except when the airlock is being used for multiple entries, then at least once per 30 da ys, by verifying seal leakage to be less than or equal to 5 scfh when the gap between the door seals is pressurized to greater than or equal to 10 psig - exception to 10 CFR 50 Appendix J) overall air lock leakage rate is less than or equal to 0.05 La when tested at greater than or equal to Pa.

c. drywell head,

d. suppression chamber access hatches, e. CRD removal hatch, f. electrical penetrations, g. TIP penetration flanges, SA flange and MC flange, h. Drywell to suppression pool vacuum breaker and associated manual isolation valves flanges and actuator seals, LSCS-UFSAR 6.2-78 REV. 13 i. Vent and purge isolation valve flanges, and packing

j. HPCS minimum flow line branch line 1(2)HP20C-2" Blind flanges

k. RCIC spectacle flange 1(2)E51-D316 blind flange half when required.

See Table 6.2-21 note 49.

l. ECCS Relief Valves Discharg ing to Suppression Pool Inbound (Containment Side) Flanges.

It should be noted that no pipe penetrations are provided with expansion bellows.

The containment penetration is an anchor point in the system, and the thermal movements have been accounted for on this basis. Therefore, no leakage rate testing of expansion bellows penetration assemblies will be required.

Test methods utilized to determine containment penetration leak rates are described as follows:

a. Equipment Access, CRD Removal, and Suppression Chamber Acess The equipment access hatch has been furnished with a double-gasketed flange and bolted dished door, as shown in Figure 3.8-34. The CRD removal and suppression chamber access hatches have been furnished with a double-gasketed flan ge and bolted door. Provision is made to test pressurize the space between the double gaskets of the door flanges and the doors.

b. Personnel Air Lock The personnel lock is constructed as a double-door, latched, welded steel vessel, as shown in Figure 3.8-33. The space between the air doors can be pressurized to peak containment pressure through the test connections provided. Each of the doors are provided with a test connection for pressurizing between the seals.

In addition, all four shaft seal assemblies are provided with a test connection to allow for indivi dual shaft seal leak test.

c. Drywell Head A double-gasketed seal and test tap, as shown in Figure 3.8-5, is provided for leak rate testing of the drywell head.

d. Electrical Penetrations

LSCS-UFSAR 6.2-79 REV. 13 Each electrical penetration, as represented in Figure 3.8-21 and listed in Table 3.8-1 (with an "E" penetr ation number), is provided with a pressure gauge to monitor leakage. The double-gasketed and O-ring

seals are provided with a test connection for leak rate testing.

e. Tip Penetration Flanges, Clean Co ndensate (MC) and Service Air (SA)

Penetrations Each TIP MC or SA penetration flange is provided with a double-gasketed seal and a test connection for type B leak testing.

f. Drywell to Suppression Pool Vacuum Breakers Each drywell to suppression pool vacuum breaker has two double-

gasketed flanges and a manual actuator O-ring and shaft seal. These seals are provided with test connections for leak testing. The Vacuum Breaker line manual isolation valves have a double-gasketed flange on the inboard or containment side provid ed with test connections for leak testing. The outboard flanges on the manual isolation valves are leak tested by pressurizing the entire vacuum breaker line and performing

soap bubble test on the outboard flan ge. The stem seal or packing of these valves will be tested either locally or by primary containment

pressurization and subsequent soap bubble inspection.

g. Vent and Purge Isolation Valves Each inboard vent and purge valve has a double-gasketed flanged seal on its containment side. These seals are provided with test connections for leak testing. The stem packing of these valves is also provided with a test connection for packing leak test. See also Table 6.2-21 Note 41.

h. HPCS Minimum Flow Line Blind Flanges One double-gasketed blind flange is installed on each of the HPCS

minimum flow line branch connections 1(2)HP20C-2". These flanges are provided with a test connection for type B leak testing.

i. RCIC Spectacle Flange 1(2)E51-D316 The installed blind flange half of spectacle flange 1(2)E51-D316 is tested by pressurizing with air the upstream RCIC full flow test return line to Condensate Storage Tank and then check for leaks at the flange upstream gasket joint. Done when required per Table 6.2-21 note 49.

LSCS-UFSAR 6.2-80 REV. 13 j. ECCS Relief Valves' Containment Side Flanges are Type B tested by one of the following methods: Test Port/Testable Gasket; Primary Containment Pressurization and subsequent soap bubble inspection; Special Test Equipment mounted over the flange thus pressurizing against the gasket.

Test pressures are given in Table 6.2-23.

The acceptance criteria for the preoperational containment penetration leakage rate test is in compliance with the criteria given in Appendix J of 10 CFR 50. The periodic test acceptance criteria is established in accordance with the LaSalle

County Station Local Leak Rate Test Program, and also is in agreement with Appendix J Option B of 10 CFR 50, NRC Regulatory Guide 1.163, Nuclear Energy Institute NEI-94-01 Rev. 0, and ANSI/ANS-56.8-1994.

6.2.6.3 Containment Isolation Valve Leakage Rate Test

Those containment isolation valves that are to receive a Type C test are so indicated in Table 6.2-21.

Test taps for leakage rate testing have been provided on the lines associated with the containment isolation valves. These taps are indicated on the valve arrangement drawings associated with Ta ble 6.2-21. The test method is as described in Appendix J of 10 CFR 50. Te st pressures are shown in Table 6.2-23.

The acceptance criteria for the leakage rate testing is given in Table 6.2-23 and the Primary Containment Leak Rate Testing Program.

6.2.6.4 Scheduling and Reporting of Periodic Tests

The periodic leakage test schedule is give n in the LaSalle County Station Leak Rate Test Program.

6.2.6.5 Special Testing Requirements The secondary containment will be tested as required by the Technical Specifications.

6.2.7 References

1. F. J. Moody, "Maximum Two-Phase Vessel Blowdown from Pipes," Topical Report APED-4827, General Electric Company, 1965.

LSCS-UFSAR 6.2-81 REV. 14, APRIL 2002

2. A. J. James, "The General Electr ic Pressure Suppression Containment Analytical Model, (NEDO-10320), April 1971.

3. A. J. James, "The General Electr ic Pressure Suppression Containment Analytical Model," April 1971, Su pplement 1, (NEDO-10320), May 1971.

4. K. V. Moore and W. H. Ratting , "RELAP 4-A Computer Program for Transient Thermal-Hydraulic Analysis, "ANCR-1127, Aerojet Nuclear

Company, December 1973.

5. F. J. Moody, "Maximum Rate of a Single Component, Two Phase Mixture," Journal of Heat Transfer, Transactions, American Society of Mechanical Engineers, Vol. 87, No. 1, February 1965.

6. I. E. Idelchik, Handbook of Hydraulic Resistance, AEC-TR-6630, 1966.

7. "RELAP 4/MOD5 A Computer Progra m for Transient Thermal- Hydraulic Analysis of Nuclear Reactors and Related Systems," ANCR-NUREG-1335, Aerojet Nuclear Company, September 1976.

8. NEI 94-01, Rev. 0, July 26, 1995, Nuclear Energy Institute Industry Guideline for Implementing Performance-Based Option of 10CFR Part 50

Appendix J.

9. ANSI/ANS 56.8-1994, American Nation al Standard for Containment System Leakage Testing Requirements.

10. GE Document EAS-49-0888, "Justification of Continued Operation With Increased Suppression Pool Temperature at LaSalle County Station,"

Revision 1, August 1988. (Proprietary)

11. Technical Specification Submittal Lette r Sections 3.6.2.1 and 4.6.2.1, dated 10-07-88.

12. Amendment 67 for Unit 1 (Facil ity Operating License NFP-11), and Amendment 49 for Unit 2 (Facility Oper ating License NFP-18), dated July 7, 1989.

13. Calc. L001799, Rev. 0, "Assessment of Containment Line Base Mat Reactor Pedestal, Downcomer Bracing, Drywell Floor & Suppression Pool Columns for Suppression Pool Temperature Increase." 14. Calc. L001800, Rev. 0, "Assessment of Containment Wall for Suppression Pool Temperature Increase" LSCS-UFSAR 6.2-81a REV. 14, APRIL 2002

15. Calc. L001810, Rev. 0, "Impact of Increase in the Suppression Pool Temperature at LaSalle on Design Basis Suppression Pool Dynamic Loads." 16. Letter from ComEd NFS dated 5-07-98, Nuclear Fuel Services Letter, NFS:BSA:98-055, dated 5-08-98, from R.W.

Tsai to G. Campbell, "Impact of Initial Suppression Pool Temperature on Hydrogen Generation" 17. Calc. 3C7-0181-003, Rev. 3, "Suppr ession Pool Temperature Transient Studies" 18. General Electric Letter Repo rt GE-NE-B13-01920-013, January 1998, "Current Suppression Pool Water Temperatures Following a Design Basis Accident for LaSalle County Station Units 1 and 2"

19. General Electric Report EAS-083-1188, "Elimination of the High Suppression Pool Temperature Limit for LaSalle County Station Units 1 & 2", dated November 1988.

20. General Electric Letter Repo rt GE-NE-T23-00762-00-01, July 1998, "Evaluation of Peak Suppression Pool Temperature with Assumption of Feedwater Coastdown and Reduced RHR Flow Rate During Long-Term Containment Cooling" 21. Letter from J. A. Benjamin (ComEd) to U. S. NRC, "Request for a Change to the Technical Specifications, 'Vacuum Relief System'" dated August 6, 1999.

22. Letter from J. A. Benjamin (Com Ed) to U. S. NRC, "Supplemental Information to Request for a Change to the Technical Specifications to Vacuum Relief System" dated November 15, 1999.

23. Letter dated December 21, 1999 from D.

M. Skay to O. D.

Kingsley, "Issuance of Amendments, approved amendment 138 for LaSalle Unit 1 and amendment 122 for LaSalle Unit 2."

24. Licensing Topical Report, "Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," NEDC-31897P-A, May 1992.

25. LaSalle County Station Power Uprate Project, Task 400, "Containment System Response," GE-NE-A1300384-02-01R1, Revision 1, October 1999 (and Task Report Changes based on Stea m Plume Analysis, GE-LPUP-332, dated 5/4/2000).

LSCS-UFSAR 6.2-82 REV. 15, APRIL 2004

26. General Electric Company, "General Electric Company Analytical Model for Loss-of Coolant Analysis in Accordance with 10CFR50 Appendix K," NEDO-20566A, September 1986.

27. ComEd letter to NRC, "Response to Request for Additional Information License Amendment Request for Power Uprate Operation," dated 3/31/2000.

28. General Electric Company, NEDO-30832, "Elimination of Limit on Local Suppression Pool Temperature for SRV Discharge with Quenchers," Class I, December 1984, (NRC approved version wi th NRC Safety Evaluation Report issued as NEDO-30832-A, Class I, May 1995).

29. General Electric Analysis of LaSalle Steam Plume Ingestion Potential, NSA 00-116, dated 3/29/2000.

30. LaSalle County Station Power Uprate Project, Task 401, "Annulus Pressurization," GE-NE-A1300384-06-01, Revision 0, June 1999.

31. Design Analysis No. L-002874, Rev. 0, "LaSalle County Station Power Uprate Project Task 400: Containment System Response (GE-NG-A1300384-02-01 R3) Revision 3".

32. EC #334017, Rev. 0, "Increased Cooling Water Temperature Evaluation to a new Maximum Allowable of 104

°F."

LSCS-UFSAR TABLE 6.2-1 (SHEET 1 OF 2) TABLE 6.2-1 REV. 14 - APRIL 2002 CONTAINMENT DESIGN PARAMETERS DRYWELL SUPPRESSION CHAMBER A. Drywell and Suppression Chamber 1. Internal design pressure, psig 45 45 2. External design pressure, psig 5 5 3. Drywell deck design differential pressure, psid a) Downward 25 25 b) Upward 5 5 4. Design temperature, °F 340 275 5. Drywell (including vents) net free volume, ft 3 229,538 6. Design leak ratio, %/day @ 45 psig 0.5 0.5 7. Suppression chamber free volume, ft 3 a) minimum 164,800 b) maximum 168,100 8. Suppression chamber water volume a) Minimum, ft 3 128,800 b) Maximum, ft 3 131,900 9. Pool cross-section area, ft 2 a) Water surface (excluding pedestal and drywell floor support columns) 4999 b) Total 5899 10. Pool depth (normal), ft 26.5

LSCS-UFSAR TABLE 6.2-1 (SHEET 2 OF 2) TABLE 6.2-1 REV. 0 - APRIL 1984 DRYWELL SUPPRESSION CHAMBER B. Vent System 1. Number of downcomers 98

2. Internal downcomer diameter, in. 23.5

3. Total vent area, ft 2* 295 4. Downcomer submergence*

12 ft 4 in. (maximum) 5. Downcomer loss factor*

5.2

• The actual limiting area is 232 ft 2 based on the opening size through the downcomer protective covers (top hats). The corresponding loss factor is 3.2.

However, since the analysis requires that the entrance losses, pipe losses and exit losses be based on a single area, the higher loss factor of 5.2 was utilized, resulting in a higher pressure and, th erefore, a more conservative analysis.

LSCS-UFSAR TABLE 6.2-2 (SHEET 1 OF 2) TABLE 6.2-2 REV. 14, APRIL 2002 ENGINEERED SAFETY SYSTEMS INFORMATION FOR CONTAINMENT RESPONSE ANALYSES (AT 3434 MWt)

CONTAINMENT ANALYSIS VALUE* FULL CAPACITY CASE A CASE B CASE CA. Drywell Spray System (RHR system)

1. Number of pumps 2 2 1 0 2. Number of lines 2 2 1 0 3. Number of headers 2 2 1 0 4. Spray flow rate, gpm/pump 6700 6700 6700 0 5. Spray thermal efficiency, %

--- --- --- --- B. Suppression Pool Spray (RHR system)

1. Number of pumps 2 2 1 0 2. Number of lines 2 2 1 0 3. Number of headers 1 1 1 0 4. Spray flow rate, gpm/pump 450 450 450 0 5. Spray thermal efficiency, %

--- --- --- --- C. Containment Cooling System (RHR system) 1. Number of pumps 2 2 1 1 2. Pump capacity, gpm/pump 7450** 7450 3. Heat exchangers

a. Type - inverted U-tube, single pass shell, multipass tubes, vertical mounting

b. Number 2 2 1 1 c. Heat transfer area, ft 2 /unit 11,000 11,000 11,000 11,000 d. Overall heat transfer coefficient, Btu/hr - ft 2 - °F 215

• Cases A, B, and C defined in Table 6.2-5. ** A supplementary evaluation has been performed for a slightly reduced RHR pump flow rate of 7200 gpm (suppression pool cooling mode); as discussed in Section 6.2.2.3.4 long term suppression pool temperature is not significantly impacted and the peak long term pool temperature does not exceed the 200

°F maximum value given in Table 6.2-5.

LSCS-UFSAR TABLE 6.2-2 (SHEET 2 OF 2)

TABLE 6.2-2 REV. 14, APRIL 2002 FULL CAPACITY CONTAINMENT ANALYSIS VALUE* CASE A CASE B CASE C e. Secondary coolant flow rate per exchanger, lb/hr 3.7x10 6 --- 3.7x10 6 --- f. Design service water temperature (CSCS) Minimum, °F 32 Maximum, °F 100 100 100 100 D. ECCS Systems:

1. High-pressure core spray (HPCS)

a. Number of pumps 1 1 1 1

b. Number of lines 1 1 1 1

c. Flow rate, gpm 6200 6200 6200 6200

2. Low-pressure core spray (LPCS)

a. Number of pumps 1 1 0 0

b. Number of lines 1 1 0 0

c. Flow rate (rated), gpm/line 6250 6250 0 0

d. Number of headers 2 2 0 0 3. Low-pressure coolant injection (LCPI) a. Number of pumps 3 3 1 1 b. Number of lines 3 3 1 1

c. Flow rate, gpm/line 7067 7067 7067 7067

4. Residual heat removal (RHR)

a. Pump flow rate:

Shell side 7450** Tube side 7400 b. Source of cooling water RHR service water c. Flow begins, seconds Manual, approximately 600 *** E. Automatic Depressurization System 1. Total number of safety/relief valves 18 2. Number actuated on ADS 7

• • • Refer to Section 6.2.2.3.6 for further discussion on the sensitivity of this time period.

• Cases A, B, and C defined in Table 6.2-5.

LSCS-UFSAR TABLE 6.2-3 (SHEET 1 OF 2) TABLE 6.2-3 REV. 14, APRIL 2002 (AT 3434 MWt)

INITIAL CONDITIONS EMPLOYED IN CONTAINMENT RESPONSE ANALYSES A. Reactor Coolant System (at 105% rated steam flow and at normal liquid levels) 1. Reactor power level, MWt 3434 2. Average coolant pressure, psig 1025

3. Average coolant temperature, °F 550

4. Mass of reactor coolant system liquid, lbm 676,700

5. Mass of reactor coolant system steam, lbm 24,900

6. Liquid plus steam energy, Btu 380 x 10 6 7. Volume of water in vessel, ft 3 11,175 8. Volume of steam in vessel, ft 3 9,640 9. Volume of water in recirculation loops, ft 3 1,030 10. Volume of steam in steamlines, ft 3 1,030 11. Volume of water in feedwater line, ft 3 20,778* 12. Volume of water in miscellaneous lines, ft 3 191 13. Total reactor coolant volume, ft 3 22,712 14. Stored water

a. Condensate storage tank, gal 350,000 b. Fuel storage pool, ft 3 50,000

• Does not represent the feedwater vo lume used in post-LOCA feedwater coastdown/injection evaluation. This evaluation is discussed in detail in Section 6.2.1.1.3.1.1 in paragraph titled, "Evaluation of Post-LOCA Feedwater Injection".

LSCS-UFSAR TABLE 6.2-3 (SHEET 2 OF 2) TABLE 6.2-3 REV. 15, APRIL 2004 B. Containment Drywell Suppression Chamber

1. Pressure, psig 0.75 0.75 2. Inside temperature, °F 135 100* 3. Outside temperature, °F 104 104 4. Relative humidity, % 20 100

5. Service water temperature (CSCS), °F 100 100 6. Water volume, ft 3 (minimum)

--- 128,800 7. Vent submergence, (maximum)

--- 12 ft 4 in.

• As discussed in Section 6.2.1.8 supplementary evaluations have been satisfactorily completed with a 105

°F initial suppression pool temperature.

LSCS-UFSAR TABLE 6.2-3A TABLE 6.2-3a REV. 17 APRIL 2008 INITIAL CONDITIONS EMPLOYED IN CONTAINMENT RESPONSE ANALYSIS (AT 3559 MWt)

A. Reactor Coolant System 1. Reactor power level, MWt 3559 2. Average coolant pressure, psig 1025

B. Containment Drywell Suppression Chamber

1. Pressure, psig 0.75 0.75

2. Inside temperature, °F 135 105

3. Relative humidity, % 20 100

4. Service water temperature (CSCS), °F (1) 100 100

5. Water volume, ft 3 (minimum) ---- 128,800* (maximum) 131,900*

6. Vent submergence, ft (minimum) ---- 11.7 (maximum) 12.33

• Conservative values used in Reference 22.

(1) Evaluated for post-accident peak of 104

°F in Reference 32.

LSCS-UFSAR TABLE 6.2-4 TABLE 6.2-4 REV. 14, APRIL 2002 MASS AND ENERGY RELEASE DATA FOR ANALYSIS OF WATER POOL PRESSURE-SUPPRESSION CONTAINMENT ACCIDENTS (AT 3434 MWt)

A. Effective accident break area (total), ft 2 3.113 Pipe ID, in.

21.686 B. Components of effective break area:

1. Recirculation line area, ft 2 2.565 2. Cleanup line area, ft 2 0.080 3. Jet pumps area, ft 2 0.468 C. Break area/vent area ratio 0.010 D. Primary system energy distribution

• 1. Steam energy, 10 6 Btu 29.6 2. Liquid energy, 10 6 Btu 355.3 3. Sensible energy, 10 6 Btu a. Reactor vessel 106.1 b. Reactor internals (less core) 58.6 c. Primary system piping 34.6 d. Fuel** 25.2 E. Assumptions used in pressure transient analysis

1. Feedwater valve closure time Instantaneous See Note 1

2. MSIV closure time (sec) 3.5 3. Scram time (sec)

< 1 4. Liquid carryover, %

100 5. Turbine stop valve closure (sec) 0.2

• All energy values except fuel are based on a 32°F datum.

• • Fuel energy is based on a datum of 285°F.

Note 1 This assumption has been supplemented for a conservative evaluation on the peak long term suppression pool temperature. This supplemental evaluation postulates the addition of feedwater mass and energy injected at time t=600 seconds after LOCA. Section 6.2.1.1.3.1.1 in the paragraph titled, "Evaluation of Post-LOCA Feedwater Injection" discusses this in further detail.

LSCS-UFSAR TABLE 6.2-5 TABLE 6.2-5 REV. 14, APRIL 2002 LOSS OF COOLANT ACCIDENT LONG TERM PRIMARY CONTAINMENT RESPONSE

SUMMARY

(AT 3434 MWt)

CASE LPCI AND/OR LPCS PUMPS

SERVICE WATER PUMPS CONTAINMENT SPRAY (gal/min)

HPCS (gal/min)

LPCI AND/OR LPCS (gal/min)

PEAK POOL TEMPERATURE (°F) ** SECONDARY PEAK PRESSURE (psig) A 3/1 4 14,134 6200 21,200/

6,250 168.4 5.3 B 1/0 2 7,067 6200 7067/0 200 9.6 C 1/0 2 0 6200 7067/0 200++ 14.2 ** Supplementary evaluations have been performed, as discussed in Section 6.2.1.8, based on an increase in the initial suppression pool temperature (from 100

°F to 105°F), the peak suppression pool bulk temperature is less than 200

°F. ++ A supplementary evaluation, for the effect on long term peak pool temperature, has been performed for the addition of feedwater mass and energy at t=600 seconds and a reduced RHR pump flow in the suppression pool cooling mode (7200 gpm versus 7450 gpm). The 200

°F peak pool temperature given above is not exceeded.

TABLE 6.2-5a REV. 16, APRIL 2006 LSCS-UFSAR TABLE 6.2-5A LOSS OF COOLANT ACCIDENT LONG TERM PRIMARY CONTAINMENT RESPONSE

SUMMARY

(AT 3559 MWt)

CASE LPCI AND/OR LPCS PUMPS

SERVICE WATER PUMPS

CONTAINMENT SPRAY (gal/min)

HPCS (gal/min)

LPCI AND/OR LPCS (gal/min)

PEAK POOL TEMPERATURE* (°F) PRIMARY PEAK SUPPRESSION CHAMBER PRESSURE (PSIG) SECONDARY SUPPRESSION CHAMBER PEAK PRESSURE (psig) C 1/0 2 0 6200 7200

• • 196.1 27.9 12.4

• See Figures 6.2-5A, 6.2-6A and 6.2-7A fo r long term containment responses vs. time.

• • RHR flow through heat exchanger (Reference 20)

LSCS-UFSAR TABLE 6.2-6 TABLE 6.2-6 REV. 14, APRIL 2002 ENERGY BALANCE FOR DESIGN-BASIS RECIRCULATION LINE BREAK ACCIDENT (AT 3434 MWt)

PRIOR TO DBA (0 sec) AT TIME OF PEAK PRESSURE DIFFERENCE (0.75 at Recirc.)

AT END OF BLOWDOWN

(~53 sec) AT TIME OF PEAK CONTAINMENT PRESSURE (~27009 sec - minimum ECCS available; ~7047 sec - all ECCS Available)

UNIT 1. Reactor coolant (vessel &

pipe inventory) 414.0 x 10 6 400 x 10 6 11.8 x 10 6 45.6 x 10 6 /41.8 x 10 6 Btu 2. Fuel and cladding 34.0 Fuel 34.8 x 10 6 32.3 x 10 6 12.8 x 10 6 4.07 x 10 6 /3.72 x 10 6 Btu Cladding 3.05 x 10 6 3.05 x 10 6 2.99 x 10 6 0.956 x 10 6 /0.904 x 10 6 Btu 3. Core internals, also reactor coolant piping pumps &

valves 91.2 x 10 6 91.2 x 10 6 91.2 x 10 6 31.4 x 10 6 /55.5 x 10 6 Btu 4. Reactor vessel metal 107.0 x 10 6 107.0 x 10 6 107.0 x 10 6 37 x 10 6 /64.4 x 10 6 Btu 5. Reactor coolant piping, pumps and valves Included in (3)

6. Blowdown enthalpy NA 546 NA NA Btu/lbm 7. Decay heat 0 .402920 x 10 6 8.802 x 10 6 1020 x 10 6 /383.0 x 10 6 Btu 8. Metal-water reaction heat 0 0 0.02 x 10 6 .471 x 10 6 /.471 x 10 6 Btu 9. Drywell structures Storage Capacitance Neglected Btu 10. Drywell air 1.52 x 10 6 1.73 x 10 6 0 1.77 x 10 6 /158 x 10 6 Btu 11. Drywell steam 0.335 x 10 6 7.41 x 10 6 25.7 x 10 6 7.06 x 10 6 /5.32 x 10 6 Btu 12. Containment air 1.17 x 10 6 1.17 x 10 6 2.77 x 10 6 1.41 x 10 6 /1.49 x 10 6 Btu 13. Containment steam 0.522 x 10 6 0.522 x 10 6 1.29 x 10 6 5.57 x 10 6 /2.86 x 10 6 Btu 14. Suppression pool water 887 x 10 6 887 x 10 6 1300 x 10 6 1770 x 10 6 /1490 x 10 6 Btu 15. Heat transferred by heat exchangers 0 0 0 752 x 10 6 /260 x 10 6 Btu NOTE 1: Results of analysis for MS and recirc line breaks are approximately the same; however, the progress of the events is more rapid for the MS break than for the recirc. Note 2: A supplementary evaluation, for the effect on long term peak pool temperature, has been performed for the addition of feedwater mass and energy injection at t=600 seconds, the total additional energy calculated due to the feedwater volume and the feedwater piping metal sensible heat is 2.07 x E08 Btu. (Ref. 18).

LSCS-UFSAR TABLE 6.2-7 TABLE 6.2-7 REV. 15, APRIL 2004 ACCIDENT CHRONOLOGY DESIGN-BASIS RECIRCULATION LINE BREAK ACCIDENT (AT 3434 MWt)

TIME (sec) ALL ECCS IN OPERATION MINIMUM ECCS AVAILABLE Vents cleared 0.824 0.824 Drywell reaches peak pressure 20.14 20.14

Maximum positive differential pressure occurs 0.831 0.831 Initiation of the ECCS 30 30 End of blowdown 52.15 52.15 Vessel reflooded ( ) 109.53 Introduction of RHR heat exchanger (approx.) 600* (approx.) 600*

Containment reaches peak secondary pressure 10,915 27,009

• Refer to Section 6.2.2.3.6 for further discussion on the sensitivity of this time period.

LSCS-UFSAR TABLE 6.2-8 TABLE 6.2-8 REV. 14 - APRIL 2002

SUMMARY

OF ACCIDENT RESULTS FOR CONTAINMENT RESPONSE TO RECIRCULATION LINE AND STEAMLINE BREAKS (AT 3434 MWt)

A. Accident Parameters RECIRCULATION LINE BREAK

• STEAMLINE BREAK 1. Peak drywell pressure, psig 39.6 32 2. Peak drywell deck differential pressure, psid 24.2 17.5 3. Time(s) of peak pressures, sec 22 11 4. Peak drywell temperature, °F 286 320 5. Peak suppression chamber pressure, psig 30.6 25 6. Time of peak suppression chamber pressure, sec 144 50 7. Peak suppression pool temperature during blowdown, °F 148** 100** 8. Peak suppression pool temperature, long term, °F 200++ 9. Calculated drywell margin, % 12

10. Calculated suppression chamber margin, %

32 11. Calculated deck differential pressure margin, %

3.2

• See Figures 6.2-2 and 6.2-5 for plots of pressures vs time.

See Figures 6.2-3 and 6.2-7 for plots of temperatures vs time.

• • As discussed in Section 6.2.1.8 supplementary evaluations have been satisfactorily completed with a 105

°F initial suppression pool temperature. ++ See Notes in Table 6.2-5.

LSCS-UFSAR TABLE 6.2-8A TABLE 6.2-8a REV. 14 - APRIL 2002

SUMMARY

OF ACCIDENT RESULTS FOR SHORT-TERM CONTAINMENT RESPONSE TO RECIRCULATION LINE BREAK (AT 3559 MWt)

A. Accident Parameters RECIRCULATION LINE BREAK

• 1. Peak drywell pressure, psig 39.9 2. Peak drywell deck differential pressure, psid 22.4 3. Peak drywell temperature, °F 286

• See Figure 6.2-2A for short-term pressure response vs time. See Figure 6.2-3A for short-term temperature response vs time.

LSCS-UFSAR TABLE 6.2-9 TABLE 6.2-9 REV. 13 SUBCMPARTMENT NODAL DESCRIPTION RECIRCULATION OUTLET LINE BREAK WITH SHIELDING DOORS NODE NUMBER DESCRIPTION VOLUME (ft 3) HEIGHT (ft) FLOW CROSS-SECTIONAL AREA (ft) BOTTOM ELEVATION (ft) INITIAL CONDITIONS TEMP, (°F) PRESS, (psia) HUMID, *(%) CALC. PEAK

PRESS DIFF, (psid) 1 Lower Reactor Skirt 100.6 5.07 18.40 755.29 550.

15.45 0.1 47.9 2 Lower Reactor Skirt 100.6 5.07 18.40 755.29 550.

15.45 0.1 48.0 3 Lower Reactor Skirt 100.6 5.07 18.40 755.29 550.

15.45 0.1 47.4 4 Lower Reactor Skirt 150.9 5.07 23.36 755.29 550.

15.45 0.1 47.9 5 Lower Reactor Skirt 150.9 5.07 23.36 755.29 550.

15.45 0.1 48.1 6 Upper Reactor Skirt 121.0 7.47 20.98 760.36 550.

15.45 0.1 47.9 7 Upper Reactor Skirt 121.0 7.47 20.98 760.36 550.

15.45 0.1 48.0 8 Upper Reactor Skirt 121.0 7.47 20.98 760.36 550.

15.45 0.1 47.5 9 Upper Reactor Skirt 181.5 7.47 25.64 760.36 550.

15.45 0.1 48.1 10 Upper Reactor Skirt 181.5 7.47 25.64 760.36 550.

15.45 0.1 47.8 11 Lower Recirc. Noz. Sect. 39.87 6.92 10.02 767.83 550.

15.45 0.1 74.2 12 Lower Recirc. Noz. Sect. 54.28 4.90 10.50 767.83 550.

15.45 0.1 47.3 13 Lower Recirc. Noz. Sect. 61.94 4.90 10.50 767.83 550.

15.45 0.1 48.2 14 Lower Recirc. Noz. Sect. 81.43 4.90 13.47 767.83 550.

15.45 0.1 48.2 15 Lower Recirc. Noz. Sect. 80.54 4.90 13.47 767.83 550.

15.45 0.1 46.4 16 Upper Recirc. Noz. Sect. 26.77 2.67 8.43 774.75 550.

15.45 0.1 72.0 17 Upper Recirc. Noz. Sect. 52.18 4.69 10.30 772.73 550.

15.45 0.1 45.2 18 Upper Recirc. Noz. Sect. 52.18 4.69 10.30 772.73 550.

15.45 0.1 40.9 19 Upper Recirc. Noz. Sect. 78.28 4.69 13.27 772.73 550.

15.45 0.1 37.7 20 Upper Recirc. Noz. Sect. 77.39 4.69 13.27 772.73 550.

15.45 0.1 37.2 21 Mid-Section 67.48 6.41 12.44 777.42 550.

15.45 0.1 39.5 22 Mid-Section 67.48 6.41 12.44 777.42 550.

15.45 0.1 39.2 23 Mid-Section 67.48 6.41 12.44 777.42 550.

15.45 0.1 36.7 24 Mid-Section 101.2 6.41 15.52 777.42 550.

15.45 0.1 36.0 25 Mid-Section 101.2 6.41 15.52 777.42 550.

15.45 0.1 35.9 26 LPCI Noz. Sect. 171.1 9.59 18.61 783.83 550.

15.45 0.1 27.6 27 LPCI Noz. Sect. 155.8 9.59 18.61 783.83 550.

15.45 0.1 27.3 28 LPCI Noz. Sect. 155.8 9.59 18.61 783.83 550.

15.45 0.1 26.7 29 LPCI Noz. Sect. 171.1 9.59 18.61 783.83 550.

15.45 0.1 26.5 30 Feedwater Noz. Sect. 155.8 8.81 17.86 793.42 550.

15.45 0.1 19.3 31 Feedwater Noz. Sect. 153.4 8.81 17.86 793.42 550.

15.45 0.1 19.0 32 Feedwater Noz. Sect. 143.9 8.81 17.86 793.42 550.

15.45 0.1 18.9 33 Feedwater Noz. Sect. 164.1 8.81 17.86 793.42 550.

15.45 0.1 19.0 34 Annulus Receiver 19.76 6.92 10.02 767.83 550.

15.45 0.1 115.1 35 Break Node 19.52 4.92 7.04 769.56 550.

15.45 0.1 322.0 36 Upper Drywell 16315. 41.0 400. 793.42 135.

15.45 15.0 -- 37 Mid-Drywell 11665. 12.1 965. 781.32 135.

15.45 15.0 -- 38 Lower Drywell 82775. 44.7 1850. 736.62 135.

15.45 15.0 --

• Relative humidity.

LSCS-UFSAR TABLE 6.2-10 TABLE 6.2-10 REV. 0 - APRIL 1984 SUBCOMPARTMENT NODAL DESCRIPTION FEEDWATER LINE BREAK WITH SHIELDING DOORS NODE NUMBER DESCRIPTION VOLUME (ft 3) HEIGHT (ft) FLOW CROSS-SECTIONAL AREA (ft) BOTTOM ELEVATION (ft) INITIAL CONDITIONS TEMP, (°F) PRESS, (psia) HUMID, *(%) CALC. PEAK PRESS DIFF, (psid) 1 Lower Reactor Skirt 150.9 5.07 23.36 755.29 550. 15.45 0.1 14.0 2 Lower Reactor Skirt 150.9 5.07 23.36 755.29 550. 15.45 0.1 14.0 3 Lower Reactor Skirt 150.9 5.07 23.36 755.29 550. 15.45 0.1 14.0 4 Lower Reactor Skirt 150.9 5.07 23.36 755.29 550. 15.45 0.1 14.1 5 Upper Reactor Skirt 181.5 7.47 23.80 760.36 550.

15.45 0.1 14.0 6 Upper Reactor Skirt 181.5 7.47 23.80 760.36 550.

15.45 0.1 13.9 7 Upper Reactor Skirt 181.5 7.47 23.80 760.36 550.

15.45 0.1 14.0 8 Upper Reactor Skirt 181.5 7.47 23.80 760.36 550.

15.45 0.1 14.1 9 Recirc. Noz. Sect. 159.7 9.59 17.83 767.83 550. 15.45 0.1 14.4 10 Recirc. Noz. Sect. 157.9 9.59 17.83 767.83 550. 15.45 0.1 14.1 11 Recirc. Noz. Sect. 157.9 9.59 17.83 767.83 550. 15.45 0.1 13.6 12 Recirc. Noz. Sect. 167.4 9.59 17.83 767.83 550. 15.45 0.1 13.4 13 Mid-Section 67.48 6.41 12.44 777.42 550. 15.45 0.1 18.2 14 Mid-Section 67.48 6.41 12.44 777.42 550. 15.45 0.1 15.5 15 Mid-Section 67.48 6.41 12.44 777.42 550. 15.45 0.1 14.0 16 Mid-Section 101.2 6.41 15.79 777.42 550. 15.45 0.1 13.5 17 Mid-Section 101.2 6.41 15.79 777.42 550. 15.45 0.1 13.3 18 LPCI Noz. Sect. 100.8 9.59 15.52 783.83 550. 15.45 0.1 17.7 19 LPCI Noz. Sect. 110.0 9.59 15.52 783.83 550. 15.45 0.1 16.1 20 LPCI Noz. Sect. 116.1 9.59 15.52 783.83 550. 15.45 0.1 14.3 21 LPCI Noz. Sect. 171.1 9.59 18.61 783.83 550. 15.45 0.1 13.0 22 LPCI Noz. Sect. 155.8 9.59 18.61 783.83 550. 15.45 0.1 12.6 23 Annulus Receiver 45.22 10.58 13.39 793.42 550.

15.45 0.1 50.8 24 Feedwater Noz. Sect. 55.63 10.58 13.39 793.42 550. 15.45 0.1 36.9 25 Feedwater Noz. Sect. 116.2 10.58 16.48 793.42 550. 15.45 0.1 21.3 26 Feedwater Noz. Sect. 131.5 10.58 16.48 793.42 550. 15.45 0.1 11.5 27 Feedwater Noz. Sect. 176.7 10.58 19.57 793.42 550. 15.45 0.1 10.5 28 Feedwater Noz. Sect. 171.8 10.58 19.57 793.42 550. 15.45 0.1 10.3 29 Break Node 16.12 4.00 5.42 796.75 550. 15.45 0.1 201.6 30 Lower Drywell 16315. 41.00 400. 793.42 135. 15.45 15.0 -- 31 Mid Drywell 11665. 12.10 965. 781.32 135. 15.45 15.0 -- 32 Upper Drywell 82775. 44.70 1850. 736.62 135. 15.45 15.0 --

• Relative humidity.

LSCS-UFSAR TABLE 6.2-11 TABLE 6.2-11 REV. 13 SUBCOMPARTMENT NODAL DESCRIPTION SIMULTANEOUS BREAK OF THE HEAD SPRAY LINE AND RPV HEAD VENT LINE IN THE HEAD CAVITY INITIAL CONDITIONS DBA BREAK CONDITIONS Volume No. Description Height, ft Cross-Sectional Area, ft 2 Volume ft 3 Temp.

°F Press.

psia Humid. *% Break Loc. Vol. No. Break Line Break Area, ft 2 Brea k Type Calc. Peak Press Diff. psid Design Peak Press Diff. psid Design Margin % 1 Head Cavity 15.57 261.5 4072. 135. 15.45 0.1 1 1RI24B

-6 + 1NB13 A-4 .163 Doubl e-ended guillo tine break 7.0 nodes 1 to 2 10.6 150 2 Drywell 79.74 2315.0 184664. 135. 15.45 0.1 3 Wetwell 33.87 5198.0 176085. 100** 15.45 0.1

• Relative humidity The peak differential pressure across the bulkhead plate (Pnode 1 - Pnode 2) for this case = 7.0 psid Design differential pressure across the bulkhead plate = 10.6 psid ** As discussed in Section 6.2.1.8 supplementary evaluations have been satisfactorily completed with a 105ºF initial suppression pool temperature.

LSCS-UFSAR TABLE 6.2-12 TABLE 6.2-12 REV. 13 SUBCOMPARTMENT NODAL DESCRIPTION RECIRCULATION LINE BREAK IN THE DRYWELL INITIAL CONDITIONS DBA BREAK CONDITIONS Volume No. Description Height, ft Cross-Sectional Area, ft 2 Volume ft 3 Temp

. °F Press

. psia Humid.* % Break Loc. Vol. No. Break Line Break Area, ft 2 Break Type Calc.

Peak Press Diff. psid Design Peak Press Diff. psid Design Margin % 1 Head Cavity 15.57 261.5 4072. 135. 15.45 0.1 2 Drywell 79.74 2315.0 177049. 135. 15.45 0.1 2 Recirculation line 3.216 Double-ended guilloti ne 9.0 10.6 118 3 Wetwell 33.87 5198.0 176085. 100** 15.45 0.1

• Relative humidity The peak differential pressure across the bulkhead plate (P node1 - P node 2) for this case = -9.0 psid

The design differential pressure across the bulkhead plate = 10.6 psid

• • As discussed in Section 6.2.1.8 supplementary evaluations have been satisfactorily completed with a 105

°F initial suppression pool temperature.

LSCS-UFSAR TABLE 6.2-13 TABLE 6.2-13 REV. 14 - APRIL 2002 SUBCOMPARTMENT VENT PATH DESCRIPTION SIMULTANEOUS BREAK OF THE HEAD SPRAY LINE AND RPV HEAD VENT LINE IN THE HEAD CAVITY VENT PATH NO FROM VOL. NODE NO. TO VOL. NODE NO. DESCRIPTION OF VENT PATH FLOW

HYDRAULIC HEAD LOSS, K CHOKED UNCHOKED AREA**

ft2 LENGTH** ft DIAMETER ft FRICTION K, ft/d TURNING LOSS, K EXPANSION, K CONTRACTION, K TOTAL 1 1 2 HVAC vents through bulkhead plate 6.12 3.76 - - - - 2.62 choked 2* 2 3 98-24 inch downcomers 295.00 70.8 19.38 - - - - 1.90 unchoked 3 0 1 Break of head spray line & RPV head vent line in head cavity 0.163 0.0 0.46 - - - - 0.00 fill

• Opened on a differential pressure of 5.2 psid

• • Length/Area is the inertial term input directly into RELAP4 / MOD5

LSCS-UFSAR TABLE 6.2-14 TABLE 6.2-14 REV. 14 - APRIL 2002 SUBCOMPARTMENT VENT PATH DESCRIPTION RECIRCULATION LINE BREAK IN THE DRYWELL VENT PATH NO FROM VOL. NODE NO. TO VOL. NODE NO. DESCRIPTION OF VENT PATH FLOW

HYDRAULIC HEAD LOSS, K CHOKED UNCHOKED AREA**ft2 LENGTH** ft DIAMETER ft FRICTION K, ft/d TURNING LOSS, K EXPANSION, K CONTRACTION, K TOTAL 1 1 2 HVAC vents without ductwork through bulkhead plate 11.12 6.12 3.76 - - - - 2.62 unchoked 2* 2 3 98-24 inch downcomers 295.00 70.8 19.38 - - - - 1.90 unchoked 3 0 2 Recirculation line break in drywell 1.00 0.00 0.46 - - - - 0.00 fill

• Opened on a differential pressure of 5.2 psid

• • Length/Area is the inertial term input directly into RELAP4 / MOD5

LSCS-UFSAR TABLE 6.2-15 REV. 0 - APRIL 1984 TABLE 6.2-15 SIMULTANEOUS BREAK OF THE HEAD SPRAY LINE AND RPV HEAD VENT LINE IN THE HEAD CAVITY INPUT DATA*

- LASALLE - HEAD CAVITY PRESSURIZATION - 3C7-0476-003 REV 0 4266-00

• PROBLEM DIMENSIONS 010001 -2 9 5 3 3 0 0 3 0 1 0 1 0 0 0 0 0 3

• PROBLEM CONSTANTS 010002 0.0 1.0

• EDIT VARIABLES 020000 AP 1 AP 2 AP 3 JW 2 AH 1 JW 1 TD 1 FD 1 AD 1

• TIME STEP DATA 030010 1 50 0 0 0.01 0.00005 2.0 030020 1 50 0 0 0.002 0.00005 3.5 030030 1 50 0 0 0.0005 0.00005 3.9 030040 1 50 0 0 0.01 0.00005 8.0 030050 1 50 0 0 0.1 0.00005 30.0

• TRIP CONTROLS 040010 1 1 0 0 20.0 0.0 040020 2 4 2 3 5.2 0.0 040030 3 1 0 0 0.0 0.0

• VOLUME DATA 050011 0 0 15.45 135. .001 4072. 15.57 0. 0 261.5 18.3 819.73 0 050021 0 0 15.45 135. .001 184664. 79.74 0. 0 2315.0 54.3 740.00 0 050031 0 0 15.45 100. .001 176085. 33.87 0. 0 5198.0 81.4 706.14 0

• JUNCTION DATA 080011 1 2 0 0 0. 11.12 819.73 .55 2.62 2.62 0 1 0 3 0. .6 -100 080021 2 3 0 1 0.0 295.000 740.00 .24 1. 9 1. 9 0 1 0 30 .6 -100. 080031 0 1 1 0 357.9 0.16267 827.52 .00 0.0 0.0 0 1 0 30. 1. -100.

• VALVE DATA CARDS 110010 -2 0 0 0. 0. 0. 0.

• FILL TABLE CARDS 130100 3 1 2 3 'LBS/SEC' 550. 1. 0. 130101 0 2200. 30. 2200.

______________________

• RELAP4/MOD5 computer code utilized for analysis.

LSCS-UFSAR TABLE 6.2-16 REV. 0 - APRIL 1984 TABLE 6.2-16 RECIRCULATION LINE BREAK INPUT DATA*

= LASALLE - HEAD CAVITY PRESSURIZATION - 3C7-0476-003 REV 0 4266-00

• RECIRCULATION LINE BREAK

• 4 HVAC INLET VENTS AVAILABLE FOR FLOW INTO HEAD CAVITY

• PROBLEM DIMENSIONS 010001 -2 9 2 3 3 0 0 3 0 1 0 1 0 0 0 0 0 3

• PROBLEM CONSTANTS 010002 0.0 1.0

• EDIT VARIABLES 020000 AP 1 AP 2 AP 3 JW 2 AH 2 JW 1 TD 1 FD 1 TD 2

• TIME STEP DATA 030010 1 50 0 0 0.005 0.00005 2.0 030020 1 50 0 0 0.01 0.00005 30.0

• TRIP CONTROLS 040010 1 1 0 0 10.0 0.0 040020 2 1 0 0 0.824 0.0 040030 3 1 0 0 0.0 0.0

• VOLUME DATA 050011 0 0 15.45 135. .001 4072. 15.57 0. 0 261.5 18.3 819.73 0 050021 0 0 15.45 135. .001 177049. 79.74 0. 0 2315.0 54.3 740.00 0 050031 0 0 15.45 100. .001 176085. 33.87 0. 0 5198.0 81.4 706.14 0

• JUNCTION DATA 080011 1 2 0 0 0. 4.92 819.73 .83 1.52 1.52 0 1 0 3 0. .6 -100. 080021 2 3 0 1 0.0 295.000 740.00 .24 1. 9 1. 9 0 1 0 3 0. .6 -100. 080031 0 2 1 0 25690. 1. 770. .00 0.0 0.0 0 1 0 3 0. 1. -100.

• VALVE DATA CARDS 110010 -2 0 0 0. 0. 0. 0.

• FILL TABLE CARDS 130100 3 4 9 1 'LBS/SEC'

• TIME FLOW ENTHALPY 130101 0.0000 22710.0 532.0 130102 0.0016 22710.0 532.0 130103 0.0017 34060.0 532.0 130104 1.5500 34060.0 532.0 130105 1.5600 27550.0 532.0 130106 1.7500 27550.0 532.0 130107 1.7600 24840.0 547.0 130108 1.9800 24840.0 547.0 130109 10.1100 24320.0 538.0 _______________________

• RELAP4/MOD5 computer code utilized for analysis.

LSCS-UFSAR TABLE 6.2-17 REV. 0- APRIL 1984 TABLE 6.2-17 MAIN STEAMLINE BREAK INPUT DATA LISTING OF INPUT DATA FOR CASE 1 1 = DATA SET 071576-2RLASALLE STUDY 3C7-0476-003 09.8.026-3.0 RELAP4 - MAIN STEAM 2

• PROBLEM DIMENSIONS 3 010001 -2 9 2 3 3 0 0 7 0 1 0 5 0 0 0 0 0

• PROB-DIM 4

• PROBLEM CONSTANTS 5 010002 0.0 1.0 6

• EDIT VARIABLES 7 020000 AP 1 AP 2 AP 3 JW 2 AH 2 JW 1 TD 1 FD 1 TD 2 8

• TIME STEPS 9 030010 1 50 0 0 0.005 0.00005 2.0 10 030020 1 50 0 0 0.01 0.00005 10.0 11

• TRIP CONTROLS 12 040010 1 1 0 0 10.0 0.0 13 040020 2 1 0 0 0.75 0.0 14 040030 3 1 0 0 0.5 0.0 15

• VOLUME DATA CARDS - - 3.7 - P8.9 16 050011 0 0 15.45 -1. 0.556 4077. 15.57 15.57 0 261.5 0. 819.73 17 050021 0 0 15.45 -1. 0.556 177049. 79.74 79.74 0 2315. 0. 740.00 18 050031 0 0 15.45 -1. 0.524 176085. 33.87 33.87 0 5198. 0. 706.14 19

• JUNCTION DATA CARDS 08XXXY - 3.10 - P 91 20 080011 2 1 0 0 0. 6.213 819.73 0.84 1.56 0. 1 0 0 0 0. 0.6 1 0 21 080021 2 3 0 1 0. 295. 740.00 0.24 1.9 0. 1 0 0 0 0 0.6 1.0 22 080031 0 2 1 0 8646.0 1. 770. 0. 0. 0. 0 0 0 0 0 0

• M-STREAM 23 080041 0 2 2 0 0. 1. 770. 0. 0. 0. 0 0 0 0 0 0

• M-STREAM 24 080051 0 2 3 0 0. 1. 770. 0. 0. 0. 0 0 0 0 0 0

• M-STREAM 25 080061 0 2 4 0 0. 1. 770. 0. 0. 0. 0 0 0 0 0 0

• M-STREAM 26 080071 0 2 5 0 0. 1. 770. 0. 0. 0. 0 0 0 0 0 0 *M-STREAM 27

• VALVE DATA CARDS 11XXX0 - - 3.16 P97 28 110010 -2 0. 0. 0. 0. 29

• FILL TABLE DATA CARDS - 13XXYY -- 3.18 P.98 30 130100 4 3 0 0 1.0 547.75 0.0 8646. 1.0 8646. 31 130101 1.01 0.0 10.0 0.0 32 130200 6 3 0 0 1.0 547.43 0.0 0.0 1.0 0.0 33 130201 1.01 920.2 4.39 1319.0 4.4 0.0 10.0 0.0 34 130300 4 3 0 0 1.0 545.55 0.0 0.0 4.39 0.0 35 130301 4.40 1319.0 10.14 2051.00 36 130400 6 3 0 0 0.0 547.67 0.0 0.0 0.99 0.0 37 130401 1.0 28390.0 4.39 27460.0 4.4 0.0 10.0 0.0 38 130500 4 3 0 0 0.0 547.08 0.0 0.0 4.39 0.0 39 130501 4.40 27460.0 10.14 24430.00 40

• LSCS-UFSAR TABLE 6.2-18 TABLE 6.2-18 REV. 14, APRIL 2002 REACTOR BLOWDOWN DATA FOR RECIRCULATION LINE BREAK (AT 3434 MWt)

TIME (sec)

STEAM FLOW (lb/sec)

LIQUID FLOW (lb/sec)

STEAM ENTHALPY (Btu/lb) LIQUID ENTHALPY (Btu/lb) 0 0 22710 1195.3 532.0 0.0016 0 22710 1195.3 532.0 0.0017 0 34060 1195.3 532.0 1.55 0 34060 1195.3 532.0 1.56 0 27550 1195.3 532.0 1.75 0 27550 1195.3 532.0 1.76 0 24840 1192.0 547.0 1.98 0 24810 1192.0 547.0 10.11 0 24320 1193.8 538.8 20.61 0 23460 1196.5 526.0 20.64 3084 11930 1196.5 526.3 25.11 2813 8872 1201.6 493.4 30.01 2382 6175 1204.5 456.6 35.01 1844 3934 1204.3 416.3 40.01 1272 2431 1201.0 374.9 46.87 139 2410 1177.4 261.3 46.94 290 0 1177.0 259.9 47.62 44 0 1173.5 248.4 47.69 0 0 1173.3 247.5

LSCS-UFSAR TABLE 6.2-18A TABLE 6.2-18a REV. 15, APRIL 2004 TABLE 6.2-18A REACTOR BLOWDOWN FOR RECIRCULATION LINE BREAK (AT 3559 MWT)

Time (sec)

Break Flow

Rate (lbm/sec)

Time (sec)

Break Flow

Enthalpy (Btu/lbm) 0 0. 0. 516.8 0.003906 3.698x10 4 5.768 535.8 0.7676 3.558x10 4 8.096 544.6 2.268 3.084x10 4 8.283 544.7 2.768 2.892x10 4 8.689 558.1 3.768 2.653x10 4 10.19 553.4 5.143 2.497x10 4 11.13 550.

8.283 2.549x10 4 11.47 774.2 9.189 2.456 x10 4 11.69 860.4 11.13 2.453 x10 4 11.88 880.6 11.47 1.466 x10 4 11.92 880.

11.6 1.160 x10 4 12.83 818.3 11.84 9.661x10 3 13.08 819.6 12.39 9.116 x10 3 14.08 789.4 12.83 9.808 x10 3 15.27 744.8 13.33 9.702 x10 3 18.27 685.5 16.27 1.071 x10 4 21.58 652.7 18.83 1.027 x10 4 24.45 639.6 24.45 8.853 x10 3 27.2 635.

32 5.568 x10 3 32 635.2

LSCS-UFSAR TABLE 6.2-19 TABLE 6.2-19 REV. 14, APRIL 2002 REACTOR BLOWDOWN DATA FOR MAIN STEAMLINE BREAK (AT 3434 MWt)

TIME (sec)

STEAM FLOW (lb/sec) LIQUID FLOW (lb/sec)

STEAM ENTHALPY (Btu/lb) LIQUID ENTHALPY (Btu/lb) 0.0 11770 0 1190.9 550.9 0.19 11600 0 1191.3 549.1 0.194 8577 0 1191.3 549.1 0.999 8369 0 1192.3 545.3 1.0 899 28450 1192.3 545.3 4.0 1169 27230 1193.4 540.8 10.1 1248 19050 1195.9 529.2 20.38 1730 14680 1200.6 501.3 30.13 1874 9762 1204.2 462.4 40.0 1545 4932 1204.0 409.6 50.0 552 3058 1192.4 322.0 55.32 8.4 253 1173.4 247.9 55.44 0 0 1173.0 246.7

LSCS-UFSAR TABLE 6.2-20 TABLE 6.2-20 REV. 14 - APRIL 2002 CORE DECAY HEAT FOLLOWING LOCA FOR CONTAINMENT ANALYSIS (AT 3334 MWT)

TIME (Seconds)

NORMALIZED CORE HEAT*

0 1.0 0.9 0.9330 2.1 0.7662

5.0 0.5005 6.93 0.3850

9.03 0.2955 15.93 0.1491 30.0 0.0471 10 2 0.0381 10 3 0.0223 10 4 0.0119 10 5 0.00668 10 6 0.00267 3 x 10 6 0.00190

• Normalized Power = 3434 MWt Includes fuel relaxation energy

LSCS-UFSAR TABLE 6.2-20A TABLE 6.2-20a REV. 15 - APRIL 2004 CORE DECAY HEAT FOLLOWING LOCA FOR CONTAINMENT ANALYSIS (AT 3559 MWt)

TIME (Seconds) NORMALIZED CORE HEAT*

0.0 1.0 1.0 0.589 4.0 0.577 10.0 0.377

20.0 0.117 40.0 0.0466 60.0 0.0421 80.0 0.0399 120.0 0.0375 1,000.0 0.0211 10,000.0 0.0108 20,000.0 0.00903 40,000.0 0.00762 80,000.0 0.00634

______________________________

• Normalized Power = 3559 MWt Includes fission energy, decay energy, fuel relaxation energy, and metal-water reaction energy

LSCS-UFSAR TABLE 6.2-21 SHEET 1 OF 49 REV. 15, APRIL 2004 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14,15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENT TYPE C TESTLENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft) M-1 TO M-4 55 Main Steam (includes drain line) Steam 26 26 1 1/2 No No No A (b) A (b) A (b) Detail (a) 1&2B21-F022A,B,C,D 1&2B21-F028A,B,C,D 1&2B21-F067A,B,C,D Inside Outside Outside Yes (Note 30) Yes (Note 30) Yes (Note 30)

N/A 11 N/A M-5 & M-6 55 Reactor Feed (includes connection to RWC) Condensate 24 24 24 4 No No No No AC (b)

AC (b) A (b)

A (b) Detail (b) 1&B21-F010A,B 1&2B21-F032A,B 1&2B21-F065A,B 1&2G33-F040 Inside Outside Outside Outside Yes Yes Yes Yes N/A N/A 43 N/A M-7 55 RHRS/Shutdown Suction Reactor Water 20 20 3/4 No No No A (b)

A (b) A(b) Detail (ah) 1&2E12-F009

1&2E12-F008

1&2E12-F460 Inside Outside Inside Yes Yes Yes N/A 8 N/A M-8 & M-9 55 (Note 28) RHRS/Shutdown Return Reactor Water 12 12 2 No No No AC (a) A (b)

A (a) Detail (d) 1&2E12-F050A,B 1&2E12-F053A,B 1&2E12-F099A,B Inside Outside Inside No (Note 28) Yes No (Note 28) N/A 3 N/A M-10 55 (Note 28) LPCS Injection Suppression Pool Water 12 12 Yes Yes AC (a) A (b) Detail (AJ) 1&2E21-F006

1&2E21-F005 Inside Outside No (Note 28) Yes N/A 3 M-11 55 (Note 28) HPCS Injection Suppression Pool Water 12 12 Yes Yes AC (a) A (b) Detail (AJ) 1&2E22-F005

1&2E22-F004 Inside Outside No (Note 28) Yes N/A 3 M-12 to M-14 55 (Note 28) RHR/LPCI Injection Suppression Pool Water 12 12 Yes Yes AC (a) A (b) Detail (AJ) 1&2E12-F041A,B,C 1&2E12-F042A,B,CInside Outside No (Note 28) Yes N/A 7 M-15 55 Steam to RCIC System (Includes Rhr Supply) Steam 10 1 10 4 Yes Yes No Yes A (b)

A (b)

A (b) A (b) Detail (e) 1&2E51-F063 1&2E51-F076 1(2)E51-D324 1&2E51-F008 Inside Inside Outside Outside Yes Yes Yes Yes N/A N/A 13 max. N/A M-16 56 Cooling Water Supply Demineralized Water 6 6 3/4 No No No A (b)

A (b) A(b) Detail (f) 1&2WR029 1&2WR179

1&2WR225 Outside Inside Inside Yes Yes Yes 4 N/A N/A M-17 56 Cooling Water Return Demineralized Water 6 6 3/4 No No No A (b)

A (b) A(b) Detail (f) 1&2WR040 1&2WR180 1&2WR226 Outside Inside Inside Yes Yes Yes 5 N/A N/A M-18 & M-19 56 RHRS/Containment Spray Suppression Pool Water 16 No No A (b)

A (b) Detail (g) 1&2E12-F017A,B 1&2E12-F016A,B Outside Outside Yes Yes N/A 11 M-20 56 Drywell Purge Air 26 26 1 1/2 1 1/2 8 No No No No No A (b) A (b)

A (b)

A (b) A (b) Detail (s)

Detail (s) Detail (s) Detail (s) 1&2VQ030 1&2VQ029

1&2VQ047

1&2VQ048 1&2VQ042 Outside Outside Outside Outside Outside Yes Yes Yes Yes Yes N/A 10 N/A 10 max. 10 max. Summary of Lines Penetrating the Primary Containment LSCS-UFSAR TABLE 6.2-21 SHEET 2 OF 49 REV. 16, APRIL 2006 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CODE CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS M-1 to M-4 AO Globe AO Globe MO Gate 1 1

1 Auto Auto Auto RM RM RM O O

O C C

C C C

C C C As is C,D,E,H,P,RM C,D,E,H,P,RM C,D,E,H,P,RM 3 to 5 3 to 5 Standard ESS 2 ESS 1 ESS 1 Note (1,20)

Note (1) Note (48)

M-5 to M-6 Swing Check U1/Swing Check U2 AO No Slam-Check MO Gate MO Gate 1 1 2 2 Process Process RM RM NA RM M M O O O O C C C O C C C C NA NA As is As is Rev. Flow B,F,Rev. FlowRM(Note 34)

RM(Note 34) Instantaneou s Instantaneou s Standard Standard NA ESS 2 ESS 1 ESS 1 Note (17)

Note (20, 53, 54) M-7 MO Gate MO Gate Relief 1 1

2 Auto Auto Process RM RM N/A C C

C O O C C C

C As is As is C A,D,U,RM A,D,U,RM N/A 40 sec 40 sec Instantaneou s ESS 2 ESS 1 N/A Note (51) 1E12-F008M-8 & M-9 No Slam-Check MO Globe MO Globe 1 1 1 Process Auto Auto NA RM RM C C C O O O C C C NA As is As is Rev. Flow A,D,U,RM A,D,F,U,RM Instantaneou s 29 sec Standard ESSA 2 ESS 1 ESS 1 Note (3)

M-10 No Slam-Check MO Gate 1 1 Process Auto NA RM C C C C O O NA As is Rev. Flow RM (Notes 31, 36) Instantaneou s Standard ESS 1 ESS 1 Note (3)

Note (51)

M-11 No Slam-Gate MO Gate 1 1 Process Auto NA RM C C C C O O NA As is Rev. Flow RM (Notes 31, 36) Instantaneou s Standard ESS 3 ESS 3 Note (3)

Note (51)

M-12 to M-14 No Slam-Gate MO Gate 1 1 Process Auto NA RM C C C C O O NA As is Rev. Flow RM (Notes 31, 36) Instantaneou s Standard Note (22)

Note (22) Note (3)

Note (51)

M-15 MO Gate MO Globe NA MO Gate 1 1 1 1 Auto Auto NA Auto RM RM NA RM O C C O O C C O O O C C As is As is NA As is D,RM D,RM NA D,RM 15 sec Standard NA Standard ESS 2 ESS 2 NA ESS 1 Note (20)

Note (20)

Note (60)

M-16 MO Gate MO Gate Relief 2 2 2 Auto Auto Process RM RM N/A O O C O O C C C C As is As is C B,F,RM B,F,RM N/A Standard Standard N/A ESS 1 ESS 2 N/A M-17 MO Gate MO Gate Relief 2 2 2 Auto Auto Process RM RM N/A O O C O O C C C C As is As is C B,F,RM B,F,RM N/A Standard Standard N/A ESS 1 ESS 2 N/A M-18 & M-19 MO Gate MO Gate 2 2 Auto Auto RM RM C C C C C C As is As is G,RM G,RM Standard Standard Note (22)

Note (22) Note (2,20,52,54) Note (2, 51, 52)

M-20 AO Butterfly AO Butterfly MO Globe MO Globe AO Butterfly 2

2 2

2 2 Auto Auto Auto Auto Auto RM RM RM RM RM C C O O C C C

C C O C C

C C C C C As is As is C B,F,Y,Z,RM B,F,Y,Z,RM B,F,Y,Z,RM B,F,Y,Z,RMB,F,Y,Z,RM 10 sec 10 sec 23 sec 23 sec 10 sec ESS 2 ESS 1 ESS 2 ESS 1 ESS 1 Note (8,20,41,46,50,54) Note (8,46)

Note (20,54))

Note (46)

LSCS-UFSAR TABLE 6.2-21 SHEET 3 OF 49 REV. 15, APRIL 2004 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14,15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft)

M-21 56 56 (Note 32) Vent from Drywell

Drywell Pressure Air

Air 26 2 26 2 3/4 No No No No No A(b) A(b) A(b)

A(b) C Detail (h)

Detail (w) 1&2VQ034 1&2VQ035 1&2VQ036 1&2VQ068 1&2CM102 Outside Outside Outside Outside Outside Yes Yes Yes Yes No N/A N/A 23 max N/A 10 max. M-21 55 (Note 33) RPV Level and Pressure Reactor Water 3/4 Yes C Detail (AB) 1B21-F571 Outside No 10 max.

M-22 55 Main Stream Drains Stream-WaterMixture 3 3 No No A(b)

A(b) Detail (c) 1&2B21-F016

1&2B21-F019 Inside Outside Yes Yes N/A 6 M-23 Spare (Unit 1) M-23 56 Combustible Gas Control Drywell Suction AIR/Vapor Mixture 4 4 Yes Yes A(b)

A(b) Detail (g) 2HG001B 2HG002B Outside Outside Yes Yes N/A 10 M-24 Spare M-25 & M-26 56 Chilled Water Supply Demineralized Water 8 8 3/4 No No No A(b) A(b) A(b) Detail (AF) 1&2VP063A,B 1&2VP113A,B 1&2VP198A,B Outside Inside Inside Yes Yes Yes 6 N/A N/A M-27 & M-28 55 Chilled Water Return Demineralized Water 8 8 3/4 No No No A(b)

A(b)

A(b) Detail (AF) 1&2VP053A,B 1&2VP114A,B 1&2VP197A,B Outside Inside Inside Yes Yes Yes 6 N/A N/A LSCS-UFSAR TABLE 6.2-21 SHEET 4 OF 49 REV. 13 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CODE CLASS PRIMARY METHOD OF ACTUATIONSECONDAR Y METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS M-21 AO Butterfly MO Globe AO Butterfly MO Globe Excess Flow Check 2 2 2 2 2 Auto Auto Auto Auto Process RM RM RM RM N/A C C C C O C C C C O C C C C O C As is C As is N/A F,B,Y,Z,RM F,B,Y,Z,RM F,B,Y,Z,RM F,B,Y,Z,RM F,B,Y,Z,RM 10 Sec 5 Sec 10 Sec 5 Sec Instantaneou s ESS 2 ESS 2 ESS 1 ESS 1 NA Note (8,20,41,46,54) Note (8,20)

Note (8,46)

Note (8)

M-21 EFCV 2 Process NA O O O NA Flow Instantaneou s NA Note (23,33)

M-22 MO Gate MO Gate 1 1 Auto Auto RM RM O O C C C C As is As is C,D,E,H,P,RM C,D,E,H,P,RM Standard Standard ESS 2 ESS 1 Note (20),(51)

Note (51) M-23 M-23 MO Gate MO Globe 2 2 RM RM M M C C C C O O As is As is RM(Note 37)

RM(Note 37) Standard Standard Note (23)

Note (23)

Note (20,54)

M-24 M-25 TO M-26 MO Gate MO Butterfly Relief 2 2 2 Auto Auto Process RM RM N/A O O C O O C C C C As is As is N/A B,F,RM B,F,RM Process Standard Standard N/A ESS 1 ESS 2 N/A Note (20)

Note (20) M-27 & M-28 MO Gate MO Butterfly Relief 2 2 2 Auto Auto Process RM RM N/A O O C O O C C C C As is As is N/A B,F,RM B,F,RM Process Standard Standard N/A ESS 1 ESS 2 N/A Note (20)

Note (20)

LSCS-UFSAR TABLE 6.2-21 SHEET 5 OF 49 REV. 17, APRIL 2008 CONTAINMEN T PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14,15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft)

M-29 55 (Note 28) RCIC RPV Head Spray (Includes RHR Head Spray) Condensate 6 6 6 6 Yes Yes Yes Yes AC(a) AC(a) A(b)

A(b) Detail (i) 1 &2E51-F066 1 &2E51-F065 1 &2E51-F013 1 &2E12-F023 Inside Outside Outside Outside No (Note 28)

No (Note 28)

Yes Yes N/A N/A 20 Max (Unit 1)10 Max (Unit 2)N/A M-30 55 Reactor Cleanup Reactor Water 6 6 No No A(b)

A(b) Detail (t) 1 &2G33-F001 1 &2G33-F004 Inside Outside Yes Yes N/A 5 M-31& M-32 NA (Note 45) Containment High Rad Detector M-33 56 Combustible Gas Control Drywell Suction Air/Vapor Mixture 4 4 Yes Yes A(b) A(b) Detail (g) 1HG001B 1HG002B Outside Outside Yes Yes N/A 10 M-33 Spare (Unit 2)

M-34 55 Standby Liquid Control Sodium Pentaborate Solution 1 1/2 1 1/2 1 1/2 No No No AC(b) C AD(b) Detail (u)

1&2C41-F007

1&2C41-F006 1&2C41-F004A,B Inside Outside Outside No (Note 62)No No (Note 62) N/A N/A 100 M-35 Spare M-36 55 Recirc. Loop Sampling Reactor Water 3/4 3/4 3/4 No No No A(b)

A(b)

A(b) Detail (ae)

1&2B33-F019

1&2B33-F020

1&2B33-F395 Inside Outside Inside Yes Yes Yes N/A 10 Max N/A M-37 56 Clean Condensate Condensate 3

3 No No A(b)

A(b) Detail (ai) 1&2MC033 1&2MC027 Outside Outside No (Note 43)

No (Note 43) N/A 4 M-38 56 Service Air Air 3 3 No No A(b)

A(b) Detail (v) 1&2SA046 1&2SA042 Outside Outside No (Note 43)

No (Note 43) N/A 4 M-39 Spare M-40A,B,C,D 55 (Note 24)

CRD Insert Condensate 1 No A Note (24) 1&2C11-D001-120 1&2C11-D001-123 Outside Outside No No 45 Max M-41A,B,C,D 55 (Note 24)

CRD Withdrawal Condensate 3/4 No A Note (24) 1&2C11-D001-121 1&2C11-D001-122 Outside Outside No No 45 Max M-42 to M-46 54 TIP Drive NA 3/8 No NA Note (18) 1&2C51-J004 Outside Yes Note (18) 2 M-47 54 Air Supply Air 3/4 No A(b) 1&2IN031 Outside Yes M-48 Spare

LSCS-UFSAR TABLE 6.2-21 SHEET 6 OF 49 REV. 14, ARPIL 2002 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CODE CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS M-29 No Slam-Check No Slam-Check MO Gate MO Globe 1 1 1 1 Process Process Auto Auto NA NA RM RM C C C C C C C C C C C C NA NA As is As is Rev. Flow Rev. Flow RM (Note 31) A,D,U,RM(Note

31) Instantaneou s Instantaneou s 15 Sec Standard ESS 1 ESS 1 ESS 1 ESS 1 Note (3)

Note (3) Note (51)

M-30 MO Gate MO Gate 1 1 Auto Auto RM RM O O O O C C As is As is B,J,RM B,J,RM < 10 sec < 10 sec ESS 2 ESS 1 Note (61)

M-31 & M-32 M-33 MO Gate MO Globe 2 2 RM RM M M C C C C O O As is As is RM(Note 37)

RM(Note 37) Standard Standard Note (23)

Note (23)

Note (20,54)

M-33

M-34 No Slam-Check No Slam-Check Explosive 1

1 1 Process Process RM NA NA NA C C C C C C C C C NA NA Rev. Flow Rev. Flow NA --

-- NA NA NA M-35 M-36 AO Globe Check AO Globe 2 2 2 Auto Process Auto RM N/A RM O C O O C O C C

C Closed N/A Closed B,C,RM Reverse Flow B,C,RM Standard Instantaneou s Standard ESS 2 N/A ESS 1 Note (9,42)

Note (9,42)

M-37 Gate Gate 2 2 M M NA NA C C C C C C NA NA NA NA NA NA NA NA Note (43)

Note (43)

M-38 Gate Gate 2 2 M M NA NA C C C C C C NA NA NA NA NA MA NA NA Note (43)

Note (43) M-39 M-40 A, B, C, D SO Gate SO Gate Note (27)

Note (27) Auto Auto RM RM C C C C C C As is As is A,RM A,RM Instantaneou s Instantaneou s Typical of 185 Typical of 185 M-41 A, B, C, D SO Gate SO Gate Note (27)

Note (27) Auto Auto RM RM C C C C C C As is As is A,RM A,RM Instantaneou s Instantaneou s Typical of 185 Typical of 185 M-42 to M-46 Solenoid Ball 2 Auto RM C C C C A,F,RM (note 31) NA NA M-47 SO Globe 2 Auto RM O O C C B,F,RM 5 sec ESS 2 M-48 Spare

LSCS-UFSAR TABLE 6.2-21 SHEET 7 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14,15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft) M-49 & M-50 56 Recric. Flow Control Valve Hydraulic Piping Hydraulic Fluid (Fyrquel) 3/4 3/4 1/2 1/2 1/2 1/2 3/4 3/4 No No No No No No No No Note (19)

Note (19)

Note (19)

Note (19)

Note (19)

Note (19)

Note (19)

Note (19) Detail (c)

Detail (c)

Detail (c)

Detail (c) Detail (c) Detail (c)

Detail (c)

Detail (c) 1&2B33-F338A,B 1&2B33-F339A,B 1&2B33-F340A,B 1&2B33-F341A,B1&2B33-F342A,B1&2B33-F343A,B 1&2B33-F344A,B 1&2B33-F345A,B Inside Outside Inside Outside Inside Outside Inside Outside No (Note 35)

No (Note 35)

No (Note 35)

No (Note 35)No (Note 35)No (Note 35)

No (Note 35)

No (Note 35) N/A N/A N/A N/A M-51 Spare M-52 55 (Note 33) RPV Level Reactor Water 3/4 Yes C Detail (AB) 2B21-F570 Outside No (Note 33) 10 Max M-53 56 Combustible Gas Control Drywell Suction Air/Vapor Mixture 4 4 Yes Yes A(b) A(b) Detail (g) 1&21HG001A 1&21HG002A Outside Outside Yes Yes N/A 10 M-54 (Unit 1) Spare M-54 (Unit 2) 56 56 56 Air Dryer Blowdown Drywell Pneumatic Comp Discharge Drywell Pneumatic Comp Suction Air Air Air 3 3 2 2 2 1/2 2 1/2 No No No No No No A(b) A(b) AC(b) A(b) A(b) A(b) Detail (g)

Detail (AL)

Detail (g) 2IN074 2IN075 2IN018 2IN017 2IN001A 2IN001B Outside Outside Outside Outside Outside Outside Yes Yes Yes Yes Yes Yes N/A 5 N/A 5

LSCS-UFSAR TABLE 6.2-21 SHEET 8 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CODE CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION(6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS M-49 & M-50 SO Globe SO Globe SO Globe SO Globe SO Globe SO Globe SO Globe SO Globe 2 2 2 2

2 2 2 2 Auto Auto Auto Auto Auto Auto Auto Auto RME RME RME RME RME RME RME RME O O O O

O O O O O O O O

O O O O C C C C

C C C C C C C C

C C C C B,F,RME B,F,RME B,F,RME B,F,RME B,F,RME B,F,RME B,F,RME B,F,RME Instantan.Instantan.Instantan.

Instantan.

Instantan.Instantan.Instantan.

Instantan. ESS 2 ESS 1 ESS 2 ESS 1 ESS 2 ESS 1 ESS 2 ESS 1 Note (35)

Note (35)

Note (35)

Note (35)

Note (35)

Note (35)

Note (35)

Note (35) M-51 M-52 EFCV 2 Process NA O O O NA Flow Instantan. NA M-53 MO Gate MO Globe 2 2 RM RM M M C C C C O O As is As is RM (Note 37)

RM (Note 37) Standard Standard Note (23)

Note (23)

Note (20,54)

M-54 (Unit 1)

M-54 (Unit 2) AO Globe AO Globe No Slam-Check AO Globe AO Globe AO Globe 2 2 2 2 2 2 Auto Auto Process Auto Auto Auto M M NA M RM RM O O O O O O O O O O O O C C C C C C C C NA C C C F,H,RM F,H,RM NA F,H,RM F,H,RM F,H,RM Standard Standard Instantan.Standard Standard Standard ESS 2 ESS 1 ESS 2 ESS 2 ESS 1

Note (28)

Note (20)

LSCS-UFSAR TABLE 6.2-21 SHEET 9 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14,15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft)

M-55 57 ADS Pneumatic Supply Nitrogen or Air 1 Yes B Detail (j) 1 & 2IN100 Outside No (Note 38) 5 M-56 55 (Note 33) Reactor Water Level Reactor Water 3/4 Yes C Detail (w) 1 &2B21-F372 Outside No (Note 33) 10 Max M-57 Spare M-58 Deleted M-59 56 (Note 58) Clean Condensate to Refueling Bellows Condensate 2 2 No No A(b)

A(b) Detail (v) 1&2FC113 1&2FC114 Outside Outside Yes Yes N/A 5 M-59 55 (Note 33) RPV Level and Pressure Reactor Water 3/4 Yes C Detail (AB) 1B21-F570 Outside No 10 Max M-60 (Unit 1) 56 Drywell Pneumatic Compressor Discharge Air 2 2 3 3 No No No No AC(b) A(b) A(b)

A(b) Detail (AL)

Detail (g) 1IN018 1IN017 1IN074 1IN075 Outside Outside Outside Outside Yes Yes Yes Yes N/A 5 M-60 (Unit 2) 57 ADS Pneumatic Supply Nitrogen or Air 1 Yes B Detail (j) 2IN101 Outside No (Note 38) 5 M-61 (Unit 1) 57 ADS Pneumatic Supply Nitrogen or Air 1 Yes B Detail (j) 1IN101 Outside No (Note 38) 5 M-61 (Unit 2) Spare M-62 (Unit 1) 56 Drywell Pneumatic Comp Discharge Air 2 1/2 2 1/2 No No A(b)

A(b) Detail (g) 1IN001A 1IN001B Outside Outside Yes Yes N/A 5 M-62 (Unit 2) Spare M-63 & M-64 55 Recirc. Pump Seal Injection Supply Condensate 3/4 3/4 No No A(a) A(a) Detail (h)

Note (25) 1&2B33-F013A,B 1&2B33-F017A,B Inside Outside Yes (Note 25)

Yes (Note 25)

N/A M-65 56 (Note 58) Reactor Well Bulkhead Drain Water 10 10 No No A(b)

A(b) Detail (V) (Unit 1 only) Detail (AD) ( Unit 2 only) 1&2FC115 1&2FC086 Outside Outside Yes Yes N/A 5 M-65 (Unit 2) 55 (Note 33) RPV Level Reactor Water 3/4 Yes C Detail (AB) 2B21-F571 Outside No (Note 33) 10 max LSCS-UFSAR TABLE 6.2-21 SHEET 10 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CLASS PRIMARY METHOD OF ACTUATIO N SECONDAR Y METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITIONPOWER FAILURE VALVE POSITION (6)ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS M-55 SO Globe 2 RM M O O O O NA InstantaneousESS 2 M-56 Excess Flow check 2 Process NA O O O NA Flow InstantaneousNA M-5 7 M-5 8 M-59 Globe Globe 2 2 M M NA NA L.C.

L.C. C C C C NA NA NA NA NA NA NA NA Note (20,54)

Note (20) M-59 EFCV 2 Process NA O O O NA Flow InstantaneousNA Note(23,33)

M-60 (Unit 1) Check AO Globe AO Globe AO Globe 2 2

2 2 Process Auto Auto Auto NA M M

M O O

O O O O

O O C C

C C NA C C

C NA B,F,RM B,F,RM B,F,RM InstantaneousStandard Standard Standard ESS 2 ESS 2 ESS 1 Note (28)

Note (28)

M-60 (Unit 2) SO Globe 2 RM M O O FO FO NA InstantaneousESS 2 M-61 (Unit 2) SO Globe 2 RM M O O FO FO NA InstantaneousESS 2 M-61 (Unit 2)

M-62 (Unit 1) AO Globe AO Globe 2 2 Auto Auto RM RM O O O O C C C C B,F,RM B,F,RM Standard Standard ESS 2 ESS 1 Note (20)

M-62 (Unit 2) M-63 & M-64 No Slam-Check No Slam-Check 2 2 Process Process NA NA O O O O C C NA NA Reverse Flow Reverse FlowInstantaneous Instantaneous NA NA M-65 Gate Gate 2 2 M M NA NA C C C C C C NA NA NA NA NA NA NA NA Note (20 ,54) (Note 20 Unit 1 only) M-65 (Unit 2) EFCV 2 Process NA O O O NA Flow InstantaneousNA

LSCS-UFSAR TABLE 6.2-21 SHEET 11 OF 49 REV. 17, APRIL 2008 CONTAINMEN T PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14, 15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft)

M-66 56 Suppression Chamber Purge Line Air 26 26 1 1/2 1 1/2 8 No No No No No A (b) A (b)

A (b)

A (b) A (b) Detail (s)

Detail (s) Detail (s) Detail (s) 1&2VQ027 1&2VQ026 1&2VQ050 1&2VQ051 1&2VQ043 Outside Outside Outside Outside Outside Yes Yes Yes Yes Yes N/A 8 7 Max. M-67 56 Suppression Chamber Vent Line Air 26 26 2 No No No A (b) A (b) A (b) Detail (h) 1&2VQ031 1&2VQ040 1&2VQ032 Outside Outside Outside Yes Yes Yes N/A 17 N/A M-68 56 (Note 28) LPCS Suction from Suppression Pool Suppression Pool Water 24 Yes B Detail (m) 1&2E21-F001 Outside No (Note 39) 2 M-69 56 (Note 28) HPCS Suction from Suppression Pool Suppression Pool Water 24 Yes B Detail (m) 1&2E22-F015 Outside No (Note 39) 5 M-70 56 (Note 28) 56 (Note 32) RHR (LPCI) Suction From Supp. Pool Supp. Pool Water Level Suppression Pool Water Supp. Pool /water 24 3/4 Yes No B C Detail (m)

Detail (w) 1&2E12-F004A

1&2CM002 Outside Outside No (Note 39)

No (Note 32) 2 10 Max. M-71 56 (Note 28) 56 (Note 32) RHR (LPCI) Suction From Supp. Pool Supp. Pool Water Level Suppression Pool Water Supp. Pool Water 24 3/4 Yes No B C Detail (m)

Detail (w) 1&2E12-F004C 1&2CM010 Outside Outside No (Note 39) No (Note 32) 2 10 Max.

LSCS-UFSAR TABLE 6.2-21 SHEET 12 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWERFAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS M-66 AO Butterfly AO Butterfly MO Globe MO Globe AO Butterfly 2

2 2

2 2 Auto Auto Auto Auto Auto RM RM RM RM RM C C O O C C C

C C O C C

C C C C C As is As is C F,B,Y,Z,RM F,B,Y,Z,RM F,B,Y,Z,RM F,B,Y,Z,RMF,B,Y,Z,RM 10 sec. 10 sec.

23 sec.

23 sec. 10 sec. ESS 2 ESS 1 ESS 1 ESS 1 ESS 1 Note(8,20,46,54)Note(8,46) Note(20, 54)

Note (46)

M-67 AO Butterfly AO Butterfly MO Globe 2 2 2 Auto Auto Auto RM RM RM C C C C C C C C C C C As is F,B,Y,Z,RM F,B,Y,Z,RMF,B,Y,Z,RM 10 sec. 10 sec. Standard ESS 2 ESS 1 ESS 2 Note (8,20,41,46, 54) Note (8, 46) Note (8,20)

M-68 MO Gate 2 RM M O O O As is RM (Note 36) Standard ESS 1 Note (20)

M-69 MO Gate 2 Auto RM O O O As is RM (Note 36) Standard ESS 3 Note (20)

M-70 MO Gate EFCV 2 2 RM Process M NA O O O O O O As is NA RM (Note 36)

Flow Standard Instantan.

Note (22)NA Note (20)

M-71 MO Gate EFCV. 2 2 RM Process M NA O O O O O O As is NA RM (Note 36)

Flow Standard Instantan.

Note (22)Na Note (20)

LSCS-UFSAR TABLE 6.2-21 SHEET 13 OF 49 REV. 17, APRIL 2008 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14,15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft)

M-72 56 (Note 28) RHR (LPCI) Suction From Supp. Pool Suppression Pool Water 24 Yes B Detail (m) 1&2E12-F004B Outside No (Note 39) 2 M-73 & M-74 56 RHR to Suppression Pool Spray Header Suppression Pool Water 4 No B Detail (z) 1&2E12-F027A,B Outside No (Note 29) 23 M-75 56 (Note 28) RCIC Pump Suction From Suppression Pool Suppression Pool Water 8 Yes B Detail (m) 1&2E51-F031 Outside No (Note 39) 2 M-76 56 (Note 28) RCIC Turbine Exhaust Steam 10 Yes Yes A (b)

A (b) Detail (o) 1&2E51-F068

1&2E51-F040 Outside Outside Yes Yes 3 N/A 56 (Note 28) LPCS Test Line Suppression Pool 14 Yes B 1&2E21-F012 Outside No (Note 29) 225 max. LPCS Min. Flow Line Water 4 Yes B 1&2E21-F011 Outside No (Note 29)

M-77 RHR Suction RV 2 Yes B 1&2E12-F088A Outside No (Note 29) 56 (Note 28) RCIC Full Flow Test Return to Supp. Pool

Suppression Pool Water 4

Yes B

Detail (AA)

1(2)E51-F362

1(2)E51-F363

1(2)E51-F022 1(2)E51-F059 Outside Outside Outside Outside Yes (Note 49)Yes (Note 49)

Yes (Note 49)Yes (Note 49) 215 max. 230 max. M-78 Spare M-79 & M-84 56 (Note 28) RHR Min. Flow Line RHR Test Line Supp. Pool Water 18 18 14 8 4 2 Yes Yes Yes Yes Yes Yes B B B B B

C Detail (q),(AG)

1&2E12-F024A,B 1&2E12-F021 1&2E12-F302 1&2E12-F064A,B,C1&2E12-F011A,B 1&2E12-F088B Outside Outside Outside Outside Outside Outside No (Note 29)No (Note 29)No (Note 29)No (Note 29)No (Note 29)No (Note 29) 300 Max. M-80 56 (Note 28) RCIC Pump Min. Flow Line Condensate 2 Yes B Detail (r) 1&2E51-F019 Outside No (Note 29) 40 M-81 56 (Note 28) RCIC Vacuum Pump Discharge Condensate 1 1/4 1 1/4 No No A (b)

A (b) Detail (r) 1&2E51-F069

1&2E51-F028 Outside Outside Yes Yes 3 N/A M-82 56 (Note 28) HPCS Test Line HPCS Min Flow Line Condensate 14 4 Yes Yes B B Detail (l) 1&2E22-F023

1&2E22-F012 Outside Outside No (Note 29)No (Note 29) 29 Max. M-83 & M-93 56 (Note 28) LPCS Safety/Relief Valve Discharge Suppression Pool 4 2 Yes Yes C C Detail (AK) 1&2E21-F018

1&2E21-F031 Outside Outside No (Note 29)No (Note 29) 125 Max. M-85 M-86 M-87 M-90 M-91 M-99 56 (Note 28) RHR Safety/Relief Valve Discharge Suppression Pool Water 2

2 2 2 2 Yes Yes Yes Yes Yes C C C C C Detail (AK) 1&2E12-F025A

1&2E12-F025B 1&2E12-F025C 1&2E12-F088C 1&2E12-F030

1&2E12-F005 Outside Outside Outside Outside Outside Outside No (Note 29)No (Note 29)No (Note 29)No (Note 29)No (Note 29) No (Note 29) 69 Max.

LSCS-UFSAR TABLE 6.2-21 SHEET 14 OF 49 REV. 16, APRIL 2006 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS M-72 MO Gate 2 RM M O O O As is RM (Note 36) Standard Note (22) Note (20) M-73 & M-74 MO Gate 2 Auto RM C C C As is G, RM 30 sec Note (22) Note (2, 20,56) M-75 MO Gate 2 Auto RM C C C As is RM (Note 36) Note (59) ESS 1 (DC)Note (20,57)

M-76 MO Gate Check 2 2 Auto Process RM NA O C O C O C As is As is RM (Note 36)

Reverse Flow Note (59) Instantan. ESS 1 Note (20,54))

M-77 MO Globe MO Gate Relief Gate Gate MO Globe MO Globe 2 2

2 2 2 2 2 RM RM Process Manual Manual Process Process M M NA NA NA RM RM C O C C C C C C O C C C C C C C

C C C C C As is As is NA NA NA As is As is Rm(Notes 31,36)

Rm(Notes 31,36)

RM(Notes 31,36)RM(Notes 31,36)

Note (47) Standard --- --- --- Note (59)

Note (59) ESS 1 ESS 1 --

-- -- ESS 1 ESS 1 Note (20)

Note (20)

Note (20)

-- Note (20,54)

-- M-78 M-79 & M-84 MO Globe MO Globe Gate MO Gate MO Gate Relief 2 2 2

2 2 2 Auto Auto M RM RM Process RM RM NA M M NA C C C O C C C C

C C C C C C

C C C C As is As is NA As is As is NA G,RM G,RM RM(Notes 31,36)

GRM(Notes31,3

6) Standard Standard -- Standard Note (50) 22 sec

-- Note (22)ESS 2 -- Note (22)ESS 1 -- Note (2 20)

Note (20)

Note (20)

Note (20)

Note (20)

Note (20) M-80 MO Globe 2 RM M C C C As is RM(Notes 31,36)7 sec ESS 1 (DC)Note (20)

M-81 MO Globe No Slam Check 2 2 RM Process M NA O C O C O C As is NA RM(Notes 31,36)Reverse Flow Note (59) Instantan. ESS 1 NA Note (20,54)

M-82 MO Globe MO Gate 2 2 Auto Auto M M C C C C C C As is As is G,RM G,RM Standard Standard ESS 3 ESS 3 Note (20)

Note (20,56) M-83 & M-93 Relief Relief 2 2 Process Process NA NA C C C C C C NA NA Process Process NA NA NA NA Note (20)

Note (20)

M-85 M-86 M-87 M-90 M-91 M-99 Relief Relief Relief Relief Relief Relief 2 2

2 2 2 2 Process Process Process Process Process Process NA NA NA NA NA NA C C

C C C C C C

C C C C C C

C C C C NA NA NA NA NA NA Process Process Process Process Process Process NA NA NA NA NA NA NA NA NA NA NA NA Note (20)

Note (20)

Note (20)

Note (20)

Note (20)

Note (20)

LSCS-UFSAR TABLE 6.2-21 SHEET 15 OF 49 REV. 17, APRIL 2008 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft) M-88 & M-89 56 (Note 28) RHR Safety/Relief Valve Discharge and H x Vent Line Steam 3/4 3/4 6 2 Yes Yes Yes Yes B B

C C Detail (p) 1&2E12-F073A,B 1&2E12-F074A,B 1&2E12-F055A,B 1&2E12-F311A,B Outside Outside Outside Outside No (Note 29)No (Note 29)No (Note 29)No (Note 29) N/A 56 Max. M-92 56 (Note 28) RCIC Safety/Relief Valve Discharge Condensate 4 No C Detail (AK) 1&2E12-F036B Outside No (Note 29) 5 M-94 56 (Note 28) HPCS Safety/Relief Valve Discharge Condensate 2 Yes C Detail (AK) 1&2E22-F014 Outside No (Note 29) 27 M-95 Spare M-96 56 Drywell Equip.

Drains Water 4 4 No No A (b)

A (b) Detail (g) 1&2RE025 1&2RE024 Outside Outside Yes Yes 10 N/A M-97 56 Drywell Equip. Drain Cooling Water 2 2 No No A (b)

A (b) Detail (g) 1&2RE029 1&2RE026 Outside Outside Yes Yes 10 N/A M-98 56 Drywell Floor Drains Water 4 4 No No A (b)

A (b) Detail (g) 1&2RF012 1&2RF013 Outside Outside Yes Yes N/A 10 M-100 56 (Note 28) SUPR CHBR N 2/O 2 1/2 No A (b) Detail (g) 1CM019A 1CM020A Outside Outside Yes Yes 60 60 LSCS-UFSAR TABLE 6.2-21 SHEET 16 OF 49 REV. 13 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS M-88 & M-89 MO Globe MO Glove Relief Relief 2 2 2 2 RM RM Process Process M M NA NA C C C C C C C C C C C C As is As is NA NA RM (Note 36)RM (Note 36)Process Process Standard Standard NA NA ESS 1 ESS 1 NA NA Note (20)

Note (20)

Note (20)

Note (20) M-92 Relief 2 Process NA C C C NA Process NA NA Note (20) M-94 Relief 2 Process NA C C C NA Process NA NA Note (20) M-95 M-96 AO Globe AO Globe 2 2 Auto Auto RM RM C C C C C C C C B,F,RM B,F,RM Standard Standard ESS 1 ESS 2 Note (20)

M-97 AO Globe AO Globe 2 2 Auto Auto RM RM C C C C C C C C B,F,RM B,F,RM Standard Standard ESS 1 ESS 2 Note (20,42,54)

M-98 AO Glove AO Glove 2 2 Auto Auto RM RM C C C C C C C C B,F,RM B,F,RM Standard Standard ESS 2 ESS 1 Note (20,42)

M-100 SOL Globe SOL Globe 2

2 Auto Auto RM RM O O O O C C C C B, F, RM B, F, RM 5 sec 5 sec ESS 1 ESS 2 Note 20

LSCS-UFSAR TABLE 6.2-21 SHEET 17 OF 49 REV. 17, APRIL 2008 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in.) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft)

M-101 56 56 (Note 28) RCIC Turbine Exhaust Breaker Line RCIC Safety/Relief Valve Discharge Air Condensate 2

2 4 Yes Yes No A (b)

A (b) C Detail (o)

Detail (AK) 1&2E51-F080

1&2E51-F086 1&2E12-F036A Outside Outside Outside Yes Yes No (Note 29) 17 NA 5 M-102 Spare M-103 NA Vacuum Breaker Air 24 Yes Exempt Detail (y) 1&2PC003C Outside No 4 M-104 56 (Note 32) 56 NA Supp. Pool Water Level Combustible Gas Control Return Vacuum Breaker Supp. Pool Water Air Vapor Mixture Air 3/4 6 6 24 No Yes Yes Yes C A (b)

A (b) Exempt Detail (w) Detail (g)

Detail (y) 1&2CM012 1&2HG005A

1&2HG006A 1&2PC003A Outside Outside Outside Outside No (Note 32)Yes Yes No 10 Max. NA 4 M-105 56 (Note 32) NA Supp. Pool Water Level Vacuum Breaker Supp. Pool Water Air 3/4 24 No Yes C Exempt Detail (w)

Detail (y) 1&2CM004 1&2PC003D Outside Outside No (Note 32)

No 10 Max. 4 M-106 NA 56 Vacuum Breaker Combustible Gas Control Return Air Air Vapor Mixture 24 6 6 Yes Yes Yes Exempt A (b)

A (b) Detail (y)

Detail (g) 1&2PC003B 1&2HG005B

1&2HG006B Outside Outside Outside No Yes Yes 4 N/A M-107 NA NA Vacuum Breaker Vacuum Breaker Air Air 24 24 Yes Yes Exempt C Detail (y)

Detail (y) 1&2PC002C

1&2PC001C Outside Outside No No 2 M-108 NA NA Vacuum Breaker Vacuum Breaker Air Air 24 24 Yes Yes Exempt C Detail (y)

Detail (y) 1&2PC002A

1&2PC001A Outside Outside No No 2 M-109 NA NA Vacuum Breaker Vacuum Breaker Air Air 24 24 Yes Yes Exempt C Detail (y)

Detail (y) 1&2PC002D

1&2PC001D Outside Outside No No 2 M-110 NA NA Vacuum Breaker Vacuum Breaker Air Air 24 24 Yes Yes Exempt C Detail (y)

Detail (y) 1&2PC002B

1&2PC001B Outside Outside No No 2 LSCS-UFSAR TABLE 6.2-21 SHEET 18 OF 49 REV. 15, APRIL 2004 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CODE CLASS PRIMARY METHOD OF ACTUATION SECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS M-101 MO Globe MO Globe Relief 2 2

2 RM RM Process M M NA O O C O O C C C

C As is As is NA F,RM (Note 36)

F,RM (Note 36)Process Note (59) Standard NA ESS 1 ESS 2 NA Note (20)

Note (20) M-102 M-103 Butterfly 2 M NA O O O NA NA NA NA Note (4, 55) M-104 EFCV MO Gate MO Gate Butterfly 2

2 2

2 Process RM RM M NA M M NA O C C O O C C O O O

O O NA As is As is NA Flow RM (Note 37)

RM (Note 37)

NA Instantan.Standard Standard NA Note (23)

Note (23)

NA NA Note (20,54)

Note (4,55)

M-105 EFCV Butterfly 2 2 Process M NA NA O O O O O O NA NA Flow NA Instantan.

NA NA NA Note (4,55)

M-106 Butterfly MO Gate MO Gate 2 2 2 M RM RM NA M M O C C O C C O O O NA As is As is NA RM (Note 37)

RM (Note 37)

NA Standard Standard NA Note (23)

Note (23) Note (4,55)

Note (20,54)

M-107 Butterfly Vacuum Breaker 2 2 M Process N/A N/A O C O C O C/O NA NA NA Pressure Differential NA NA NA ESS1 ESS2 Note (4,55)

Note (4) M-108 Butterfly Vacuum Breaker 2 2 M Process NA NA O C O C O C/O NA NA NA Pressure Differential NA NA NA ESS1 ESS2 Note (4,55)

Note (4) M-109 Butterfly Vacuum Breaker 2 2 M Process NA NA O C O C O C/O NA NA NA Pressure Differential NA NA NA ESS1 ESS2 Note (4,55)

Note (4) M-110 Butterfly Vacuum Breaker 2 2 M Process NA NA O C O C O C/O NA NA NA Pressure Differential NA NA NA ESS1 ESS2 Note (4,55)

Note (4)

LSCS-UFSAR TABLE 6.2-21 SHEET 19 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TESTLENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft) I-1A, B, C, D, E, F ---

--- --- --- ---

--- --- --- --- --- --- I-2 55 (Note 26) RPV Level and Pressure Reactor Water 3/4 Yes C Detail (w) 1&2B21-F374 Outside No (Note 33) 10 max. I-3 --- --- --- --- --- --- --- --- --- --- 10 max. I-4A 55 (Note 26) 55 (Note 33) RPV Level and Pressure Backfill Reactor Water Reactor Water 3/4 1/2 Yes No C C(b) Detail (w)

Detail (ac) 1&2B21-F376 1&2C11-F423G/

1&2C11-F422G Outside Outside No (Note 33)

Yes (Note 33) 10 max. 10 max I-4B, C, D, E ---

--- --- --- ---

--- --- --- --- --- 10 max. I-4F 56 SUPR CHBR/DW Oxygen Monitor (Unit 1) or Drywell Humidity Monitor (Unit 2) Air 3/4 3/4 No No A (b)

A (b) Detail (g) 1&2CM017A

1&2CM018A Outside Outside Yes Yes 10 max.

10 max. I-5A 55 (Note 26) 55 (Note 33) RPV Level and Pressure Backfill Reactor Water

Reactor Water 3/4 1/2 Yes No C C (b) Detail (w)

Detail (ac) 1&2B21-F359 1&2C11-F423B/

1&2C11-F422B Outside Outside No (Note 33)

Yes (Note 33) 10 max.

18 max. I-5B, C, D, E ---

--- --- --- ---

--- --- --- --- --- 10 max. I-5F 56 Drywell Tritium Grab Sample (Unit 1) or Drywell Humidity Monitor (Unit 2) Air 3/4 3/4 No No A (b)

A (b) Detail (g) 1&2CM017B

1&2CM018B Outside Outside Yes Yes 10 max. 10 max I-6 55 (Note 26) RPV Level and Pressure Reactor Water 3/4 Yes C Detail (w) 1&2B21-F355 Outside No (Note 33) 10 max.

I-7 55 (Note 26) 55 (Note 33) RPV Level and Pressure Backfill Reactor Water Reactor Water 3/4 1/2 Yes No C C (b) Detail (w)

Detail (ac) 1&2B21-F361 1&2C11-F423D/

1&2C11-F422D Outside Outside No (Note 33)

Yes (Note 33) 10 max. 13 max I-8A 55 (Note 26) 55 (Note 33) RPV Level and Pressure Backfill Reactor Water Reactor Water 3/4 1/2 Yes No C C (b) Detail (w)

Detail (ac) 1&2B21-F378 1&2C11-F423F/

1&2C11-F422F Outside Outside No (Note 33)

Yes (Note 33) 10 max. 54 max. I-8B, C, F ---

--- --- --- ---

--- --- --- --- --- --- I-8D 56 Drywell Pressure Air 3/4 No C Detail (w) 1&2VQ061 Outside No (Note 32) 10 max.

LSCS-UFSAR TABLE 6.2-21 SHEET 20 OF 49 REV. 13 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CLASS PRIMARY METHOD OF ACTUATION SECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS I-1A,B,C,D,E,F -- -- -- -- -- -- -- -- -- -- -- Spare I-2 Excess Flow Check 2 Process NA O O O NA Flow Instantaneou s NA I-3 -- -- -- -- -- -- -- -- -- -- -- Spare I-4A Excess Flow Check Checks 2 2 Process Process NA NA O O O C O C NA NA Flow Flow Instantaneou s Instantaneou s NA NA Note (33) I-4B,C,D,E -- -- -- -- -- -- -- -- -- -- -- Spare I-4F SO Globe SO Globe 2 2 Auto Auto RM RM O O O O C C C C B,F,RM B,F,RM 5 sec.

5 sec. ESS 2 ESS 1 Note (20)

I-5A Excess Flow Check Checks 2 2 Process Process NA NA O O O C O C NA NA Flow Flow Instantaneou s Instantaneou s NA NA Note (33) I-5B,C,D,E -- -- -- -- -- -- -- -- -- -- -- Spare I-5F SO Globe SO Globe 2 2 Auto Auto RM RM O O O O C C C C B,F,RM B,F,RM 5 sec.

5 sec. ESS 2 ESS 1 Note (20)

I-6 Excess Flow Check 2 Process NA O O O NA Flow Instantaneou s NA I-7 Excess Flow Check Checks 2 2 Process Process NA NA O O O C O C NA NA Flow Flow Instantaneou s Instantaneou s NA NA Note (33) I-8A Excess Flow Chk Checks 2 2 Process Process NA NA O O O C O C NA NA Flow Flow Instantaneou s Instantaneou s NA NA Note (33) I-8B,C,F -- -- -- -- -- -- -- -- -- -- -- Spare I-8D Excess Flow Check 2 Process NA O O O NA Flow Instantaneou s NA

LSCS-UFSAR TABLE 6.2-21 SHEET 21 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14, 15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TESTLENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft) I-8E 57 (Note 44) RPV Head Seal Leak Detection Air 3/4 No --- Detail (j) 1&2E31-F303 Outside No 10 max. I-9a 55 (Note 26) RPV Level and Pressure Reactor Water 3/4 Yes C Detail (w) 1&2B21-F370 Outside No (Note 33) 10 max. I-9B, C ---

--- --- --- ---

--- --- --- --- --- 10 max. I-9D, E, F 57 (Note 44) ADS Accumulator Pressure Air 3/4 Yes B Detail (j) 1&2B21-F342D, V, SOutside No 10 max. I-10A & B 55 (Note 26) RPV Level and Pressure Reactor Water 3/4 3/4 Yes Yes C C Detail (w)

Detail (w) 1&2B21-F363 1&2B21-F353 Outside Outside No (Note 33)

No (Note 33) 10 max.

10 max. I-10C & D 55 (Note 26) RCIC Steam Flow 3/4 3/4 Yes Yes C C Detail (w)

Detail (w) 1&2B21-F415B

1&2B21-F415A Outside Outside No (Note 33)

No (Note 33) 10 max. 10 max I-10E & F ---

--- --- --- ---

--- --- --- --- --- 10 max. I-11A 56 Primary Cont. Air Sample Air Air 1/2 1/2 No No A (b)

A (b) Detail (g) 1&2CM031 1&2CM032 Outside Outside Yes Yes 10 max.

10 max. I-11B 56 (Note 28) Post LOCA Containment Monitoring Air 1/2 1/2 1/2 Yes No No B A (b)

A (b) Detail (k) Detail (g)

Detail (g) 1&2CM022A 1&2CM029 1&2CM030 Outside Outside Outside No (Note 40)Yes Yes 10 max. NA 10 max. I-12A 55 RPV Level and Pressure Reactor Water 3/4 Yes --- Detail (w) 1&2B21-F357 Outside No (Note 33) 10 max.

I-12B, C, E, F 57 (Note 44) ADS Accumulator Pressure Air 3/4 Yes B Detail (j) 1&2B21-E342E, R, U, C Outside No 10 max. I-12D --- --- --- --- --- --- --- --- --- --- ---

I-13 56 (Note 32) Drywell Pressure Air 3/4 Yes C Detail (w) 1&2B21-F382 Outside No (Note 32) 10 max. I-14A, B, C, D, E, F --- --- --- --- --- --- --- --- --- --- 10 max. I-15A, B, C, D 55 (Note 26) Steam Flow Steam 3/4 3/4 3/4 3/4 Yes Yes Yes Yes C C C C Detail (w) Detail (w)

Detail (w) Detail (w) 1&2B21-F328B 1&2B21-F327B

1&2B21-F327A 1&2B21-F328A Outside Outside Outside Outside No (Note 33) 10 max.

10 max.

10 max. 10 max. I-15 E & F 55 (Note 26) RWCU Flow Reactor Water 3/4 3/4 No No C C Detail (w)

Detail (w) 1&2G33-F312A

1&2G33-F312B Outside Outside No (Note 33)

No (Note 33) 10 max.

10 max. I-16A 55 (Note 26) RHR Line Integrity Reactor Water 3/4 Yes C Detail (w) 1&2E12-F315 Outside No (Note 33) 10 max. I-16B & C --- --- --- --- --- --- --- --- --- --- 10 max.

LSCS-UFSAR TABLE 6.2-21 SHEET 22 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS I-8E Globe 2 Manual NA O O O NA -- -- NA I-9A Excess Flow Check 2 Process NA O O O NA Flow InstantaneousNA I-9B,C -- -- -- -- -- -- -- -- -- -- -- Spare I-9D,E,F Manual 2 Manual NA O O O NA -- -- -- I-10A & B Excess Flow Chk Excess Flow Chk 2 2 Process Process NA NA O O O O O O NAl NA Flow Flow Instantaneous

Instantaneous NA NA I-10C & D Excess Flow Chk Excess FlowChk 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-10E & F -- -- -- -- -- -- -- -- -- -- -- Spare I-11A SO Globe SO Globe 2 2 Auto Auto RM RM O O O O C C C C B,F,RM B,F,RM 5 sec.

5 sec. ESS 2 ESS 1 Note (20)

I-11B SO Globe SO Globe SO Globe 2 2

2 Auto Auto Auto RM RM RM C/O O O C O O O C C O C C RM (Note 37)B,F,RM B,F,RM 5 sec.

5 sec.

5 sec. ESS 1 ESS 2 ESS 1 Note (20)

Note (20)

I-12A Excess Flow Check 2 Process NA O O O NA Flow InstantaneousNA I-12B,C,E,F Manual 2 Manual NA O O O NA -- -- -- I-12D -- -- -- -- -- -- -- -- -- -- -- Spare I-13 Excess Flow Check 2 Process NA O O O NA Pressure InstantaneousNA I-14A,B,C,D,E -- -- -- -- -- -- -- -- -- -- -- Spare I-15A,B,C,D Excess Flow Chk Excess Flow Chk Excess Flow Chk Excess Flow Chk 2 2 2 2 Process Process Process Process NA NA NA NA O O O O O O O O O O O O NA NA NA NA Flow Flow Flow Flow Instantaneous Instantaneous Instantaneous Instantaneous NA NA NA NA I-15E & F Excess Flow Chk Excess Flow Chk 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-16A Excess Flow Check 2 Process NA O O O NA Flow InstantaneousNA I-16B & C -- -- -- -- -- -- -- -- -- -- -- Spare

LSCS-UFSAR TABLE 6.2-21 SHEET 23 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in.) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14)

VALVE ARRANGEMENT FIGURE 6.2-32 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft) I-16D & E 55 (Note 26) RCIC Steam Flow Steam 3/4 3/4 Yes Yes C C Detail (w) 1&2B21-F413B

1&2B21-F413A Outside Outside No (Note 33)

No (Note 33) 10 Max.

10 Max. I-16F 55 (Note 26) LPCS/LPCI P Reactor Water 3/4 Yes C Detail (w) 1&2E21-F304 Outside No (Note 33) 10 Max.

I-17A 55 (Note 26) Jet Pump Pressure Reactor Water 3/4 No C Detail (w) 1&2B21-F344 Outside No (Note 33) 10 Max. I-17B,C,D,E,F --- --- --- -- -- --

-- -- -- --- 10 Max.

I-18 56 (Note 32) Drywell Pressure Air 3/4 Yes -- Detail (w) 1&2B21-F365 Outside No (Note 32) 10 Max.

I-19A I-19B I-19C I-19D I-19E I-19F 55 (Note 26) Jet Pump Flow Reactor Water 3/4 3/4 3/4 3/4 3/4 3/4 No No No No No No C C

C C C C Detail (w)

Detail (w)

Detail (w)

Detail (w) Detail (w) Detail (w) 1&2B21-F443 1&2B21-F439 1&2B21-F437 1&2B21-F441 1&2B21-F445A 1&2B21-F447 Outside Outside Outside Outside Outside Outside No (Note 33)

No (Note 33)

No (Note 33)

No (Note 33)No (Note 33)No (Note 33) 10 Max.

10 Max.

10 Max.

10 Max. 10 Max. 10 Max. I-20A I-20B I-20C I-20D I-20E I-20F 55 (Note 26) Jet Pump Flow Reactor Water 3/4 3/4 3/4 3/4 3/4 3/4 No No No No No No C C C C

C C Detail (w)

Detail (w) Detail (w)

Detail (w)

Detail (w) Detail (w) 1&2B21-F455A 1&2B21-F451 1&2B21-F449 1&2B21-F453 1&2B21-F445B 1&2B21-F455B Outside Outside Outside Outside Outside Outside No (Note 33)

No (Note 33)No (Note 33)

No (Note 33)

No (Note 33)No (Note 33) 10 Max.

10 Max. 10 Max.

10 Max.

10 Max. 10 Max. I-21A,B,C,D,E,F --- --- --- --- -- - --- --- --- --- 10 Max. I-22A & D 55 (Note 26) Recirc. Pump Seal Press. Reactor Water 3/4 3/4 No No C C Detail (w)

Detail (w) 1&2B33-F319A

1&2B33-F317A Outside Outside No (Note 33)

No (Note 33) 10 Max.

10 Max. I-22B & C 55 (Note 26) Recirc. Pump Flow Reactor Water 3/4 3/4 3/4 3/4 No No No No C C C C Detail (x) Detail (x)

Detail (x) Detail (x) 1&2B33-F313C 1&2B33-F313D 1&2B33-F311C 1&2B33-F311D Outside Outside Outside Outside No (Note 33)No (Note 33)

No (Note 33)No (Note 33) 10 Max. 10 Max.

10 Max. 10 Max.

LSCS-UFSAR TABLE 6.2-21 SHEET 24 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CODE CLASS PRIMARY METHOD OF ACTUATION SECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS I-16D & E Excess Flow Check Excess Flow Check 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-16F Excess Flow Check 2 Process NA O O O NA Flow InstantaneousNA I-17A Excess Flow Check 2 Process NA O O O NA Flow InstantaneousNA I-17B,C,D,E,F --- -- --- --- --- --- Spare I-18 Excess Flow Check 2 Process NA O O O NA Pressure InstantaneousNA I-19A I-19B I-19C I-19D I-19E I-19F Excess Flow Check Excess Flow Check Excess Flow Check Excess Flow Check Excess Flow Check Excess Flow Check 2 2 2 2 2 2 Process Process Process Process Process Process NA NA NA NA NA NA O O O O O O O O O O O O O O O O O O NA NA NA NA NA NA Pressure Pressure Pressure Pressure Pressure Pressure Instantaneous Instantaneous Instantaneous Instantaneous Instantaneous Instantaneous NA NA NA NA NA NA I-20A I-20B I-20C I-20D I-20E I-20F Excess Flow Check Excess Flow Check Excess Flow Check Excess Flow Check Excess Flow Check Excess Flow Check 2 2 2 2 2 2 Process Process Process Process Process Process NA NA NA NA NA NA O O O O O O O O O O O O O O O O O O NA NA NA NA NA NA Pressure Pressure Pressure Pressure Pressure Pressure Instantaneous Instantaneous Instantaneous Instantaneous Instantaneous Instantaneous NA NA NA NA NA NA I-21A,B,C,D,E,F --- --- -- -- -- -- -- --- --- -- Spare I-22A & D Excess Flow Check Excess Flow Check 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-22B & C Excess Flow Check Excess Flow Check Excess Flow Check Excess Flow Check 2 2 2 2 Process Process Process Process NA NA NA NA O O O O O O O O O O O O NA NA NA NA Flow Flow Flow Flow Instantaneous Instantaneous Instantaneous Instantaneous NA NA NA NA LSCS-UFSAR TABLE 6.2-21 SHEET 25 OF 49 REV. 13 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in.) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14)

VALVE ARRANGEMENT FIGURE 6.2-32 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft) I-22E & F 55 (Note 26) Recirc. Pump P Reactor Water 3/4 3/4 No No C C Detail (w)

Detail (w) 1&2B33-F315A

1&2B33-F315B Outside Outside No (Note 33)

No (Note 33) 10 Max.

10 Max. I-23A --- --- --- --- -- - --- --- --- --- 10 Max.

I-23B 55 (Note 26) Recirc. Pump Suction Press. Reactor Water 3/4 No C Detail (w) 1&2B33-F301A Outside No (Note 33) 10 Max. I-23C & D 55 (Note 26) Recirc. Pump Flow Reactor Water 3/4 3/4 3/4 3/4 No No No No C C

C C Detail (x)

Detail (x) 1&2B33-F307C 1&2B33-F307D 1&2B33-F305C 1&2B33-F305D Outside Outside Outside Outside No (Note 33)

No (Note 33)

No (Note 33)

No (Note 33) 10 Max.

10 Max.

10 Max.

10 Max. I-23E & F 55 (Note 26) RHR Shutdown Flow Reactor Water 3/4 3/4 Yes Yes C C Detail (w)

Detail (w) 1&2E12-F359B

1&2E12-F359A Outside Outside No (Note 33)

No (Note 33) 10 Max.

10 Max. I-24A,B,C,D,E,F -- --- --- --- -- --

-- --- --- --- 10 Max. I-25A & B 55 (Note 26) RHR Line Integrity Reactor Water 3/4 3/4 Yes Yes C C Detail (w)

Detail (w) 1&2E12-F319

1&2E12-F317 Outside Outside No (Note 33)

No (Note 33) 10 Max.

10 Max. I-25C, D, E, F ---

--- --- --- --- - -- --- --- --- 10 Max. I-26 56 (Note 32) Drywell Press. Air 3/4 Yes C Detail (w) 1&2B21-F367 Outside No (Note 33) 10 Max. I-27A & D 55 (Note 26) Recirc. Pump Flow Reactor Water 3/4 3/4 3/4 3/4 No C C C C Detail (x)

Detail (x) 1&2B33-F307A 1&2B33-F307B 1&2B33-F305A

1&2B33-F305B Outside Outside Outside Outside No (Note 33)No (Note 33)No (Note 33)

No (Note 33) 10 Max. 10 Max. 10 Max.

10 Max.

LSCS-UFSAR TABLE 6.2-21 SHEET 26 OF 49 REV. 13 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CODE CLASS PRIMARY METHOD OF ACTUATION SECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS I-22E & F Excess Flow Check Excess Flow Check 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-23A --- - --- -- - - - -- ---- --- -- Spare I-23B Excess Flow Check 2 Process NA O O O NA Flow Instantaneous NA I-23C & D Excess Flow Check Excess Flow Check Excess Flow Check Excess Flow Check 2 2 2 2 Process Process Process Process NA NA NA NA O O O O O O O O O O O O NA NA NA NA Flow Flow Flow Flow Instantaneous Instantaneous Instantaneous Instantaneous NA NA NA NA I-23E & F Excess Flow Check Excess Flow Check 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-24A,B,C,D,E,F --- - --- --- - - - -- ---- --- -- Spare I-25A & B Excess Flow Check Excess Flow Check 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-25C, D, E, F --- - --- -- - - - -- -- --- -- Spare I-26 Excess Flow Check 2 Process NA O O O NA Pressure Instantaneous NA I-27A & D Excess Flow Check Excess Flow Check Excess Flow Check Excess Flow Check 2 2 2 2 Process Process Process Process NA NA NA NA O O O O O O O O O O O O NA NA NA NA Pressure Pressure Pressure Pressure Instantaneous Instantaneous Instantaneous Instantaneous NA NA NA NA

LSCS-UFSAR TABLE 6.2-21 SHEET 27 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14,15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TESTLENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft) I-27B & C 55 (Note 26) RHR Shutdown Flow Reactor Water 3/4 3/4 Yes C C Detail (w)

Detail (w) 1&2E12-F360A

1&2E12-F360B Outside Outside No (Note 33)

No (Note 33) 10 Max.

10 Max. I-27E&F 55 (Note 26) Recirc. Pump Seal Press. Reactor Water 3/4 3/4 No No C C Detail (w) 1&2B33-F317B 1&2B33-F319B Outside Outside No (Note 33)

No (Note 33) 10 Max.

10 Max. I-28A 55 (Note 26) Recirc. Pump Suction Press. Reactor Water 3/4 No C Detail (w) 1&2B33-F301B Outside No (Note 33) 10 Max. I-28B & C 55 (Note 26) Recirc. Pump P Reactor Water 3/4 3/4 No No C C Detail (w)

Detail (w) 1&2B33-F315D 1&2B33-F315C Outside Outside No (Note 33)

No (Note 33) 10 Max.

10 Max. I-28D & E 55 (Note 25) Recirc. Pump Flow Reactor Water 3/4 3/4 3/4 3/4 No No No No C C C C Detail (x)

Detail (x) 1&2B33-F313A 1&2B33-F313B 1&2B33-F311A 1&2B33-F311B Outside Outside Outside Outside No (Note 33)

No (Note 33)No (Note 33)No (Note 33) 10 Max.

10 Max. 10 Max. 10 Max. I-28F 55 (Note 26) RPV Drain Flow Reactor Water 3/4 No C Detail(w) 1&2G33-F309 Outside No (Note 33) 10 Max. I-29A, D, E, F 55 (Note 26) Steam Flow Steam 3/4 3/4 3/4 3/4 No No No No C C C C Detail(w) Detail(w)

Detail(w) Detail(w) 1&2B21-F326D 1&2B21-F325D 1&2B21-F325C 1&2B21-F326C Outside Outside Outside Outside No (Note 33)No (Note 33)

No (Note 33)No (Note 33) 10 Max. 10 Max.

10 Max. 10 Max. I-29B 55 (Note 26) Core P Reactor Water 3/4 Yes C Detail(w) 1&2B21-F350 Outside No (Note 33) 10 Max.

I-29C 55 (Note 26) RPV Bottom Head Drain Flow Reactor Water 3/4 No C Detail(w) 1&2B21-F346 Outside No (Note 33) 10 Max. I-30A & B 55 (Note 26) RPV/HPCS P Reactor Water 3/4 3/4 No No C C Detail(w)

Detail(w) 1&2B21-F348

1&2E22-F304 Outside Outside No (Note 33)

No (Note 33) 10 Max.

10 Max. I-30C, D, E, F 57 (Note 44) MSIV Accumulator Pressure Air 3/4 No B Detail(j) 1&2B21-F329A,B,C,D Outside No 10 Max.

I-31A I-31B I-31C I-31D I-31E I-31F 55 (Note 26) Jet Pump Flow Reactor Water 3/4 3/4 3/4 3/4 3/4 3/4 No No No No No No C C C C

C C Detail(w)

Detail(w) Detail(w)

Detail(w)

Detail(w) Detail(w) 1&2B21-F471 1&2B21-F469 1&2B21-F473 1&2B21-F465B 1&2B21-F475B 1&2B21-F475A Outside Outside Outside Outside Outside Outside No (Note 33)

No (Note 33)No (Note 33)

No (Note 33)

No (Note 33)No (Note 33) 10 Max.

10 Max. 10 Max.

10 Max.

10 Max. 10 Max

LSCS-UFSAR TABLE 6.2-21 SHEET 28 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS I-27B & C EFC EFC 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-27E & F EFC EFC 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-28A EFC 2 Process NA O O O NA Flow Instantaneous NA I-28B & C EFC EFC 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-28D & E EFC EFC EFC EFC 2 2 2 2 Process Process Process Process NA NA NA NA O O O O O O O O O O O O NA NA NA NA Flow Flow Flow Flow Instantaneous InstantaneousInstantaneousInstantaneous NA NA NA NA I-28F EFC 2 Process NA O O O NA Flow Instantaneous NA I-29A,D,E,F EFC EFC EFC EFC 2 2 2 2 Process Process Process Process NA NA NA NA O O O O O O O O O O O O NA NA NA NA Flow Flow Flow Flow InstantaneousInstantaneous InstantaneousInstantaneous NA NA NA NA I-29B EFC 2 Process NA O O O NA Flow Instantaneous NA I-29C EFC 2 Process NA O O O NA Flow Instantaneous NA I-30A & B EFC EFC 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneous Instantaneous NA NA I-30C,D,E,F Manual 2 Manual NA O O O NA -- -- --

I-31A I-31B I-31C I-31D I-31E I-31F EFC EFC EFC EFC EFC EFC 2 2

2 2 2 2 Process Process Process Process Process Process NA NA NA NA NA NA O O

O O O O O O

O O O O O O

O O O O NA NA NA NA NA NA Flow Flow Flow Flow Flow Flow Instantaneous Instantaneous Instantaneous InstantaneousInstantaneousInstantaneous NA NA NA NA NA NA EFC = Excess Flow Check

LSCS-UFSAR TABLE 6.2-21 SHEET 29 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14,15)

VALVE ARRANGEMENT FIGURE 6.2-31 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft)

I-32A I-32B I-32C I-32D I-32E I-32F 55 (Note 26) Jet Pump Flow Reactor Water 3/4 3/4 3/4 3/4 3/4 3/4 No No No No No No C C

C C C C Detail (w)

Detail (w)

Detail (w) Detail (w) Detail (w)

Detail (w) 1&2B21-F465A 1&2B21-F467 1&2B21-F463 1&2B21-F459 1&2B21-F457 1&2B21-F461 Outside Outside Outside Outside Outside Outside No (Note 33)

No (Note 33)

No (Note 33)No (Note 33)No (Note 33)

No (Note 33) 10 Max.

10 Max.

10 Max. 10 Max. 10 Max.

10 Max. I-33 56 (Note 32) Drywell Pressure Air 3/4 Yes C Detail (w) 1&2B21-F380 Outside No (Note 33) 10 Max. I-34A, D, E, F 55 (Note 26) Steam Flow Steam 3/4 3/4 3/4 3/4 Yes Yes Yes Yes C C C C Detail (w)

Detail (w) Detail (w) Detail (w) 1&2B21-F328D 1&2B21-F328C 1&2B21-F327C 1&2B21-F327D Outside Outside Outside Outside No (Note 33)

No (Note 33)No (Note 33)No (Note 33) 10 Max.

10 Max. 10 Max. 10 Max. I-34B & C ---

--- --- --- --- - --- --- --- --- --- I-35 56 (Note 28)

56 Post LOCA Containment Monitoring HRSS Sampling

Air Air 1/2 1/2 Yes No B

A(b) A(b) Detail (k)

Detail (g) Detail (g) 1&2CM023B

1&2CM085 1&2CM086 Outside

Outside Outside No (Note 40)

Yes Yes 10 Max.

10 Max. 10 Max. I-36 56 (Note 28) Post LOCA Containment Monitoring Air 1/2 1/2 1/2 Yes No No B A (b)

A (b) Detail (k) Detail (g)

Detail (g) 1&2CM024A 1&2CM027 1&2CM028 Outside Outside Outside No (Note 40)Yes Yes 10 Max. Not Applicable10 Max. I-37A, B, C, D 55 (Note 26) Steam Flow Steam 3/4 3/4 3/4 3/4 Yes Yes Yes Yes C C C C Detail (w) Detail (w)

Detail (w) Detail (w) 1&2B21-F325A 1&2B21-F326A

1&2B21-F325B 1&2B21-F326B Outside Outside Outside Outside No (Note 33)No (Note 33)

No (Note 33)No (Note 33) 10 Max. 10 Max.

10 Max. 10 Max. I-37E & F ---

--- --- --- ---

--- --- --- --- --- 10 Max. I-38 & 39 NA Supp. Chamber Air 1 1/4 No

--- --- ---

--- ---

--- ---

--- ---

--- 10 Max.

10 Max.

LSCS-UFSAR TABLE 6.2-21 SHEET 30 OF 49 REV. 14, APRIL 2002 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS I-32A I-32B I-32C I-32D I-32E I-32F EFC EFC EFC EFC EFC EFC 2 2 2 2

2 2 Process Process Process Process Process Process NA NA NA NA NA NA O O O O

O O O O O O

O O O O O O

O O NA NA NA NA NA NA Flow Flow Flow Flow Flow Flow Instantaneous InstantaneousInstantaneous Instantaneous InstantaneousInstantaneous NA NA NA NA NA NA I-33 EFC 2 Process NA O O O NA Pressure InstantaneousNA I-34A,D,E,F EFC EFC EFC EFC 2 2 2 2 Process Process Process Process NA NA NA NA O O O O O O O O O O O O NA NA NA NA Pressure Pressure Pressure Pressure Instantaneous InstantaneousInstantaneousInstantaneous NA NA NA NA I-34B,C -- -- -- -- -- -- -- -- -- -- -- Spare I-35 SO Globe SO Globe SO Globe 2 2 2 RM Manual Manual N/A N/A N/A C/O C C C C C O C/O C/O O C C RM --- --- 5 sec. --- --- ESS 2 N/A N/A I-36 SO Globe SO Globe SO Globe 2 2

2 RM Auto Auto N/A RM RM C/O O O C O O O C C O C C RM B,F,RM B,F,RM 5 sec.

5 sec.

5 sec. ESS 1 ESS 2 ESS 1 Note (20)

I-37A,B,C,D EFC EFC EFC EFC 2 2 2 2 Process Process Process Process NA NA NA NA O O O O O O O O O O O O NA NA NA NA Flow Flow Flow Flow Instantaneous InstantaneousInstantaneousInstantaneous NA NA NA NA I-37E&F -- -- -- -- -- -- -- -- -- -- -- Spare I-38 & 39

-- -- --

-- --

-- --

-- --

-- --

-- --

-- --

-- --

-- --

-- --

-- RTDs are provided through these connections

EFC = Excess Flow Check

LSCS-UFSAR TABLE 6.2-21 SHEET 31 OF 49 REV. 13 CONTAINMENT PENETRATION NUMBER NRC GDC LINE ISOLATED FLUID CONTAINED LINE SIZE (in.) ESF SYSTEM (NOTE 21)THROUGH LINE LEAKAGE CLASSIFICATION (NOTE 14)

VALVE ARRANGEMENT FIGURE 6.2-32 VALVE NUMBER LOCATION WITH RESPECT TO CONTAINMENTTYPE C TEST LENGTH OF PIPE FROM CONTAINMENT TO OUTERMOST VALVE (ft) I-40,41, 42,43 56 (Note 32) Supp. Pool Water Level Supp. Pool Water 3/4 3/4 3/4 3/4 3/4 3/4 3/4 3/4 No No No No No No No No Exempt Exempt Exempt Exempt Exempt Exempt Exempt Exempt Detail (v) Detail (v)

Detail (v) Detail (v) Detail (v)

Detail (v)

Detail (v) Detail (v) 1&2CM039 1&2CM040 1&2CM041 1&2CM042 1&2CM043 1&2CM044 1&2CM045 1&2CM046 Outside Outside Outside Outside Outside Outside Outside Outside No (Note 32)No (Note 32)

No (Note 32)No (Note 32)No (Note 32)

No (Note 32)

No (Note 32)No (Note 32) 10 Max. 10 Max.

10 Max. 10 Max. 10 Max.

10 Max.

10 Max. 10 Max. I-44 & 46 -- Supp. Pool Water Temp. -- 1 1/4 1 1/4 -- -- -- 10 Max.

10 Max. I-45 56 (Note 28) Drywell Air Sampling Post LOCA Cont. Mont. Drywell Humidity Sampling Air 1 No No Yes No No No No A (b)

A (b) B A(b)

A(b)

A(b)

A(b) Detail (g)

Detail (k) Detail (g)

Detail (g) 1&2CM034 1&2CM033 1&2CM025A 2CM020A 2CM019A 1&2CM020B

1&2CM019B Outside Outside Outside Outside Outside Outside Outside Yes Yes No (Note 40)Yes Yes Yes Yes 10 Max.

10 Max.

10 Max. 10 Max. 10 Max.

10 Max.

10 Max. I-47 56(Note 28)

56 Post LOCA Containment Monitoring

HRSS Sampling Air

Air 1 1/4

1/2 Yes

No B

A(b) A(b) Detail (w)

Detail (g) Detail (g) 1&2CM026B

1&2CM089 1&2CM090 Outside

Outside Outside No(Note 40)

Yes Yes 10 Max

10 Max. 10 Max.. I-48 & 49 56 (Note 32) Supp. Pool Water Level Supp. Pool Water 1 1/4 1 1/4 No No C C Detail (w)

Detail (w) 1&2E22-F341

1&2E22-F342 Outside Outside No(Note 32)

No(Note 32) 10 Max.

10 Max. I-50 56 (Note 28) 56 Post LOCA Containment Monitoring HRSS Sampling Air Air 1/2 1/2 Yes No B A(b) A(b) Detail (k)

Detail (g) Detail (g) 1&2CM021B

1&2CM085 1&2CM086 Outside

Outside Outside No (Note 40)

Yes Yes 10 Max.

10 Max. 10 Max.

LSCS-UFSAR TABLE 6.2-21 SHEET 32 OF 49 REV. 13 CONTAINMENT PENETRATION NUMBER VALVE TYPE ASME SECTION III CODE CLASS PRIMARY METHOD OF ACTUATIONSECONDARY METHOD OF ACTUATION NORMAL VALVE POSITIONSHUTDOWN VALVE POSITION POST ACCIDENT POSITION POWER FAILURE VALVE POSITION (6) ISOLATION SIGNAL VALVE CLOSURE TIME (7) POWER SOURCE REMARKS I-40,41, 42,43 Globe Globe Globe Globe Globe Globe Globe Globe 2 2

2 2 2 2 2

2 Manual Manual Manual Manual Manual Manual Manual Manual N/A N/A N/A N/A N/A N/A N/A N/A C C

C C C C C

C C C

C C C C C

C C C

C C C C C

C NA NA NA NA NA NA NA NA Flow Flow Flow Flow ---

---

---

--- ---

---

---

--- --- ---

---

--- NA NA NA NA NA NA NA NA I-44 & 46 RTDs are provided through these connecti I-45 SO Globe SO Globe SO Globe SO Globe SO Globe SO Globe SO Globe 2 2 2 2 2

2 2 Auto Auto Auto Auto Auto Auto Auto RM RM RM RM RM RM RM O O C/O O O

O O O O C O O

O O C C O C C

C C C C O C C

C C B,F,RM B,F,RM RM (Note 37)B,F,RM B,F,RM B,F,RM B,F,RM 5 sec. 5 sec. 5 sec. 5 sec.

5 sec.

5 sec. 5 sec. ESS 2 ESS 1 ESS 1 ESS 2 ESS 1 ESS 2 ESS 1 (Note 20)

(Note 20)

(Note 20)

I-47 SO Globe

SO GLOBE

SO GLOBE 2

2 2 Auto

Manual Manual RM

N/A N/A C/O

C C C

C C O

C/O C/O O

C C RM(Note37)

---

--- 5 sec.

---

--- ESS 2

N/A N/A

I-48 & 49 Excess Flow Check Excess Flow Check 2 2 Process Process NA NA O O O O O O NA NA Flow Flow Instantaneou s Instantaneou s NA NA I-50 SO Globe

SO Globe SO Globe 2

2 2 Auto

Manual Manual RM

N/A N/A C/O

C C C

C C O

C/O C/O O

C C RM (Note37)

---

--- 5 sec.

---

--- ESS 2

N/A N/A LSCS-UFSAR TABLE 6.2-21 SHEET 33 OF 49 REV. 14, APRIL 2002 SIGNAL DESCRIPTION A Reactor vessel low water level level 3 - (A scram occurs at this level also. This is the higher of the two low water level signals.)

B Reactor vessel low low water level level 2 - (The RCIC and HPCS systems are initiated at this level also. (This is the lower of the two low water level signals.)

C High radiation - Main steam

D Line break - High area temperat ure or very high system flow.

E Main condenser low vacuum.

F High drywell pressure.

G Reactor vessel low low low water level (Level 1) or high drywell pressure (Emergency Core Cooling System are started).

H Reactor vessel low low low water level (Level 1)

J Line break in cleanup system - high space temperature.

M Line break in RHR shutdown and head cooling (high space temberature).

P Low main steamline pressure at inlet turbine (RUN mode only).

U High reactor vessel pressure - close RHR shutdown cooling valves and head cooling valves.

Y High radiation, fuel pool ventilation exhaust.

Z High radiaion, reactor building ventilation exhaust.

RM Remote manual switch from control room. (All regular Class A and Class B isolation valves are capable of remote manual

operation from the control room.)

RME Remote manual switch from Auxilia ry Electric Equipment Room.

Note - position indication also available in Control Room in group summary position indicator lights.

LSCS-UFSAR TABLE 6.2-21 SHEET 34 OF 49 REV. 15, APRIL 2004 These notes are keyed by number to correspond to numbers in parenthesis in Table 6.2-21.

1. Main steam isolation valves require that both solenoid pilots be de-energized to close valves. Accu mulator air pressure plus spring force together close valves when both pilots are de-energized. Voltage failure at only one pilot does not cause valve closure. The valves are designed to fully close in less than 5 seconds.

2. Suppression pool spray (1(2)E12-F027A/B) and suppression pool cooling valves (1(2)E12-F024A/B) have interlocks that allow them to be manually reopened after automatic closure. This setup permits suppression pool spray, for high drywell pressure conditions, and/or suppression pool water cooling. The drywell spray valves (1(2)E12-F016A/B, 1(2)E12-F017A/B), do not receive any automatic closure signals.

3. Testable check valves are provided with an air operator for remote opening with zero differential pre ssure across the valve seat. These valves will close on reverse flow even though the test switches may be positioned for open. The valves open when pump pressure exceeds reactor pressure even though the test switch may be closed. The testable feature has been eliminated from the Division 1, 2, and 3, ECCS testable check valves.

4. In the normal configuration the lines are considered to be an extension of primary containment. If a vacuum breaker valve is inoperable, the butterfly valve will be closed to prevent bypass leakage. If a vacuum breaker valve is subsequently removed, a blind flange will be added, and the flang e and butterfly valve will form the containment boundary. The vacuum breaker valves will be leakage tested as part of the periodic low pressure suppression bypass leakage test. The acceptance limits are based on the allowable suppression bypass capability of the containment.

5. A-c motor-operated valves required for isolation functions are powered from the a-c standby power buses. D-c operated isolation

valves are powered from the station batteries.

6. All motor-operated isolation valves remain in the last position upon failure of valve power. All air-operated valves close on motive air failure except the VQ Butterfly valves which require their solenoid valves to be deenergized.

LSCS-UFSAR TABLE 6.2-21 SHEET 35 OF 49 REV. 13 7. The standard operating times for power actuated valves based on actual stem travel shall be less than or equal to 110% of the nominal values below: Motor-operated Air-Operated Gate valves 12 in./min Not applicable Globe valves 4 in./min 4 in./min Butterfly valves 30 - 90 seconds 0 - 10 seconds

8. Reactor building vent exhaust high radiation signal "z" and fuel pool ventilation exhaust high radiation signal "Y" are generated by two trip units; this requires one unit at high trip or both units at downscale (instrument failure trip), in order to initiate isolation.

9. Valves can be opened or closed by remote manual switch for operating convenience during any mode of reactor operation except when an automatic signal is present.

10. Normal status position of valve (open or closed) is the position during normal power operation of the reactor (see "Normal Status" column).

11. Deleted.

12. Deleted.

13. Deleted.

14. Categories indicated are in accordance with ASME Section XI Article IWV-2000. The types of leakage tests are as follows: (a) water test and (b) air test. Exempt valves are those used for testing, draining, venting, maintenance or operational convenience.

15. The leakage criteria for these valves is specified in 10 CFR 50 Appendix J and the LaSalle Primary Containment Leak Rate Testing Program. 16. Deleted.

17. The outboard check valves on the feedwater return lines are provided with an air operator for testing the valves to ensure that the disks are not frozen in the open position. The actuator moves the disk partially into the flow stream, but is not capable of completely closing the valve against flow. The feedwater valve actuator is used to apply seating force to the valve for ensuring leaktightness at low differential pressures. The actuator will be exercised to assure operability prior to leak testing.

LSCS-UFSAR TABLE 6.2-21 SHEET 36 OF 49 REV. 14, APRIL 2002

18. The TIP drive guide tubes provide a sealed path for the flexible drive cable of the TIP probes. The TIP tubing seals the TIP system from the reactor coolant and forms a leak tight boundary designed for reactor coolant pressure boundary conditions. The shear valve is provided to cut the cable in the event that the drive cable cannot be withdrawn, and the

ball provides the guide tubes with shut-off capability.

The LaSalle TIP system design specifications require that the maximum leakage rate of the ball and shear valves shall be in accordance with the

Manufacturers Standardization Society (Hydrostatic Testing of Valves). The ball valves are 100% leak tested to the following criteria by the manufacturer:

Pressure 0 - 62 psig Temperature 340°F Leak Rate 10-3 cm 3 /s A statistically chosen sample of the shear valves is tested by the manufacturer to the following criteria:

Pressure 0 - 125 psig Temperature 340°F Leak Rate 10

-3 cm 3 /sec STP.

The shear valves have explosive squibs and require testing to destruction. They cannot therefore be 100% tested nor can they be tested in accordance with 10 CFR 50 Appendix J requirements after installation.

Isolation is accomplished by a seis mically qualified solenoid-operated ball valve, which is normally closed. Ball valve position is indicated in the control room. The ball valve is periodically leak tested in accordance with the LaSalle 10 CFR 50 Appendix J Program and the acceptable leakage limits for these valves are in accordance with the Appendix J program.

When the TIP system cable is inserted, the ball valve of the selected tube opens automatically so that the probe and cable may advance. A maximum of four valves may be opened at any one time to conduct calibration, and any one guide tube is used, at most, a few hours per year.

If closure of the line is required during calibration, a signal causes the cable to be retracted and the ball valve to close automatically after completion of cable withdrawal. If a TIP cable fails to withdraw or a ball valve fails to close, each line is equipped with an explosive shear valve.

LSCS-UFSAR TABLE 6.2-21 SHEET 37 OF 49 REV. 14, APRIL 2002 If a failure occurs, the shear valve would be manually actuated from the Main Control Room to shear the TIP cable and isolate the penetration.

Because the TIP shear valve requires testing to destruction, it is not tested in accordance with 10 CFR 50 Appendix J, but instead is tested as specified in Technical Specification. The Technical Specification verifies continuity of the explosive charge and batch sampling testing of the explosive squib charges, with replacement of the explosive squib before expiration of the shelf-life and operating life. A statistical sample of the shear valves are leak tested in the manufacturers shop to ensure that the leakage limits conform to the design specification limits of 10-3 cm 3/sec.

19. The hydraulic lines are sealed pipe designed for 2000 psig operating pressure.

20. Test pressure is not in the same direction as the pressure existing when the valve is required to perform the safety function as required by Appendix J to 10 CFR 50. Either m anufacturers' test data, site test results or justification (e.g., reverse te st pressure tending to lift disk from seat) will be available on site to verify that testing in the reverse direction will provide either equivalent or more conservative results.

LSCS-UFSAR TABLE 6.2-21 SHEET 38 OF 49 REV. 14, APRIL 2002

21. Although the valves listed may be included in the containment isolation system which is an ESF system, a "yes" designation is given only for those valves in systems where the parent system containing the valve is an ESF system.

22. The valves associated with RHR "A" loop are powered from ESS1 sources. The valves associated with RHR "B" and "C" loops are supplied from ESS2 power sources.

23. The power source for the valves associated with penetrations M-23 (Unit 2), M-33 (Unit l) and M-106 is ESS1. The power source for the valves associated with penetrations M-53 and M-104 is ESS2. This arrangement was used to maintain redundancy of function for the

combustible gas control system. The valves are closed during normal plant operation, and are open only for periodic testing and following a LOCA. 24. Criterion 55 concerns those lines of the reactor coolant pressure boundary penetrating the primary reactor containment. The control rod drive (CRD) insert and withdraw lines are not part of the reactor coolant pressure boundary. The basis to which the CRD lines are designed is commensurate with the safety importance of isolating these lines. Since these lines are vital to the scram function, their operability is of utmost concern. In the design of this system, it has been accepted practice to omit automatic valves for isolation purposes, as this introduces a possible failure mechanism. As a means of providing positive actuation, manual shutoff valves (1&2C11D001-101 and -102) are used. The charging water, drive water and cooling water headers are provided with a check valve (1&2C11D001-115, -137 and -138) within the hydraulic control unit (HCU), a Seismic Category I module, and the normally closed solenoid valves (1&2C11D001-120, -121, -122 and -123). These valves will prevent any direct flow away from containment. These valves are shown on Sheet 3 of Drawing M-100 (Unit 1) and M-146 (Unit 2).

If an insert line fails, a ball check valve provided in each drive is designed to seal off the broken line by using reactor pressure to shift the ball check valve to the upper seat. This feature also prevents any direct flow away from the primary containment.

LSCS-UFSAR TABLE 6.2-21 SHEET 39 OF 49 REV. 14, APRIL 2002 When the HCU's are pressurized, leaks resulting from degraded piping integrity would be observed by the Op erators on their daily rounds. In addition, several indicators in the control room, such as temperature and pressure of CRD cooling water or dryw ell sump pump operation, indicates whether leakage is excessive. The maximum leakage expected at this penetration is 3 gpm when the RPV is still pressurized (about 1000 psi). This leakage also assumes a single active failure of a check valve inside the HCU. After the reactor vessel is depressurized, the CRD leakage will decrease to about 0.5 gpm. It may also be said that leakage monitoring of the CRD insert and withdraw lines is provided by the overall type A leakage rate test. Since the RPV and nonseismic portions of the CRD system are vented during the performance of the Type A test, any leakage from these lines would be included in the total Type A test leakage.

The flowout of the CRD is restricted through the HCU performance test requirements to ensure that HCU leak age does not exceed 0.2 gpm. The maximum leakage expected for these penetrations is 0.2 gpm per HCU. If a single failure is assumed, the maximum leakage would be 3 gpm. Seismic tests have demonstrated the seal integrity of the CRD system. Maximum leakage following these tests did not exceed 3 gpm.

The system design criteria are as follows:

Seismic Category Quality Group Classification Quality Assurance Classification Valves; insert and withdraw I B I Insert and withdraw line

piping I B I The CRD insert and withdraw lines are compatible with the criteria intended by 10 CFR 50, Appendix J for Type C testing, since the acceptance criterion for Type C testing allows demonstration of fluid leakage rates by associated bases. The maximum leakage expected has been factored in with the total allowable containment penetration leakage and determined to be acceptabe.

LSCS-UFSAR TABLE 6.2-21 SHEET 40 OF 49 REV. 13 25. The recirculation pump seal water line extends from the recirculation pump through the drywell and connects to the CRD supply line outside the primary containment. The seal water line forms a part of the reactor coolant pressure boundary; therefore, the consequences of failing this line have been evaluated. This evaluation shows that the consequences of breaking this line are less severe than failing an instrument line. Therefore, the two check valves in series provide sufficient isolation capability for postulated failure of this line.

These lines are high-pressure lines coming from the discharge of the CRD pumps to the recirculation pump seals. They are provided with a check valve inside the containment and a check valve outside the containment.

The inside and outside check will receive a Type C local leak test with water as the testing mechanism during refueling outages.

26. See Note 33.

LSCS-UFSAR TABLE 6.2-21 SHEET 41 OF 49 REV. 15, APRIL 2004

27. The Hydraulic Control Unit (HCU) is a factory-assembled engineered module of valves, tubing, piping, and stored water which controls a single control rod drive by the application of precisely timed sequences of pressures and flows to accomplish slow insertion or withdrawal of the control rods for power control, and rapid insertion for reactor scram.

Although the hydraulic control unit, as a unit, is field installed and connected to process piping, many of its internal parts differ markedly from process piping components because of the more complex functions they

must provide.

Thus, although the codes and standards invoked by Groups A, B, C and D pressure integrity quality levels clearly apply at all levels to the interfaces between the HCU and the connecting conventional piping components (e.g.,

pipe nipples, fittings, simple hand valves, etc.), it is considered that they do not apply to the specialty parts (e.g., solenoid valves, pneumatic components, and instruments). The HCU shutoff (isolation) valves are

Quality Group B.

The design and construction specifications for the HCU do invoke such codes and standards as can be reasonably applied to individual parts in developing required quality levels, but these codes and standards are supplemented with additional requirements for these parts and for the remaining parts and details. For example, 1) all welds are penetrant tested (PT), 2) all socket welds are inspected for gaps between pipe and socket bottom, 3) all welding is performed by qualified welders, and 4) all work is done per written procedures. Quality Group D is generally applicable because the codes and standards involk ed by that group contain clauses which permit the use of manufacturer's standards and proven design techniques which are not explicitly defined within the codes of Quality Group A, B, or C. This is supplemented by the QC techniques.

28. These lines have been evaluated to an acceptable alternative design basis other than that specifically listed in GD C 55 and 56. This alternate basis is found in SRP 6.2.4.II.6, and the

LSCS-UFSAR TABLE 6.2-21 SHEET 42 OF 49 REV. 13 evaluation to the criteria specified therein is as follows:

a. All lines are in engineered safety feature or engineered safety featured-related systems.

b. System reliability can readily be seen to be greater when only a single valve is provided, since the addition of another valve in series provides an additional potential point of failure, and, in the case of relief valve discharge lines, the installation of an additional valve is actually prohibited by the ASME Code.

c. The systems are closed outside containment.

d. A single active failure of these ESF systems can be accommodated.

e. The systems outside containment are protected from missiles consistent with their classification as ESF systems.

f. The systems are designed to Seismic Category I standards.

g. The systems are classified as Safety Class 2.

h. The design ratings of these systems meet or exceed those specified for the primary containment.

i. The leaktightness of these systems is assured by normal surveillance, inservice testing and leak detection monitoring.

j. The single valve on these lines is located outside containment.

29. These lines are always filled with water on the outboard side of the containment thereby forming a water seal. They are maintained at a pressure that is always higher t han primary containment pressure by water leg pumps; thus, precluding any outleakage from primary containment. However, even if outleakage did occur it would be into an ESF system which forms a closed loop outside primary containment. Thus, any leakage from primary containment would return to primary containment through this closed loop.

LSCS-UFSAR TABLE 6.2-21 SHEET 43 OF 49 REV. 17, APRIL 2008 These valves are under continuous le akage test because they are always subjected to a differential pressure acting across the seat. Leakage through these valves is continuously monitore d by the pressure switches in the pump discharge lines, which have a low alarm setpoint in the main control room.

Even though a special leakage test is not merited on these valves for the reasons discussed above, a system leakage test will be performed and compared to an acceptance limit based on site boundary dose considerations.

30. The leakages through the Main Steamline valves will not be included in establishing the acceptance limits for the combined leakage in accordance with the 10 CFR 50, Appendix J, Type B and C tests. The NRC granted exemption to 10 CFR 50, Appendix J, for not including MSIV leakage in the Type A, B, or C acceptance criteria. This exemption is based on the use of the MSIV Isolated Condenser Leakage Treatment Method discussed in Section 6.8, and associated analyses.

31. Although only one isolation valve signal is indicated for these valves, the valves also receive automatic signals from various system operational parameters. For example, the ECCS pump minimum flow valves close

automatically when adequate flow is ac hieved in the system; the ECCS test lines close automatically on receipt of an accident signal. Although these signals are not considered isolation signals; and are therefore, excluded from this table, there are other system operation signals that control these valves to ensure their proper position for safe shutdown. Reference to the logic diagrams for these valves indicates which other signals close these valves. 32. To satisfy the requirements of General Design Criterion 56 and to perform their function, these instrument lines have been designed to meet the requirements of Regulatory Guide 1.11 (Safety Guide 11).

These lines are Seismic Category I and terminate in instruments that are Seismic Category I. They are provided with manual isolation valves and excess flow check valves.

LSCS-UFSAR TABLE 6.2-21 SHEET 44 OF 49 REV. 14, APRIL 2002 The integrity of these lines is to be tested during the Type "A" Test. These lines and their associated instruments are to be pressurized to P

a. Surveillance inspections are performed to ensure that the leaktight integrity of these lines and their associated instruments. Additional inservice inspection is included in the Technical Specifications. This inservice inspection verifies the function of the excess flow check valves.

Isolation is provided by the excess flow check valve. In the event of a line rupture downstream of the check valv e and a containment pressure above 2 psig this valve would close to limit the amount of leakage.

33. To perform their function and to satisfy the requirements of General Design Criterion 55, these instrument lines have been designed to meet the requirements of Regulatory Guide 1.11 (Safety Guide 11).

These lines are Seismic Category I and terminate in instruments that are Seismic Category I. They-are provided with flow-restricting orifices, manual isolation valves, and excess flow check valves.

The flow-restricting orifice is sized to assure that in the event of a postulated failure of the piping or component, the potential offsite exposure would be substantially below the guidelines of 10 CFR 100.

Isolation is provided by the excess flow check valve. In the event of a line rupture downstream of the check valves, this valve would close to limit the amount of leakage.

The integrity of these lines are tested during the Type "A" Test. Surveillance inspections are performed to ensure the leaktight integrity of these lines and their associated instruments. Additional inservice testing is included in the Technical Specifications. This inservice inspection verifies the function of the excess flow check valves.

For Unit 1 Penetrations M-21 and M-59, and Unit 2 Penetrations M-52 and M-65 reference leg backfill lines have been installed to comply with NRC

Bulletin 93-03. These lines tap into the reference legs outboard of the excess flow check valves. Two safety related, Seismic Category I, check valves provide the boundary between the non-safety related CRD system and the safety related reference leg. These two check valves also form part of the boundary that will be checked by surveillance inspections in accordance with Check Valve Monitoring and Preventative Maintenance Program.

For Penetrations I-4A, I-5A, I-7 and I-8A, reference leg backfill lines have been installed to comply with NRC Bulletin 93-03. These lines tap into reference lines 1(2)NB10A-3/4", 1(2)NB12A-3/4", 1(2)NB23A-3/4" and 1(2)NB25A-3/4" between the containment penetration and the manual isolation valve/excess flow check valve combination. This makes these lines part of the reactor coolant pressure boundary. This location was chosen to prevent the mispositioning of the manual isolation valve (while the injection line is LSCS-UFSAR TABLE 6.2-21 SHEET 45 OF 49 REV. 13 functioning) from over pressurizing all the instruments on the instrument panel. Two safety related, Seismic Category I, check valves in series act as the outboard containment isolation valves. These two valves also provide the boundary between the non-safety related CRD system and the safety related reference leg as well as form part of the boundary that will be checked by surveillance inspections in accordance with Check Valve Monitoring and Preventative Maintenance Program.

34. These valves are provided for long-term leaktightness only. Feedwater check valves in each line provide immediate isolation. These MO valves are remote manually closed from the control room upon indication of loss of feedwater flow. Therefore, no additional isolation signals are required.

35. Penetrations M-49 and M-50 contain lines for the hydraulic control of the reactor recirculation flow control valves. The hydraulic fluid in these lines is used to position the flow control valves.

Three of four lines of each penetration in this system are under a constant pressure test during normal plant operations due to its high operating pressure of 1800 psig. The fourth line of each penetration in this system is a seal leakage return line back to the HPU Reservoir. Any leakage from this system would be limited to hydraulic fluid which fills these lines and is independent of the containment atmosphere.

In order to perform Type C leakage tests on the isolation valves associated with this system, the system would have to be disabled and the hydraulic

fluid drained. This is detrimental to the proper operation of the system in that possible damage could occur in establishing the test condition or restoring the system to normal.

Therefore, these hydraulic isolatio n valves are exempted from Type C testing.

36. The feedback information available to the plant operator which enables him to determine when the valves with only a "Remote Manual (RM)" closure should be closed is summarized as follows:

a. Leak detection information, as described in Subsection 7.6.2.2 is available to enable the operator to determine the location of a leak or line failure, and close the isolation valve associated with that line.

b. RPV level information is available to the operator to ascertain whether the flow is actually reaching the RPV.

c. Suppression pool water level information would also identify the occurrence of a line failure or leakage.

LSCS-UFSAR TABLE 6.2-21 SHEET 46 OF 49 REV. 17, APRIL 2008

37. These valves are required to open on signals B and F during the post-LOCA conditions. They remain closed during all other plant operating states, except cold shutdown. Therefore, there is no reason to provide them with any isolation signal other than remote manual.

38. The ADS supply lines are maintained at a minimum pressure of 160 psig at all times. Leakage in these lines is monitored by pressure instrumentation which alarms in the main control room on low pressure. Therefore, these lines are always under a continuous leak test, and a specific local leak rate test (Type C) will not be performed. The intent of the requirement is satisfied however, by the system design itself.

39. The ECCS and RCIC suction lines are no rmally filled with water on both the inboard and outboard side of containm ent, thereby forming a water seal to the containment environment. The valves are open during post-LOCA conditions to supply a water source for the ECCS pumps. Since a break in an ECCS line need not be considered in conjunction with a DBA, the only possible situation requiring one of these valves to be closed during a DBA is an unacceptable leakage in an ECCS. However, because these ECCS systems are constantly monitored for excessive leakage, this is not a credible event for design.

40. These valves are required to open and remain open following a LOCA to allow the containment air to be sampled. They are part of a system which constitutes a closed loop outside of the containment and will be open during Type A testing. Therefore there is no reason to perform a Type C test on these valves.

LSCS-UFSAR TABLE 6.2-21 SHEET 47 OF 49 REV. 14, APRIL 2002

41. The inboard flange of these butterfly valves has been provided with a double O-ring type gasket with a leakoff test connection provided between the O-rings. This permits the performance of a Type B leak rate test on this non-welded containment boundary, in addition to the Type C leak test on the valve seats.

42. These valves are capable of being manually overrided by applying jumpers to the isolation logic when a containment isolation signal is present, in order to

obtain reactor coolant sample at the High Radiation Sample System Panels under post-accident conditions.

43. These penetrations are provided with removable spools outboard of the outboard isolation valve. During operation these lines will be blind flanged using a double O-ring and Type B leak tested. In addition, the packing of these isolation valves will be soap-bubble tested to ensure insignificant or no leakage at containment test pressure.

44. These lines have been evaluated to an acceptable alternate design basis other than that specifically listed in GDC 57. This alternate basis is found in SRP 6.2.4.II.6.a.

45. High Radiation Detectors (1&2 RE-CM011 and 1&2 RE-CM017) have been installed in Containment Penetrations M-31 & M-32. These detectors are mounted in steel sleeves which protrude into the Primary Containment at diverse locations, so as to view a larger segment of the containment atmosphere, maintain accessibility for maintenance and calibration, and to minimize exposure during maintenance and calibration. The Containment Penetration is Seal Welded on the inside of the containment and Blind Flanged on the outside of the Containment.

46. These valves are provided with plugged Tees between the solenoid valve and the air cylinder for applying air pressure to the air cylinder using an air bellows hand pump for opening the valve, if instrument air is not available.

47. These valves have different closure time.

1E21-F012 Closure time - less than or equal to 40 seconds 2E21-F012 Closure time - sl ower than standard (see below)

48. These valves have a slower than standard stem speed, but operate faster than the Tech Spec requirement. The valves' stroke time has been evaluated

and is acceptable.

LSCS-UFSAR TABLE 6.2-21 SHEET 48 OF 49 REV. 14, APRIL 2002

49. In Test Mode 1 the RCIC System is aligned to take suction from the Condensate Storage Tank (CST) and the full flow test return line is aligned to the CST.

Valves E51-F362 and E51-F363 will become primary containment isolation valves. In Test Mode 2 the RCIC System is aligned to take suction from the Suppression Pool (SP). Valves E51-F362 and E51-F363 will no longer be containment isolation valves. Valves E51-F022, and E51-F059 will become containment isolation valves and spectacle flange E51-D316 (blind side) will be a containment isolation boundary.

50. General Electric Specification 22A2817AK Rev. 6 states that the maximum operating time for valves 1(2)E12-F064 A/B/C is eight seconds. The intent is to insure that RHR pump minimum flow requirements are met. The downstream orifice becomes the limiting device before the valve fully opens. An evaluation (NTS 373-201-98-CAQ05833.00) concluded as long as the minimum flow valves pass the required minimum flow in 8 seconds or less, the GE specification

requirements are met.

51. These valves are subject to bonnet pressure locking. The reactor side valve discs have vent holes drilled in them to prevent pressure accumulation in the bonnet.

52. Exempt Change DCPs 9500254, 255, 256, and 257 change the Valve Closure time for the 1E12-F017B, 17A, 16B, and 16A valves from approximately 75 seconds to approximately 95 seconds. Exempt Change E01-2-94-934A, B, C and D change the Valve Closure time for the 2E12-F016A, B and 2E12-F017A, B valves from approximately 75 seconds to approximately 95 seconds. These are no longer in the standard operating time range for a motor operated gate valve.

53. Exempt Changes E01-1-94-433 and E01-2-94-939-E changed the valve closure times for the 1G33-F040 and 2G33-F040 valves, respectively, from approximately 21 seconds to 39 seconds. This is no longer in the standard operating time range for a motor operated gate valve.

54.

The stem packing of these inboard primary containment isolation valves (located outside primary containment) is not tested for leakage during Type C Local Leak Rate Testing. The packing itself is either local leak rate tested via test port or subjected to pressure and subsequent ly soap bubble tested during primary containment pressurization on a periodic basis in accordance with 10 CFR 50 Appendix J and the LaSalle Station Leak Rate Test Program.

55. The Vacuum Breaker line manual isolation valves have a double-gasketed flange on the inboard or containment side provided with test connections for leak testing. The outboard flanges on the manual isolation valves are leak tested by pressurizing the entire vacuum breaker line and performing a soap bubble test on the outboard flange. The stem seal or packing of these valves will be tested either locally or by primary containment pressurization and subsequent soap bubble inspection.

LSCS-UFSAR TABLE 6.2-21 SHEET 49 OF 49 REV. 17, APRIL 2008

56. This valve is subject to bonnet pressure locking. The non-containment side valve disc has a vent hole drilled in it to prevent pressure accumulation in the bonnet.

57. This valve is subject to bonnet pressure locking. The bonnet of this valve has a hole drilled in it discharging th rough piping and isolation valves to allow manual venting of the bonnet.

58. These lines have been evaluated to an acceptable alternative design basis other than that specifically listed in GDC 56 and SRP 6.2.4.II. NRC approval of this design is found in the LaSalle Safety Evaluation Report (SER), NUREG 0519 Section 22.2.II.E.4.2.

59. These valves are monitored by the IST/MOV program as implemented by Subsection ISTC of ASME OM Code 2001 Edition through 2003 Addenda, and Code Case OMN-1 "Alternative Rules for Pressure and Inservice Testing of Certain Electric Motor Operated Valve Assemblies in Light Water Reactor Power Plants".

60. Valves 1(2)E51-F064 have been replaced by spectacle flanges 1(2)E51-D324.

61. In response to Generic Letter 96-06, a hole exists in the inboard disc at the inboard containment isolation valve to prevent thermal over-pressurization of the penetration.

62. Penetration M-34 contains the Standby Liquid Control System Injection line.

The Standby Liquid Control System (SBLC) Line enters the reactor vessel below the core plate. Under post LOCA conditions, the reflooding capability of the jet pumps will always assure the core to be two-third s covered. This provides assurance that the SBLC line will always be water filled post-LOCA. Thus, the SBLC line is not a potential primary containment atmospheric pathway either during or following a Design Basis Accident (DBA). Type C testing is not required on boundaries that do not constitute potential primary containment atmospheric pathways during and following a DBA. Thus, it is not required to Type C test any of the containment isolation valves in that pathway.

The SBLC line including valves 1&2C41-F007 and 1&2C41-F004A,B will be hydrostatically tested on a periodic basis to insure their leak tight integrity and evaluated against the leakage requirements of Technical Specifications SR 3.6.1.3.11.

LSCS-UFSAR TABLE 6.2-22 (SHEET 1 OF 2) TABLE 6.2-22 REV. 14, APRIL 2002 PARAMETERS USED TO DETERMINE HYDROGEN CONCENTRATION

1.

Reactor power 3,559 MWt

2.

Number of assemblies 764

3.

Total Zr mass in active clad/assembly 101 lb

4. Zirconium clad mass 77,187 lb

5. Fraction of Zr clad reacted 0.945%

6. Drywell free volume 229,538 ft 3 7. Suppression chamber volume 165,100 ft 3 8. Drywell initial temperature 135° F

9. Drywell initial pressure 0.75 psig

10. Drywell initial relative humidity 20%

11. Suppression chamber initial temperature 105° F**

12. Suppression chamber initial pressure 0.75 psig

13. Suppression chamber initial relative humidity 100%

14.

Thermal recombiner capacity 125 scfm

LSCS-UFSAR TABLE 6.2-22 (SHEET 2 OF 2)

TABLE 6.2-22 REV. 13 15. The guidelines as set forth in Regulatory Guide 1.7 were followed:

a) 50% of the halogens and 1%

of the solids present in the core are intimately mixed with the coolant water.

b) 25% of the halogens plate out on surfaces in the containment.

c) All noble gases and 25% of the halogens are released from the core to the containment atmosphere.

d) All other fission products remain in the fuel rods.

e) G(H 2)*is 0.5 molecules/100eV

f) G(O 2)*is 0.25 molecules/100eV

g) The following percentage of fissio n product radiation energy is absorbed by the coolant:

Percentage Radiation Type Location of Source 0% Beta Fuel Rods 100% Beta Coolant 10% Gamma Fuel Rods 100% Gamma Coolant

• For water, borated water, and borated alkaline solutions. ** As discussed in Section 6.2.1.8 supplementary evaluations have been satisfactorily completed with a 105

°F initial suppression pool temperature. (Reference 14)

LSCS-UFSAR TABLE 6.2-23 TABLE 6.2-23 REV. 14, APRIL 2002 CONTAINMENT LEAKAGE TESTING LEAK RATES at Pa (%/24 hours)

TYPE OF TEST PER APPENDIX J OF 10 CFR 50 DESCRIPTION OF TEST CALCULATED PEAK PRESSURE Pa (psig) MAXIMUM ALLOWABLE (La) DESIGN (Ld) TEST PRESSURE Pt (psig)

A Integrated Leak Rate 39.9 0.635(3) 0.5 (6)

B Local Penetration Leakage Rate 39.9 (1) (1) (6)

C Local Containment Isolation Valve Leakage Rate 39.9 (1)(2) 0.1 SCFH per inch of

nominal valve size at 50 psig (6) - MSIV Leakage Rate 39.9 (5) 100 scfh 25 (4)

(1) The combined leakage rate of all penetrations and valves exclusive of MSIV leakage subject to Type B and C tests shall be less than 0.60 La, as specified in Appendix J to 10 CFR 50.

(2) See Table 6.2-21, Note 15.

(3) Exclusive of the MSIV leakage rates.

(4) Exemption of 10 CFR 50, as stated in III C.3 of Appendix J.

(5) The sum of all four main steam lines shall be less than 400 SCFH. Any MSIV exceeding the proposed limit will be repaired and retested to meet a leakage rate of less than 25 SCFH.

(6) Test pressure shall be, as a minimum, equal to Pa. Variance in test pressure shall be in accordance with ANSI/ANS 56.8-1994.

LSCS-UFSAR TABLE 6.2-24 (SHEET 1 OF 2) TABLE 6.2-24 REV. 0 SUBCOMPARTMENT VENT PATH DESCRIPTION RECIRCULATION OUTLET LINE BREAK WITH SHIELDING DOORS HEAD LOSS, K VENT PATH NO. FROM VOL. NODE NO. TO VOL. NODE NO. DESCRIPTION OF VENT PATH FLOW AREA*

(ft 2) LENGTH (ft) (L/A) (ft

-1) HYDRAULIC DIAMETER (ft) FRICTION LOSS, K f TURNIN G LOSS, K bl EXPANSION AND CONTRACTION, K g TOTAL 1 1 2 unchoked 14.86 5.98 0.40 4.05 - 0.10 0.14 0.24 2 2 3 unchoked 14.86 5.98 0.40 4.05 - 0.10 0.14 0.24 3 3 4 unchoked 14.86 7.48 0.50 4.05 - 0.12 0.28 0.40 4 4 5 unchoked 14.86 8.97 0.60 4.05 - 0.14 0.28 0.42 5 6 7 choked 20.19 6.04 0.30 4.40 - 0.06 0.16 0.22 6 7 8 choked 20.19 6.04 0.30 4.40 - 0.06 0.16 0.22 7 8 9 choked 20.19 7.55 0.38 4.40 - 0.07 0.32 0.39 8 9 10 unchoked 20.19 9.06 0.45 4.40 - 0.09 0.32 0.41 9 35 34 choked 7.04 2.50 0.30 2.42 - 0.85 0.00 0.85 10 34 11 choked 10.02 3.19 0.32 2.95 - 0.03 0.32 0.35 11 11 12 choked 7.47 4.78 0.64 2.70 - 0.56 0.00 0.56 12 12 13 choked 7.09 6.37 0.90 2.70 - 0.52 0.32 0.84 13 13 14 unchoked 7.09 7.96 1.13 2.70 - 0.53 0.32 0.85 14 14 15 unchoked 7.09 9.55 1.35 2.70 - 1.00 0.64 1.64 15 11 17 choked 2.11 4.78 2.26 2.70 - 0.05 0.00 0.05 16 16 17 choked 3.87 6.37 1.46 2.20 - 0.07 0.31 0.38 17 17 18 unchoked 6.79 6.37 0.94 2.70 - 0.52 0.31 0.83 18 18 19 unchoked 6.79 7.96 1.17 2.70 - 0.54 0.31 0.85 19 19 20 unchoked 6.79 9.55 1.41 2.70 - 1.01 0.62 1.63 20 21 22 unchoked 9.83 6.35 0.65 3.00 - 0.06 0.30 0.36 21 22 23 choked 9.83 6.35 0.65 3.00 - 0.06 0.30 0.36 22 23 24 unchoked 9.83 7.93 0.81 3.00 - 0.07 0.60 0.67 23 24 25 unchoked 9.83 9.52 0.97 3.00 - 0.08 0.60 0.68 24 26 27 unchoked 14.68 9.52 0.65 3.25 - 0.98 0.30 1.28 25 27 28 unchoked 14.68 9.52 0.65 3.25 - 0.08 0.60 0.68 26 28 29 unchoked 14.68 9.52 0.65 3.25 - 0.98 0.30 1.28 27 30 31 unchoked 13.49 9.52 0.71 3.20 - 0.97 0.30 1.27 28 31 32 unchoked 13.49 9.52 0.71 3.20 - 0.53 0.60 1.13 29 32 33 unchoked 13.49 9.52 0.71 3.20 - 0.97 0.30 1.27 30 6 1 unchoked 18.40 6.27 0.33 5.80 0.03 0.00 0.00 0.03, 0.03** 31 7 2 unchoked 18.40 6.27 0.33 5.80 0.03 0.00 0.00 0.03, 0.03** 32 8 3 unchoked 18.40 6.27 0.33 5.80 0.03 0.00 0.00 0.03, 0.03** 33 9 4 unchoked 23.36 6.27 0.22 5.80 0.03 0.00 0.00 0.03, 0.03** 34 10 5 unchoked 23.36 6.27 0.22 5.80 0.03 0.00 0.00 0.03, 0.03** 35 34 6 choked 3.61 7.20 1.40 3.70 0.01 0.00 1.12 1.13, 0.90** 36 11 6 choked 3.61 7.20 1.40 3.70 0.01 0.00 1.12 1.13, 0.90** 37 12 7 unchoked 7.22 6.19 0.62 3.70 0.01 0.00 1.12 1.13, 0.90** 38 13 8 unchoked 7.22 6.19 0.62 3.70 0.01 0.27 1.12 1.40, 1.17** 39 14 9 unchoked 10.84 6.19 0.41 3.70 0.01 0.00 1.12 1.13, 0.90** 40 15 10 unchoked 10.84 6.19 0.41 3.70 0.01 0.00 1.12 1.13, 0.90** 41 12 17 unchoked 8.56 4.80 0.56 3.70 0.01 0.45 0.00 0.46 42 13 18 unchoked 8.56 4.80 0.56 3.70 0.01 0.45 0.00 0.46 LSCS-UFSAR TABLE 6.2-24 (SHEET 2 OF 2)

TABLE 6.2-24 REV. 0 HEAD LOSS, K VENT PATH NO. FROM VOL.

NODE NO. TO VOL.

NODE NO. DESCRIPTION OF VENT PATH FLOW AREA* (ft 2) LENGTH (ft) (L/A) (ft

-1) HYDRAULIC DIAMETER (ft) FRICTION LOSS, K f TURNING LOSS, K bl EXPANSION

AND CONTRACTION, K g TOTAL 43 14 19 unchoked 12.84 4.80 0.37 3.70 0.01 0.45 0.00 0.46 44 15 20 unchoked 11.65 4.80 0.41 3.70 0.01 0.43 0.00 0.44 45 34 16 choked 5.94 4.80 0.94 3.70 0.03 0.00 0.00 0.03 46 11 16 unchoked 5.94 4.80 0.94 3.70 0.03 0.85 0.00 0.88 47 16 21 choked 7.72 4.54 0.44 3.70 0.01 0.00 0.66 0.67 48 17 22 choked 7.72 5.55 0.59 3.70 0.02 0.00 0.66 0.68 49 18 23 unchoked 7.72 5.55 0.59 3.70 0.02 0.00 0.66 0.68 50 19 24 unchoked 11.57 5.55 0.40 3.70 0.02 0.00 0.66 0.68 51 20 25 unchoked 11.57 5.50 0.40 3.70 0.02 0.00 0.66 0.68 52 21 26 choked 7.72 8.00 0.80 3.90 0.03 0.27 0.66 0.96 53 22 26 choked 3.86 8.00 1.60 3.90 0.03 0.35 0.66 1.04 54 22 27 choked 3.86 8.00 1.60 3.90 0.03 0.35 0.66 1.04 55 23 27 choked 7.72 8.00 0.80 3.90 0.03 0.00 0.66 0.69 56 24 28 unchoked 11.57 8.00 0.54 3.90 0.03 0.27 0.66 0.96 57 25 29 unchoked 11.57 8.00 0.54 3.90 0.03 0.28 0.66 0.97 58 26 30 choked 11.57 9.20 0.60 3.90 0.03 0.31 0.66 1.00 59 27 31 choked 11.57 9.20 0.60 3.90 0.03 0.35 0.66 1.04 60 28 32 choked 11.57 9.20 0.60 3.90 0.03 0.28 0.66 0.97 61 29 33 choked 11.57 9.20 0.60 3.90 0.03 0.31 0.66 1.00 62 30 36 choked 9.27 - 1.05 - 0.01 0.00 0.74 0.75 63 31 36 choked 13.90 - 0.70 - 0.02 0.00 1.67 1.69 64 32 36 choked 13.90 - 0.70 - 0.02 0.00 1.67 1.69 65 33 36 choked 9.27 - 1.05 - 0.01 0.00 0.74 0.75 66 33 36 choked 2.04 - 1.05 - - - - 1.72 67 32 36 choked 0.68 - 3.39 - - - - 1.71 68 31 36 choked 2.10 - 1.11 - - - - 1.71 69 30 36 choked 1.77 - 1.25 - - - - 1.72 70 36 37 unchoked 400. - 0.06 - - - - 0.05 71 29 37 choked 1.39 - 1.50 - - - - 1.73 72 28 37 choked 0.71 - 3.30 - - - - 1.71 73 27 37 choked 0.71 - 3.30 - - - - 1.71 74 26 37 choked 1.39 - 1.50 - - - - 1.71 75 37 38 unchoked 965. - 0.03 - - - - 0.05 76 20 38 choked 1.25 - 1.97 - - - - 1.71 77 19 38 choked 1.07 - 2.20 - - - - 1.71 78 18 38 choked 0.71 - 3.30 - - - - 1.71 79 17 38 choked 0.71 - 3.30 - - - - 1.71 80 15 38 choked 1.25 - 1.97 - - - - 1.71 81 14 38 choked 1.07 - 2.20 - - - - 1.71 82 13 38 choked 1.47 - 1.50 - - - - 1.71 83 12 38 choked 0.71 - 3.30 - - - - 1.71 84 11 38 choked 0.71 - 3.30 - - - - 1.71 85 35 38 choked 1.08 - 2.43 - - - - 1.71 86 0 35 choked 1.00 - 0.00 - - - - 0.00

• Minimum cross-sectional area. **Loss coefficient for reverse flow.

LSCS-UFSAR TABLE 6.2-25 (SHEET 1 OF 2)

TABLE 6.2-25 REV. 0 - APRIL 1984 SUBCOMPARTMENT VENT PATH DESCRIPTION FEEDWATER LINE BREAK WITH SHIELDING DOORS HEAD LOSS, K VENT PATH NO. FROM VOL. NODE NO. TO VOL. NODE NO. DESCRIPTION OF VENT PATH FLOW AREA* (ft 2) LENGTH (ft)

(L/A) (ft

-1) HYDRAULIC DIAMETER (ft)

FRICTION LOSS, K f TURNING LOSS, K bl EXPANSION AND CONTRACTION, K E

TOTAL 1 1 2 unchoked 14.86 8.97 0.60 4.05 - 0.15 0.14 0.29 2 2 3 unchoked 14.86 8.97 0.60 4.05 - 0.15 0.28 0.43 3 3 4 unchoked 14.86 8.97 0.60 4.05 - 0.15 0.14 0.29 4 5 6 unchoked 20.19 9.06 0.45 4.40 - 0.09 0.16 0.25 5 6 7 unchoked 20.19 9.06 0.45 4.40 - 0.09 0.32 0.41 6 7 8 unchoked 20.19 9.06 0.45 4.40 - 0.09 0.16 0.25 7 9 10 unchoked 13.88 9.55 0.69 3.10 - 1.00 0.31 1.31 8 10 11 unchoked 13.88 9.55 0.69 3.10 - 0.65 0.62 1.27 9 11 12 unchoked 13.88 9.55 0.69 3.10 - 1.00 0.31 1.31 10 13 14 unchoked 9.83 6.35 0.65 3.00 - 0.06 0.45 0.51 11 14 15 unchoked 9.83 6.35 0.65 3.00 - 0.06 0.45 0.51 12 15 16 unchoked 9.83 7.80 0.81 3.00 - 0.08 0.30 0.38 13 16 17 unchoked 9.83 9.52 0.97 3.00 - 0.09 0.30 0.39 14 18 19 unchoked 14.68 6.35 0.44 3.25 - 0.49 0.30 0.79 15 19 20 unchoked 14.68 6.35 0.44 3.25 - 0.53 0.30 0.83 16 20 21 unchoked 14.68 6.35 0.54 3.25 - 0.51 0.00 0.51 17 21 22 unchoked 14.68 6.35 0.65 3.25 - 0.55 0.30 0.85 18 29 23 choked 5.42 2.50 0.40 2.52 - 0.85 0.00 0.85 19 23 24 choked 16.19 3.17 0.20 3.20 - 0.03 0.30 0.33 20 24 25 choked 16.19 4.76 0.30 3.20 - 0.05 0.00 0.05 21 25 26 unchoked 16.19 6.35 0.40 3.20 - 0.73 0.60 1.33 22 26 27 unchoked 16.19 7.93 0.50 3.20 - 0.74 0.60 1.34 23 27 28 unchoked 16.19 9.52 0.60 3.20 - 0.09 0.30 0.39 24 5 1 unchoked 23.80 6.27 0.26 5.80 - 0.00 0.00 0.03 25 6 2 unchoked 23.80 6.27 0.26 5.80 - 0.00 0.00 0.03 26 7 3 unchoked 23.80 6.27 0.26 5.80 0.03 0.00 0.00 0.03 27 8 4 unchoked 23.80 6.27 0.26 5.80 0.03 0.00 0.00 0.03 28 9 5 unchoked 10.84 8.53 0.54 3.70 0.02 0.26 0.85 1.13, 1.28** 29 10 6 unchoked 10.84 8.53 0.54 3.70 0.02 0.26 0.85 1.13, 1.28** 30 11 7 unchoked 10.84 8.53 0.54 3.70 0.02 0.26 0.85 1.13, 1.28** 31 12 8 unchoked 10.84 8.53 0.54 3.70 0.02 0.26 0.85 1.13, 1.28**

LSCS-UFSAR TABLE 6.2-25 (SHEET 2 OF 2)

TABLE 6.2-25 REV. 0 - APRIL 1984 HEAD LOSS, K VENT PATH NO. FROM VOL. NODE NO. TO VOL. NODE NO. DESCRIPTION OF VENT PATH FLOW AREA* (ft 2) LENGTH (ft)

(L/A) (ft

-1) HYDRAULIC DIAMETER (ft)

FRICTION LOSS, K f TURNING LOSS, K bl EXPANSION AND CONTRACTION, K E

TOTAL 32 13 9 unchoked 7.22 8.00 0.83 3.70 0.02 0.31 0.63 0.96 33 14 9 unchoked 3.61 8.00 1.66 3.70 0.02 0.31 0.63 0.96 34 14 10 unchoked 3.61 8.00 1.66 3.70 0.02 0.31 0.63 0.96 35 15 10 unchoked 7.22 8.00 0.83 3.70 0.02 0.31 0.63 0.96 36 16 11 unchoked 10.84 8.00 0.56 3.70 0.02 0.31 0.63 0.96 37 17 12 unchoked 10.84 8.00 0.56 3.70 0.02 0.36 0.63 1.01 38 18 13 choked 7.71 8.00 0.80 3.90 0.02 0.00 0.66 0.68 39 19 14 choked 7.71 8.00 0.80 3.90 0.02 0.35 0.66 1.03 40 20 15 unchoked 7.71 8.00 0.80 3.90 0.02 0.28 0.66 0.96 41 21 16 unchoked 11.57 8.00 0.54 3.90 0.02 0.29 0.66 0.97 42 22 17 unchoked 11.57 8.00 0.54 3.90 0.02 0.28 0.66 0.96 43 23 18 choked 3.86 10.08 1.94 3.90 0.04 0.00 0.66 0.70 44 24 18 choked 3.96 10.08 1.94 3.90 0.04 0.00 0.66 0.70 45 25 19 choked 7.71 10.08 0.97 3.90 0.04 0.28 0.66 0.98 46 26 20 choked 7.71 10.08 0.97 3.90 0.04 0.30 0.66 1.00 47 27 21 unchoked 11.57 10.08 0.65 3.90 0.04 0.29 0.66 0.99 48 28 22 unchoked 11.57 10.08 0.65 3.90 0.04 0.27 0.66 0.97 49 23 30 choked 1.54 - 3.60 - 0.01 0.00 1.60 1.61 50 24 30 choked 3.86 - 1.30 - 0.02 0.00 1.05 1.07 51 25 30 choked 7.71 - 1.06 - 0.02 0.00 1.97 1.99 52 26 30 choked 7.71 - 1.06 - 0.02 0.00 1.97 1.99 53 27 30 unchoked 9.27 - 0.79 - 0.01 0.00 2.39 2.40 54 28 30 unchoked 11.57 - 0.65 - 0.02 0.00 1.80 1.82 55 29 30 choked 0.68 - 3.96 - - - - 1.71 56 28 30 choked 0.68 - 3.96 - - - - 1.71 57 27 30 unchoked 1.36 - 1.98 - - - - 1.71 58 26 30 unchoked 1.36 - 1.70 - - - - 1.73 59 25 30 unchoked 0.68 - 3.96 - - - - 1.71 60 30 31 unchoked 400. - 0.06 - - - - 0.05 61 22 31 choked 0.71 - 3.86 - - - - 1.71 62 21 31 unchoked 1.39 - 1.70 - - - - 1.73 63 20 31 unchoked 0.68 - 2.98 - - - - 1.74 64 19 31 unchoked 1.42 - 1.93 - - - - 1.71 65 31 32 unchoked 965. - 0.03 - - - - 0.05 66 12 32 choked 2.89 - 0.90 - - - - 1.71 67 11 32 choked 2.50 - 1.17 - - - - 1.71 68 10 32 unchoked 2.50 - 1.17 - - - - 1.71 69 9 32 unchoked 2.14 - 1.29 - - - - 1.71 70 0 32 choked 1.0 - 0.0 - - - - 0.0

• Minimum cross-sectional area. ** Loss coefficient for reverse flow.

LSCS-UFSAR TABLE 6.2-26 TABLE 6.2-26 REV. 0 - APRIL 1984 MASS AND ENERGY RELEASE RATE DATA RECIRCULATION OUTLET LINE BREAK (For Biological Shield Pressurization Analysis)

BREAK AREA 2.753 ft 2 TIME (sec)

LIQUID MASS FLOW RATE (lb m/sec) STEAM MASS FLOW RATE (lb m/sec) LIQUID ENTHALPY (Btu/lb m) STEAM ENTHALPY (Btu/lb m) TOTAL MASS RELEASE RATE (lb m/sec) TOTAL ENERGY RELEASE RATE (Btu/sec) 0.0 0. 0. 527.4 1195.9 0. 0.

0.0020 742. 0. 527.4 1195.9 742. 3.92 x 10 5 0.0040 2388. 0. 527.4 1195.9 2388. 1.26 x l0 6 0.0060 4958. 0. 527.4 1195.9 4958. 2.62 x 10 6 0.0080 8926. 0. 527.4 1195.9 8926. 4.71 x l0 6 0.0100 14162. 0. 527.4 1195.9 14162. 7.47 x 10 6 0.0173 36184. 0. 527.4 1195.9 36184. l.91 x 10 6 0.0194 36184. 0. 527.4 1195.9 36184. 1.91 x 10 7 0.0194 18324. 0. 527.4 1195.9 18324. 9.67 x 10 6 0.0220 21146. 0. 527.4 1195.9 21146. 1.12 x 10 7 0.0240 22890. 0. 527.4 1195.9 22890. 1.21 x 10 7 0.0260 24294. 0. 527.4 1195.9 24294. l.28 x 10 7 0.0280 25222. 0. 527.4 1195.9 25222. 1.33 x 10 7 0.0300 25730. 0. 527.4 1195.9 25730. 1.36 x 10 7 0.0310 25770. 0. 527.4 1195.9 25770. 1.36 x 10 7 5.0 25770. 0. 527.4 1195.9 25770. 1.36 x 10 7 LSCS-UFSAR TABLE 6.2-27 TABLE 6.2-27 REV. 0 - APRIL 1984 MASS AND ENERGY RELEASE RATE DATA FEEDWATER LINE BREAK (For biological shield pressurization analysis)

BREAK AREA 1.538 ft TIME (sec)

LIQUID MASS FLOW RATE (lb m/sec) STEAM MASS FLOW RATE (lb m/sec) LIQUID ENTHALPY (Btu/lb m) STEAM ENTHALPY (Btu/lb m) TOTAL MASS RELEASE RATE (lb m/sec) TOTAL ENERGY RELEASE RATE (Btu/sec) 0.0 14,197. 0. 397.8 1190. 14,197. 5.65 x 10 6 0.00105 14,197. 0. 397.8 1190. 14,197. 5.65 x 10 6 0.00106 21,599. 0. 397.8 1190. 21,599. 8.60 x 10 6 1.0 21,599. 0. 397.8 1190. 21,599. 8.60 x 10 6 LSCS-UFSAR TABLE 6.2-28 (SHEET 1 OF 8) PRIMARY CONTAINMENT ISOLATION VALVES TABLE 6.2-28 REV. 14, APRIL 2002 VALVE FUNCTION AND NUMBER VALVE GROUP(a) MAXIMUM ISOLATION TIME (Seconds) A. AUTOMATIC ISOLATION VALVES 1. Main Steam Is olation Valves 1(2)B21-F022A, B, C, D 1(2)B21-F028A, B, C, D 1 5* 2. Main Steam Line Drain Valves 1(2)B21-F016 1(2)B21-F019 1(2)B21-F067A, B, C, D 1 15 15 23 3. Reactor Coolant System Sample Line Valves (b) 1(2)B33-F019 1(2)B33-F020 3 5 4. Drywell Equipment Drain Valves 1(2)RE024 1(2)RE025 1(2)RE026 1(2)RE029 2 20 20 15 15 5. Drywell Floor Drain Valves 1(2)RF012 1(2)RF013 2 20 6. Reactor Water Cleanup Suction Valves 1(2)G33-F001(c) 1(2)G33-F004 5 10 7. RCIC Steam Line Valves 1(2)E51-F008(d) 1(2)E51-F063 1(2)E51-F076 8 20 15 15 8. Containment Vent and Purge Valves 1(2)VQ026 1(2)VQ027 1(2)VQ029 1(2)VQ030 1(2)VQ031 1(2)VQ032 1(2)VQ034 1(2)VQ035 1(2)VQ036 1(2)VQ040 1(2)VQ042 1(2)VQ043 1(2)VQ047 1(2)VQ048 1(2)VQ050 1(2)VQ051 1(2)VQ068 4 10 10 10 10 10 5 10 5 10 10 10 10 5 5 5 5 5 9. RCIC Turbine Exhaust Vacuum Breaker Line Valves 1(2)E51-F080 1(2)E51-F086 9 N/A LSCS-UFSAR TABLE 6.2-28 (SHEET 2 OF 8) PRIMARY CONTAINMENT ISOLATION VALVES TABLE 6.2-28 REV. 14, APRIL 2002 VALVE FUNCTION AND NUMBER VALVE GROUP(a) MAXIMUM ISOLATION TIME (Seconds) A. AUTOMATIC ISOLATION VALVES (CONTINUED) 10. Containment Monitoring Valves 2 5 1(2)CM017A,B 1(2)CM0l8A,B 1(2)CM019A,B 1(2)CM020A,B 1(2)CM021B (f ) 1(2)CM022A (f) 1(2)CM025A (f) 1(2)CM026B(f) 1(2)CM027 1(2)CM028 1(2)CM029 1(2)CM030 1(2)CM031 1(2)CM032 1(2)CM033 1(2)CM034

11. Drywell Pneumatic Valves 1(2)IN001A and B 1(2)IN017 1(2)IN074 1(2)IN075 1(2)IN031 10 10 10 10 2 30 22 22 22 5 12. RHR Shutdown Cooling Mode Valves 1(2)E12-F008 1(2)E12-F009 1(2)E12-F023 1(2)E12-F053A and B 6 40 40 90 29 13. Tip Guide Tube Ball Valves (Five Valves) 1(2)C51-J004 7 N/A 14. Reactor Building Closed Cooling Water System Valves 1(2)WR029 1(2)WR040 1(2)WR179 1(2)WR180 2 30 15. Primary Containment Chilled Water Inlet Valves 1(2)VP113A and B 1(2)VP063A and B 2 90 40 16. Primary Containment Chilled Water Outlet Valves 1(2)VP053A and B 1(2)VP114A and B 2 40 90 LSCS-UFSAR TABLE 6.2-28 (SHEET 3 OF 8) PRIMARY CONTAINMENT ISOLATION VALVES TABLE 6.2-28 REV. 13 88 VALVE FUNCTION AND NUMBER VALVE GROUP(a) MAXIMUM ISOLATION TIME (Seconds) A. AUTOMATIC ISOLATION VALVES (CONTINUED) 17. Recirc. Hydraulic Flow Control Line Valves 1(2)B33-F338 A and B 1(2)B33-F339 A and B 1(2)B33-F340 A and B 1(2)B33-F341 A and B 1(2)B33-F342 A and B 1(2)B33-F343 A and B 1(2)B33-F344 A and B 1(2)B33-F345 A and B 2 5 18. Feedwater Testable Check Valves 1(2)B21-F032 A and B 2 N/A B. MANUAL ISOLATION VALVES

1. 1(2)FC086 N/A 2. 1(2)FC113 N/A 3. 1(2)FC114 N/A 4. 1(2)FC115 N/A 5. 1(2)MC027 (h) N/A 6. 1(2)MC033 (h) N/A 7. 1(2)SA042 (h) N/A 8. 1(2)SA046 (h) N/A 9. 1(2)CM039 N/A 10. 1(2)CM040 N/A 11. 1(2)CM041 N/A 12. 1(2)CM042 N/A 13. 1(2)CM043 N/A 14. 1(2)CM044 N/A 15. 1(2)CM045 N/A 16. 1(2)CM046 N/A 17. 1(2)CM085 N/A 18. 1(2)CM086 N/A 19. 1(2)CM089 N/A 20. 1(2)CM090 N/A C. EXCESS FLOW CHECK VALVES

1. 1(2)B21-F374

2. 1(2)B21-F376

3. 1(2)B21-F359

4. 1(2)B21-F355

5. 1(2)B21-F361

6. 1(2)B21-F378

7. 1(2)B21-F372

8. 1(2)B21-F370

9. 1(2)B21-F363

10. 1(2)B21-F353

11. 1(2)B21-F415A, B

12. 1(2)B21-F357

LSCS-UFSAR TABLE 6.2-28 (SHEET 4 OF 8) PRIMARY CONTAINMENT ISOLATION VALVES TABLE 6.2-28 REV. 13 VALVE FUNCTION AND NUMBER VALVE GROUP(a) MAXIMUM ISOLATION TIME (Seconds) C. EXCESS FLOW CHECK VALVES (CONTINUED) 13. 1(2)B21-F382

14. 1(2)B21-F328A, B, C, D

15. 1(2)B21-F327A, B, C, D

16. 1(2)B21-F413A, B

17. 1(2)B21-F344

18. 1(2)B21-F365

19. 1(2)B21-F443

20. 1(2)B21-F439

21. 1(2)B21-F437

22. 1(2)B21-F441

23. 1(2)B21-F445A, B

24. 1(2)B21-F453

25. 1(2)B21-F447

26. 1(2)B21-F455A, B

27. 1(2)B21-F451

28. 1(2)B21-F449

29. 1(2)B21-F367

30. 1(2)B21-F326A, B, C, D

31. 1(2)B21-F325A, B, C, D

32. 1(2)B21-F350

33. 1(2)B21-F346

34. 1(2)B21-F348

35. 1(2)B21-F471

36. 1(2)B21-F473

37. 1(2)B21-F469

38. 1(2)B21-F475A, B

39. 1(2)B21-F465A, B

40. 1(2)B21-F467

41. 1(2)B21-F463

42. 1(2)B21-F380

43. 1(2)G33-F312A, B

44. 1(2)G33-F309

45. 1(2)E12-F315

46. 1(2)E12-F359A, B

47. 1(2)E12-F319

48. 1(2)E12-F317

49. 1(2)E12-F360A, B

50. 1(2)E21-F304

51. 1(2)E22-F304

52. 1(2)E22-F341

LSCS-UFSAR TABLE 6.2-28 (SHEET 5 OF 8) PRIMARY CONTAINMENT ISOLATION VALVES TABLE 6.2-28 REV. 13 VALVE FUNCTION AND NUMBER VALVE GROUP(a) MAXIMUM ISOLATION TIME (Seconds) C. EXCESS FLOW CHECK VALVES (CONTINUED) 53. 1(2)E22-F342 54. 1(2)B33-F319A, B 55. 1(2)B33-F317A, B 56. 1(2)B33-F313A, B, C, D

57. 1(2)B33-F311A, B, C, D

58. 1(2)B33-F315A, B, C, D

59. 1(2)B33-F301A, B 60. 1(2)B33-F307A, B, C, D

61. 1(2)B33-F305A, B, C, D

62. 1(2)CM004 63. 1(2)CM002 64. 1(2)CM012 65. 1(2)CM010 66. 1(2)VQ061 67. 1(2)B21-F457 68. 1(2)B21-F459 69. 1(2)B21-F461 70. 1(2)CM102 71. 1(2)B21-F570 72. 1(2)B21-F571 D. OTHER ISOLATION VALVES 1. Deleted

2. Reactor Feedwater and RWCU System Return 1(2)B21-F010A, B 1(2)B21-F065A, B 1(2)G33-F040

3. Residual Heat Removal/Low Pressure Coolant Injection System 1(2)E12-F042A, B, C 1(2)E12-F016A, B 1(2)E12-F017A, B 1(2)E12-F004A, B, C 1(2)E12-F027A, B 1(2)E12-F024A, B 1(2)E12-F021 1(2)E12-F302 1(2)E12-F064A, B, C 1(2)E12-F011A, B 1(2)E12-F088A, B, C 1(2)E12-F025A, B, C 1(2)E12-F030 1(2)E12-F005

LSCS-UFSAR TABLE 6.2-28 (SHEET 6 OF 8) PRIMARY CONTAINMENT ISOLATION VALVES TABLE 6.2-28 REV. 15, APRIL 2004 VALVE FUNCTION AND NUMBER VALVE GROUP (a) MAXIMUM ISOLATION TIME (Seconds) D. OTHER ISOLATION VALVES (CONTINUED) 3. Residual Heat Removal/Low Pressure Coolant Injection System (Continued) 1(2)E12-F073A, B 1(2)E12-F074A, B 1(2)E12-F055A, B 1(2)E12-F036A, B 1(2)E12-F311A, B

4. Low Pressure Core Spray System 1(2)E21-F005 1(2)E21-F001 1(2)E21-F012 1(2)E21-F011 1(2)E21-F018 1(2)E21-F031

5. High Pressure Core Spray System 1(2)E22-F004 1(2)E22-F015 1(2)E22-F023 1(2)E22-F012 1(2)E22-F014

6. Reactor Core Isolation Cooling System 1(2)E51-F013 1(2)E51-F069 1(2)E51-F028 1(2)E51-F068 1(2)E51-F040 1(2)E51-F031 1(2)E51-F019 1(2)E51-F059(i) 1(2)E51-F022(i) 1(2)E51-F362(j) 1(2)E51-F363(j)

7. Post LOCA Hydrogen Control 1(2)HG001A, B 1(2)HG002A, B 1(2)HG005A, B 1(2)HG006A, B

LSCS-UFSAR TABLE 6.2-28 (SHEET 7 OF 8) PRIMARY CONTAINMENT ISOLATION VALVES

TABLE 6.2-28 REV. 15, APRIL 2004 VALVE FUNCTION AND NUMBER VALVE GROUP (a) MAXIMUM ISOLATION TIME (Seconds) D. OTHER ISOLATION VALVES (CONTINUED) 8. Standby Liquid Control System 1(2)C41-F004A, B 1(2)C41-F006 1(2)C41-F007

9. Reactor Recirculation Seal Injection 1(2)B33-F013A, B 1(2)B33-F017A, B

10. Drywell Pneumatic System 1(2)IN018 1(2)IN100 1(2)IN101

11. Reference Leg Backfill 1(2)C11-F422B 1(2)C11-F422D 1(2)C11-F422F 1(2)C11-F422G 1(2)C11-F423B 1(2)C11-F423D 1(2)C11-F423F 1(2)C11-F423G

12. Control Rod Drive Insert Lines 1(2)C11-D001-120 1(2)C11-D001-123

13. Control Rod Drive Withdrawal Lines 1(2)C11-D001-121 1(2)C11-D001-122

14. RHR Shutdown Cooling 1(2)E12-F460

15. Reactor Coolant System Sample Line Valve 1(2)B33-F395

16. Reactor Building Closed Cooling Water 1(2)WR225/226

17. Primary Containment Chilled Water Inlet Valve 1(2)VP198A/B

18. Primary Containment Chilled Water Outlet Valve 1(2)VP197A/B

19. Containment Monitoring System 1(2)CM023B 1(2)CM024A

• But 3 seconds. a) See Technical Specification for isolation signal(s) that operates each valve group. b) May be opened on an intermittent basis under administrative control. c) Not closed by SLCS actuation. d) Deleted.

LSCS-UFSAR TABLE 6.2-28 (SHEET 8 OF 8) PRIMARY CONTAINMENT ISOLATION VALVES TABLE 6.2-28 REV. 13 e) Not closed by Trip Functions 4a, c, d, e or f of Technical Specification 3.3.2, Table 3.3.2-1. f) Opens on an isolation signal.

g) Also closed by drywell pressure-high signal h) These penetrations are provided with removable spools outboard of the outboard isolation valve. During operation, these lines will be blind flanged using a double O-ring. i) If valves 1(2)E51-F362 and 1(2)E51-F363 are lock ed closed and acceptably leak rate tested, then valves 1(2)E51-F059 and 1(2)E51-F022 are not considered to be primary containment isolation valves and are not required to be leak rate tested. j) Either the 1(2)E51-F362 or the 1(2)E51-F363 valve may be open when the RCIC system is in the standby mode of operation, and both valves may be open during operation of the RCIC system in the full flow test mode, providing that:

(1) valve 1(2)E51-F022 is acceptably leak rate tested, and (2) valve 1(2)E51-F059 is deactivated, locked closed and acceptably leak rate tested, and (3) the spectacle flange, installed immediately downstream of the 1(2)E51-F059 valve, is closed and acceptably leak rate tested.

LSCS-UFSAR 6.3-1 REV. 13 6.3 EMERGENCY CORE COOLING SYSTEMS 6.3.1 Design Bases

The objective of the emergency core coolin g systems (ECCS), in conjunction with the containment, is to limit the release of radioactive materials following a loss-of-coolant accident so that resulting radiation exposures are within the guideline values given in published regulations.

Safety design bases for the emergency core cooling systems are given in the following subsections.

6.3.1.1 Summary Description of the Emergency Core Cooling System The emergency core cooling system (ECCS) consists of a high-pressure core spray (HPCS) system, a low-pressure core spray (LPCS) system, a low-pressure coolant injection (LPCI) system, and an automatic depressurization system (ADS).

The HPCS consists of a single, motor-driven pump and associated piping, valves, controls and instrumentation. The system is designed to pump water over the entire range of operating pressures, and thus can spray water into the reactor vessel even if the reactor pressure remains near normal operating levels. For small breaks which do not result in rapid vessel depressurization, the HPCS maintains the proper reactor water level and depressurizes the vessel.

The HPCS sprays the top surface of the core until sufficient water accumulates in the vessel to reflood the core. Water is in jected into the vessel through nozzles in a circular sparger above and around the periphery of the core.

The LPCS is a loop similar to, but independent of, the HPCS. The low pressure system is designed to provide protection in case of larger breaks which would rapidly depressurize the reactor vessel. Like the HPCS, water from the LPCS enters the vessel through nozzles in a circu lar sparger located above and around the core periphery. The LPCS limits the maximu m cladding temperature and cools it to saturation upon flooding the core. This system acts to protect the core for intermediate and large breaks, and is a ssisted by the HPCS and ADS for small breaks. The LPCI is capable of delivering a large flood of water into the core to refill the vessel once it depressurizes. It consists of three residual heat removal subsystem pumps, each of which injects water into the vessel through its own separate piping and penetrations. The function of this system is to cool the core by flooding, thereby cooling the cladding to saturation after a LOCA. The LPCI acts to protect the core for intermediate or large breaks, and is assisted by the HPCS and ADS for small breaks.

LSCS-UFSAR 6.3-2 REV. 15, APRIL 2004 Because the spraying and flooding systems can draw water from the suppression pool, they have a continuous supply of water. Water and steam from the vessel which would be lost through a postulated pipe break are collected in the suppression pool. Likewise, water pumped by the ECCS and lost through a break would also accumulate in the suppression pool.

The ADS utilizes 7 of the 13 safety/relief va lves (Unit 2 has a total of 13 valves). These are activated as a backup to the HPCS to reduce vessel pressure in case of breaks for which depressurization is required, so that flow from the LPCI and LPCS can enter the vessel in time to cool th e core and limit cladding temperature.

6.3.1.1.1 Range of Cool ant Ruptures and Leaks The emergency core cooling systems provide adequate core cooling in the event of any size break or leak in the nuclear system process barrier up to and including the design-basis break and the double-ended recirculation line break.

6.3.1.1.2 Fission Product Decay Heat In the event of a loss-of-coolant accide nt, the emergency core cooling systems remove both residual stored heat and radioactive decay heat from the reactor core at a rate that limits the maximum fuel cl adding temperature to a value less than the 10 CFR 50 limit of acceptability of 2200

° F. The amount of heat to be removed is discussed in Section 6.2.

6.3.1.1.3 Reactivity Required for Cold Shutdown The reactor is designed to be in the cold shutdown condition with the control rod of highest reactivity worth fully withdrawn and all other control rods fully inserted.

Refer to Subsection 4.3.2 for a complete discussion.

6.3.1.2 Functional Requirement Design Bases

a. Emergency core cooling systems are provided with sufficient capacity, diversity, reliability, and redundancy to cool the reactor core under all design-basis accident conditions.

b. Emergency core cooling systems are initiated automatically by conditions that sense the potential inadequacy of the normal core cooling.

c. The emergency core cooling sy stems are capable of startup and operation regardless of the avail ability of offsite power supplies and the normal generating system of the plant.

LSCS-UFSAR 6.3-3 REV. 13

d. Action taken to effect containment integrity does not negate the ability to achieve core cooling. All ECCS pumps are designed to operate without benefit of containment back pressure.

e. The components of the emergency core cooling systems within the reactor vessel are designed to withstand the transient mechanical loadings during a loss-of-coolant accident so that the required core cooling flow is not restricted.

f. The equipment of the emergency core cooling systems can withstand the physical effects of a loss-of-coolant accident so that the core can be effectively cooled. Such effects considered are missiles, fluid jets, pipe whip, high temperature, pressure, humidity, and seismic acceleration.

g. To provide a reliable supply of water for the emergency core cooling systems, the prime source of liquid for cooling the reactor core after a loss-of-coolant accident is a stored source located within the containment. The source is located so that a closed cooling water path is established during emergency core cooling systems operation.

6.3.1.3 Reliability Requirements Design Bases

The flow rate and sensing networks of each emergency core cooling system are testable during reactor shutdown. All active components are testable during normal operation of the nuclear system.

6.3.2 System Design

The ECCS, containing four separate subsyste ms, is designed to satisfy the following performance objectives:

a. to prevent fuel cladding fragmentation for any mechanical failure of the nuclear boiler system up to, and including, a break equivalent to the largest nuclear boiler system pipe;

b. to provide this protection by at least two independent, automatically actuated cooling systems;

c. to function with or without external (offsite) power sources; and

d. to permit testing of all ECCS by acceptable methods including, wherever practical, testing during power plant operations.

LSCS-UFSAR 6.3-4 REV. 14, APRIL 2002 The aggregate of these emergency core cooling systems is designed to protect the reactor core against fuel cladding dama ge (fragmentation) across the entire spectrum of line break accidents.

The power for operation of the ECCS is from regular a-c power sources. Upon loss of the regular power, operation is from onsite standby a-c power sources. Standby sources have sufficient diversity and capacity so that all ECCS requirements are satisfied. The HPCS is powered from one a-c supply bus. The LPCS and one LPCI are powered from a second a-c supply bus and the two remaining LPCI are powered from a third and separate a-c supply bus. The HPCS has its own diesel generator as its alternate power supply. The LPCS and one LPCI loops switch to one site backup power supply and the other two LPCI loops switch to a second site backup power supply.

All systems start automatically. The star ting signal comes from at least two independent and redundant sensors of drywell pressure and low reactor vessel water level. Refer to Subsection 7.3.1.

2 for a complete discussion of the ECCS instrumentation and starting and control logic.

Further discussion of the integrated performance of the ECCS is presented in Subsection 6.3.3.7. The bounds within which system parameters must be maintained and the acceptable inoperable components are discussed in the Technical Specifications.

6.3.2.1 Schematic Piping and Instrumentation Diagrams Piping and instrumentation diagrams fo r the subsystems and components which constitute the ECCS are provided and are re ferenced under the discussion of that subsystem or component.

6.3.2.2 Equipment and Component Descriptions

6.3.2.2.1 High-Pressure Core Spray (HPCS) System The high-pressure core spray (HPCS) system consists of a single motor-driven pump located outside the primary containment and associated system piping, valves, controls and instrumentation. The system is designed to operate from normal offsite auxiliary power or from a standby diesel-generator supply if offsite power is not available. The piping and instrumentation diagram (P&ID) for the HPCS is shown in Drawing Nos. M-95 and M-141. The HPCS system process diagram is shown in Figure 6.3-1.

The principal HPCS equipment is located outside the primary containment. Suction piping is provided from the suppression pool. The suppression pool water LSCS-UFSAR 6.3-5 REV. 13 source assures a closed cooling water supply for extended operation of the HPCS system. After the HPCS injection piping enters the vessel, it divides and enters the shroud at two points near the top of the shroud. A semicircular sparger is attached to each outlet. Nozzles are spaced around the spargers to spray the water radially over the core and into the fuel assemblies. The HPCS in jection piping is provided with an isolation valve on each side of the containment barrier. Remote controls for operating the valves and diesel generator are provided in the plant control room.

The controls and instrumentation of the HPCS system are described, illustrated, and evaluated in detail in Chapter 7.0.

The HPCS system is designed to cool the reactor core sufficiently to prevent fuel cladding temperatures from exceeding the 10 CFR 50 limit of 2200

° F following any break in the nuclear system piping. The system is designed to pump water into the reactor vessel over a wide range of pressures.

For small breaks that do not result in rapid reactor depressurization, the syst em maintains reactor water level and depressurizes the vessel. For large breaks the HPCS system cools the core by a spray. If a loss-of-coolant accident should occur, a low water level signal or a high drywell pressure signal initiates a reactor scram, the HPCS and its support equipment. The HPCS flow automatically stops when a high water level in the reactor vessel is signaled. The HPCS system also serves as a backup to the RCIC system in the event the reactor becomes isolated from the main condenser during operation and feedwater flow is lost.

If normal auxiliary power is not available, the HPCS pump motor is driven by its own onsite power source. The HPCS standby power source is discussed in Section 8.3.

The HPCS system vessel pressure versus flow characteristic assumed in LOCA analyses is shown in Figure 6.3-2.

Figure 6.3-10 shows the minimum required pump head for HPCS system in order to meet the LOCA analyses assumptions. When the system is started, initial flow rate is established by primary system pressure. As vessel pressure decreases, flow will increase. When vessel pressure reaches 200 psid (differential pressure between the reactor vessel and the suction source) the system reaches rated core spray flow. The HPCS motor size is based on peak horsepower requirements.

The elevation of the HPCS pump is below the water level of the suppression pool. This assures a flooded pump suction. Pu mp NPSH requirements are met even with the containment at atmospheric pressure by providing adequate suction head and suction line size. The HPCS pump characteristics, head, flow, horsepower, and required NPSH are shown in Figure 6.3-3.

LSCS-UFSAR 6.3-6 REV. 14, APRIL 2002 If the HPCS line should break outside th e containment, a check valve in the line inside the drywell will prevent loss of reactor water outside the containment. The HPCS pump and piping are positioned to av oid damage from the physical effects of design-basis accidents, such as pipe whip, missiles, high temperature, pressure, and humidity.

To assure continuous core cooling, signal s to isolate the containment do not operate any HPCS valves which could affect flow to the reactor pressure vessel.

The HPCS equipment and support structures are designed in accordance with Seismic Category I criteria (Chapter 3.0). The system is assumed to be filled with water for seismic analysis.

6.3.2.2.2 Automatic Depressurization System (ADS)

If the RCIC and HPCS cannot maintain the reactor water level, the automatic depressurization system, which is independent of any other ECCS, reduces the reactor pressure so that flow from LPCI and LPCS systems enters the reactor vessel in time to cool the core and limit fuel cladding temperature.

The automatic depressurization system employs nuclear system pressure relief valves to relieve high-pressure steam to the suppression pool. The design, number, location, description, and evaluation of the pressure relief valves are discussed in detail in Subsection 5.2.2.

4.1. The operation of the ADS is discussed in Subsection 7.3.1.2.2. The piping and instrument diagram (P&ID) for the ADS is shown in Drawings M-55 and M-116.

6.3.2.2.3 Low-Pressure Core Spray (LPCS) System

The low-pressure core spray system consists of a centrifugal pump that can be powered by normal auxiliary power or the standby a-c power system; a spray sparger in the reactor vessel above the co re (separate from the HPCS sparger);

piping and valves to convey water from the suppression pool to the sparger; and associated controls and instrumentation.

Drawing Nos. M-94 and M-140 show the P&ID for the low-pressure core spray system, and Figure 6.3-4 shows the process diagram for the low-pressure core spray system.

When low water level in the reactor vessel or high pressure in the drywell is sensed, with reactor vessel pressure low enough, the low-pressure core spray system automatically sprays water into the top of th e fuel assemblies to cool the core. This action is initiated in conjunction with other ECCS subsystems soon enough, and at a sufficient flow rate to maintain the fuel cladding temperature below 2200

° F. (The low-pressure coolant injection system starts from the same signals and operates independently to achieve the same objective by flooding the reactor vessel.)

LSCS-UFSAR 6.3-7 REV. 13 The low-pressure core spray system protects the core in the event of a large break in the nuclear system and when the HPCS is unable to maintain reactor vessel water level. Such protection extends to a small break in which the ADS or HPCS has operated to lower the reactor vessel pressu re to the operating range of the LPCS. The system vessel pressure versus flow characteristic assumed for LOCA analyses is shown in Figure 6.3-5. Figure 6.3-11 shows the minimum required pump head for the LPCS system in order to m eet the LOCA analyses assumption.

The LPCS pump receives power from an a-c power bus having standby power source backup supply. The pump motor and associ ated automatic motor-operated valves for the LPCS and one LPCI loop receive a-c power from the same bus, while another bus provides a-c power for equipment on the other two LPCI loops (Section 8.3).

The low-pressure core spray pump and all motor-operated valves can be operated individually by manual switches located in the control room. Operating indication is provided in the control room by a flowmeter and valve indicator lights.

To assure continuity of core cooling, signals to isolate the containment do not operate any low-pressure core spray system valves which could affect flow to the reactor pressure vessel.

The LPCS injection check valve is the only low-pressure core spray equipment in the containment required during a loss-of-coolant accident that requires consideration for the high temperature and humidity environment in the drywell resulting from the accident. The valve actuates on flow through the pipeline, independent of any external signal. The actuator is provided only for local repositioning. Thus, neither the normal nor accident environment in the drywell affects the operability of the low-pressure core spray equipment for the accident.

The LPCS system piping and support structures are designed in accordance with Seismic Category I criteria (Chapter 3.0). The system is assumed to be filled with water for seismic analysis.

LPCS flow passes through a motor-operated pump suction valve that is normally open. This valve can be closed by a remote manual switch (located in the control room) to isolate the LPCS system from the suppression pool should a leak develop in that system. This valve is located in the core spray pump suction line as close to the suppression pool penetration as practical. Because the LPCS conveys water from the suppression pool, a closed loop is established for the spray water escaping from the break.

The LPCS pump is located in the reacto r building below the water level in the suppression pool to assure positive pump suction. Pump NPSH requirements are met with the containment at atmospheric pressure. A pressure gauge is provided to indicate the suction head. The LPCS pump characteristics are shown in Figure 6.3-6.

LSCS-UFSAR 6.3-8 REV. 13 6.3.2.2.4 Low-Pressure Coolant Injection (LPCI) Subsystem

The low-pressure coolant injection subsys tem is one of the independent operating subsystems of the RHR system. The LPCI su bsystem is actuated by low water level in the reactor or high pressure in the drywell. The subsystem, in conjunction with other ECC subsystems, is required to flood the core before fuel cladding temperature reaches 2200

° F and then to maintain water level.

LPCI operation provides protection to the core for a large break in the nuclear system in addition to the LPCS and HPCS. Protection provided by LPCI also extends to a small break in which the ADS or HPCS have reduced the reactor vessel pressure to the LPCI operating range. The vessel pressure versus flow characteristic assumed in the LOCA anal yses for the LPCI pumps is shown in Figure 6.3-7. Figure 6.3-12 shows the minimum required pump head for the LPCI system in order to meet the LOCA analyses assumptions.

Figure 6.3-8 shows the schematic process diagram (and process data) of the RHR system. The LPCI subsystem uses the three RHR motor-driven centrifugal pumps to convey water from the suppression pool to the reactor vessel through three separate nozzles. The RHR pumps receiv e power from a-c power buses having standby power source backup supply. Tw o RHR pump motors and the associated automatic motor-operated valves receive a-c power from one bus, while the LPCS

pump and the other RHR pump motor and valves receive power from another bus (Section 8.3).

The pump, piping, control and instrumentation of the LPCI loops are separated and protected so that any single physical event, or missiles generated by rupture of any pipe in any system within the drywell, cannot make all loops inoperable.

To assure continuity of core cooling, signals to isolate the primary containment do not operate any RHR system valves which interfere with the LPCI mode of operation.

The LPCI injection check valves on each LPCI line are the only LPCI components in the drywell required to actuate during a loss-of-coolant accident that require consideration for the high temperature and humidity environment in the drywell resulting from the accident. The valves actuate on flow through the pipeline, independent of any external signal. The actuator is provided only for local repositioning. Thus, neither the normal nor accident environment in the containment affects the operability of the low-pressure coolant injection equipment for the accident.

LSCS-UFSAR 6.3-9 REV. 15, APRIL 2004 Using the suppression pool as the source of water for LPCI establishes a closed loop for recirculation of LPCI water escaping from the break. LPCI pumps and equipment are described in detail in Subsecti on 5.4.7, which also describes the other functions served by the same pumps if not needed for the LPCI function. The portions of the RHR required for accident protection are designed in accordance with Seismic Category I criteria (Chapter 3.0). The piping and instrument diagram (P&ID) for the LPCI is shown in Drawings M-96 and M-142.

6.3.2.2.5 ECCS Discharge Line Fill System

One design requirement of any core cooling system is that cooling water flow to the reactor vessel be initiated rapidly when the system is called on to perform its function. This quick start system characteristic is provided by quick opening valves, quick start pumps, and standby a-c power source. The lag between the signal pump start and the initiation of flow into the RPV can be minimized by always keeping the core cooling pump discharge lines full. If these lines were empty when the systems were called for, the large momentum forces associated with accelerating fluid into a dry pipe could ca use physical damage to the piping. The ECCS discharge line fill system maintains the pump discharge lines in a filled condition.

Since the ECCS discharge lines are elevated above the suppression pool, check valves are provided near the pumps to prevent back flow from emptying the lines into the suppression pool. Past experience has shown that these valves will leak slightly, producing a small back flow that will eventually empty the discharge piping. To ensure that this leakage from the discharge lines is replaced and the lines are always kept filled, a water leg pump is provided for each ECCS division.

The power supply to these pumps is classi fied as essential when the main ECCS pumps are deactivated. Indication is provided in the control room as to whether these pumps are operating, and ESF system status lights indicate low discharge lines pressure. The piping and instrume nt diagram (P&ID) for the ECCS is shown on the P&IDs for HPCS, LPCS, and LPCI.

6.3.2.2.6 ECCS Pumps NPSH The ECCS pump specifications are such th at the NPSH requirements for HPCS, LPCS and LPCI are met with the containment at atmospheric pressure and the suppression pool at saturation temperature. Calculations were performed to evaluate ECCS NPSH requirements post DBA-LOCA. The calculations used the

following conservative inputs:

1. Maximum ECCS pump flow - unthrottled system, reactor pressure at 0 psid, maximizing suction friction losses and NPSH required.

LPCI pump - 8100 gpm LPCS pump - 8100 gpm HPCS pump - 7000 gpm LSCS-UFSAR 6.3-10 REV. 13

2. Increased clean, commercial steel piping friction losses to account for potential aging effects, thus maximizing suction losses. An absolute roughness of

0.0005 ft was used (vs. 0.00015 ft. for clea n pipe), resulting in an increase in calculated head loss of about 22 percent.

3. To account for strainer plugging, the he ad loss across the debris bed formed on the stacked disk replacem ent strainers installed at the suction of the ECCS pumps due to accumulation of insulation debris and miscellaneous fibrous and particulate matter debris produced as a result of a LOCA is determined. This head loss is added to the head loss associated with a clean strainer.

4. Containment conditions used in the analysis are containment at atmospheric pressure and the suppression pool at saturation temperature (212F).

5. A minimum suppression pool elevation of 695' 11-1/2" is used. This includes a worst-case post-LOCA drawdown of 43 inches.

6. NPSH Required values for the ECCS pumps are taken from the vendor pump curves. With respect to the pump suction inlet centerline, the NPSH Required is: LPCI pump - 14.0 ft. @8100 gpm LPCS pump - 2.0 ft. @8100 gpm HPCS pump - 5.0 ft. @7000 gpm The calculations determined that adequate NPSH exists to meet ECCS pump requirements post LOCA for all ECCS pumps.

Additionally, adequate margin exists to ensure that flashing does not occur in any of the ECCS pump suction lines post-LOCA.

LSCS-UFSAR 6.3-11 REV. 17, APRIL 2008 ECCS PUMP NPSH AND FLASHING M ARGINS FOR LIMITING SUPPRESSION POOL CONDITIONS Pump Pump Flow Rate (gpm) Strainer Margin for NPSH (ft.) Strainer Margin for Flashing (ft.) Clean Strainer Head Loss 1 (ft.) Head Loss due to post-LOCA debris 2 (ft.) NPSH Margin (ft.) RHR/LPCI 8100 5.4 12.4 0.71 3.6 1.1 LPCS 8100 17.6 12.6 0.71 3.6 8.3 HPCS 7000 14.0 11.6 0.53 3.6 7.4 1 0.76 feet @8400 gpm 2 Maximum value (@8100 gpm, Unit 2) 6.3.2.2.7 Design Pre ssures and Temperatures The design pressures and temperatures at various points in the system, during each of the several modes of operation of the ECC subsystems, can be obtained from the miscellaneous information blocks on the fo llowing process diagrams: Figure 6.3-1 for the HPCS, Figure 6.3-4 for the LPCS, and Figure 6.3-8 for the LPCI.

The operational characteristics of the ADS valves are presented in Subsection 5.2.2.

6.3.2.2.8 Coolant Quantity

With reference to the Mark II containment at LaSalle County Station Units 1 and 2, the HPCS system normally takes suction from the suppression pool which contains a minimum of 128,800 cubic feet of water.

The LPCS and LPCI systems also take suction from the suppression pool for their source of water.

The CSCS equipment cooling water system source (cooling lake) which provides the ultimate heat sink for cooling the suppression pool during the recovery from a DBA has sufficient capacity to accept heat from the suppression pool and prevent it from exceeding 200

° F.

6.3.2.2.9 Pump Characteristics Pump characteristic curves and the pump power requirements for all ECCS pump are shown in Figures 6.3-3, 6.3-6, and 6.3-9. Pump power requirements are given in Chapter 8.0.

LSCS-UFSAR 6.3-12 REV. 13 6.3.2.2.10 Heat Exchanger Characteristics

There are no heat exchangers in the closed cooling water path associated with the emergency core cooling subsystems. The heat exchangers in the RHR system are discussed in Section 6.2.

6.3.2.2.11 ECCS Flow Diagrams

A schematic diagram and the flow rates and pressures of the various ECCS subsystems can be obtained from the follo wing process diagrams: Figure 6.3-1, High-Pressure Core Spray System; Figure 6.3-4, Low-Pressure Core Spray System; and Figure 6.3-8, Residual Heat Removal System. (The RHR process diagrams show the low-pressure coolant injection system.) These parameters are presented for several modes of operation, including loss-of-coolant accident and test conditions.

6.3.2.2.12 Relief Valves and Vents

The ECC subsystems contain relief valves to protect the components and piping from inadvertent overpressure conditions.

The HPCS system has one relief valve on the discharge side of the pump downstream of the check valve to relieve thermally expanded fluid:

Nominal relief setting: 1500 psig.

HPCS suction side relief valve:

Nominal relief setting: 100 psig Capacity: > 10 gpm, 10% Accumulation.

The LPCS system pump discharge relief valve:

Nominal relief setting: 550 psig Capacity: 100 gpm, 10% Accumulation.

LPCS suction side relief valve:

Nominal relief setting: 100 psig Capacity: > 10 gpm, 10% Accumulation.

LSCS-UFSAR 6.3-13 REV. 14, APRIL 2002 The LPCI system pump discharge relief valve (one for each of three pumps):

Nominal relief setting: 500 psig.

6.3.2.2.13 Motor-Operated Va lves and Controls (General)

Motor-operated valves are used in the RHR, HPCS, and LPCS emergency core cooling (ECC) systems; they are also used in the RCIC, feedwater, recirculation, reactor water cleanup (RWCU), standby gas treatment, standby liquid control, main

steam, and hydrogen recombiner systems.

In addition, motor-operated valves are installed on various primary and secondary containment isolation lines, certain sample lines for containment sampling in the post-LOCA condition, and other lines as indicated in Table 6.3-9.

Valve motor operators in these safety systems are provided with thermal overload protection devices. To ensure that the thermal overloads will not prevent the motor-operated valves from performing their safety-related functions under emergency conditions, the thermal overload protection devices are either bypassed under accident conditions or have sufficiently high trip setpoints to prevent inadvertent trips during valve operatio n per Regulatory Guide 1.106, Rev. 1. Thermal overload bypass circuits are normally installed on the safety-related motor-operated valves that are required to operate during or immediately following an accident such as the primary containment automatic isolation, emergency core

cooling, and RCIC system valves. Ther mal overload bypass circuits are not installed on the hydrogen recombiner valves since these valves are not required to be operated until several hours after the accident has occurred. In addition, these valves are normally closed and are prov ided with only a remote manual control system. For the valves equipped with thermal overlo ad bypass circuits, the thermal overload protection is either (1) normally in the circuit but automatically bypassed whenever

any safety-related use of the valve is initiated, or (2) continuously bypassed and temporarily placed in the circuit via a test switch when the motors are undergoing periodic surveillance or maintenance testing.

To prevent the valve motors from bein g damaged during normal operation or surveillance testing when the thermal overloads are not bypassed, the thermal overloads are set to trip the valve motor operators during locked rotor conditions. A schematic or typical thermal overload bypass arrangement is shown in Figure 6.3-47 and a list of motor-operated valves which have their thermal overload protection bypassed during an accident condition is given in Table 6.3-9.

For the hydrogen recombiner motor-operated valves, the thermal overloads are always in the circuit. However, setting calculations based on IEEE-741-1990 demonstrate that the thermal overloads for these valves will not inadvertently trip LSCS-UFSAR 6.3-14 REV. 13 during required valve operation. The trip setpoints of these thermal overloads have been verified to account for the uncertainti es due to the ambient temperature at the location of the overload device following an accident and the inaccuracies in the device trip characteristics.

Further information on motor-operated valves and controls is provided in Subsection 6.2.4.

6.3.2.2.14 Process Instrumentation

Multiple instrumentation is available to th e operator in the control room to assist him in assessing the post-LOCA conditions.

Basically, these indications are two varieties: those which indicate the pressures, temperatures and level in the reactor vessel and in the containment; and those that

provide indication of operation of the ECCS, position of valves and circuit breakers and flows of ECCS systems.

The most significant instruments in the first category would be:

a. reactor vessel level, b. reactor vessel pressure, c. containment pressure, d. containment temperature, e. suppression pool level, and

f. suppression pool temperature, and in the category of ECCS:

a. LPCI flow, b. LPCS flow, and

c. HPCS flow, Other available instrumentation is listed in the P&ID included with the description

of the above system in Chapters 5.0 and 6.

0. Discussion of instrumentation also appears in some detail in Chapter 7.0.

LSCS-UFSAR 6.3-14a REV. 14, APRIL 2002 6.3.2.2.15 Scram Discharge System Pipe Break In August 1981, the U. S. Nuclear Regu latory Commission published NUREG-0803, "Generic Safety Evaluation Report regarding integrity of BWR Scram System Piping". This document addressed the possibility of Scram Sy stem pipe breaks outside the primary containment. Specifically, a generic BWR probabilistic risk assessment in that document indicated that the postulated Scram Discharge Volume (SDV) event is not a dominant contributor to the probability of core damage. However, NRC guidance in Chapter 5 of NUREG-0803 required that certain plant specific issues be addressed by BWR owners. These plant specific issues included (1) Piping Integrity, (2) Mitigation Capability, and (3) Environmental Qualification.

LaSalle Station has addressed the plant-specific recommendations of NUREG-0803 in the response to NRC per Reference 34. The plant-specific evaluation established that even with the postulated break in the Scram Discharge System piping, the LaSalle leak detection equipment and the Station Operating Procedures will guide the Reactor Operators to prompt and successful mitigation of the event with equipment that is qualified for safe shutdown, adequate core cooling, and capable of maintaining secondary containment integrity.

LSCS-UFSAR 6.3-15 REV. 15, APRIL 2004 6.3.2.3 Applicable Codes and Classification All piping systems and components (pumps , valves, etc.) for the ECCS comply with the applicable codes, addenda, code cases, and errata in effect at the time the equipment is procured. See Tables 3.

2-1, 3.2-2, 3.2-3 and 3.2-4 for code requirements pertaining to components and systems. Tables 3.2-1, 3.2-2, and 3.2-3 list code editions in effect at the ti me of original equipment procurement.

The piping and components of the ECCS subsystems within the containment and out to and including the pressure retainin g injection valve are Class I. All other piping and components are Class 2, 3, or non-Code as indicated on the system P&ID. Subsection NA, NB, NC and ND of the Code apply to the ECCS.

The equipment and piping of the ECCS, in order to meet specified seismic capabilities, are designed to the requirements of Seismic Category I. This class includes all structures and equipments essential to the safe shutdown and isolation of the reactor, or the failure or damage of which could result in undue risk to the health and safety of the public.

6.3.2.4 Materials Specif ications and Compatibility Refer to Table 5.2-7, Reactor Coolant Pressure Boundary Materials (Section 5.2) for a presentation of the specifications which generally apply to the selection of materials used in the emergency core cooling system. Nonmetallic materials such as lubricants, seals, packings, paints and primers, insulation, as well as metallic materials, etc., are selected as a result of an engineering review and evaluation for compatibility with other materials in the system and the surroundings with concern for chemical, radiolytic, mechanical, and nuclear effects.

Materials used in or on the emergency core cooling system are reviewed and evaluated with regard to radiolytic and pyrolytic composition and attendant effects

on safe operation of the ECCS. For example, guidance on the use of fluoro carbon plastic (Teflon) is provided to address IGSCC and FME concerns associated with use of Teflon. Only inorganic thermal in sulation, which does not decompose due to radiation or temperature, is used in these environments. All paints used are suitable for the temperature conditions anticipated for their service. Additional information is presented in Section 6.1.

6.3.2.5 System Reliability As applied to the ECCS, availability is defi ned as the probability that the system is operable when required. The ECCS avail ability is a function of the component system test intervals and the failure ra tes of the component parts used in the systems. The component parts used in the ECCS have low failure rates, as evidenced by historical field operating ex perience. The ECCS availability required

LSCS-UFSAR 6.3-16 REV. 14, APRIL 2002 to assure adequate plant safety is established as a system design requirement. System availability is evaluated to assure adherence to the availability design requirement, the periodic surveillance test intervals, and allowable repair times for inoperable systems. When applicable, analyses are performed by the methods outlined in Reference 1. The levels of redundancy, diversity, and surveillance requirements combine to yield a high order of system availability.

ECCS analyses to determine peak core temp eratures are based on the most limiting single failures, assuming no offsite power is available. The analyses demonstrate that the ECCS function is sufficient to meet the Appendix K criteria. The analyses do not consider various minimum combinations of the remaining systems, following a postulated single failure, which are suff icient to meet the Appendix K criteria.

6.3.2.6 Protection Provisions

The emergency core cooling system piping and components are protected against damage from movement, from thermal stre sses, from the effects of the LOCA and the safe shutdown earthquake.

The component supports which protect agai nst damage from movement and from seismic events are discussed in Subsection 5.4.14. The methods used to provide assurance that thermal stresses do not cause damage to the ECCS are described in Subsection 3.9.1.

The ECCS are protected against the effects of pipe whip, which might result from piping failures up to and including the LO CA. This protection is provided by separation, pipe whip restra ints, or energy absorbing materials if required. One of these three methods will be applied to prov ide protection against damage to piping and components of the ECCS which otherwis e could result in a reduction of ECCS effectiveness to an unacceptable level.

The ECCS piping and components located ou tside the reactor building are protected from internally and externally generated missiles by the reinforced concrete structure of the ECCS pump rooms. In addi tion, the watertight construction of the ECCS pump rooms, when required, protects against mass flooding.

6.3.2.7 Provisions for Performance Testing

High-Pressure Core Spray System

a. A full flow test line is provided to route water from and to the suppression pool without entering the reactor pressure vessel.

b. Instrumentation is provided to indicate system performance during normal test operations.

LSCS-UFSAR 6.3-17 REV. 14, APRIL 2002 c. All motor-operated valves are capable of manual operation either local or remote for test purposes with the exception of valves E22-F010 and E22-F011. Valves E22-F001, E22-F010, and E22-F011 are no longer consid ered part of the design basis for the HPCS System.

d. System relief valves are re movable for bench testing during plant shutdown.

e. Drains are provided to leak test the major system valves.

Low-Pressure Core Spray System

a. A full flow test line is provided to route water from and to the suppression pool without entering the reactor pressure vessel.

b. A provision exists to crosstie to the RHR Shutdown Cooling suction line to utilize reactor quality water when testing the pump discharge into the reactor pressure vessel during normal plant shutdown. Utilization of this crosstie is optional as testing can be performed with suction from the Suppression Pool.

c. Instrumentation is provided to indicate system performance during normal and test operations.

d. All motor-operated valves and check valves are capable of operation for test purposes.

e. Relief valves are removable for bench testing during plant shutdown.

Low-Pressure Coolant Injection System

a. A discharge test line is provided for each of the three pump loops to route suppression pool water back to the suppression pool without entering the reactor pressure vessel.

b. A suction test line supplying re actor grade water, is provided to test loop "C" discharge into the reactor pressure vessel during normal plant shutdown.

c. Instrumentation is provided to indicate system performance during normal and test operations.

d. All motor-operated valves, air-operated valves, and check valves are capable of manual operation for test purposes.

LSCS-UFSAR 6.3-18 REV. 17, APRIL 2008 e. Shutdown lines taking suction from the reactor system water are provided for loops "A" and "B" to test pump discharge into the reactor pressure vessel during normal plant shutdown and to provide for shutdown cooling.

f. All relief valves are removable for bench testing during plant shutdown.

6.3.2.8 Manual Actions

The initiation of the ECCS is completely au tomatic. No operator action is assumed for at least 10 minutes after initiation.

As shown elsewhere in this section, something less than 4 minutes is required to reflood the core following the design-basis accident. The length of time required is a function of the size and location of the break and the location of the postulated single failure, if any. A time sequence of events for these oper ations is given in Table 6.3-3.

The design evaluations are all based on these rather long operator delays, and indicate considerable safety margin is still available.

6.3.3 ECCS Performance Evaluation

The performance of the ECCS is evaluated through application of the 10 CFR 50 Appendix K evaluation models and then showing conformance to the acceptance criteria of 10 CFR 50.46 (References 1, 19, 20, 40 and 41 for GE fuel and References 11, 12, 13, 14, 15 and 46 for FANP fuel) provide a complete description of the methods used to perform the calculations. These methods are summarized herein. A summary description of the loss-of-coolant accident results are also provided herein. LOCA Analysis for Power Up rate to 3489 MWt was performed in References 18, 20, 33, and 42 for GE fuel and References 16 and 47 for FANP fuels.

The information provided herein is applicable to the current licensing basis LOCA analyses from References 18, 33, 16, 42 and 47.

The information provided herein is applicable to the initial LOCA analysis, unless

otherwise noted.

The ECCS performance is evaluated for th e entire spectrum of break sizes for postulated LOCA's. The accidents, as listed in Chapter 15.0, for which ECCS operation is required are:

a. 15.2.8 feedwater piping break; LSCS-UFSAR 6.3-19 REV. 17, APRIL 2008 b. 15.6.4 spectrum of BWR steam system piping failures outside of containment; and

c. 15.6.5 loss-of-coolant accidents.

Chapter 15.0 provides the radiological consequences of the above listed events.

6.3.3.1 ECCS Bases for Technical Specifications

The maximum average planar linear heat generation rates calculated in this performance analysis provide the basis fo r technical specifications designed to ensure conformance with the acceptance criteria of 10 CFR 50

.46. Minimum ECCS functional requirements are specified in Su bsections 6.3.3.4 and 6.3.3.5, and testing requirements are discussed in Subsection 6.3.4. Limits on minimum suppression pool water level are discussed in Section 6.2.

6.3.3.2 Acceptance Criteria for ECCS Performance

The applicable acceptance criteria, extracted from 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Sy stems for Light-Water-Cooled Nuclear Power Reactors," are listed, and for each criterion applicable parts of Subsection 6.3.3, where conformance is demonstrated, are indicated. A detailed description of the methods used to show compliance are shown in References 11, 19, 20 and 46.

Criterion 1; Peak Cladding Temperature

"The calculated maximum fuel element cladding temperature shall not exceed 2200°F." Conformance to Criterion 1 is shown in Tables 6.3-6a, 6.3-6i and 6.3-8.

Compliance with criterion 1 for GE fuels is demonstrated in References 18, 33 and

42. Criterion 2: Maximum Cladding Oxidation

"The calculated total local oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidatio n." Conformance to Criterion 2 is shown in Tables 6.3-6a, 6.3-6i and 6.3-8. Compliance with criterion 2 for GE fuels is demonstrated in References 18,33 and 42.

Criterion 3: Maximum Hydrogen Generation

"The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical LSCS-UFSAR 6.3-20 REV. 16, APRIL 2006 amount that would be generated if all the metal in the cladding cylinder surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react." Conformance to Criterion 3 is shown in Tables 6.3-6a, 6.3-6i, an d 6.3-8. Compliance with criterion 3 for GE fuels is demo nstrated in References 18,33 and 42.

Criterion 4: Coolable Geometry "Calculated changes in core geometry shall be such that the core remains amenable to cooling." As described in Reference 1, Se ction III, conformance to Criterion 4 is demonstrated by conformance to Criteria 1 and 2. Compliance with criterion 4 for GE fuels is demonstrated in References 18,33 and 42.

Criterion 5: Long-Term Cooling "After any calculated successful initial op eration of the ECCS, the calculated core temperature shall be maintained at an acceptably low value; and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core." Conformance to Criterion 5 is demonstrated generically for General Electric BWR's in Reference 20, Section III.A. Briefly summarized, when the core refloods shortly following the postulated LOCA, the fuel rods will return quickly to saturation temperature over their entire length. For large pipe breaks the heat flux in the core will eventually be inadequate to maintain a two-phase water flow over the entire length of the core. The static water level inside the core shroud is approximately that of the jet pump suctions.

When at least one spray system is available long-term, the upper third of the core will remain wetted by the core spray water as in non-jet pump BWRs, and there will be no further perforation or metal-water reaction.

6.3.3.3 Single-Failure Considerations

The functional consequences of potential operator errors and single failures, (including those which might cause any manually controlled electrically operated valve in the

ECCS to move to a position which could adversely affect the ECCS) and the potential for submergence of valve motors in the ECCS are discussed in Subsection 6.3.2.5 and Tables 6.3-5, 6.3-6. Tabl e 6.3-6 shows that all potential single failures can be identified as no more severe than one of the following failures:

a. Low-pressure coolant injection (LPCI), emergency diesel-generator, which powers two LPCI pumps. Fo r example, failure of one LPCI pump or one LPCI injection valve is less severe than the diesel-generator failure which disables two LPCI pumps.

b. Low-pressure core spray (LPCS) emergency diesel-generator, which powers one LPCI pump and one LPCS pump.

LSCS-UFSAR 6.3-21 REV. 16, APRIL 2006 c. High-pressure core spray (HPCS).

d. One automatic depressuri zation system (ADS) valve.

It is, therefore, only necessary to consider each of the above single failures in the emergency core cooling system performance analyses. For large breaks, failure of one of the diesel generators is, in general, the most severe failure. For small breaks, the HPCS is the most severe failure. The systems of the ECCS which remain operational after these fa ilures are shown in Table 6.3-6.

For the LOCA evaluation model which covers the entire spectrum of break sizes (large breaks to small breaks), failure of the HPCS ECCS subsystem in Division 3 due to failure of its associated diesel generator is, in general, the most severe failure. The remaining operable ECCS subsystems, which include one spray subsystem, provide the capability to adequately cool the core, under near-term and long-term conditions, and prevent excessive fuel damage. For all LOCA analyses, only six ADS valves are assumed to functi on. An additional analysis has been performed which assumes five ADS valves function, however, in this analysis all

low pressure and high pressu re ECCS subsystems are also assumed to be available.

A single failure in the ADS (one ADS valve) has no effect in larg e breaks. Only six of the seven available ADS valves were assumed operable in the LOCA analyses to support one safety/relief valve out-of-servi ce operation. One ADS valve from the 6 valves modeled in the LOCA analyses was assumed to fail for the single failure evaluation as shown in Table 6.3-6.

6.3.3.4 System Performance During the Accident In general, the system response to an accident can be described as follows:

a. receiving an initiation signal;

b. a small lag time (to open all valves and have the pumps up to rated speed); and

c. finally, the ECCS flow entering the vessel.

Key ECCS actuation setpoints and time delays for all the emergency core cooling systems are provided in Table 6.3-2 for the GE LOCA analysis and in Table 6.3-2a for the FANP LOCA analysis.

The flow delivery rates analyzed in Subsec tion 6.3.4 can be determined from the head-flow curves and the pressure versus time plots discussed in Subsection 6.3.3.7. Simplified piping and instrumentation and functional control diagrams for the LSCS-UFSAR 6.3-21a REV. 17, APRIL 2008 ECCS are provided in Subsection 6.3.2.

The operational sequence of ECCS for the DBA is shown in Table 6.3-3 for the GE LOCA analysis. T able 6.3-7a shows the operational sequence of ECCS for the Reference 17 ATRIUM-9B DBA analysis.

Table 6.3-7b shows the operational sequence for the limiting recirculation break from the FANP ATRIUM- 9B LOCA analysis.

Operator action is not required for ECCS operation, except as a monitoring function, during the short-term cooling period following the LOCA. During the short-term cooling period, the operator will take action as specified in Subsection 6.2.2.3 to place the containment cooling system into operation.

LSCS-UFSAR 6.3-22 REV. 17, APRIL 2008 6.3.3.5 Use of Dual F unction Components for ECCS With the exception of the LPCI system, th e systems of the ECCS are designed to accomplish only one function: to cool the reactor core following a loss of reactor coolant. To this extent, components or portions of these systems (except for pressure relief) are not required for operation of other systems which have emergency core cooling functions, or vice versa. Because either the ADS initiating signal or the overpressure signal opens th e safety-relief valve, no conflict exists.

The LPCI subsystem is configured from the RHR pumps and some of the RHR valves and piping. When th e reactor water level is low, the LPCI subsystem (line up) has priority through the valve control logic over the other RHR subsystems for containment cooling. Immediately following a LOCA, the RHR system is directed to the LPCI mode. When the RHR shutdown cooling mode is utilized, the transfer to the LPCI mode must be remote manually initiated.

6.3.3.6 Limits on ECC System Parameters

The limits on the ECC system parameters ar e identified in Subsections 6.3.3.2, 6.3.3.7.3 and 6.3.3.7.4.

Any number of components in any given syst em may be out of service, up to and including the entire system. The maximu m allowable out-of-service time is a function of the level of redundance and the specified test intervals.

6.3.3.7 ECCS Analysis for LOCA 6.3.3.7.1 GE LOCA Analysis Procedures and Input Variables

The procedures approved for LOCA analysis conformance calculations are described in detail in References 1, 19 and 40. These procedures were used in the calculations enumerated in Subsection 6.3.3. For co nvenience, the four computer codes are briefly described below. The interfaces between the codes are shown schematically in Figures II-2a, II-2b, and II-2c in the "Documentation of Evaluation Models," Section II.A of Reference 1. The majo r interfaces are briefly noted below.

Short-Term Thermal-Hydraulic Model (LAMB)

The LAMB code is a model which is used to analyze the short-term thermodynamic and thermal-hydraulic behavior of the coolant in the vessel during a postulated LOCA. In particular, LAMB predicts the core flow, core inlet enthalpy and core pressure during the early stages of the reactor vessel blowdown. For a detailed description of the model and a discussion regarding sources of input to the model, refer to the "LAMB Code Documentation," Section II.A.3 of Reference 1.

LSCS-UFSAR 6.3-23 REV. 17, APRIL 2008 Transient Critical Power Model (SCAT)

The SCAT code is used to evaluate the short-term thermal-hydraulic response of the coolant in the core during a postulated LOCA. SCAT receives input from LAMB and analyzes the convective heat transfer process in the thermally limited fuel bundle. For a detailed description of the model and a discussion regarding sources of input to the model, refer to the "SCAT Code Documentation," Section II.A.4 of Reference 1.

Long-Term Thermal-Hydraulic Model and Refill/Reflood Model (SAFE/REFLOOD)

The SAFE/REFLOOD code is a model which is used to analyze the long-term thermodynamic behavior of the coolant in the vessel. The SAFE/REFLOOD code calculates the uncovery and reflooding of the core and the duration of spray cooling and (for small breaks) the peak cladding temperature.

For a detailed description of the model and a discussion regarding sources of input to the model, refer to the "SAFE code and REFLOOD code documentation," Sections II.A.1 and II.A.2 of Reference 1.

Core Heatup Model (CHASTE)

The CHASTE code solves the transient heat transfer equations for specific axial planes of each fuel bundle type for la rge breaks. CHASTE receives input from SCAT, SAFE and REFLOOD and calculates cladding temperatures and local cladding oxidation during the entire LOCA transient. For a detailed description of the CHASTE model and a discussion regarding sources of input, refer to the "CHASTE code documentation," Section II.A.5 of Reference 1.

The significant input variables used by the Initial LOCA codes are listed in Table 6.3-2.

Core Heatup Model (GESTR-LOCA)

The GESTR-LOCA code is used to initialize the fuel stored energy and fuel rod fission gas inventory at the onset of a postulated LOCA for input to SAFER. GESTR-LOCA also initializes the transient pellet-cladding gap conductance for input to both SAFER and SCAT.

Long-term System Response (SAFER)

This code is used to calculate the long-term system response of the reactor for reactor transients over a complete spectrum of hypothetical break sizes and

locations. SAFER is compatible with the GESTR-LOCA fuel rod model for gap LSCS-UFSAR 6.3-24 REV. 17, APRIL 2008 conductance and fission gas release. SAFER tracks, as a function of time, the core water level, system pressure response , ECCS performance, and other primary thermal-hydraulic phenomena occurring in th e reactor. SAFER realistically models all regimes of heat transfer which occur inside the core during the event, and it provides the outputs as a function of time for heat transfer coefficients and PCT.

The significant input variables used by GESTR-LOCA and SAFER are presented in Table 4-1 and Figure 3-1 in Reference 8.

SAFER/GESTR LOCA Model Code Descriptions Results of extensive LOCA experiment al programs since 1974 have clearly demonstrated the large conservatisms th at the SAFE/RELOAD LOCA models have with respect to modeling the vessel inventor y, inventory distribution and core heat transfer. A new thermal-hydraulic model (SAFER) and a new fuel rod thermal-mechanical model (GESTR-LOCA) have been developed to provide more realistic calculations for LOCA analyses. The SAFER and GESTR-LOCA models are summarized below and discussed in detail in References 19, 40, 43 and 44. As with the SAFE/REFLOOD LOCA models (des cribed above for initial core), SAFER/GESTR-LOCA is applicable to prepressurized fuel. Non-pressurized fuel calculations results in conservative limits with respect to pressured fuel.

Realistic Thermal-Hydraulics Model (SAFER)

SAFER replaces the combination of the SAFE and REFLOOD ECCS performance evaluation models discussed above for initial cores.

The SAFER code employs a heatup model with a simplified radiation heat transfer correlation to calculate PCT and local maximum oxidation, which CHASTE heatup calculation discussed above. The PCT and local maximum oxidation fraction from SAFER can be used directly.

Best Estimate fuel Rod Thermal Mechanical Model (GESTR-LOCA)

The GESTR-LOCA model has been developed to provide best-estimate predictions of the thermal performance of GE nuclear fuel rods experiencing variable power histories. For ECCS analyses, the GESTR-LO CA model is used to initialize the fuel stored energy and fuel rod fission gas invent ory at the onset of a postulated LOCA.

Details of the GESTR-LOCA models are provided in Reference 19.

Transition Boiling Transition Model (TASC)

TASC replaces the SCAT boiling transition model discussed above. The TASC model is used to evaluate the short-term thermal-hydraulic response of the coolant LSCS-UFSAR 6.3-24a REV. 17, APRIL 2008 in the core during a postulated loss-of-coolant accident. In particular, the convective heat transfer response in the thermally limiting fuel bundle is analyzed during the transient. For a detailed description of the model and a discussion regarding sources of input to the model refer to Reference 45.

SAFER/GESTR-LOCA Model Application Methodology

Using the SAFER/GESTR-LOCA models, the LOCA events are analyzed with nominal values of inputs and correlations. A calculation is performed in conformance to Appendix K and checked for consistency with generic statistical upper bound analyses that encompass modeling uncertainties in SAFER/GESTR-LOCA and uncertainties

related to plant parameters.

6.3.3.7.1.2 FANP LOCA Analysis Procedures and Input Variables

The procedures approved for LOCA analysis conformance calculations are described in detail in References 11 and 46. These proc edures were used in the calculations enumerated in Section 6.3.3. The EXEM BWR as described in Reference 11 employs four major computer codes to evaluate the system and fuel response during all phases of a LOCA. For convenience thes e four computer codes are br iefly described below. The interface between the codes are shown sche matically in References 11 and 46. The major interfaces are briefly noted below.

Blowdown Analysis (RELAX)

The RELAX code is a model which is used to calculate the system thermal-hydraulic response during the blowdown phase of the LOCA. In RELAX the core is represented by an average core channel to determine the properties of the coolant in the vessel. In

particular, RELAX predicts the upper and lo wer plenum boundary conditions for the hot channel analysis along with the core average conditions at the time of rated spray for initialization of the FLEX analysis. For a detailed discussion regarding sources of input to the model refer to the References 12 and 46.

Refill/Reflood Analysis (FLEX) (Reference 16, ATRIUM-9B and Reference 37 ATRIUM-10 Analysis)

The FLEX code is a model used to analyze the system hydraulic response during a postulated LOCA from the time of rated spray to the time of hot node reflood. The

principal result of FLEX is the prediction of time for hot node reflood. FLEX also provides a prediction of reactor vessel c oolant inventory during the ECCS injection period. FLEX provides the time of hot node re flood and the time of bypass reflood to the HUXY analysis. For a detailed description of the model and a discussion regarding sources of input to the model, refer to Reference 12.

LSCS-UFSAR 6.3-24b REV. 16, APRIL 2006 Heatup Analysis (HUXY)

The HUXY code is a model used to perfor m the heatup calculations for the entire postulated LOCA accident. HUXY predicts the thermal response of each fuel rod in one LSCS-UFSAR 6.3-25 REV. 17, APRIL 2008 axial plane of the hot channel assembly. Until time of rated spray HUXY uses RELAX calculated hot channel heat transfer coefficients. After the time of rated spray and prior to hot node reflood, HUXY uses Appendix K spray heat transfer coefficients for the fuel rods and the water canister. After the time of hot node reflood, HUXY uses Appendix K reflood heat transfer coefficients. The principal results of the HUXY heatup analysis are the peak clad temperature and the percent local oxidation of the fuel cladding. For a detailed description of the model and a discussion regarding sources of input to the model, refer to References 13 and 14.

Fuel Parameters (RODEX2)

The RODEX2 code is a model which predicts fuel parameters used as input to the blowdown and heatup analysis both for the system and hot channel analyses.

RODEX2 predicts the fuel stored energy, the pellet-clad gap, the pellet-clad gap heat transfer coefficient, and fission gas invent ory. These calculations are based on the initial conditions of the system at the onset of a postulated LOCA event. For a detailed description of the model and a discussion regarding sources of input to the model, refer to Reference 15.

6.3.3.7.2 Accident Description A detailed description of the Initial LOCA calculation methodology is provided in References 1, 19 and 40. The SAFER/GE STR LOCA analysis is summarized in Reference 18, 33, 35 and 42. The FANP LOCA analysis is summarized in Reference 16, 17, 37, 47 and 48. For convenience, a short description of the major events during a design-basis accident (DBA) is included here.

Immediately after the postulated double-ended recirculation line break, vessel pressure and core flow begin to decrease. The initial pressure response is governed by the closure of the main steam isolation valves and the relative values of energy added to the system by decay heat and energy removed from the system by the initial blowdown of fluid from the downcomer. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceas es to pump almost immediately because it has lost suction. The pump in the intact loop coasts down relatively slow. This pump coastdown governs the core flow response for the next several seconds. When the jet pump suctions uncover, calculated core flow decreases to near zero. When the recirculation pump suction nozzle uncovers, the energy release rate from the break increases significantly and the pressure begins to decay more rapidly. As a result of the increased rate of vessel pressure loss, the initially subcooled water in the lower plenum saturates and flashes up through the core, increasing the core flow. This low plenum flashing continues at a reduced rate for the next several seconds.

Heat transfer rates on the fuel cladding (Figure 6.3-20) during the early stages of the blowdown are governed primarily by the core flow response. Nucleate boiling continues in the high power plane until shortly after jet pump uncovery. Boiling transition follows shortly after the core flow loss that results from jet pump LSCS-UFSAR 6.3-26 REV. 17, APRIL 2008 uncovery. Film boiling heat transfer rates then apply, with increasing heat transfer resulting from the core flow increase during the lower plenum flashing period. Heat transfer then slowly decreases until the high power axial plane uncovers. At that time, convective heat transfer is assumed to cease.

Water level inside the shroud (Figure 6.3-17) remains high during the early stages of the blowdown because of flashing of the water in the core. After a short time, the level inside the shroud has decreased to uncover the core. Several seconds later the ECCS is actuated. As a result the vessel wa ter level begins to increase. Some time later, the lower plenum is filled, and the core is subsequently rapidly recovered.

The cladding temperature at the high power plane (Figure 6.3-29) increases initially because nucleate boiling is not maintained even though, the heat input decreases and the sink temperature decreases. A rapid, short duration cladding heatup follows the time of boiling transition when film boiling occurs and the cladding temperature approaches that of the fuel.

The subsequent heatup is slower, being governed by decay heat and core spray heat transfer. Finally, the heatup is

terminated when the core is recovere d by the accumulation of ECCS water.

6.3.3.7.3 Break Spectrum Calculations A complete spectrum of postulated break sizes and location is considered in the evaluation of ECCS performance. The gene ral analytical procedures for conducting break spectrum calculations are discusse d in References 11 and 46 for the FANP fuel and Reference 19 for GE fuel. For ease of reference, a summary of all figures and tables presented in subsection 6.3.3 is shown in Table 6.3-4. All figures and tables for the LaSalle specific SAFER/

GESTR-LOCA analysis are presented in References 18, 33 and 42. All figures and tables for the LaSalle specific FANP-LOCA analysis are presented in References 17, 36 and 48.

A complete break spectrum for GE fuel wa s evaluated in Reference 8. However, with the relaxation of certain ECCS parameters (i.e. HPCS injection valve stroke time increased from 14 to 28 seconds; LPC I and LPCS injection valve stroke time increased from 20 to 40 seconds), parts of the break spectrum calculations were repeated in Reference 18 to confirm the limiting case. The LOCA analysis for

Power Uprate to 3489 MWt was performed in Reference 35. A summary of the current SAFER/GESTR-LOCA results of the break spectrum calculations is shown in tabular form in Table 6.3-8. A summa ry of the FANP LOCA results for the break spectrum calculations for ATRIUM-9B fuel is shown in tabular form in Tables 6.3-8a and 6.3-8b. Results for ATRIUM-10 fuel are given in References 36 and 48. Conformance to the acceptance criteria (PCT < 2200 oF, local clad oxidation < 17% and a core wide metal water reaction < 1%) is demonstrated. Details of calculations for specific breaks are included in subs equent paragraphs. The LOCA analysis for GE14 fuel was performed in Reference 42.

LSCS-UFSAR 6.3-27 REV. 17, APRIL 2008 6.3.3.7.4 Large Recirculation Line Break Calculations 6.3.3.7.4.1 GE Fuel LOCA Analysis Larg e Recirculation Line Break Calculations Important results from the GE LOCA analyses of the DBA (double ended guillotine break of the recirculation suction line with a single failure of the HPCS diesel generator) are shown in Figures C-3a, C-3b, C-3c, and C-3d of Reference 18. These figures are not included in this section because GE considers this information proprietary and will not release them for use in a public domain document. The following results are shown in Reference 18 for the DBA LOCA:

For the GE LOCA analyses:

a) Water level as a function of time from SAFER. (Figure C-3a) b) Reactor vessel pressure as a function of time from SAFER. (Figure C-3b) c) Fuel rod convective heat transfer coefficient as a function of time from SAFER. (Figure C-3d) d) Peak cladding temperature as a functi on of time from SAFER. (Figure C-3c)

This case is the limiting break from the break spectrum calculations and defines the licensing basis PCT for the GE 8x8 NB fuel.

The maximum local oxidation and peak cl adding temperature from the GE LOCA (SAFER/GESTR) analysis of the DBA as well as other break sizes, single failures and break locations are shown in Table 6.3-8. Figures identified above are shown in Reference 18 (3323 MWs), they are not show n in the UFSAR because GE considers this information proprietary and will not re lease them for use in a public domain document. Power uprate results are shown in Reference 33 and the GE 14 Results are shown in Reference 42.

A "Unit Status Sheet", which tracks the changes in PCT after each 10CFR50.46 submittal is maintained by Nuclear Fuels.FANP Fuel LOCA Analysis Large Recirculation Line Break Calculations FANP performed LOCA break spectrum analyses for ATRIUM-9B and ATRIUM-10 fuel types (References 17 and 36). In addition, the Reference 48 ATRIUM-10 analysis is being applied to both Unit 1 and Unit 2. The limiting large break for ATRIUM-9B fuel is the 1.0 double-ended guillotine break of the recirculation suction piping with a single failure of th e LPCS diesel generator. The limiting large break for the ATRIUM-10 fuel analysis of Reference 36 is the 2.0 square feet split break of the recirculation suction piping with a single failure of the LPCS diesel generator. For the Reference 48, EXEM BWR-2000 analysis for ATRIUM-10, the limiting case is the double-ended guillotine break with 0.8 discharge co-efficient with the LPCI diesel generator single failure.

LSCS-UFSAR 6.3-27a REV. 17, APRIL 2008 Important results from the FANP fuel LO CA analyses of the limiting large break (1.0 double ended guillotine break of the recirculation suction piping with a single failure of the LPCS diesel generator) for ATRIUM-9B fuel are shown in Figures 6.3-13 through 6.3-29. Similar plots for ATRIUM-10 fuel can be found in References 36 and 48. These results from Reference 17 are:

a) Upper plenum pressure as a function of time during blowdown from RELAX.

LSCS-UFSAR 6.3-28 REV. 17, APRIL 2008 b) Total Break Flow as a function of time during blowdown from RELAX.

c) Core inlet flow as a function of time during blowdown from RELAX.

d) Core outlet flow as a function of time during blowdown from RELAX.

e) Lower downcomer mixture level as a function of time during blowdown from RELAX. f) Lower plenum liquid mass as a function of time during blowdown from RELAX.

g) Hot channel high power node quality as a function of time during blowdown from RELAX.

h) Hot channel high power node heat transfer coefficient as a function of time during blowdown from RELAX.

i) System pressure as a function of time from FLEX.

j) Lower plenum mixture level as a function of time during refill/reflood from FLEX. k) Relative entrainment as a function of time during refill/reflood from FLEX.

l) Core entrained liquid flow as a function of time during refill/reflood from FLEX.

m) ADS flow as a function of ti me during blowdown from RELAX.

n) LPCI flow as a function of ti me during blowdown from RELAX.

o) LPCS flow as a function of ti me during blowdown from RELAX.

p) HPCS flow as a function of time during blowdown from RELAX.

q) Peak cladding temperature as a function of time from HUXY.

The limiting large break for FANP ATRIUM-9B fuel is not the overall limiting break from the break spectrum analysis.

The small break case as described in Section 6.3.3.7.6.2 is the lim iting case for the licensing basis for FANP ATRIUM-9B fuel. Therefore, the large break results are not the basis for the ATRIUM-9B MAPLHGR limits. The ATRIUM-9B MAPLHGR limits are determined from small break analysis and they are given in Section 6.3.3.7.6.2.

LSCS-UFSAR 6.3-29 REV. 17, APRIL 2008 The MAPLHGR limits currently in the LaSalle Station's COLR for ATRIUM-9B fuel remain valid because they were the bo unding MAPLHGR values used in the SPC LOCA analysis and are conservative. The bundle specific, exposure dependent MAPLHGR limits for LaSalle Station's current fuel cycle are presented in the COLR. (Reference 21) 6.3.3.7.5 Deleted.

6.3.3.7.6 Small Recirculation Line Break Calculations 6.3.3.7.6.1 GE Fuel LOCA Analysis Small Recirculation Line Break Calculations Important results from the GE LOCA analysis of the small break (0.08 recirculation piping suction break with a single failure of the HPCS diesel generator) are shown in Figures B-1, B-2, B-3, and B-4 of Reference 42, for GE 14 fuel. These figures are not included in this section because GE considers this information proprietary and will not release them for use in a public domain document. The following results are shown in Reference 42 for the 0.08 small break LOCA:

For the GE LOCA analyses:

a) Water level as a function of time from SAFER. (Figure B-1) b) Reactor vessel pressure as a function of time from SAFER. (Figure B-2) c) Fuel rod convective heat transfer coefficient as a function of time from SAFER. (Figure B-4) d) Peak cladding temperature as a functi on of time from SAFER. (Figure B-3)

The limiting large break GE 14 fuel is not the overall limiting break from the break spectrum analysis. The small break case is described in Section 6.3.3.7.6.1 is the limiting case for the licensing Basis for GE 14 fuel.

6.3.3.7.6.2 FANP Fuel LOCA Analysis Small Recirculation Line Break Calculations FANP performed LOCA break spectrum analyses for ATRIUM-9B and ATRIUM-10 fuel types (References 17, 36 and 48). The limiting small break for ATRIUM-9B fuel is the 1.1 square feet break of the re circulation discharge piping with a single failure of the HPCS diesel generator. The limiting small break for the Reference 36 ATRIUM-10 fuel analysis is the 1.0 square feet break of the recirculation suction piping with a single failure of the HPCS diesel generator.

The PCT for the limiting small break for each fuel type bounds the PCT for the large breaks for the Reference 17 analysis of ATRIUM-9B and the Reference 36 analysis of ATRIUM-10. Therefore, th e MAPLHGR limits were determined from LSCS-UFSAR 6.3-29a REV. 17, APRIL 2008 the limiting small break analysis. The MAPLHGR limits for each fuel type were determined in References 16 and 37 and are given in Tables 6.3-6a and 6.3-6i. For the Reference 48 EXEM BWR-2000 break spectrum analysis for ATRIUM-10 fuel, the small break results are less limiting than those of the large break case identified in Section 6.3.3.7.4.2. For the limiting large break/single failure combination, the ATRIUM-10 EXEM BWR-2000 heatup analysis (Reference 47) yielded lower PCT and oxidation faction results than the ATRIUM-10 results of Reference 37. Table 6.3-6j summarizes the licensing basis results from the Reference 47 ATRIUM-10 analysis, which is being applied to both Unit 1 and Unit 2. The bundle specific, exposure dependent MAPLHGR limits for LaSalle Station's current fuel cycle are presented in the COLR (Reference 21).

LSCS-UFSAR 6.3-30 REV. 17, APRIL 2008 Important results from the FANP LOCA analysis of the small break yielding the highest cladding temperature for ATRIUM

-9B fuel are shown in Figures 6.3-30 through 6.3-46. Similar plots for ATRIUM-10 fuel can be found in References 37 and 48. These results from Reference 16 are as follows:

a) Upper plenum pressure as a function of time during blowdown from RELAX.

b) Total Break Flow as a function of time during blowdown from RELAX.

c) Core inlet flow as a function of time during blowdown from RELAX.

d) Core outlet flow as a function of time during blowdown from RELAX.

e) Lower downcomer mixture level as a function of time during blowdown from RELAX. f) Lower downcomer liquid mass as a func tion of time during blowdown from RELAX. g) Hot channel high power node quality as a function of time during blowdown from RELAX.

h) Hot channel high power node heat transfer coefficient as a function of time during blowdown from RELAX.

i) System pressure as a function of time from FLEX.

j) Lower plenum mixture level as a function of time during refill/reflood from FLEX. k) Relative entrainment as a function of time during refill/reflood from FLEX.

l) Core entrained liquid flow as a function of time during refill/reflood from FLEX.

m) ADS flow as a function of ti me during blowdown from RELAX.

n) LPCI flow as a function of ti me during blowdown from RELAX.

o) LPCS flow as a function of ti me during blowdown from RELAX.

p) HPCS flow as a function of time during blowdown from RELAX.

q) Peak cladding temperature as a function of time from HUXY.

LSCS-UFSAR 6.3-31 REV. 17, APRIL 2008 6.3.3.7.7 Calculations For Other Break Locations 6.3.3.7.7.1 GE Fuel LOCA Analysis Ca lculations for Other Break Locations

GE analyzed four non-recirculation break locations to determine the limiting non-recirculation line break and whether or not the results of this break were bound by the limiting recirculation line break. Th ese breaks are the HPCS line break, the feedline break, the main steamline break inside containment, and the steamline break outside of containment. The main steamline break outside containment (see Section 6.3.3.7.8.1) was determined to be the limiting non-recirculation line break in Reference 8. Reference 8 also shows that the HPCS line break, the feedline break, and the main steamline break inside cont ainment result in no cladding heatup beyond the initial cladding temperatur

e. For these reasons no other non-recirculation line breaks needed to be examined in References 18, 33, and 42.

6.3.3.7.7.2 FANP Fuel LOCA Analysis Calculations for Other Break Locations

FANP also analyzed non-recirculation line breaks in References 17 and 36. These included breaks in HPCS and LPCI. Addi tional breaks (main steamline, feedwater line, reactor water cleanup line and shutdown cooling lines) were dispositioned in References 17 and 36 as non-limiting. Refe rences 17 and 36 show that breaks inside containment are less limiting than breaks outside containment. The most limiting

non-recirculation line breaks are the HPCS and the LPCI line breaks, of which the HPCS line break with a single failure of the LPCI diesel generator is the most limiting. See Table 6.3-8b for a summary of the non-recirculation line break results for ATRIUM-9B fuel. Results for ATRIUM-10 fuel are given in Reference 36.

For the Reference 48 EXEM BWR-2000 brea k spectrum analysis for ATRIUM-10 fuel, the limiting case of the HPCS line break was analyzed. The worst single failure for this case is the LPCS diesel generator. The ECCS line brea ks are nonlimiting.

6.3.3.7.8 Steamline Br eak Outside Containment

Any break outside the primary containment in a line which connects directly to the reactor pressure vessel will initiate ADS action if conditions as described in subsection 7.3.1.2.2.3 are met. Therefore, given th e LOCA assumptions of no feedwater or RCIC, and assuming the failure of HPCS if the main steamline isolation valves (MSIV) close and the break becomes is olated or is too small to depressurize the vessel to below the shutoff head of th e low-pressure ECC systems, then actuation of the ADS is necessary to reduce the ve ssel pressure so that the low-pressure ECC systems can terminate the transient. This will occur automatically after the time delay bypass of high drywell pressure.

The outside steamline break is a representative analysis of this class of breaks, since a large amount of vessel inventory is lost through the broken steamline before the MSIV's can isolate the break. All these type s of breaks have the same characteristic LSCS-UFSAR 6.3-32 REV. 17, APRIL 2008 sequence of events once the MSIV's close culminating in automatic ADS actuation and subsequent vessel reflooding by the low-pressure ECC systems.

6.3.3.7.8.1 GE Fuel Steamline Break Outside Containment Analysis A GE outside steamline break analysis was investigated assuming automatic ADS

action 12 minutes after RPV level reaches level 1. A complete set of results using the small-break method is provided in Fi gures D-5a through D-5d of Reference 18. These figures are not included in this section because GE considers this information proprietary and will not release them for use in a public domain document. The steamline break outside containment anal ysis for Power Uprate to 3489 MWt was performed in Reference 33. The peak cladding temperature predicted is far below the 2200 o F limit. Table 6.3.7 lists the sequen ce of events associated with this break. 6.3.3.7.8.2 FANP Fuel Steamline Break Outside Containment Analysis Main Steam Line Breaks outside containment are inherently less challenging to fuel limits than MSLB inside containm ent. For MSLB outside containment, isolation valve closure will terminate break flow prior to the loss of significant inventory and the core will remain covered. The FANP analysis (References 17, 36 and 48) dispositions the steamline break inside containment by showing that the consequences of the steamline break on the core are bound by the recirculation line break analyses. The consequences of the steamline break are far from limiting with respect to 10 CFR 50.46 acceptance criteria. The accident does not result in a significant challenge to the fuel limits.

The high heat transfer during blowdown period and the rapid initiation of the low pressure ECCS lead to the predicted PCT hundreds of degrees less than the limiting re circulation line break. In many cases there is no heatup of the fuel during a steamline break. Although a steamline break may be limiting with respect to reactor vess el, containment, or radiological limits, these analyses are not significantly impacted by fuel or core design characteristics.

6.3.3.8 LOCA Analysis Conclusions 6.3.3.8.1 Errors and Changes Affecting The LOCA Analyses A new LOCA analysis (Reference 42) was performed for GE Fuel to support the introduction of GE 14 fuel for LaSalle Units 1 and 2 . There is no other type of GE fuel in the LaSalle Unit 1 and 2 core.

The GE LOCA analysis in support of GE 14 fuel incorporated all known errors and the licensing basis PCT for the GE 14 fuel is 1460 °F. All known errors and issues have been incorporated in the GE LOCA analysis (Reference 42).

The analysis of record for FANP ATRIUM-9B fuel (Reference 16) was performed in March 1999 to support the introduction of ATRIUM-9B fuel into the Unit 2 Cycle 8 core. The PCT is 1807

°F and it was reported in th e May 1999 10 CFR50.46 letter.

The subsequent letter in February 2002 report ed changes to PCT due to code errors, which increased PCT by 18

°F. The June 2000 10CFR50.46 annual letter reported LSCS-UFSAR 6.3-32a REV. 17, APRIL 2008 no assessments due to errors or plant changes. The June 2001 10CFR 50.46 annual letter reported assessments due to FANP code errors, Unit 2 Cycle 9 reload fuel and Unit 2 LPCS riser leakage. The June 2002 10CFR 50.46 annual letter reported assessments due to incorrect pellet dish volume terms in RDX2LSE fuel swelling calculation, reconciliation of RODEX2-2A numerical iteration scheme, incorrect HUXY gadolinia conductivity model, incorrect calculation start time for the BULGEX code, incorrect constant used in the rupture temperature calculation, incorrect Zircaloy heat of reaction, Unit 1 Cycle 10 reload fuel and the ATRIUM-9B exposure extension. These assessments resulted in a net PTC change of 2

°F. The June 2003 10CFR 50.46 annual letter report ed assessments due to incorrect calculation of inertia terms for recirculation pump discharge break junctions, Unit 2 Cycle 10 reload fuel, Unit 2 jet pump riser leakage and Unit 1 mid-cycle reload that resulted in a net PCT increase of 5

°F. For the March 2004 10CFR 50.46 report several assessment and error were reported but there was no net change in the PCT. Therefore, the PCT for ATRIUM-9B is 1832

°F. For the March 2006 10CFR 50.46 annual report there was no assessment nor any error reported for the GE14 and ATRIUM-9B or ATRIUM-10 fuel and hence there was no impact on the PCT.

Reference 37 shows that the PCT for ATRI UM-10 fuel is 1807 F. The ATRIUM-10 fuel LOCA analysis were reanalyzed in Refe rence 47. The Reference 47 analysis is applicable to all ATRIUM-10 fuel in both Un it 1 and 2, and thus the licensing basis PCT is 1729 F. For Unit 1 Cycle 12, there will be no ATRIUM-9B fuel, and the Reference 47 analysis for ATRIUM-10 fuel is being applied. That analysis shows the PCT to be 1729

°F, which is therefore the licensing basis PCT for FANP fuel in Unit 1.

A "Unit Status Sheet", which tracks the changes in PCT after each 10CFR50.46 submittal is maintained by Nuclear Fuels.

LSCS-UFSAR 6.3-33 REV. 15, APRIL 2004 This page is intentionally blank due to information deleted as a result of Revision 15, April 2004.

LSCS-UFSAR 6.3-34 REV. 17, APRIL 2008 6.3.3.9.1 GE Fuel LOCA Analysis Conclusions Having shown compliance with the applicable acceptance criteria of Subsection 6.3.3.2, it is concluded that the ECCS equipment will perform its function in an acceptable manner and meet all of the 10 CFR 50.46 acceptance criteria, given operation at or below the maximum average planar linear heat generation rates for GE fuels given in the COLR. The licensing basis PCT is in the most recent 10CFR50.46 report on each unit's NRC dock et. As stated in Reference 42, the licensing basis PCT for the GE 14 fuel is 1460

°F.

A "Unit Status Sheet", which tracks the errors or changes which affect any of the LOCA analyses and the current licensing basis PCT is maintained by Nuclear Fuels.

6.3.3.9.2 AREVA Fuel LOCA Analysis Conclusions

Having shown compliance with the applicable acceptance criteria of Subsection 6.3.3.2, it is concluded that the ECCS equipment w ill perform its function in an acceptable manner and meet all of the 10 CFR 50.46 acce ptance criteria, given operation at or below the maximum average planar linear he at generation rates for AREVA fuels given in the COLR, Reference 21. The licensing basis PCT for AREVA ATRIUM-9B fuel is 1832 °F. This number is based upon the ATRIUM-9B LOCA analysis (Reference 16) plus the arithmetic sum of all PCT changes due to errors or changes to the ATRIUM-9B LOCA analysis. Further details on the PCT changes due to errors or changes to the ATRIUM-9B LOCA analysis may be found in section 6.3.3.8.1.

The licensing basis PCT for AREVA ATRIUM-10 fuel from Reference 37 is 1807 F.

The Reference 47 analysis being applied to the Unit 1 Cycle 12 shows the ATRIUM-10 fuel licensing bases PCT is 1729

°F. Further details on the PCT changes due to errors or changes to the ATRIUM-10 LOCA analysis may be found in section 6.3.3.8.1. The ATRIUM-10 fuel LOCA analysis were reanalyzed in Reference 47. The licensing basis PCT for AREVA ATRIUM-10 fuel from Reference 47 is 1729 F.

Since there is no ATRIUM-9B fuel in either Unit 1 or Unit 2, the current licensing basis PCT is 1729 F, and is applicabl e to both Unit 1 and Unit 2.

A "Unit Status Sheet", which tracks the errors or changes which affect any of the LOCA analyses and the current licensing basis PCT, is maintained by Nuclear Fuels.

6.3.3.10 MSIV Closure Change from Reactor Water Level 2 to Level 1

By letter dated March 6, 1987 (Referen ce 7), CECo submitted a LOCA safety evaluation to justify changing the MSIV water level isolation setpoint. Previously, the most limiting LOCA, the one that results in the highest peak cladding temperature and determines the maximum average planar linear heat generation LSCS-UFSAR 6.3-35 REV. 16, APRIL 2006 rate (MAPLHGR) limit, wa s the recirculation suction line break DBA. ECCS calculations were performed using the NRC staff approved codes, SAFE, REFLOOD and CHASTE. The effects of the proposed lower setpoint for large, intermediate and small break LOCAs were considered.

CECo stated that large and intermediate LOCA events would not be affected by the setpoint change. For these events, there would be a rapid depressurization and inventory loss within the reactor vessel result ing in a fast actuation of the MSIVs. It was concluded that the lower MSIV setpoint would not significantly increase the reactor core inventory loss, the total core uncovery time or subsequent fuel heatup, or the radiation release to the environment. Thus, the setpoint change would not affect the consequences of design basis accidents. The NRC Staff accepted the findings.

For a small break LOCA there is a potential of initiation of MSIV closure at the proposed lower level setpoint which results in raising the peak cladding temperature (PCT). This event was analyzed. The results show that increase in PCT is less than 30

°F. The highest small break LOCA PCT would be substantially less than 2200

°F limit. The results of the LOCA analyses show that the DBA remains unchanged. Therefore, the MAPLHGR will not be changed. The NRC found this acceptable.

6.3.4 Tests and Inspections

Each active component of the emergency core cooling systems that is provided to operate in a design-basis accident is designed to be tested during normal operation of the nuclear system.

The HPCS, ADS, LPCI, and LPCS loops are tested periodically to assure that the emergency core cooling systems will operate.

Preoperational tests of the emergency core cooling systems were conducted during the final stages of plant construction prio r to initial startup (Chapter 14.0 of the FSAR). These tests assure correct functioning of all controls, instrumentation, pumps, piping, and valves. System reference characteristics, such as pressure differentials and flow rates, are documented following the preoperational tests and are used to establish the limits of acceptability for measurements obtained in subsequent operational tests.

During plant operations, the pumps valves, piping, instrumentation, wiring, and other components outside the drywell can be inspected visually at any time.

Components inside the drywell can be inspected when the drywell is open for access. When the reactor vessel is open, the spargers and other internals can be inspected.

Testing frequencies of most ECCS components are correlated with testing frequencies of the associated controls and instrumentation. When a pump or valve LSCS-UFSAR 6.3-36 REV. 16, APRIL 2006 control is tested, the operability of that pump or valve and its associated instrumentation is tested by the same action. The portions of the emergency core cooling systems requiring primary system pressure integrity are designed to specifications for in-service inspection.

A design flow functional test of the HPCS over the operating pressure and flow range is performed during normal plant operation by pumping water from the suppression pool and back through the full flow test return line to the suppression pool. The suction valve from the suppression pool is normally open and the discharge valve to the reactor remains closed.

The HPCS test conditions are tabulate d on the HPCS process flow diagram, Figure 6.3-1. If an initiation signal occurs while the HPCS is being tested, the system returns to the operating mode.

The HPCS can be tested at full flow with suppression pool water at any time except when the reactor vessel water level is low.

Each LPCI loop can be tested during reactor operation. The test conditions are tabulated in Figure 6.3-8. This test does not inject cold water into the reactor because the injection line valves are closed.

To test an LPCI pump at rated flow, the test line valve to the suppression pool is opened, the pump suction valve from the suppression pool is opened (this valve is

normally open), and the pumps are started using the remote/manual switches in the control room. Correct operation is determined by observing the instruments in the control room.

The LPCI injection check valve inside the drywell is tested by monitoring flow into the reactor vessel during surveillance testing.

If an initiation signal occurs during the test, the LPCI system returns to the operating mode. The valves in the test bypass lines are closed automatically to assure that the LPCI pump discharge is correctly routed to the reactor vessel.

Similarly, the LPCS pump and valves are tested periodically during reactor operations. With the injection valve closed and the return line open to the suppression pool, full flowing pump capability is demonstrated. The injection valve and the LPCS injection check valve are tested in a manner similar to that

previously discussed for the LPCI valves. The system test conditions during reactor shutdown are shown on the LPCS system process diagram, Fi gure 6.3-4. The portion of the LPCS outside the drywell is inspected for leaks during tests. Controls and instrumentation tests are described in Subsection 7.3.1.2.3.

LSCS-UFSAR 6.3-37 REV. 13 6.3.5 Instrumentation Requirements Design details, including redundancy and logic, of the instrumentation of the ECCS are discussed in Subsection 7.3.1.

6.3.5.1 HPCS Actuation Instrumentation The HPCS is automatically actuated by the following sensed variables: reactor vessel low water level, or drywell high pressure.

In addition, the HPCS can be manually actuated from the control room.

6.3.5.2 ADS Actuation Instrumentation The ADS is automatically actuated by the fo llowing sensed variables: reactor vessel low water level and drywell high pressures. The drywell high pressure signal is not required for auto initiation if the drywell pressure bypass timer (DPBT) times out. Another time delay allows the logic to reset or the operator to bypass automatic blowdown if conditions have corrected themselves or the signals are erroneous. A manual switch may be used to inhibit ADS action if necessary. For further discussion see subsecti on 7.3.1.2.2.3.

In addition, the ADS can be manually actuated from the control room.

6.3.5.3 LPCS Actuation Instrumentation

The LPCS is automatically actuated by the following sensed variables: reactor vessel low water level, or drywell high pressure.

In addition the LPCS can be manually actuated from the control room.

6.3.5.4 LPCI Actuation Instrumentation

The LPCI is automatically actuated by th e following sensed variables: reactor vessel low water level, or drywell high pressure. Reactor vessel low water level or drywell high pressure also stops other modes of RHR system operation so that LPCI

is not inhibited.

In addition, the LPCI can be manually actuat ed from the control room. Subsection 7.3.1.3.2.3 discusses conformance to IEEE-279 and other applicable regulatory

requirements for the ECCS instrumentation and controls.

LSCS-UFSAR 6.3-38 REV. 17, APRIL 2008

6.3.6 References

1. "Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50 Appendix K," NEDO

-20566-A, September 1986.

2. "Documentation of the Reanalys is Results for the Loss-of-Coolant Accident (LOCA) of Lead and Non-Lead Plants," letter from Darrell G.

Eisenhut (NRC) to E. D. Fuller (GE) June 30, 1977.

3. "Safety Evaluation for General Electric ECCS Evaluation Model Modifications," letter from K. R. Goller (NRC) to G. G. Sherwood (GE), April 12, 1977.

4. "Request for Approval for Use of Loss-of-Coolant Accident (LOCA) Evaluations Model Code REFLOOD05," letter from A. J. Levine (GE) to D. B. Vassalo (NRC), March 14, 1977.

5. "General Electric (GE) Loss-o f-Coolant Accident (LOCA) Analysis Model Revisions - Core Heatup Co de CHASTE05," letter from A. J.

Levine (GE) to D. F.

Ross (NRC), January 27, 1977.

6. Quadrex Document QUAD-1-83-008 Analysis reported MSIV Design Modification for LaSalle County Station, prepared by Quadrex

Corporation, August 24, 1983.

7. Letter dated March 6, 1987 from C.

M. Allen (CECo NLA) to H. R. Denton (NRC) concerning MSIV Level Setpoint Change from Level 2 to Level 1.

8. GE Document, "SAFER/GESTR-LOCA, Loss-of-Coolant Accident Analysis, LaSalle County Statio n Units 1 & 2," NEDC-31510P, December 1987.

9. Errata and Addenda Sheet No. 2, dated January 1989, for GE Document NEDC-31510P.

10. LaSalle Administrative Technical Requirements

11. "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model," ANF-91-048(P)(A) and Supplement 1, Advanced Nuclear Fuels Corporation, January 1993.

LSCS-UFSAR 6.3-39 REV. 16, APRIL 2006

12. "Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," XN-NF 19(P)(A) Volumes 2, 2A, 2B, 2C, Exxon Nuclear Company, Inc., September 1982.

13. "HUXY: A Generalized Multirod Heatup Code with 10 CFR 50 Appendix K Heatup Option - User's Manual," XN-CC-33(A) Revision 1, Exxon Nuclear Company, Inc., November 1975.

14. "BULGEX: A Computer Code to Determine the Deformation and the Onset of Bulging of Zircaloy Fuel Rod Cladding," XN-74-21 Revision 2, and XN-NF-27 Revision 2, Exxon Nuclear Company, Inc., December 1974.

15. "RODEX2 Fuel Rod Thermal Mechanical Response Evaluation Model," XN-NF-81-58(P)(A) Revision 2 Supplements 1 and 2, Exxon Nuclear

Company, Inc., March 1984.

16. "LaSalle LOCA ECCS Analysis MAPLHGR Limits for ATRIUM-9B Fuel,"

EMF-2175(P), Siemens Power Corpor ation, Revision 0, March 1999.

17. "LOCA Break Spectrum Analysis for Lasalle Units 1 and 2," EMF-2174(P), Siemens Power Corporation, Revision 0, March 1999.

18. GE Document, "LaSalle County Station Units 1 and 2 SAFER/GESTR-LOCA Loss-Of-Coolant Accident Analysis," NEDC-32258P, General

Electric Company, October 1993.

19. "The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-Of-Coolant Accident, Volume I, GESTR-LOCA - A Model for the

Prediction of Fuel and Thermal Perf ormance, Volume II, SAFER - Long Term Inventory Model for BWR Loss-Of-Coolant Analysis, Volume III, SAFER/GESTR Application Methodol ogy, NEDE-23785-1-P-A, February 1985 and Volume III, Supplement 1, Revision 1, "Additional Information for Upper Bound PCT Calculation," March 2002.

20. "General Electric Company Analytical Model For Loss-Of-Coolant Analysis in Accordance With 10CFR50 Appendix K," NEDO-20566A, General Electric Company, September 1986.

21. Core Operating Limits Report (COLR) for LaSalle County Station, Latest Revision.

22. "RPV Bottom Drain Impact on LOCA Analysis," Letter From P. D.

Knecht (GE) to Robert Tsai (ComEd) dated March 13, 1994.

LSCS-UFSAR 6.3-40 REV. 15, APRIL 2004

23. "Reporting of Changes and Errors in ECCS Evaluation Models," Letter R. J. Reda (GE) to R. C. Jo nes Jr. (NRC) dated June 28, 1996.

24. "Reporting of Changes and Errors in ECCS Evaluation Models," Letter R. J. Reda (GE) to R. C. Jone s Jr. (NRC) dated February 20, 1996.

25. "Reporting of Changes and Errors in ECCS Evaluation Models," Letter R. J. Reda (GE) to R. C. Jones Jr. (NRC) dated December 15, 1995.

26. "LaSalle County Nuclear Station Unit 1 ECCS Flow Uncertainty Evaluation," NEDC-32835P, Dated June 1998.

27. "LaSalle County Nuclear Power Station Jet Pump Riser Safety Evaluation, Evaluation of Rise r Leakage Impact," GENE-A1300439-00-02P, Dated March 1999.

28. "LaSalle Units 1 and 2 Principal LOCA Analysis Parameters," EMF-95-041, Revision 2, Siemens Power Corporation, Dated April 2001.

29. Letter, D. C. Serell (GE) to R. E. Parr, "Revised LaSalle 1 and 2 OPL-4 Form," Dated August 27, 1987.

30. "LaSalle County Nuclear Power Station Jet Pump Riser Safety Evaluation, Evaluation of Surveilla nce Monitoring Parameters," GE-NE-A13-00439-00-01P, Dated February 1999.

31. Letter, D. Garber (SPC) to R.J.

Chin (ComEd) "10CFR50.46 Reporting for the LaSalle Units", DE6:99:129, May 6, 1999.

32. "LaSalle Unit 1 Cycle 9 Relo ad Analysis", EM F-2276, Rev. 1, October, 1999.

33. GE document GE-NE-208-21-1093, "Engineering Evaluation Requirements for the LaSalle County Station Units 1 and 2 SAFER-GESTR Loss of Coolant Accident Analysis with ECCS Relaxations," dated November 1993.

34. Letter, C. E. Sargent (ComEd) to A. Schwencer (NRC) "LaSalle County Station Units 1 and 2 Response to NUREG-0803, NRC Docket Nos. 50-373 and 50-374", January 21, 1982 (SEAG Number 00-000505).

35. "LaSalle County Station Power Uprate Project, Task 407, ECCS Performance," GE-NE-A1300384-39-01, Revision 1, September 1999.

LSCS-UFSAR 6.3-40a REV. 17, APRIL 2008

36. "LaSalle Units 1 and 2 LOCA Brea k Spectrum Analysis for ATRIUM-10 Fuel", EMF-2639(P), Revision 0, Framatome ANP, November 2001.

37. "LaSalle Units 1 and 2 LOCA-ECCS Analysis MAPLHGR Limit for ATRIUM-10 Fuel", EMF-2641(P), Revision 0, Framatome ANP, November 2001.

38. "Responses to Exelon Comments - Extended Exposure for ATRIUM-9B Fuel", Framatome ANP Letter DEG:01:

136, D. Garber to F. Trikur, September 6, 2001.

39. "ATRIUM-9B Exposure Extension MAPLHGR Analysis Results for LaSalle Units 1 and 2," Framatome ANP Letter DEG:02:024, D.

Garber to F. Trikur, January 22, 2002.

40. The GESTR-LOCA and SAFER Models for the Evaluation of Loss-of-Coolant Accident: Volume III, Supplement I, Additional Information for Upper Bound PCT, and NEDE-23785P-A March 2002.

41. Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Models, NEDC-32950P January 2000.

42. GE Document GE-NE-0000-0022-8684-R2, Exelon LaSalle Units 1 and 2 SAFER/GESTR Loss-of-Coolant Acci dent Analyses for GE14 Fuel, November 2006.

43. SAFER Model for Evaluation of Loss-of-Coolant Accidents for Jet Pump and Non-Jet Pump Plants, Volumes I and II, NEDC-30996P-A, October 1987.

44. Compliance of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model, NEDC-32950P, January 2000 as reviewed by letter

from S.A. Richards (NRC) to J.F.

Klapproth (GE), "General Electric Nuclear Energy (GENE) Topical Reports NEDC-32950P and NEDC-32084P Acceptability Review," May 24, 2000.

45. NEDC-32084 P-A, Revision 2, "TASC-03A A Computer Program For Transient Analysis of a Single Channel," July 2002.

46. EMF-2361(P)(A) Revision 0, "EXEM BWR-2000 ECCS Evaluation Model," Framatome ANP May 2001.

47. EMF-3231(P) Revision 0, "LaS alle Units 1 and 2 EXEM-BWR-2000 LOCA-ECCS Analysis MAPLHER Limit for ATRIUM-10 Fuel," November 2005.

LSCS-UFSAR 6.3-40b REV. 17, APRIL 2008

48. EMF-3230(P) Revision 0, "LaS alle Units 1 and 2 EXEM-BWR-2000 LOCA Break Spectrum Analysis for Analysis for ATRIUM-10 Fuel," November 2005.

LSCS-UFSAR TABLE 6.3-1 TABLE 6.3-1 REV. 13

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LSCS-UFSAR TABLE 6.3-2 (SHEET 1 of 5)

Significant Input Variables Used In the GE Loss-Of-Coolant Accident Analyses TABLE 6.3-2 REV. 16, APRIL 2006 A. Plant Parameters Units Nominal Analysis Value Core Thermal Power MWt 3722 3797 % of Rated Core Thermal Power

% 106.7 108.8 Core Flow lbm/hr 108.5 x 10 6 108.5x10 6 Vessel Steam Dome

Pressure psia 1050 1053 Source of Information: Reference 42.

• Based on licensed power of 3489 Mwt.

LSCS-UFSAR TABLE 6.3-2 (SHEET 2 of 5)

Significant Input Variables Used In the GE Loss-Of-Coolant Accident Analyses TABLE 6.3-2 REV. 15, APRIL 2004 B. Emergency Core Cooling System Parameters Low Pressure Coolant Injection System Initiating Signals Units Analysis Value Vessel pressure at which flow may commence psid (vessel to drywell) 200 Minimum rated flow at vessel pressure gpm (3 pumps, 2 pumps, 1 pump) psid (vessel to drywell) 17961, 11974, 5987 20 System Head-flow

Delivery characteristics (3 pumps) psid/gpm 200/0 20/17961 Low water level

or High drywell pressure Inches referenced to instrument zero psig -161.5 (Level 1)*

2.5 Maximum allowable time

delay from initiating signal to pump capable of speed and injection valve full open (assuming vessel pressure permissive is satisfied) sec 60.0 Maximum Allowable

Injection Valve Stroke Time ** sec 40.0 Pressure at which

injection valve may open psig 435.0 Source of Information: Reference 36 and used only in that analysis. See Reference 42 for GE14 LOCA analysis

• Analytical Setpoint is approximately equal to top of active fuel ** No flow is assumed until the injection valve is full open

LSCS-UFSAR TABLE 6.3-2 (SHEET 3 of 5)

Significant Input Variables Used In the GE Loss-Of-Coolant Accident Analyses TABLE 6.3-2 REV. 16, APRIL 2006 Low Pressure Core Spray System Vessel pressure at which flow may commence psid (vessel to drywell) 255 Minimum rated flow at vessel pressure gpm psid (vessel to drywell) 5600 122 System Head-flow Delivery characteristics psid/gpm 255/0 122/5600 0/7000 Initiating Signals Units Analysis Value Low water level or High drywell pressure Inches referenced to instrument zero psig -161.5 (Level 1)*

2.5 Maximum allowed (runout) flow gpm 7000 Maximum allowable time delay from initiating signal to pump capable of speed and injection valve

full open (assuming vessel pressure permissive is satisfied) sec 60.0 Maximum Allowable

Injection Valve Stroke Time ** sec 40.0 Pressure at which injection valve may open psig 435.0 Source of Information: Reference 33 and used only in that analysis. See Reference 42 for GE14 LOCA analysis

• Analytical Setpoint is approximately equal to top of active fuel ** No flow is assumed until the injection valve is full open LSCS-UFSAR TABLE 6.3-2 (SHEET 4 of 5)

Significant Input Variables Used In the GE Loss-Of-Coolant Accident Analyses TABLE 6.3-2 REV. 16, APRIL 2006 High Pressure Core Spray System Vessel pressure at which flow may commence psid (vessel drywell) 1160 Minimum flow at vessel

pressure gpm psid (vessel drywell) 750 @ 1130

5400 @ 200 System Head-flow Delivery characteristics psid/gpm 1160/0 1130/750 200/5400 0/5400 Initiating Signals Units Analysis Value Low water level

or High drywell pressure Inches referenced to instrument zero psig -97.9 (Level 2)*

2.5 Maximum Allowable

Injection Valve Stroke Time ** sec 28.0 Maximum allowable time

delay from initiating signal to rated flow available and injection valve full open**

sec 41 Source of Information: Reference 33 and used only in that analysis. See Reference 42 for GE14 LOCA analysis

• Analytical Setpoint is approximately equal to 5.3 feet above top of active fuel ** No flow is assumed until the injection valve is full open LSCS-UFSAR TABLE 6.3-2 (SHEET 5 of 5)

Significant Input Variables Used In the GE Loss-Of-Coolant Accident Analyses TABLE 6.3-2 REV. 16, APRIL 2006 Automatic Depressurization System Total Number of valves installed 7 Number of valves used in

analysis 6 Minimum flow capacity of any six valves at vessel pressure lb/hr psig (at vessel pressure) 5.17 x 10 6 1146 Initiating Signals Units Analysis Value Low water level

and High drywell pressure Inches referenced to instrument zero psig -161.5 (Level 1)*

2.5 Delay time from all

initiating signals completed to the time valves are open sec 120 Low water level

and Maximum Time Delay Inches referenced to instrument zero sec -161.5 (Level 1)*

720 Source of Information: Reference 33, and used only in that analysis. See Reference 42 for GE14 LOCA analysis C. Fuel Parameters Fuel Type GE14 GE8x8NB (GE9B) Fuel Bundle Geometry 10x10 8 x 8 Number of Fuel Rods 92 60 Source of Information: Reference 33 and 42

• Analytical Setpoint is approximately equal to top of active fuel

LSCS-UFSAR TABLE 6.3-2a (SHEET 1 of 5)

Significant Input Variables Used In the FANP Loss-Of-Coolant Accident Analyses TABLE 6.3-2a REV. 17, APRIL 2008 A. Plant Parameters Units Analysis Value Core Thermal Power MWt 3796.44 % of Rated Core Thermal Power

% 102* Vessel Steam Output LBm/hr 16.57 x 10 6 Corresponding Percent

of Rated Steam Flow percent 102*

Core Flow lbm/hr 113.9x10 6 Corresponding Percent of Rated Core Flow percent 105 Vessel Steam Dome

Pressure psia 1050 Maximum Recirculation Line Break Area for

DEG ft 2 5.072 Source of Information: References 28, 16, 37 and 47

• Based on an uprated power of 112%

LSCS-UFSAR TABLE 6.3-2a (SHEET 2 of 5)

Significant Input Variables Used In the FANP Loss-Of-Coolant Accident Analyses TABLE 6.3-2a REV. 17, APRIL 2008 B. Emergency Core Cooling System Parameters Low Pressure Coolant Injection System Vessel pressure at which flow may commence psid (vessel to drywell) 200 Minimum rated flow at vessel pressure gpm (3 pumps, 2 pumps, 1 pump) psid (vessel to drywell) 17961, 11974, 5987 20 System Head-flow

Delivery characteristics (3 pumps) psid/gpm 200/0 20/17961 Initiating Signals Units Analysis Value Low-Low-Low water level or High drywell pressure Inches referenced to instrument zero psig -161.5 (Level 1)**

2.5 Low Pressure System response time from detection of LOOP sec 60.0 Maximum Allowable

Injection Valve Stroke Time *** sec 40.0 Pressure at which

injection valve may open psig 435.0 Minimum Flow Valve Opening Time Closing Time Max Bypass Line Flow per Pump Closure Setpoint sec sec gpm gpm 15.0 15.0 870.0 2463.0 Source of Information: References 28

• • Analytical Setpoint is approximately equal to top of active fuel *** Flow is assumed to increase linearly over the entire valve stroke LSCS-UFSAR TABLE 6.3-2a (SHEET 3 of 5)

Significant Input Variables Used In the FANP Loss-Of-Coolant Accident Analyses TABLE 6.3-2a REV. 17, APRIL 2008 Low Pressure Core Spray System Vessel pressure at which flow may commence psid (vessel to drywell) 255 Minimum rated flow at vessel pressure gpm psid (vessel to drywell) 5600 122 System Head-flow Delivery characteristics psid/gpm 255/0 122/5600 0/7000 Initiating Signals Units Analysis Value Low-Low-Low water level or High drywell pressure Inches referenced to instrument zero psig -161.5 (Level 1)**

2.5 Maximum (runout) flow gpm 7000 Low Pressure System response time from detection of LOOP sec 60.0 Maximum Allowable

Injection Valve Stroke Time *** sec 40.0 Pressure at which

injection valve may open psig 435.0 Minimum Flow Valve Opening Time Closing Time Max Bypass Line Flow per Pump Closure Setpoint sec sec gpm gpm 7.0**** 7.0**** 950.0 2121.0 Source of Information: References 28

• • Analytical Setpoint is approximately equal to top of active fuel *** Flow is assumed to increase linearly over the entire valve stroke

• • • • For Unit 2 Minimum Flow Valve opening and closing time. For Unit 1 there is no requirement LSCS-UFSAR TABLE 6.3-2a (SHEET 4 of 5)

Significant Input Variables Used In the FANP Loss-Of-Coolant Accident Analyses TABLE 6.3-2a REV. 16, APRIL 2006 High Pressure Core Spray System Vessel pressure at which flow may commence psid (vessel to pump

suction) 1160 Minimum rated flow at vessel pressure gpm psid (vessel to pump suction) 750 @ 1130 5400 @ 200 System Head-flow

Delivery characteristics psid/gpm 1160/0 1130/750 200/5400 0/5400 Initiating Signals Units Analysis Value Low water level or High drywell pressure Inches referenced to instrument zero psig -97.9 (Level 2)**

2.5 Maximum Allowable

Injection Valve Stroke Time *** sec 28.0 Maximum allowable time delay from LOOP to pumps capable of rated flow available and injection valve full open sec 46 Minimum Flow Valve

Max Bypass Line Flow per Pump

Closing Setpoint

gpm gpm 1350.0 1948.0 Source of Information: References 28

• • Analytical Setpoint is approximately equal to 5.3 feet above top of active fuel *** Flow is assumed to increase linearly over the entire valve stroke LSCS-UFSAR TABLE 6.3-2a (SHEET 5 of 5)

Significant Input Variables Used In the FANP Loss-Of-Coolant Accident Analyses TABLE 6.3-2a REV. 15, APRIL 2004 Automatic Depressurization System Total Number of valves installed 7 Number of valves used in

analysis 6 Minimum flow capacity of any six valves at vessel pressure lb/hr psig (vessel pressure) 5.17 x 10 6 1150 Initiating Signals Units Analysis Value Low-Low-Low water level and High drywell pressure Inches referenced to instrument zero psig -161.5 (Level 1)**

2.5 Delay time from all

initiating signals completed to the time valves are open sec 120 Low-Low-Low water

level and Maximum Time Delay ft above Top of Active Fuel sec -161.5 (Level 1)**

720 C. Fuel Parameters

Fuel Type ATRIUM-9B Fuel Bundle Geometry 9 x 9 Number of Fuel Rods 72 Fuel Type ATRIUM-10 Fuel Bundle Geometry 10 x 10 Number of Fuel Rods 83 full length rods 8 part length rods Source of Information: References 28, 16 and 37

• • Analytical Setpoint is approximately equal to top of active fuel LSCS-UFSAR TABLE 6.3-3 (SHEET 1 of 2) TABLE 6.3-3 REV. 17, APRIL 2008 OPERATIONAL SEQUENCE OF EMERGENCY CORE COOLING SYSTEMS FOR DESIGN-BASIS ACCIDENT ANALYSIS 1 (The information in this table is historical; please refer to Appendix A of Reference 33 and Reference 42 for GE14 Fuel.) The sequence of events for the limiting small break is provided in Appendix B of Reference 42.

TIME(sec) EVENTS 0 Design-basis loss-of-coolant accide nt assumed to start; normal auxiliary power assumed to be lost.

0 Drywell high pressure 2 and reactor low water level reached. All diesel generators signaled to start; scram; HPCS, LPCS, LPCI signaled to start on high drywell pressure.

t 16 Reactor low-low water level reached. HPCS receives second signal to start.

t 27 Reactor low-low-low water level reached. Main steam isolation valve close. Second signal to start LPCI and LPCS; auto-depressurization sequence begins.

(t 1+13) HPCS diesel generators ready to load; energize HPCS pump motor, open HPCS injection valve.

(t 2+13) Division 1 and 2 diesel generators ready to load; start to close containment isolation valves.

(t 1+41) HPCS injection valve open and pump at design flow, which completes HPCS startup; LPCS and LPCI (RHR "C") pumps at

rated speed.

t 3 28 Low pressure permissive fo r LPCS & LPCI injection valve (t 3+40) 68 LPCI and LPCS pumps at rated flow, LPCS and LPCI injection valves open, which completes the LPCI and LPCS startups.

~150 Core effectively reflooded assuming worst single failure; heatup terminated.

>10 min. Operator shifts to containment cooling.

LSCS-UFSAR TABLE 6.3-3 (SHEET 2 of 2)

TABLE 6.3-3 REV. 14, APRIL 2002 NOTES: 1. For the purpose of all but the next to last entry on this table, all ECCS equipment is assumed to function as designed. Performance analysis calculations consider the effects of single equipment failures. (See Subsections 6.3.2.5 and 6.3.3.3.) The re circulation suction line break DBA with limiting HPCS EDG failure case, using Appendix K assumptions, is used.

2. Credit is taken in LOCA an alyses for ECCS start on high drywell pressure signal.

Source of information: Refere nce 33 analysis results from GE.

LSCS-UFSAR TABLE 6.3-4 (SHEET 1 of 2)

KEY TO FIGURES AND TABLES IN SECTION 6.3 TABLE 6.3-4 REV. 17, APRIL 2008 Figures Applicable to Specific Breaks Large Recirculation Line Breaks Small Recirculation Line Breaks Other Break Locations GE 1.0 DEG Suction SF-HPCS/DG 6.3.3.7.4.1 AREVA 1.0 DEG Suction SF-LPCS/DG 6.3.3.7.4.2 GE 0.08 ft 2 Suction SF-HPCS/DG 6.3.3.7.6.1 AREVA 1.1 ft 2 Discharge SF-HPCS/DG 6.3.3.7.6.2 GE MSLB Outside Containment 6.3.3.7.7 Reactor Vessel Pressure C-3b* N/A B-2 N/A D-5b* Water Level C-3a* N/A B-1 N/A D-5a* Heat Transfer Coefficient C-3d* N/A B-4 N/A D-5d* Peak Cladding Temperature C-3c* 6.3-29 B-3 6.3-46 D-5c* Upper Plenum Pressure N/A 6.3-13 N/A 6.3-30 N/A Total Break Flow N/A 6.3-14 N/A 6.3-31 Core Inlet Flow N/A 6.3-15 N/A 6.3-32 N/A Core Outlet Flow N/A 6.3-16 N/A 6.3-33 N/A Lower Downcomer Mixture Level N/A 6.3-17 N/A 6.3-34 N/A Lower Downcomer Liquid Mass N/A 6.3-18 N/A 6.3-35 N/A Hot Channel High Power Node

Quality N/A 6.3-19 N/A 6.3-36 N/A Hot Channel High Power Node Heat Transfer Coefficient N/A 6.3-20 N/A 6.3-37 N/A System Pressure N/A 6.3-21 N/A 6.3-38 N/A Lower Plenum Mixture Level N/A 6.3-22 N/A 6.3-39 N/A Relative Entrainment N/A 6.3-23 N/A 6.3-40 N/A Core Entrained Liquid Flow N/A 6.3-24 N/A 6.3-41 N/A ADS Flow N/A 6.3-25 N/A 6.3-42 N/A LPCI Flow N/A 6.3-26 N/A 6.3-43 N/A LPCS Flow N/A 6.3-27 N/A 6.3-44 N/A HPCS Flow N/A 6.3-28 N/A 6.3-45 N/A

LSCS-UFSAR TABLE 6.3-4 (SHEET 2 of 2)

KEY TO FIGURES AND TABLES IN SECTION 6.3 TABLE 6.3-4 REV. 17, APRIL 2008

• These figures are shown in Reference 18 (3323 MWs), they are not shown in the UFSAR because GE considers this information proprietary and will not release them for use in a public domain document. Power uprate results are shown in Reference 33 an d the GE14 results in Reference 42.

Input Variables - Tabl es 6.3-2 and 6.3-2a Operation Sequence of ECCS for GE DBA - Table 6.3-3 Peak Cladding Temperature, Maximum Local Oxidation, and MAPLHGR vs. Exposure for FANP fuel - Table 6.3-6a, Table 6.3-6i and Table 6.3-6j Summary of GE LOCA Analys is Results - Table 6.3-8 Summary of SPC LOCA Analysis Resu lts - Table 6.3-8a and Table 6.3-8b Single Failure Analysis - Table 6.3-1

LSCS-UFSAR TABLE 6.3-5 (SHEET 1 of 6) TABLE 6.3-5 REV. 16, APRIL 2006 ECCS SINGLE VALVE FAILURE ANALYSIS SYSTEM VALVE POSITION FOR NORMAL PLANT OPERATION CLOSED OPEN CONSEQUENCES OF VALVE FAILURE ASSUMED TOGETHER WITH DESIGN-BASIS (DBA) LOCA REMAINING ECCS COOLANT DELIVERY SYSTEMS High-pressure core spray (HPCS)

Suppression pool suction E22-F015 X If MO valve fails to remain open during system operation, HPCS will no longer function. LPCS + 3 LPC1 loops Drains and pressure test connections on suction line E22-F019 E22-F017/E22-F308 E22-F339/E22-F340

X X X If these manual valves are placed in the incorrect open position, part of the flow could be diverted to locations other than the RPV. However, since all connections, except that for E22-F019, have two valves that must be left open before flow is diverted, and the leak detection system would alarm, three failures would be required for this improper position to result and go undetected. In the case of E22-F019, two failures would be required. LPCS + 3 LPCI loops +

partial HPCS Minimum flow E22-F012 X If MO valve fails to open, HPCS pump may overheat and fail. If valve fails to reclose, approximately 10% of system flow returns to suppression pool LPCS + 3 LPCI loops 90% HPCS + LPCS +3 LPCI loops Condensate tank suction to HP Core Spray E22-F001 (MO) E22-F302 (Manual)

E22-F030/E22-F309 (Pressure test connection)

X X X Valves are isolated from HPCS System by means of blind flange. Failure will have no effect on HPCS operation. HPCS + LPCS + 3 LPCI loops Test return to suppression pool E22-F023 X If MO valve is open on start of LOCA, auto close signal recloses valve. If valve fails to remain closed during system operation, approximately 90% of HPCS flow returns to suppression poo1. HPCS will no longer function.

HPCS + LPCS + 3 LPCI loops. LPCS + 3 LPCI loops Abandoned test return to condensate tank E22-F010 E22-F011

X X If these valves are placed in the incorrect open position, part of flow could be diverted to other locations than RPV. However, valves are closed and handwheels are removed. LPCS +3 LPCI loops + partial

HPCS LSCS-UFSAR TABLE 6.3-5 (SHEET 2 of 6) TABLE 6.3-5 REV. 13 SYSTEM VALVE POSITION FOR NORMAL PLANT OPERATION CLOSED OPEN CONSEQUENCES OF VALVE FAILURE ASSUMED TOGETHER WITH DESIGN-BASIS (DBA) LOCA REMAINING ECCS COOLANT DELIVERY SYSTEMS Injection valve E22-F004 X If MO valve fails to remain open, HPCS will no longer function. LPCS + 3 LPCI loops Maintenance valve E22-F038 X This manual valve is located in the discharge line inside the drywell, and if closed, would result in blocking system injection. Since the valve has position (open/closed) indication in the control room, two error/failures would be required for blockage of system flow to result (i.e., valve is placed in wrong position and operator fails to take corrective action, or position indicating lights do not properly function. LPCS + 3 LPCI loops Water leg valves E22-F026 E22-F034 E22-F006 E22-F033 X X X X These manual valves must be in the position shown to ensure that the discharge line remain filled, thus avoiding water hammer on pump start. Improper positioning would result in a pressure switch/alarm indicating the discharge line is not filled.

Therefore, two failures (valve in improper position and switch/alarm failure) must occur before the error goes undetected. LPCS + 3 LPCI loops Drains, vents and pressure test connections on discharge lines E22-F003/E22-F031 E22-F021/E22-F022 E22-F348/E22-F347 E22-F349/E22-F350 X

X X X These manual valves are normally closed, connected in series, and located on the pump discharge line. Both valves in each group must be open before water is diverted from the normal discharge path. Also, as in the case of valves F030 and F033 above, improper position would be detected by the Leak Detection System(i.e., 3 failures required for improper position to result and go undetected. LPCS + 3 LPCI loops +

partial HPCS Low-pressure core spray (LPCS)

Suppression pool suction E21-F001 X If valve fails to remain open during system operation, LPCS will no longer function.

HPCS + LPCS + 3LPCI loops. HPCS + 3 LPCI loops

LSCS-UFSAR TABLE 6.3-5 (SHEET 3 of 6) TABLE 6.3-5 REV. 13 SYSTEM VALVE POSITION FOR NORMAL PLANT OPERATION CLOSED OPEN CONSEQUENCES OF VALVE FAILURE ASSUMED TOGETHER WITH DESIGN-BASIS (DBA) LOCA REMAINING ECCS COOLANT DELIVERY SYSTEMS Drains, vents and pressure test connections on suction line E21-F008 E21-F327/E21-F328 E21-F334/E21-F335 E21-F329/E21-F330 E21-F331/E21-F332

X X X X X If these manual valves are incorrectly placed in the open position, the leak detection system would alarm. In addition, all connections except E21-F008 require that two valves in series be left in an incorrect position before suction flow is affected. Thus, three failures would be required for the improper valve positions to result in flow loss, except in the case of E21-F008 which requires two failures. HPCS + 3 LPCI loops +

partial LPCS Test return line E21-F012 X If MO valve is open on start of LOCA, auto close signal recloses valve.

If valve fails to remain closed during system operation, approximately 90% of LPCS flow returns to suppression pool. LPCS will no longer function.

HPCS + LPCS + LPCI loops.

HPCS + 3 LPCI loops Injection valve E21-F005 X If MO valve fails to remain open, LPCS will no longer function. HPCS + 3 LPCI loops Maintenance Valve E21-F051 X Since this manual valve has position indication in the control room, the valve would have to be in the wrong position (closed) and the position indication fail in order for injection blockage to occur; a malfunction requires 2 failures. HPCS + 3 LPCI loops Minimum flow E21-F011 (MO)

E21-F052 (Manual)

X X If valves are not open, LPCS pump may overheat and fail. If valve E21-F011 fails to close approximately 10% of system flow returns to suppression pool. HPCS + 3 LPCI loops For E21-F011 failure to close, HPCS + 90% LPCS + 3 LPCI loops. Drain, vent and pressure test connections on discharge line E21-F325/E21-F326 E21-F025/E21-F305 E21-F013/E21-F014 E21-F321/E21-F322 X

X X X Incorrect position could degrade injection flow. Since both manual valves are in the same drain line, both valves would have to be in the wrong position in order for injection flow to degrade; a malfunction requires 2 failures. HPCS + 3 LPCI loops +

partial LPCS

LSCS-UFSAR TABLE 6.3-5 (SHEET 4 of 6) TABLE 6.3-5 REV. 15, APRIL 2004 SYSTEM VALVE POSITION FOR NORMAL PLANT OPERATION CLOSED OPEN CONSEQUENCES OF VALVE FAILURE ASSUMED TOGETHER WITH DESIGN-BASIS (DBA) LOCA REMAINING ECCS COOLANT DELIVERY SYSTEMS Water leg Valves E21-F004 E21-F032 E21-F034 E21-F035 X X X X These manual valves must be in the indicated position to ensure discharge line remains filled. Since a low pressure alarm indicates a fill system failure, both sensor and valve position would have to be incorrect in order for the failure to go undetected. Two failures would be required HPCS + 3 LPCI loops Low-pressure coolant injection (LPCI) LPCI loop A Suppression pool suction E12-F004A X If valve fails to remain open during system operation, LPCI loop will no longer function HPCS + LPCS + 3 LPCI loops. HPCS + LPCS + 2 LPCI loops.

Minimum flow E12-F064A (MO)

E12-F018A (Manual)

X X If valves are not open, LPCI pump may overheat and fail. If valves E12-F064A fails to close approximately 10% of loop, flow returns to suppression pool HPCS + LPCS + 2 LPCI loops. HPCS + LPCS + 2 For E12-F064A failure to close. LPCI loops + 90%

LPCI loop. Test return line E12-F024A X If MO valve is open on start of LOCA, auto close signal recloses valve. If valve fails to remain closed during system operation approximately 90% of loop flow returns to suppression pool. LPCI loop will no longer function.

HPCS + LPCS + 3 LPCI loops. HPCS + LPCS + 2 LPCI loops. Drain, vent and pressure test connections on the suction line E12-F370A/E12-F369A E12-F397/E12-F398 E12-F356A/E12-F379A

E12-F071A/E12-F070 X

X X X If these manual valves are in the incorrect position, part of the flow could be diverted. However, all connections are provided with two valves in series, and the leak detection system would alarm. Thus, three failures must be postulated for the incorrect condition to go undetected. HPCS + LPCS + 2 LPCI loops + partial LPCIA LSCS-UFSAR TABLE 6.3-5 (SHEET 5 of 6) TABLE 6.3-5 REV. 13 SYSTEM VALVE POSITION FOR NORMAL PLANT OPERATION CLOSED OPEN CONSEQUENCES OF VALVE FAILURE ASSUMED TOGETHER WITH DESIGN-BASIS (DBA) LOCA REMAINING ECCS COOLANT DELIVERY SYSTEMS Low-pressure coolant injection (LPCI) (cont'd)

LPCI loop A

Heat exchanger bypass E12-F048A X

No effect. LPCI flow will be through heat exchanger. Heat exchanger pressure drop will not degrade loop flow.

HPCS + LPCS + 3 LPCI loops. Injection valve(s) E12-F042A X If MO valve fails to remain open, LPCI loop will no longer function. HPCS + LPCS + 2 LPCI loops. Maintenance valve E12-F092A (Manual) E12-F098A (Manual)

X X The valve E12-F092A with position indication in the main control room. Therefore, for this valve to be incorrectly positioned (closed), a failure of this indication as well as incorrect valve positioning (two failures) must be assumed. Valve E12-F098A could block LPCI flow if left in the incorrect (closed) position. HPCS + LPCS + 2 LPCI loops. Water leg valves E12-F085A X This manual valve must be open to ensure a filled discharge line. Incorrect positioning would be detected and alarmed in the control room by a pressure switch signal on low pressure. Thus, two failures would be required in order for valve to be incorrectly positioned.

HPCS + LPCS + 2 LPCI Drains, vents and pressure test connections on discharge line E12-F361A/E12-F362A E12-F363A/E12-F364A E12-F385A/E12-F386A E12-F080A/E12-F081A E12-F060A/E12-F075A E12-F367/E12-F368 E12-F372A/E12-F371A E12-F056A/E12-F057A E12-F321A/E12-F322A E12-F086/E12-F389 E12-F063A/E12-F388A X

X X X X X X X X X X All connections are double valved; therefore, two valves in series would have to be in an incorrect position before any flow would be diverted. In addition, the low pressure alarm would be sounded in the control room since the water leg pump would not maintain the line filled, and leak detection alarms would also be triggered by leakage into the areas.

Therefore, four failures must be postulated before any adverse effects on the system could go undetected. HPCS + LPCS + 2 LPCI +

Partial LPCI A

LSCS-UFSAR TABLE 6.3-5 (SHEET 6 of 6) TABLE 6.3-5 REV. 13 SYSTEM VALVE POSITION FOR NORMAL PLANT OPERATION CLOSED OPEN CONSEQUENCES OF VALVE FAILURE ASSUMED TOGETHER WITH DESIGN-BASIS (DBA) LOCA REMAINING ECCS COOLANT DELIVERY SYSTEMS LPCI (Cont'd)

Combustible gas control cooling water supply E12-F312A X This MO valve, if left in the incorrect position, could divert flow away from LPCI. However, position indication is provided in the main control room. HPCS + LPCS + 2 LPCI + partial LPCI A LPCI Loop A Head spray E12-F023 X This MO valve in an incorrect position (open) would be closed by an isolation signal if LPCI were activated. In addition, position indication is provided in the main control room, and the flow diverted would be sprayed into the RPV head. HPCS + LPCS + 2 LPCI +

partial LPCI A Loops B and C are identical to Loop A except for the following instances: 1) No heat exchanger bypass valve (E12-F048) exists for Loop C; however, it is provided for Loop B. 2) No combustible gas control cooling water cross-tie exists for Loop C. 3) No head spray line exists for either Loop B or C.

4) The following additional connections and valves exist on Loop C and not on Loops A or B Suppression pool cleanup suction lines E12-F303 E12-F402

X X These manual valves located in branch lines off the LPCI suction are also provided with a normally blind flanged connection. A spool piece can be added during plant shutdown to clean-up the suppression pool. Therefore, both the valve and blind flange would have to be incorrect before flow could be diverted.

HPCS + LPCS + LPCI A&B Water leg valves E12-F082

E12-F380 X

X Pressure switches are provided to alarm at low pressure if the water leg pumps are not maintaining the proper fill in the

lines. HPCS + LPCS + LPCI A

LSCS-UFSAR TABLE 6.3-6 TABLE 6.3-6 REV. 13 SINGLE FAILURES CONSIDERED FOR ECCS ANALYSIS

Assumed Failure (1) Remaining ECCS (2)

HPCS D/G LPCS + 3 LPCI + ADS (3)

LPCS D/G HPCS + 2 LPCI + ADS (3)

LPCI D/G HPCS + LPCS + LPCI + ADS(3) ADS HPCS + LPCS + 3 LPCI + 5 ADS valves

(1) Other postulated failures are not specifically considered because they result in at least as much ECCS capacity as one of the above assumed failures. (2) Systems remaining, as identified in th is table, are applicable to all non-ECCS line breaks. For a LOCA from an ECCS line break, the remaining systems are those listed for the recirculation line break, less the ECCS in which the break is assumed. (3) The analysis was performed assuming only 6 of the 7 ADS Valves were functional. This was done to support operation with one SRV out-of-service.

In the case of a single failure of the ADS, only 5 ADS valves were assumed.

LSCS-UFSAR TABLE 6.3-6a TABLE 6.3-6a REV. 17, APRIL 2008 ATRIUM-9B MAPLHGR Analysis Results Average Planar Exposure (GWd/MTU)

MAPLHGR (kW/ft) PCT (F)1 Local Cladding Oxidation (%)

2 0 13.5 1807 0.68 5 13.5 1792 0.63 10 13.5 1758 0.55 15 13.5 1709 0.47 20 13.5 1726 0.72 25 13.0 1686 0.59 30 12.5 1652 0.45 35 12.0 1640 0.45 40 11.5 1592 0.31 45 11.0 1557 0.24 50 10.5 1520 0.19 55 10.0 1474 0.15 60 9.5 1412 0.11 61.1 3 9.39 1396 0.10 64.3 9.07 -- -- 65 9.0 1384 0.16 Core average metal-water reaction is <0.16% at all exposures.

Source: EMF-2175(P) (Reference 16)

Footnotes:

1 All LOCA PCT evaluations are tracked by Nuclear Fuels and reported to the NRC. 2 Reference 32 documents that the peak lo cal cladding oxidation is changed to 0.79% due to limiting PCT change.

3 The exposure limit has been extend ed to 64.3 GWd/MTU with a MAPLHGR limit of 9.07 kW/ft (Reference 38). Note that the analyses that support the ATRIUM-9B exposure extension were actually performed for 65 GWd/MTU. However, the ATRIUM-9B fuel cannot be operated past 64.3 GWd/MTU (Reference 39).

LSCS-UFSAR TABLE 6.3-6i TABLE 6.3-6i REV. 17, APRIL 2008 ATRIUM-10 MAPLHGR Analysis Results Average Planar Exposure (GWd/MTU)

MAPLHGR (kW/ft) PCT (F) Local Cladding Oxidation

(%) 0 12.5 1729 0.48 5 12.5 1648 0.33 10 12.5 1567 0.21 15 12.5 1578 0.22 20 12.1 1546 0.19 25 11.7 1519 0.16 30 11.2 1493 0.14 35 10.8 1464 0.11 40 10.4 1428 0.09 45 9.9 1399 0.08 50 9.5 1365 0.07 55 9.1 1327 0.05 60 8.3 1243 0.03 65 7.4 1163 0.02 67 7.1 1130 0.02 Core average metal-water reaction is <<0.16% at all exposures.

Source: EMF-3231(P) (Reference 47)

Note:

1. ALL LOCA PCT evaluations are tracked by Nuclear Fuel Management and reported to the NRC.

LSCS-UFSAR TABLE 6.3-6j TABLE 6.3-6j REV. 17, APRIL 2008 Limiting ATRIUM-10 LOCA Analysis Break Characteristics and Results (Applied to Unit 1 and Unit 2)

Location Recirculation suction pipe Type / size Double-ended guillotine / 0.8 discharge coefficient Single failure Low-pressure coolant injection diesel generator Maximum MAPLHGR 12.5 (kW/ft) Peak cladding temperature 1729

(°F) Local cladding oxidation 0.48 (max %) Total hydrogen generated <<0.16

• (% of total hydrogen possible)

Source: EMF-3231(P) (Reference 47)

• Planar average MWR for the peak power plane is < 16% which indicates a CMWR significantly less than 0.16%.

LSCS-UFSAR TABLE 6.3-7 TABLE 6.3-7 REV. 14, APRIL 2002 SEQUENCE OF EVENTS FOR STEAMLINE BREAK OUTSIDE CONTAINMENT (The information in this table is historical; please refer to Appendix A from GE proprietary document GE-NE-208-21-1093, "Engineering Evaluation Requirements for the LaSalle County Station Units 1 and 2 SAFER-GESTR Loss of Coolant Accident Analysis with ECCS Relaxations," dated November 1993.)

TIME (sec) EVENT 0 Guillotine break of one main steamline outside primary containment.

~0.5 High steamline flow signal initiates closure of main

steamline isolation valve.

<1.0 Reactor begins scram.

5.5 Main steamline isolation valves fully closed.

~60 RCIC and HPCS would initiate on low water level (RCIC considered unavailable, HPCS assumed single failure, and therefore, may not be available).

~6 Safety relief valves open high vessel pressure. The valves open and close to maintain vessel pressure at approximately 1100 psi.

~300 Reactor water level above core begins to drop slowly due to loss of steam through the safety valves. Reactor

pressure still at approximately 1100 psi.

~1150 ADS auto initiates after 10 minute drywell pressure bypass timer plus the existing 2 minute initiation delay. Vessel depressurizes rapidly.

~1350 Low-pressure ECC systems initiated. Reactor fuel uncovered partially.

~1400 Core effectively reflooded and cladding temperature heatup terminated. No fuel rod failure.

LSCS-UFSAR TABLE 6.3-7a TABLE 6.3-7a REV. 15, APRIL 2004 Event Times for FANP Limiting Large Break LOCA 1.0 DEG Pump Suction SF-LPCS/DG for ATRIUM-9B Fuel Event Time (Seconds)

Initiate Break 0.0 Initiate Scram 0.6 Feedwater Flow Stops 0.5 MSIV Fully Closed 5.0 Low-Low Water Level 8.3 Low-Low-Low Water Level 9.5 Jet Pump Uncovers 10.8 Recirculation Suction Uncovers 14.7 Lower Plenum Flashes 17.1 HPCS Valve Starts to Open 13.0 HPCS Pump at Rated Speed 41.0 HPCS Flow Starts 41.0 LPCS Valve Starts to Open NA LPCS Pump at Rated Speed NA LPCS Flow Starts NA LPCI Valve Starts to Open 46.6 LPCI Pump at Rated Speed 60.0 LPCI Flow Starts 63.5 End of Blowdown (Rated Spray) 80.4 ADS Valve Opens 129.5 Start of Reflood 116.6 Core Reflood 125.2 Depressurization Ends

>150.0 Peak Cladding Temperature Occurs 125.2 Source: EMF-2174(P)

LSCS-UFSAR TABLE 6.3-7b TABLE 6.3-7b REV. 15, APRIL 2004 Event Times for FANP Limiting LOCA 1.1 ft 2 Pump Discharge SF-HPCS/DG for ATRIUM-9B Fuel Event Time (Seconds)

Initiate Break 0.0 Initiate Scram 0.6 Feedwater Flow Stops 0.5 MSIV Fully Closed 5.0 Low-Low Water Level 13.0 Low-Low-Low Water Level 15.4 Jet Pump Uncovers 18.4 Recirculation Suction Uncovers 28.9 Lower Plenum Flashes 34.3 HPCS Valve Starts to Open NA HPCS Pump at Rated Speed NA HPCS Flow Starts NA LPCS Valve Starts to Open 97.9 LPCS Pump at Rated Speed 65.0 LPCS Flow Starts 133.6 LPCS MFV Closed 147.2 LPCI Valve Starts to Open 97.9 LPCI Pump at Rated Speed 65.0 LPCI Flow Starts 144.0 LPCI MFV Closed (End of RELAX Calculation) 183.5 End of Blowdown (Rated Spray) 160.5 ADS Valve Opens 135.4 Start of Reflood 196.2 Core Reflood 203.1 Depressurization Ends

>250.0 Peak Cladding Temperature Occurs 203.1 Source: EMF-2175(P)

LSCS-UFSAR TABLE 6.3-8

SUMMARY

OF RESULTS OF SAFER/GESTR-LOCA ANALYSIS (10CFR50 Appendix K)

TABLE 6.3-8 REV. 17, APRIL 2008 LASALLE 1 & 2 SPECIFIC BREAK SPECTRUM Fuel Type: GE14 Break Size Break Location Single Failure 1st PCT 2nd PCT DBA Suction HPCS/DG 1019 / 1009 1258 / 1394 Suction LPCS/DG 1019 / 1009 1210 / 1301 Suction LPCI/DG 1019 / 1009 1214 / 1231 0.08 ft 2 Suction HPCS/DG 1032 993 0.1 ft 2 Suction HPCS/DG N/A 1446 MSLB Outside Containment HPCS/DG N/A 659 Limiting Break 0.08ft 2 Recirculation Suction Line Break Limiting ECCS Failure HPCS Diesel Generator Failure Peak Cladding Temperature (Licensing

Basis) 1460°F Maximum Local Oxidation

< 1.0% Core-Wide Metal-Water Reaction <0.1%

____________________________________ Notes: (1) First PCT is the PCT due to early boiling transition and lowering of water level before lower plenum flashing, and the second PCT is the

PCT after ECC systems inject. (2) Deleted (3) Core-Wide Metal-Water Reaction <0.1% for all cases.

(4) There is no early boiling transition for break areas less than 1.0 ft

2. Therefore, N/A is used for the first PCT and the value in the second PCT column is the peak PCT for the entire transient. (5) Based on Reference 42 for GE14 Fuel (6) The licensing basis PCT is in the most recent 10 CFR 50.46 report on each unit's NRC docket.

LSCS-UFSAR TABLE 6.3-8a Summary of Results of FANP Fuel (HUXY) LOCA Analysis*

(Sheet 1 of 2)

LaSalle 1 & 2 Specific Break Spectrum (Recirculation Line Break)

TABLE 6.3-8a REV. 15, APRIL 2004 Fuel Type: ATRIUM-9B Break Size Break Location Break Type ** Single Failure PCT (°F)* DBA Suction DEG LPCS/DG 1669 Suction DEG LPCI/DG 1661 Suction DEG HPCS/DG 1648 Discharge DEG LPCS/DG 1494 Discharge DEG LPCI/DG 1452 Discharge DEG HPCS/DG 1567 Suction DES LPCS/DG 1644 Suction DES LPCI/DG 1643 Suction DES HPCS/DG 1625 Discharge DES LPCS/DG 1505 Discharge DES LPCI/DG 1466 Discharge DES HPCS/DG 1567 80% DBA Suction DEG HPCS/DG 1636 Suction DES HPCS/DG 1621 Discharge DEG HPCS/DG 1565 Discharge DES HPCS/DG 1567 60% DBA Suction DEG LPCS/DG 1580 Suction DEG HPCS/DG 1582 Discharge DEG LPCS/DG 1474 Discharge DEG HPCS/DG 1562 Suction DES LPCS/DG 1625 Suction DES HPCS/DG 1615 Discharge DES LPCS/DG 1490 Discharge DES HPCS/DG 1564 ________________________

• Source EMF-2174(P)(Reference 17) ** For DEG breaks, the discharge coefficient and full break area are used in the analyses. For split breaks (DES), size is the fraction of twice pipe cross-section area.

The licensing basis PCT is in the most recent 10CFR 50.46 report on each unit's NRC docket.

LSCS-UFSAR TABLE 6.3-8a Summary of Results of FANP Fuel (HUXY) LOCA Analysis*

(Sheet 2 of 2)

LaSalle 1 & 2 Specific Break Spectrum (Recirculation Line Break)

TABLE 6.3-8a REV. 15, APRIL 2004 Break Size Break Location Break Type ** Single Failure PCT (°F)* 40% DBA Suction DEG HPCS/DG 1561 Suction DES HPCS/DG 1475 Discharge DEG HPCS/DG 1563 Discharge DES HPCS/DG 1567 1.6 ft 2 Suction N/A LPCS/DG 1491 Suction N/A HPCS/DG 1479 Discharge N/A LPCS/DG 1461 Discharge N/A HPCS/DG 1573 1.0 ft 2 Suction N/A LPCS/DG 1396 Suction N/A LPCI/DG 1431 Suction N/A HPCS/DG 1594 Discharge N/A LPCS/DG 1404 Discharge N/A LPCI/DG 1432 Discharge N/A HPCS/DG 1728 1.1 ft 2 Discharge N/A HPCS/DG 1737 1.2 ft 2 Discharge N/A HPCS/DG 1717 0.4 ft 2 Suction N/A LPCS/DG 1251 Suction N/A HPCS/DG 1387 Discharge N/A LPCS/DG 1363 Discharge N/A HPCS/DG 1611 0.1 ft 2 Suction N/A LPCS/DG 716 Suction N/A HPCS/DG 1317 Discharge N/A LPCS/DG 1035 Discharge N/A HPCS/DG 1429 ________________________

• Source EMF-2174(P)(Reference 17) ** For DEG breaks, the discharge coefficient and full break area are used in the analyses. For split breaks (DES), size is the fraction of twice pipe cross-section area.

The licensing basis PCT is in the most recent 10CFR 50.46 report on each unit's NRC docket.

LSCS-UFSAR TABLE 6.3-8b Summary of Results of FANP Fuel (HUXY) LOCA Analysis*

LaSalle 1 & 2 Specific Break Spectrum (Non-Recirculation Line Break)

TABLE 6.3-8b REV. 15, APRIL 2004 ATRIUM-9B Break Location Single Failure PCT (oF)Maximum Local Metal Water Reaction (%)

Core Average Metal Water Reaction (%) HPCS Line LPCS DG 1386 0.06

<1.0 HPCS Line ADS Valve 1019 0.00

<1.0 LPCI Line HPCS DG 1263 0.03

<1.0 LPCI Line LPCS DG 1188 0.02

<1.0 _______________________________

• Source EMF-2174(P) (Reference 17)

The licensing basis PCT is in the most recent 10CFR 50.46 report on each unit's NRC docket.

LSCS-UFSAR TABLE 6.3-9 (Sheet 1 of 4) TABLE 6.3-9 REV. 13 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION VALVE NUMBER BYPASS DEVICE (Continuous, Accident Conditions, or None) SYSTEM(S) AFFECTED a. 1VG001 Accident Conditions 1VG003 Accident Conditions 2VG001 Accident Conditions 2VG003 Accident Conditions SBGTS b. 1(2)VP113A Accident Conditions 1(2)VP113B Accident Conditions 1(2)VP114A Accident Conditions 1(2)VP114B Accident Conditions 1(2)VP053A Accident Conditions 1(2)VP053B Accident Conditions 1(2)VP063A Accident Conditions 1(2)VP063B Accident Conditions Primary containment chilled water coolers c. 1VQ038* Accident Conditions 1(2)VQ032 Accident Conditions 1(2)VQ035 Accident Conditions 1(2)VQ047 Accident Conditions 1(2)VQ048 Accident Conditions 1(2)VQ050 Accident Conditions 1(2)VQ051 Accident Conditions 1(2)VQ068 Accident Conditions 1VQ037* Accident Conditions 2VQ037* Accident Conditions 2VQ038* Accident Conditions Primary containment vent and purge system

d. 1(2)WR179 Accident Conditions 1(2)WR180 Accident Conditions 1(2)WR040 Accident Conditions 1(2)WR029 Accident Conditions RBCCW system

e. 1(2)B21 - F067A Accident Conditions 1(2)B21 - F067B Accident Conditions 1(2)B21 - F067C Accident Conditions 1(2)B21 - F067D Accident Conditions 1(2)B21 - F019 Accident Conditions 1(2)B21 - F016 Accident Conditions 1(2)B21 - F020 Continuous 1(2)B21 - F068 Continuous 1(2)B21 - F070 Continuous 1(2)B21 - F069 Continuous 1(2)B21 - F071 Continuous 1(2)B21 - F072 Continuous 1(2)B21 - F073 Continuous 1(2)B21 - F418A Continuous 1(2)B21 - F418B Continuous Main steam system

• These valves have thermal overload bypass for accident conditions from both Unit 1 and Unit 2 LSCS-UFSAR TABLE 6.3-9 (Sheet 2 of 4) TABLE 6.3-9 REV. 13

VALVE NUMBER BYPASS DEVICE (Continuous, Accident Conditions, or None) SYSTEM(S) AFFECTED

f. 1(2)B21 - F065A Continuous 1(2)B21 - F065B Continuous Main feedwater system

g. 1(2)E21 - F001 Continuous 1(2)E21 - F005 Accident Conditions 1(2)E21 - F011 Accident Conditions 1(2)E21 - F012 Accident Conditions LPCS system

h. 1(2)C41 - F001A Accident Conditions 1(2)C41 - F001B Accident Conditions SBLCS i. 1(2)G33 - F001 Accident Conditions 1(2)G33 - F004 Accident Conditions 1(2)G33 - F040 Continuous RWCU j. 1(2)E12 - F052A Accident Conditions 1(2)E12 - F064A Accident Conditions 1(2)E12 - F087A Accident Conditions 1(2)E12 - F004A Continuous 1(2)E12 - F047A Continuous 1(2)E12 - F048A Accident Conditions 1(2)E12 - F003A Continuous 1(2)E12 - F026A Accident Conditions 1(2)E12 - F068A Continuous 1(2)E12 - F073A Continuous 1(2)E12 - F074A Continuous 1(2)E12 - F011A Accident Conditions 1(2)E12 - F024A Accident Conditions 1(2)E12 - F016A Accident Conditions 1(2)E12 - F017A Accident Conditions 1(2)E12 - F027A Accident Conditions 1(2)E12 - F004B Continuous 1(2)E12 - F047B Continuous 1(2)E12 - F048B Accident Conditions 1(2)E12 - F003B Continuous 1(2)E12 - F068B Continuous 1(2)E12 - F073B Continuous 1(2)E12 - F074B Continuous 1(2)E12 - F026B Accident Conditions 1(2)E12 - F011B Accident Conditions 1(2)E12 - F024B Accident Conditions 1(2)E12 - F006B Continuous 1(2)E12 - F016B Accident Conditions 1(2)E12 - F017B Accident Conditions 1(2)E12 - F042B Accident Conditions 1(2)E12 - F064B Accident Conditions 1(2)E12 - F093 Continuous 1(2)E12 - F021 Accident Conditions 1(2)E12 - F004C Continuous RHR system

LSCS-UFSAR TABLE 6.3-9 (Sheet 3 of 4) TABLE 6.3-9 REV. 14, APRIL 2002

VALVE NUMBER BYPASS DEVICE (Continuous, Accident Conditions, or None) SYSTEM(S) AFFECTED 1(2)E12 - F052B Accident Conditions 1(2)E12 - F087B Accident Conditions 1(2)E12 - F099B Accident Conditions 1(2)E12 - F099A Accident Conditions 1(2)E12 - F008 Accident Conditions 1(2)E12 - F009 Accident Conditions 1(2)E12 - F040A Accident Conditions 1(2)E12 - F040B Accident Conditions 1(2)E12 - F049A Accident Conditions 1(2)E12 - F049B Accident Conditions 1(2)E12 - F053A Accident Conditions 1(2)E12 - F053B Accident Conditions

j. (cont'd) 1(2)E12 - F006A Continuous 1(2)E12 - F023 Accident Conditions 1(2)E12 - F027B Accident Conditions 1(2)E12 - F042A Accident Conditions 1(2)E12 - F042C Accident Conditions 1(2)E12 - F064C Accident Conditions 1(2)E12 - F094 Continuous RHR system

k. 1(2)E51 - F086 Accident Conditions 1(2)E51 - F022 Accident Conditions 1(2)E51 - F068 Continuous 1(2)E51 - F069 Continuous 1(2)E51 - F080 Accident Conditions 1(2)E51 - F046 Accident Conditions 1(2)E51 - F059 Accident Conditions 1(2)E51 - F063 Accident Conditions 1(2)E51 - F019 Accident Conditions 1(2)E51 - F031 Continuous 1(2)E51 - F045 Accident Conditions 1(2)E51 - F008 Accident Conditions 1(2)E51 - F010 Accident Conditions 1(2)E51 - F013 Accident Conditions 1(2)E51 - F076 Accident Conditions 1(2)E51 - F360 Accident Conditions RCIC system

l. 1(2)E22 - F004 Accident Conditions 1(2)E22 - F012 Accident Conditions 1(2)E22 - F015 Continuous 1(2)E22 - F023 Accident Conditions HPCS system

LSCS-UFSAR TABLE 6.3-9 (Sheet 4 of 4)

TABLE 6.3-9 REV. 13 VALVE NUMBER BYPASS DEVICE (Continuous, Accident Conditions, or None) SYSTEM(S) AFFECTED

m. 1(2)HG001A None 1(2)HG001B None 1(2)HG002A None 1(2)HG002B None 1(2)HG005A None 1(2)HG005B None 1(2)HG006A None 1(2)HG006B None 1(2)HG003 None 2(1)HG009 None 2(1)HG018 None 1(2)HG025 None 1(2)HG026 None 1(2)HG027 None 1(2)E12-F312A None 1(2)E12-F312B None Hydrogen recombiner system

LSCS-UFSAR 6.4-1 REV. 15, APRIL 2004 6.4 HABITABILITY SYSTEMS Habitability systems are designed to ensure habitability inside the control and the auxiliary electric equipment (AEE) rooms for both Units 1 and 2 during all normal and abnormal station operating conditions including the post-LOCA requirements, in compliance with Criterion 19 of 10 CFR 50, Appendix A. The habitability systems cover all the equipment, supplies, and procedures related to the control and auxiliary electric equipment so that control room operators are safe against postulated releases of radioactive materials, noxious gases, smoke, and steam.

Adequate sanitary facilities and medica l supplies are provided to meet the requirements of operating personnel during and after the accident. Adequate food and water storage in the control room are also provided for operators during the accident. In addition, the environment of the control and auxiliary electric equipment rooms is maintained in order to ensure the integrity of the contained safety-related controls and equipment, during all the station operating conditions.

6.4.1 Design Bases The design bases of the habitability systems upon which the functional design is established, are summarized as follows:

a. Independent HVAC systems are provid ed for the control room envelope and the auxiliary electric equipment room envelope which contains the remote shutdown panels and consists of auxiliary electric equipment room Unit 1 and Unit 2.

b. The control and auxiliary electric equipment rooms are occupied continuously on a year-round basis. The occupancy of the operating personnel is assured for a minimum period of 30 days, after a design-basis accident (DBA).

c. The habitability systems are designed to support a minimum of 5 people during normal and abnormal station operating conditions. The control room is supplied by three separate and independent breathing

air subsystems which are each comp rised of three 300 cubic foot air cylinders with appropriate pressure regulators, low pressure alarms and face masks. Two of these subsystems are for the Unit 1 and Unit 2 control room operators, while the third system supplies a manifold in the control room which can support the senior reactor operator, the control room technical adviser, and a third user as de emed necessary. All three subsystems are designed to provide a minimum of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of

breathing air for each user.

d. Sanitary facilities and medical supplies for minor injuries are provided for the control room. In addition, food and bottled water for a day (at least three meals) are stored in th e control room for a minimum of 10 people. This food is for use in accident conditions when access to the control room with food and water would be limited by dose rates.

LSCS-UFSAR 6.4-2 REV. 14, APRIL 2002 e. The radiological effects on the control and auxiliary electric equipment rooms that could exist as a conseque nce of any accident described in Chapter 15.0 are considered in the design of the habitability system.

f. The design includes provisions to preclude the effect of noxious gas and smoke from inside or outside the plant.

g. In addition to the subsystems me ntioned in (c) above, carts containing self-contained breathing air syst ems, e.g., Air-Paks, are located immediately outside the control room. These portable carts are intended for emergency use.

Each Air-Pak has a nominal 1/2 hour air breathing bottle which is rechargeable. These carts contai n adequate spares to provide necessary replacement bottles. A self-contained recharging system is provided for refilling expended air bo ttles on a timely basis to assure

an adequate air supply to emergency personnel.

At least ten total air paks are dedi cated for fire brigade use and are located where brigade members can readily obtain them. These air

packs are also rechargeable to assure adequate air supply to the fire brigade. h. The habitability systems are designed to operate effectively during and after a DBA such as a LOCA with the simultaneous loss of offsite power, design-basis earthquake, or failure of any one of the HVAC system components.

i. Radiation monitors, ammonia, and ionization detectors continuously monitor the air supply from the control room and AEE room outside

air inlets (see Figure 6.4-2). The detection of high radiation, ammonia, or smoke is alarmed in the control room. Related protection functions are simultaneously initiated for high radiation or smoke. Pressure differential indicators are provided in the control room and AEE room to monitor the pressure differential between control/AEE room and surrounding areas respectively.

Outdoor air and individual room temperature indicators are provided

for the control room HVAC system and the AEE room HVAC system.

j. Each control room and AEER HVAC subsystem has a supply air filter

unit that contains a charcoal filter unit, called the recirculation filter. Each filter unit consists of a pr e-filter and a normally bypassed charcoal filter. Upon detection of smoke in the return ductwork, the LSCS-UFSAR 6.4-3 REV. 14, APRIL 2002 charcoal filter is automatically placed in service. After validation of a high ammonia concentration in the air intake, the charcoal filter will be manually placed on line. Upon detection of high radiation, the Operator must manually place the charcoal filter on-line within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of detection to maintain the control room and AEER doses to within GDC 19 limits.

6.4.2 System Design 6.4.2.1 Definition of Control Room Envelope Habitability systems on LaSalle County Station consist of the control room envelope and the AEE room envelope. A ventilation barrier is provided between the two envelopes by supplying independent HVAC systems as described in Subsection

9.4.1.

6.4.2.2 Ventilation System Design

The detailed ventilation system design is presented in Subsection 9.4.1.

All the components are designed to perform their function during and after the design basis earthquake except for the electric heating equipment, which is supported to stay in position, but may not function.

All components are protected from internally and externally generated missiles. A layout diagram of the control and AEE room s, showing doors, corridors, stairways, shield walls and the equipment layout is given in Figure 6.4-1.

The description of controls, instruments, radiation, smoke, and ammonia monitors for the control/AEE room HVAC systems is incl uded in Subsections 7.2. and 7.3.4.3. The locations of outside air intakes and potential sources of radioactive and toxic

gas releases are indicated in Figure 6.4-2.

A detailed description of the emergency makeup air filter trains is presented in Subsection 6.5.1.

6.4.2.3 Leaktightness

The control room ductwork was leak te sted during start-up and the leakage through the isolation dampers was determined from vendor data. All cable pans and duct LSCS-UFSAR 6.4-4 REV. 14, APRIL 2002 penetrations are sealed. Approximately 1500 cfm of outside air is introduced in the control room envelope to maintain approximately 1/8 in. H 20 positive pressure with respect to adjacent areas and approximatel y 2500 cfm of outside air is introduced in the AEE room envelope to maintain approximately 1/8 in. H 20 positive pressure.

During isolation of the control room or AEE room, due to the pres ence of toxic gases in the intake stream, the outside air dampers are shut.

6.4.2.4 Interaction with Other Zones and Pressure-Containing Equipment The control room envelope is surrounded by the auxiliary building offices (elevation 768 feet). These offices are served by an independent HVAC system as described in Subsection 9.4.3. There is a ventilati on barrier between the control room and auxiliary building office HVAC systems th rough concrete wall construction and leaktight doors.

The control room envelope is isolated fr om the turbine building through leaktight double doors.

6.4.2.5 Shielding Design

The shielding for the control and AEE r ooms is designed so that the doses experienced by control room personnel during normal operation and during design-basis accidents are as low as reasonably achievable (ALARA). However, the main function of the shielding is to protect occu pants from the radiation associated with a LOCA.

During normal operation the control an d AEE rooms are shielded from radiation sources in reactor water, steam processing equipment, station vent stack, and in the calibration facility. The sources, shielding, areas affected, and the dose rates

are given in Table 6.4-1.

The design-basis accident which requires excessive radiation protection for the

control and AEE rooms is the LOCA. Th e radiation sources due to a LOCA are distributed throughout the containment and the environment surrounding the control and AEE rooms as specified in Chapter 15.0. The shielding design and doses are based on airborne, cloud, and plate out sources given in Table 6.4-2. The location of the sources is shown in Figure 6.4-3.

The shielding reduces the radiation dose ra tes inside the control room (from outside sources) to levels where the accumulated dose is a small fraction of the limit specified in Criterion 19 of 10 CFR 50 Appendix A.

The shielding arrangement for the contro l and AEE rooms is presented in Figure 6.4-1, the sources and accident doses are given in Table 6.4-2, and the LOCA LSCS-UFSAR 6.4-5 REV. 14, APRIL 2002 shielding model is shown in Figure 6.4-3. Exposure of control room personnel due to airborne radiation inside the contro l room is discussed in Chapter 15.0.

The shielding which protects the control and AEE rooms during normal operation is directly associated with the radiation sources, i.e is not part of the control and AEE rooms shielding, which provides additional radiation protection. Table 6.4-1 lists the sources, total shielding thickness, and calculated dose rates during normal

operation.

6.4.3 System Operational Procedures

During normal plant operation, the mixture of recirculated air and outside air for the control room HVAC system is filtered by high-e fficiency, water and fire resistant glass fiber filters. The contro l room HVAC system is started through a remote control switch located in the control room. The sequence of operation is given in Chapter 7.0.

To remove any noxious gases, odors, and smoke from the control room environs, a bank of charcoal absorber beds is provided with each control room air handling

equipment train. These charcoal beds, located downstream of high-efficiency filters, are normally bypassed. If noxious gases are detected in the control room environs (outside air intake), the control room HVAC system is manually put in the recirculation mode, by which the outside air intake dampers are closed and the recirculation air from the control room system is routed through the charcoal absorber banks by operation of the handswitch controls provided on the main control board.

On the smoke detection signal in the return duct, the supply air to the control room HVAC system is automatically routed through the charcoal absorber and annunciated on the main control board.

The operator may continue to route the system supply air through the charcoal abs orber for smoke removal, or depending on the condition of the outside air, may manually bypass the charcoal absorber and

purge the system with outside air. Prio r to manually placing the HVAC systems in purge, e.g., maximum outside and exhaust air, the operator shall align the supply air through the charcoal absorber.

In the event of high radiation detection fr om the outside air intake of the control room HVAC system, the radiation monito ring system automatically shuts off normal and maximum outside air supply, and maximum exhaust air to the system. The minimum outside air requirement is routed through the emergency makeup air filter train and fan (for removal of radioact ive particulates and iodine), before being supplied to the system.

Two emergency makeup air filter trains and fans are provided, each capable of handling minimum requirements of outside ai r for the system. In the event of high LSCS-UFSAR 6.4-5a REV. 13 radiation levels, each train is sized to process 4000 cfm of outside air, providing 1500 cfm to the control room HVAC system and 2500 cfm to the auxiliary electric equipment room HVAC system. Each train contains a supply air filter, which must LSCS-UFSAR 6.4-6 REV. 14, APRIL 2002 be placed on-line within the first four hours of an accident to maintain CR doses within GDC 19 values. The emergency make up air filter units are described in detail in Subsection 6.5.1.

6.4.4 Design Evaluation The control room HVAC system is designed to maintain a habitable environment and to ensure the operability of all the components in the control

room under all the station operating cond itions. The system is provided with redundant equipment to meet the single failure criteria. The redundant equipment is supplied with separate essential power sources and is operable during loss of offsite power. The power supply and control and instrumentation meet IEEE-279 and IEEE-308 criteria. All the HVAC equipment except heating

are designed for Seismic Category I.

The likelihood of an equipment fire affecting control room habitability is minimized because early ionization detection is assured, fire fighting apparatus is available, and filtration and purging capability are provided.

The following provisions are made to minimize fire and smoke hazards inside the control room and damage to nuclear safety- related circuits:

a. Most electrical wiring and equi pment are surrounded by, or mounted in metal enclosures.

b. The nuclear safety-related circuit s for redundant divisions (including wiring) are physically segregated by space or fire part itions to allow only isolated damage to electrical equipment.

c. Cables used throughout the co ntrol room are flame retardant.

d. Structural and finish materials (i ncluding furniture) for the control room and interconnecting areas have been selected on the basis of fire resistant characteristics. Structur al floors and interior walls are of reinforced concrete. Interior partitions incorporate metal, masonry, or

gypsum dry walls on metal joists. The control room ceiling, door frames, and doors are metallic. Wood trim is not used.

The air distribution in the control room is designed to supply air into the occupied area and exhaust through the cont rol panels. In the event of smoke or products of combustion in the control panels, the ionization detection system alerts the operator and automatically positions dampers to pass all the supply air delivered to the conditioned spaces through a normally bypassed absorber for smoke and odor absorption. A manual override is also provided for this function as well as the ability to introduce 100% ou tside air to purge the control room as may be necessary.

LSCS-UFSAR 6.4-7 REV. 14, APRIL 2002 Two redundant ammonia detectors are provided at each outside air intake duct to the control room HVAC system. Upon dete ction of ammonia in the outside air, a control room annunciator alarms. The intake dampers can be manually closed, and the control room HVAC system operated in 100% recirculation mode, thus routing the recirculating air through its charcoal absorbers.

Four radiation monitor channels (A, B, C, and D) are provided to detect high

radiation at each outside air intake to the control room HVAC system. These monitor channels alarm the control room upon detection of high radiation. The emergency makeup air filter trains, designed to remove radioactive particulates and absorb radioactive iodine from the minimum quantity of outside air, are automatically started upon high radiation signals from two-out-of-four radiation monitor channels. The four monitor channels are divided into two trip systems.

High radiation signals from Monitor channels A and B or C and D will start the emergency makeup filter train for each intake.

The emergency makeup air filter trains, recirculation filters, and control room shielding are designed to limit the occupational dose below levels required by

Criterion 19 of 10 CFR 50 Appendix A.

The introduction of the minimum quantity of outside air to maintain the control room and other areas served by the co ntrol room HVAC system at a positive pressure with respect to surrounding potentially contaminated areas, at all the station operating conditions except when the system is in recirculation mode, precludes infiltration of unfiltered air into the control room.

The physical location of two redundant outside air intakes provides the option of drawing makeup air to the control room HYAC system from either of them depending upon the lesser contamination level, during and after a LOCA. It is possible that due to outside wind direction after a LOCA, one of the air intakes may

not have any contaminants, while the othe r intake may have contaminants. The former may be utilized for makeup air in th e control room. This provides additional security towards maintaining the habitabilit y of the control room. The radiological

consequences due to radioactivity drawn into the control room or AEER are provided in section 15.6.5.5.

6.4.5 Testing and Inspection The control room HVAC system and its co mponents are thoroughly tested in a program consisting of the following:

a. factory and component qualification tests, b. onsite preoperational testing, and LSCS-UFSAR 6.4-8 REV. 13

c. onsite subsequent periodic testing.

Written test procedures establish minimum acceptable values for all tests. Test results are recorded as a matter of performance record, thus enabling early detection of faulty performance.

All equipment is factory inspected and tested in accordance with the applicable equipment specifications, codes, and qu ality assurance requ irements. System ductwork and erection of equipment is insp ected during various construction stages for quality assurance. Construction tests are performed on all mechanical components and the system is balanced for the design airflows and system operating pressures. Controls, interlocks , and safety devices on each system are cold checked, adjusted, and tested to ensure the proper sequence of operation.

The inplace HEPA and Charcoal filter testing acceptance criteria, and the decontamination efficiency for the emergency makeup unit comply with the values listed in Reg. Guide 1.52, Revision 2.

6.4.6 Instrumentation Requirements All the instruments and contro ls for the control room H VAC system are electric or pneumatic.

a. Each redundant control room H VAC system has a local control panel and each is independently controlled. Important operating functions are controlled and monitored from the main control room.

b. Instrumentation is provided to monitor important variables associated with normal operation. Instrument s to alarm abnormal conditions are provided in the control room.

c. A radiation detection system (instrument range 0.10 to 10,000 mr/hr.)

is provided to monitor the radiatio n levels at the system outside air intakes and inside the control room. A high radiation signal is

alarmed on the main control board.

d. The ammonia detection system is pr ovided to detect the presence of ammonia at outside air intakes. Ammonia detection is annunciated locally and in the main control room.

e. The ionization detection is provided in the outside and return air path from associated areas. Ionization detection is annunciated locally and on the main control board via the fire detection control panel.

LSCS-UFSAR 6.4-9 REV. 13

f. The control room HVAC system is designed for automatic environmental control with the manual starting of fans. The refrigeration equipment has a manual auto switch.

g. A fire protection water spray system is provided to each charcoal adsorber / absorber bed.

h. The various instruments of the co ntrol system are described in detail in Chapter 7.0.

i. The emergency makeup air filter train airflow rate and upstream HEPA filter differential pressure is transmitted to the main control board, recorded, and alarmed.

LSCS-UFSAR TABLE 6.4-1 TABLE 6.4-1 REV. 0 - APRIL 1984 DOSE RATES IN THE CONTROL AND AUXILIARY ELECTRIC EQUIPMENT (AEE) ROOMS DURING NORMAL OPERATIONS COMPONENT SOURCE RADIATION AREAS AFFECTED TOTAL SHIELD THICKNESS (INCHES)* CALCULATED DOSE RATE (mr/hr) RWCU pump Reactor water Direct gamma Control room AEE room 56 42 <0.1

<0.2 Skyshine Reactor steam Scattered gamma Control room Computer room 30 12 <0.1

<0.5 Main steam tunnel Reactor steam Direct gamma Control room AEE room 56 56 <0.5

<0.5 Station vent stack Off-gas Direct gamma Control room 40 <0.1 Feedwater pump Reactor steam Direct gamma Computer room 48 <0.1 Calibration facility C S-137 Direct gamma AEE room 24 <0.1

___________________________

• Thickness is given in inches of ordinary concrete with density of 140 pounds per cubic foot LSCS-UFSAR TABLE 6.4-2 TABLE 6.4-2 REV. 14, APRIL 2002 SHINE DOSE EXPERIENCED BY CONTROL ROOM PERSONNEL FOLLOWING LOSS-OF-COOLANT ACCIDENT

• SOURCE SOURCE DISTRIBUTION

SHIELD MODEL

ACTUAL SHIELD***

MAXIMUM DOSE RATE (R/hr) ACCUMULATED**

DOSE (rem)

1. Primary Containment a. Airborne b. Plate out 100% Nobles, 50% Halogens, 1% Particulates evenly distributed 100% on west side f 72 in. R.B. + 56 in. wall 72 in. R.B. + 56 in. wall

.6 x 10 -7 4 x 10-1 4 x 10-6 2 x 10-8 2. Reactor Building a. Airborne b. Plate out A c. Refueling floor plate out B 0.5% per day from 1 above evenly distributed 87% on west side f 13% on west side f 56 in. wall, 36 in. ceiling 56 in. wall, 48 in. ceiling 2 x 10-5 1 x 10-5 1.2 x 10-3 2.2 x 10-3 <1 x 10-2 7.3 x 10-1 3. SGTS Filter Unit 100% Halogens, particulates from 2a 36 in. R.B. +

56 in. wall 124 in. R.B. +

56 in. wall 2 x 10-9 8 x 10-6 4. Exhaust Clouds

a. External to stations

b. Airborne adjacent to control room Exhaust from 3, 100% Nobles, 10% Halogens 40 in. wall

24 in. wall

40 in. wall

24 in. wall 2 x 10-4 <1 x 10-7 8 x 10-3 <1 x 10-5 5. Control Room Air Intake Filter Unit Exhaust from 3 100% Nobles, 10% Halogens 24 in. ceiling 36 in. ceiling 2 x 10

-2 1.5 x 10-1 Total(rem):<9.4 X 10-1 leak rate of a 0.005/day <1 2 leak rate of 0.00635/day

• Due to sources outside the control room an average /Q was used to calculate the sources on the control room intake filter; more than 2/3 of this value is due to fumigation.

• • For calculation purposes, the duration of the LOCA was chosen to be 30 days. No credit was taken for containment spray or mixing in the secondary containment. The filter efficiency for the SGTS filter units is 99% for halogens and 99.95%, including filter bank bypass for particulates.

• • • Thickness of ordinary concrete with density of 140 pounds per cubic foot.

f 50% of the available halogens particulates are plated out as indicated.

Note 1: The doses due to radioactivity drawn into the Control Room and Auxiliary Electric Equipment Room are given in section 15.6.5.5. Note 2: This table was developed based upon the original source term used in the DBA LOCA analysis. The source term has been r evised, but this table is conservative; and the resultant dose is negligible compared to the GDC 19 limits.

LSCS-UFSAR 6.5-1 REV. 13 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5.1 Engineered Safety Feature (ESF) Filter Systems

The following filtration systems which ar e required to perf orm safety-related functions are provided:

a. Standby gas treatment system: Th is system is utilized to reduce halogen and particulate concentrations in gases leaking from the primary containment and which are potentially present in the secondary containment (reactor building) following the accident.

b. Control room and Auxiliary Electric Equipment Room (AEE Room)

HVAC emergency makeup air filter uni ts and recirculation filters: These systems are utilized to clean the outside air of halogen and

particulates, which are potentially present in outside air following an accident, before introducing air into the control room or AEER HVAC system.

6.5.1.1 Design Bases 6.5.1.1.1 Standby Gas Treatment System

a. The standby gas treatment system is designed to automatically start in response to any one of the following signals:

1. high pressure in Unit 1 or Unit 2 drywell,

2. low-water level in Unit 1 or Unit 2 reactor,

3. high radiation in exhaust air from over the fuel handling pools in the reactor building for either Unit 1 or Unit 2, 4. high radiation in the ventila tion exhaust plenum for reactor building for either Unit 1 or Unit 2, and

5. manual activation from the main control room.

b. The radioactive gases leaking from the primary containment and which are potentially present in the secondary containment after a LOCA are treated in order to remove particulate and radioactive and nonradioactive forms of iodine to limit the offsite dose to the guidelines of 10 CFR 100.

LSCS-UFSAR 6.5-2 REV. 15, APRIL 2004 c. The capability of one SGTS train to draw down the pressure in the secondary containment to -0.25 in. H 2O, and to maintain that secondary containment pressure, is verified on a staggered basis in accordance with Technical Specifications.

d. Any primary containment leakage (except that which is treated by the MSIV-ICLTM) will be contained within the secondary containment free air volume and will only reach the outside after passing through the SGTS. The secondary containm ent inleakage is determined by utilizing published leakage data for applicable building construction and incorporating known leakage values for piping, electrical, and duct penetrations at pressure control bo undaries. The SGTS flow rate is approximately equal to the total free air volume of the reactor buildings for both Units 1 and 2 evac uated at a rate of one per day.

The design flow rate through the SGTS also accounts for volumetric expansion of both reactor building air volumes due to temperature rises as equipment residual heat is released after ventilation and process system shutdown.

e. The secondary containment leakage is calculated in the following manner: 1. Assume laminar flow through small cracks, thus Q = K P where: P is the pressure differential across the secondary containment boundary; Q is the airflow rate (leakage); and K is the loss coefficient.

LSCS-UFSAR 6.5-3 REV. 13 2. Take a secondary containment leak rate of 4000 ft 3/min at still wind conditions with -0.25 inch (H 2O) differential pressure between the outdoor ambient condition and the in-containment pressure.

3. Assume the manufacturer's certified leak test results on the siding for the reactor building.

4. Accept the air leakage test results contained in "Conventional Building for Reactor Containment," NAA-SR-10100.

f. Two full-capacity standby gas treatment system equipment trains and associated dampers, ducts, instruments, and controls are provided.

g. Each train is sized and specified for the worst conditions, treating incoming air-steam mixtures saturated at 150° F containing fission products and incoming particulates released from primary containment at the Tech. Spec. le akage rate as determined in accordance with Regulatory Guid e 1.3 and T1D-14844. The design

nominal volume rate for each train was established at 4000 cfm.

h. Each equipment train contains the amount of charcoal required to absorb the inventory of fission pr oducts leaking from the primary containment, based on a one unit LOCA.

i. Each train is designed with the proper air heaters, demister, and prefilters needed to assure the optimum gas conditions entering the high-efficiency particulate air (HEPA) and charcoal filters. The air heater is sized to reduce air entering at 150° F, 100% relative humidity to a maximum 70% relative humidity. The demister is specified to remove any entrained moisture in the airstream.

j. A standby cooling air fan is provided for each equipment train to remove heat generated by fission product decay on the HEPA filters and charcoal adsorbers afte r shutdown of the train.

The standby cooling air fan is conservatively sized to remove approximately 7700 Btu/hr of heat (generated by instantaneous deposition of iodine, on a HEPA filter bank and charcoal adsorbers) with less than a 50° F rise in cooling air temperature. This will limit the air temperature in the SGTS to 200° F maximum to prevent possible desorption and fire. Charcoal desorption temperature is given in ORNL-NSIC-65. No credit is taken for equipment or environment heat sink. Reactor building cooling air is routed through the shutdown train and exhausted to the atmosphere via the plant vent stack.

LSCS-UFSAR 6.5-4 REV. 15, APRIL 2004 k. The SGTS exhibits a removal effi ciency of no less than 99% on radioactive and nonradioactive forms of iodine and no less than 99.95%, including filter bank bypa ss on all particulate matter 0.3 micron and larger in size. The particulate removal efficiency is predicated on the use of 99% particulate removal efficiency. The physical property of new charcoal purchased shall meet requirements specified in Table 5-1 of AN SI/ASME N509-1980. Performance requirement shall be as specified in Table 5-1 of ANSI/ASME N509-1980 with penetration less than 0.5% as tested per ASTM D3803-1989. The charcoal is cont ained in gasketless, all welded construction adsorbers to preclude bypass of the charcoal and to ensure the highest removal efficiencies on methyl iodine.

The exhaust air from each SGTS is routed through a seismically supported duct and is an elevated release at an elevation of 1080 feet

above mean sea level, approximately 186 feet 8 inches above the highest structure. The discharge air velocity from the SGTS vent exhaust pipe is approximately 1270 fpm. This high point release provides effluent dispersion ratios sufficient to meet this requirement of 10 CFR 100.

l. The SGTS is designed with redundancy to meet single failure criteria.

m. The power supplies meet IEEE 308 criteria and ensure uninterrupted operation in the event of loss of no rmal a-c power. The controls meet IEEE 279.

n. The SGTS is designed to Seismic Category I requirements.

o. The SGTS is designed to permit pe riodic testing and inspection of the principal system components described in the following subsections.

6.5.1.1.2 Emergency Ma keup Air Filter Units:

a. The emergency makeup air filter unit is designed to start automatically and provide outside air to the control room and auxiliary electric equipment room HVAC system s in response to any one of the following signals:

1. high radiation signal from the radiation monitors installed in outside air intake louvers for the control room and auxiliary

electric equipment room HVAC systems; and

2. manual activation from the main control room.

b. The T1D-14844 source model in conjunction with approved methods is used to calculate the quantity of activity released as a result of an LSCS-UFSAR 6.5-5 REV. 14, APRIL 2002 accident and to determine inlet concentrations to the emergency makeup air filter train. See section 15.6.5.5 for additional details.

c. The capacity of the emergency makeup air filter units is based on the air quantity required to maintain th e rooms served by the control room HVAC and auxiliary electric equipment room HVAC systems at a minimum of 1/8 inch H2O positive pressure with respect to adjacent areas. d. Two full capacity emergency makeup air filter units and associated dampers, ducts, and co ntrols are provided.

e. Each unit is designed with the proper air heaters, demister, and prefilters needed to assure the optimum air conditions entering the high-efficiency particulate air (HEPA) and charcoal filters.

f. The emergency makeup air filter unit exhibits a removal efficiency of no less than 95% on radioactive and nonradioactive forms of iodine and no less than 99.95%, including filter bank bypass on all particulate matter 0.3 micron and larger in size.

g. The emergency makeup air filter unit is designed to meet single failure criteria.

h. The power supplies meet IEEE 308 criteria and ensure uninterrupted operation in the event of loss of no rmal a-c power. The controls meet IEEE 279.

i. The emergency makeup air filter units are designed to Seismic Category I requirements.

j. The emergency makeup air filter units are designed to permit periodic testing and inspection of principal system components described in the following subsections.

k. Each control room and AEER HVAC subsystem has a supply air filter unit that contains a charcoal filter unit, called the recirculation filter. Each filter unit consists of a pr e-filter and a normally bypassed charcoal filter. Upon detection of smoke in the return ductwork, the charcoal filter is automatically placed in service. After validation of a high ammonia concentration in the air intake, the charcoal filter will be manually placed on line. Upon detection of high radiation, the Operator must manually place the charcoal filter on-line within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of detection to maintain the control room and AEER doses to

within GDC 19 limits.

LSCS-UFSAR 6.5-6 REV. 15, APRIL 2004 6.5.1.2 System Design

6.5.1.2.1 Standby Gas Treatment System

a. The schematic design of the SGTS is shown in Drawing No. M-89. Nominal size of principal system components are listed in the Table 6.5-1.

b. The SGTS is automatically or ma nually started to treat air exhausted from either reactor building. Two completely redundant parallel process systems are provided, each having a nominal capacity of 4000 ft3/min (at 150° F).

As indicated on the schematic in Drawing No. M-89, each process system may be considered as an installed spare. The process systems have separate equipment trains, isolation valves, power feeds, controls, and instrumentation. Two full capacity redundant standby gas treatment system equipment trains are provided. One equipment train is located in the Unit 1 reactor building and the other equipment train is located in the Unit 2 reactor building. The suction and discharge side of both trains are headered together so that either of the trains can treat the air from both reactor buildings. Each SGTS equipment train and damper on the suction and discharge side of corresponding trains are powered by electrical essential Division 2 of the related unit. Either secondary containment isolation power signal starts both equipment trains and activates both alarms in the main control room. The operator then shuts down one of the standby gas treatment system equipment trains af ter ensuring that at least one of the two redundant trains is operating.

The intake connections used for the standby gas treatment system are located on reactor building Units 1 and 2 floor elevation 820 feet 0 in. No redundant duct system component is located within 20 feet of its counterpart in areas where credible internal missiles or pipe whips might compromise redundancy.

c. Each SGTS has the following components:

1. A primary fan for inducing the air from the spaces listed previously and discharging it through the filter train and common discharge pipe for elevated release to atmosphere. The fan performance and motor selection are based on the worst

environmental conditions inside the reactor building. The flow and pressures are listed in Table 6.5-1.

LSCS-UFSAR 6.5-6a REV. 15, APRIL 2004

2. A standby cooling air fan is sized to dissipate heat generated by fission product decay on the filters. The 200 ft 3 /min flow capacity limits the maximum temperature in the train to 200° F

for 150° F entering air temperature.

The fan is used only after train shutdown and when the electric heater and primary fan are not operating.

3. A demister which removes any entrained water droplets and moisture to minimize water loading on the prefilter. The LSCS-UFSAR 6.5-7 REV. 15, APRIL 2004 demister meets qualification requirements similar to those in MSAR 71-45 and is in UL Class I.

4. A single stage electric heater is sized to reduce the humidity of the airstream to at least 70% relative humidity for the worst inlet conditions. An analysis of heater capabilities for various

entering saturated air conditions ranging from 65° F to 150° F yields a peak heating requirement of 47,000 Btu/hr at 95° F entering air temperature. A 23-kW heater is provided.

5. A prefilter, UL listed, all-gla ss media, exhibiting no less than 85% efficiency based on ASHRAE atmospheric dust spot test.

6. A high-efficiency particulate air (HEPA) filter, water resistant, capable of removing 99.95% minimum of particulate matter

which is 0.3 micron or larger in size. The filter is designed to be fire resistant. Four, 1000-ft /min elements are provided. All elements are fabricated in accordance with Military Specification MIL-F-51068, MI L-F-51079 and UL-586. The elements are size 5 with IIB element frame material. Gasket material will be SCE 43 per ASTM D1056. Testing of the HEPA filter banks is described in Subsection 6.5.1.4.

7. A charcoal adsorber capable of removing not less than 99% of radioactive and nonradioactive forms of iodine. The charcoal

adsorber is a gasketless, welded seam type, filled with impregnated coconut shell charcoal. The bank holds a total of approximately 5800 pounds of charcoal.

The charcoal specification requires an ignition temperature test

and a methyl iodide test on each batch of charcoal supplied. In addition, model tests or previous qualification test data were required to demonstrate the e ffectiveness of the bed design before construction of the actual beds. Test data proving uniform packing density of charcoal in beds was also required.

Ten test canisters are provided for each adsorber. These canisters contain the same depth of the same charcoal as is in the adsorber. The canisters are mounted, so that a parallel flow path is created between each canister and the adsorber. Periodically one of the canisters is removed and laboratory LSCS-UFSAR 6.5-8 REV. 14, APRIL 2002 tested to reverify the adsorbent efficiency. Two deluge valves in parallel connected to the station fire protection system are mounted outside of the charcoal adsorber. The charcoal bed is provided with a high temperature detector. The detector sensing high adsorber temperature will actuate an alarm in the main control room. High temperature alarms are nominally set

at 310 °F. Manual charcoal deluge valves are operated locally and then solenoid operated valves are operated from the control room. The normally manual closed isolation valves upstream of the deluge valve in all cases require local actions to initiate water flow.

8. A high efficiency particulate filter identical to the one described in item 6 previously is provided to trap charcoal fines which may be entrained by the airstream.

d. Flow control valves are utilized upstream to regulate flow through the train. The train upstream static pressure will fluctuate between +1 and -1 inches water gauge.

e. Full-size access doors to each filter compartment are provided in the equipment train housing. Access d oors are provided with transparent portholes to allow inspection of components without violating the train integrity.

f. The housing is of all welded construction, heavily reinforced.

g. Interior lights with external light switches, are provided between all train components to facilitate inspec tion, testing, and replacement of components.

h. Filter frames are in accordance with recommendations of Section 4.3 of ORNL-NSIC-65.

i. The height of release of the standby gas treatment system vent to the atmosphere is at elevation 1080 f eet (186 feet 8 inches above the highest structure on the station).

6.5.1.2.2 Emergency Ma keup Air Filter Units

a. The emergency makeup air filter units work in conjunction with the control room and auxiliary electric equipment room HVAC system as described in Subsection 9.4.1. The nominal size of principal system components is listed in Table 9.4-1.

LSCS-UFSAR 6.5-9 REV. 14, APRIL 2002

b. In the event of high radiation detection in the outside air intakes of the control room HVAC system, the radiation monitoring system automatically shuts off normal outside air supply to the system and routes the outside air through the emergency makeup air filter train and fan (for removal of radioactive particulates and iodine), before

being supplied to the control room and auxiliary electric equipment room HVAC systems.

c. Two emergency makeup air filter trains and fans are provided, each capable of handling 4000 cfm nomi nal of outside air, providing approximately 1500 cfm to the control room HVAC system and approximately 2500 cfm to the auxiliary electric equipment room

HVAC system.

d. Each emergency makeup air filter unit is comprised of the following components in sequence:

1. A demister which removes any entrained water droplets and moisture to minimize water drop lets and water loading of the prefilter. The demister will me et qualification requirements similar to those in Mine Safe ty Appliance Research (MSAR) report 71-45 and will be UL Class I.

2. A single stage electric heater, sized to reduce the humidity of the airstream to at least 70% relati ve humidity for the worst inlet conditions. An analysis of heater capacities for various entering saturated air conditions ranging from - 10° F to 95° F yields a peak heating requirement of 60, 000 Btu/hr at 95° F. A 20-kW heater is provided.

3. A prefilter, UL listed, all gla ss media, exhibiting no less than 85% efficiency based on ASHRAE Standard 52.2 method of testing.

4. A high-efficiency particulate (HEPA) filter, water resistant, capable of removing 99.97% minimum of particulate matter which is 0.3 micron or larger in size. The filter is designed to be fire resistant, as may be required after consideration of heat generation from postulated deposit of fission products. Four 1000 cfm elements are provided. A ll elements are fabricated in accordance with Military Sp ecification MIL-F-51068, MIL-F-51079, and UL-586.

LSCS-UFSAR 6.5-10 REV. 15, APRIL 2004 5. A charcoal adsorber capable of removing not less than 95% of radioactive forms of iodine is provided. The charcoal absorber is an all welded gasketless type fi lled with impregnated coconut shell charcoal. The charcoal adsorber beds hold approximately 650 pounds of charcoal.

The bed dimensions are so designed that the air has at least 0.25 seconds of residence time through the charcoal. The physical property of new charcoal purchased shall meet requirements specified in Tabl e 5-1 of ANSI/ASME N509-1980. Performance requirement shall be as specified in Table 5-1 of ANSI/ASME N509-1980 with penetration less than 0.5% as tested per ASTM D3803-1989.

The charcoal specification requires an ignition temperature test and a methyl iodine test on each batch of charcoal supplied.

Ten test canisters are provided for the charcoal adsorber. These canisters contain the same depth of the same charcoal as in the charcoal adsorber. The canisters are so mounted that a parallel flow path is created between each canister and the charcoal adsorber. Thus, the charcoal in the canisters is subjected to the same contaminants as the charcoal in the bed. Periodically, one of the canisters is removed and laboratory tested to reverify the absorbent efficiency.

Two deluge valves connected to the station fire water system are mounted adjacent to each charcoal adsorber. Manual charcoal deluge valves are operated locally. The normally closed manual isolation valves upstream of the solenoid deluge valve, in all cases, require local actions to in itiate water flow. The deluge system will spray the adsorber compartment and thereby

precluding the chance of an adsorber fire.

6. A high-efficiency particulate filter identical to the one described in item 4 is provided to trap charcoal fines which are entrained by the airstream.

7. A fan induces the air from the intake louvers and the makeup air filter train and discharges it to the suction side of the control room air handling equipment train. The fan performance is

based on the maximum density and worst pressure condition, when it is inducing -10° F air from the outdoors and the makeup air filter train, containing filters which operate at no less than LSCS-UFSAR 6.5-11 REV. 15, APRIL 2004 twice their clean pressure drop.

8. Full size access doors adjacent to each filter are provided in the equipment train housing. Access doors are provided with transparent portholes to allow inspection and maintenance of components without violating the train integrity. Spacing between filter sections is bas ed on ease of maintenance considerations.

9. The housing is an all welded construction, heavily reinforced, and built to tight leakage requirements.

10. Interior lights with external light switches are provided between all train components to facilit ate inspection, testing, and replacement of components.

6.5.1.2.3 Supply Air Filter Unit Recirculation Filter Each control room and AEER HVAC subsyste m has a supply air filter unit that contains a charcoal filter unit, called the recirculation filter. Each filter unit consists of a pre-filter and a normally bypassed cha rcoal filter. Upon detection of smoke in the return ductwork, the charcoal filter is automatically placed in service. After validation of a high ammonia concentration in the air intake, the charcoal filter will be manually placed on line. Upon detectio n of high radiation, the Operator must manually place the charcoal filter on-line within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of detection to maintain the control room and AEER doses to within GDC 19 limits.

6.5.1.3 Design Evaluation

6.5.1.3.1 Standby Gas Treatment System

The Standby Gas Treatment System (SGTS) is designed to preclude direct exfiltration of contaminated air from either reactor building following an accident or abnormal occurrence which could result in abnormally high airborne radiation in the

secondary containment. Equipment is powered from essential buses and all power circuits will meet IEEE 279 and IEEE 308. Redundant components are provided

where necessary to ensure that a single failure will not impair or preclude system operation. A standby gas treatment system failure analysis is presented in Table 6.5-2. An analysis was performed to determine the SGTS equipment capacity, based on the total inleakages to the secondary containm ent for both Units 1 and 2, while all the areas in the secondary containment are maintained at 0.25-inch water gauge negative. The secondary containment ai r pressure will begin to increase and approach 0 in. H 2O (i.e., rises from initial -0.25 in. H 2O to 0 in. H 2O) due to LSCS-UFSAR 6.5-11a REV. 15, APRIL 2004 inleakage into the secondary containment during post-LOCA and at times when SGTS is started. The secondary containment air pressure begins to decrease exponentially at the time the SGTS reaches its full capacity. As required by the Technical Specifications, within 300 seco nds the secondary containment pressure will be reduced to -0.25 in. H 2O with the SGTS at full ca pacity (see Figure 6.3-80). During this time period, the pressure di fference is always negative (assuming 0 wind speed); therefore, only inleakage from the outside atmosphere can occur.

LSCS-UFSAR 6.5-12 REV. 17, APRIL 2008 6.5.1.3.2 Emergency Ma keup Air Filter Units

The emergency makeup air filter units work in conjunction with the control room and auxiliary electric equipment room HVAC systems to maintain habitability in the control room and auxiliary equipment rooms. The design evaluation is given in Subsection 6.4.4.

6.5.1.4 Tests and Inspections

6.5.1.4.1 Standby Gas Treatment System

a. The SGTS and its components are thoroughly tested in a program consisting of the following:

1. factory and component qualification tests,

2. onsite preoperational testing, and

3. onsite periodic testing.

Written test procedures establish minimum acceptable values for all tests. Test results are recorded as a matter of performance record, thus enabling early detection of depleted performance.

b. The factory and component qualification tests consist of the following:

1. equipment train housing - a leak test +2.0 psig internal pressure, and magnetic particle or liquid penetrant testing per Section III of ASME Boiler and Pressure Vessel Code of all welds which could cause bypass le akage around HEPA filters or adsorber beds;

2. demister - qualification test or objective evidence to demonstrate compliance with specified design criteria;

3. HEPA filters - elements tested individually by applicable inspection and testing methods;

4. HEPA filter frames - liquid penetrant test per ASME B&PV Code Section III of all welds which could cause bypass leakage around HEPA filters.

5. adsorbent beds - model test of bed or objective evidence to demonstrate flow pressure characteristics, channeling effects; LSCS-UFSAR 6.5-13 REV. 15, APRIL 2004

6. adsorbent - qualification test s for ignition temperature and methyl iodine removal efficiency test;

7. fans - tested in accordance with the latest revision of AMCA Standard 210 "Air Moving and Conditioning Association Test Code for Air Moving Devices," to establish characteristic curves, etc.; 8. heater - uniform temperature test, high temperature cutout test, and adjacent equipment temperature test;

9. prefilter - objective evidence or certification that ASHRAE efficiency specified is attained; and

10. valves - shop tests demonstrating leaktightness, closure times.

c. The onsite preoperational tests are discussed in Section 14.1 of the FSAR. d. Onsite periodic testing - Operating personnel are trained and required to make surveillance checks. Th ese checks shall include visual inspection and periodically running the equipment trains for performance testing as outlined in the Technical Specifications.

6.5.1.4.2 Emergency Ma keup Air Filter Units

a. The emergency makeup air filter unit and its components were thoroughly tested in a program consisting of the following:

1. factory and component qualification tests, 2. onsite preoperational testing, and

3. onsite subsequent periodic testing.

Written test procedures establish minimum acceptable values for all tests. Test results are recorded as a matter of performance record, thus enabling early detection of faulty performance.

b. The factory and component qualification tests consisted of the following:

1. Filter Train Housing LSCS-UFSAR 6.5-14 REV. 17, APRIL 2008 a) leak test at design internal pressure, and b) magnetic particle or liquid penetrant testing per Section III of ASME Boiler and Pressure Vessel Code of all welds which could cause bypass leakage around HEPA filters or

absorber bed.

2. Demister qualification test or objective evidence to demonstrate compliance with specified design criteria.

3. Prefilter objective evidence or certification that ASHRAE efficiency

specified were attained.

4. HEPA Filters elements tested individually in accordance with applicable inspection and testing methods.

5. HEPA Filter Frames liquid penetrant testing per ASME B&PV Code Section III of all welds which could cause bypass le akage around HEPA filters or adsorber bed.

6. Adsorbent Beds model test of bed or objective evidence to demonstrate flow pressure characteristics, channeling effects.

7. Adsorbent qualification tests for ignition temperature and methyl iodine removal efficiency test.

8. Fans were tested in accordance with the latest revision of AMCA Standard 210 "Air Moving and Conditioning Association Test Code for Air Moving Devices," to establish characteristic curves, etc.

LSCS-UFSAR 6.5-15 REV. 14, APRIL 2002

9. Heater a) uniform temperature test, b) high-temperature cutout test, and c) adjacent equipment temperature test.

10. The onsite preoperational testin g as described in Chapter 14.0 of the FSAR.

11. Onsite subsequent periodic testing as described in the Technical Specifications.

6.5.1.5 Instrumentation Requirements

a. Differential pressure indicators are provided to measure the pressure drop across each filter. Pressure differential across the upstream HEPA filter is transmitted to the main control board, recorded, and alarmed on high-pressure differential.

b. Each adsorber bed is provided with high-temperature detectors. The temperature detector actuates an alarm in the control room when the increase in adsorber temperature is beyond a preset value.

c. Manual charcoal deluge valves ar e operated locally. The normally closed manual isolation valves upstre am of the solenoid deluge valve, in all cases, require local actions to initiate water flow. The deluge system will spray the adsorber compartment and thereby precluding the chance of an adsorber fire.

d. All power-operated isolation valves are supplied with position switches to provide positive indication on the main control board.

e. High-temperature cutouts are prov ided as an integral part of the single stage electric heaters. Loca l temperature indication is provided upstream and downstream of the electric heaters.

f. Flow signals are transmitted to the main control board for indication recording and are used as an input to a flow control valve provided upstream of each equipment train.

g. Remote manual operation is provided on the main control board for each fan, and each deluge valve.

LSCS-UFSAR 6.5-16 REV. 14, APRIL 2002 6.5.1.6 Materials

a. All component material is capable of a service life of 40 years normal operation plus 6 months post-LOCA at the maximum cumulative radiation exposure without any adverse effects on service, performance, or operation. A ll materials of construction are compatible with the radiation exposure set forth. This includes but is not limited to all metal components , seals, gaskets, lubricants, and finishes, such as paints, etc. The integrated dose following the once-in-a-lifetime post-LOCA, uses the valu es given in UFSAR Section 3.11.

b. Care is taken to avoid the use of any compounds or other chemicals during fabrication or production that contain chlorides or other constituents capable of inducing stress corrosion in stainless steels which are used in the adsorber bed.

c. Pressure and temperature - All components, including the housings, shall be designed in accordance with the applicable pressure and temperature conditions.

d. All filter unit gaskets and seal pads are closed-cell, ozone resistant, oil-resistant neoprene or silicone-rubber sponge, Grade SCE-43 in accordance with ASTM D1056.

e. Only adhesives as listed and approved under AEC Health and Safety Bulletin 306, dated March 31, 1971, covering Military Specification MIL-F-51068C, dated June 8, 1970, an d all the latest amendments and modifications are used.

f. The organic compounds included in the filter train are as follows:

1. charcoal;

2. the binder in the HEPA filter media (the total weight of media per filter element is approximat ely 4 pounds, or a total of 32 pounds per equipment train);

3. adhesive used in HEPA filter s - approximately 1 liquid quart of fire-retardant neoprene adhesive is used to manufacture each HEPA filter;

4. neoprene gaskets used on HEPA filters and o-rings are used on the charcoal filter sample canisters; and LSCS-UFSAR 6.5-17 REV. 13

5. the binder in the glass pads used in the demister section (this is a phenolic compound).

6.5.2 Containment Spray Systems The containment spray systems are descri bed in Section 6.3. The containment spray systems are not required for fissions product removal.

6.5.3 Fission Product Control System

The standby gas treatment system (SGTS) is used to control the cleanup of fission products from the containment following an accident and is described in detail in Subsection 6.5.1.

6.5.4 Ice Condenser as a Fi ssion Product Cleanup System

Not applicable.

LSCS-UFSAR TABLE 6.5-1 (SHEET 1 OF 4) TABLE 6.5-1 REV. 13 STANDBY GAS TREATMENT SYSTEM COMPONENTS

NAME OF EQUIPMENT TYPE, QUANTITY AND NOMINAL

CAPACITY (per component)

A. Filter Unit

1. Equipment Numbers 1VG01S, 2VG01S

2. Type Package

3. Quantity 2

4. Components of Each Unit

a. Fan

Type Centrifugal

Quantity 1 Drive Direct Capacity (ft 3/min) 4000 (nominal)

Static Pressure (in. H 2O) 14.8 b. Demister

Type Impingement Quantity 1 Bank with 4 elements Static resistance

clean (in. H 2 O) 0.95 dirty (in. H 2 O) 1.7

c. Heater

Type Electric, sheathed, single stage

LSCS-UFSAR TABLE 6.5-1 (SHEET 2 OF 4) TABLE 6.5-1 REV. 17, APRIL 2008 NAME OF EQUIPMENT TYPE, QUANTITY AND NOMINAL

CAPACITY (per component)

Quantity 1 Capacity (kW) 23

Accessories Overload cutout

d. Prefilter

Type High Efficiency Quantity 1 Bank With 4 Elements Efficiency (per ASHRAE) Dust Spot Test) 90%

Static resistance clean (in. H 2 O) 0.35 dirty (in. H 2 O) 1 e. HEPA Filters Type Absolute High Efficiency

Quantity 4 Elements per Bank. Two Banks per Train Media Glass Fiber, Waterproof, Fire Resistant

Bank Efficiency (% with 0.3 micron particles) 99.97 (Purchased) 99.95 (Operational Requirement)

Static Resistance clean (in. H 2 O) 0.7 dirty (in. H 2 O) 2 LSCS-UFSAR TABLE 6.5-1 (SHEET 3 OF 4) TABLE 6.5-1 REV. 15, APRIL 2004 NAME OF EQUIPMENT TYPE, QUANTITY AND NOMINAL

CAPACITY (per component)

f. Charcoal Adsorber Bed Type Vertical gasketless

Quantity 8 - 8 in. thick Media Impregnated Charcoal

Iodine Removal Efficiency (%) 99 (Operational Requirement) 99 (Operational Requirement)

Quantity of Media (lb) 5800

Depth of Bed (in.) 8 Residence Time for 8 in. bed (sec) 2.0

Static Resistance (in. H 2O) 4.6 g. Standby Cooling Air Fan Type Centrifugal Quantity 1

Drive Direct Capacity (ft 3/min) 200 Static Pressure (in. H 2O) 5 LSCS-UFSAR TABLE 6.5-1 (SHEET 4 OF 4) TABLE 6.5-1 REV. 13

NAME OF EQUIPMENT TYPE, QUANTITY AND NOMINAL

CAPACITY (per component)

B. Secondary Containment Isolation

Dampers

1. Equipment Numbers 1VQ037, 1VQ038 2VQ037, 2VQ038 1VR04YA&B, 1VR05YA&B 2VR04YA&B, 2VR05YA&B

2. Type Special

3. Quantity 8

4. Operator Air Cylinder

5. Diameter (in.) 72 LSCS-UFSAR TABLE 6.5-2 TABLE 6.5-2 REV. 0 - APRIL 1984 STANDBY GAS TREATMENT SYSTEM EQU1PMENT FAILURE ANALYSIS COMPONENT FAILURE FAILURE DETECTED BY ACTION Primary Fan Motor Burnout, Drive Shaft Break, etc. Flow Monitor - Low-Flow Switch Main Control Board Alarm. Redundant train started after its isolation valves are positioned properly. Operating train is then shut down. Electric Heating Coil Element Overheat High Temperature

Protection Circuit on Coil Main Control Board Indication. Redundant train started after its isolation valves are positioned properly. Operating train is then shut down.

Standby Cooling

Fan No Startup Results In High Charcoal Adsorber Temperature Temperature Switch If temperature switch trips, then alarm sounds in main

control room (Station operator manually actuates deluge valves). Redundant train started after its isolation valves are positioned properly. Operating train is then shut down.

Flow Control

Valve Fails Open Flow Monitor -

High-Flow Switch Main Control Board Alarm. Redundant train started after its isolation valves are positioned properly. Operating train is then shut down.

Flow Control

Valve Fails Shut Flow Monitor - Low-Flow Switch Main Control Board Alarm. Redundant train started after its isolation valves are positioned properly. Operating train is then shut down. Isolation Valve Fails Open None - Redundant valves or backflow dampers provided as

required.

Fails Shut Flow Monitor - Low-

Pressure Switch Main Control Board Alarm. Redundant train started after its isolation valves are positioned properly. Operating train is then shut down. HEPA Filter High Particulate

Loading High P Alarms Main Control Board Alarm. Redundant train started after its isolation valves are positioned properly. Operating train is then shut down.

Duct Destruction by

Equipment Missile or Flailing Pipe Flow Monitor - High-Flow Switch Main Control Board Alarm. Redundant train started after its isolation valves are positioned properly. Operating train is then shut down. Deluge Valve Fails Closed No Detection None required, two valves provided to flood bed.

LSCS-UFSAR 6.6-1 REV. 17 APRIL 2008 6.6 INSERVICE INSPECTION OF ASME CODE CLASS 2 AND 3 COMPONENTS 6.6.1 Components Subject to Examination

All ASME Class 2 components (pressure vessels, piping, pumps, and valves) are inservice inspected according to ASME, B&PVC, Section XI, Subsection IWC, with appropriate addendum(s). The main steamlines (four) are inspected from the first outside containment isolation valve to the turbine stop valves. Inspection requirements are the same as for ASME Class 2 components.

All ASME Class 3 components (pressure vessels, piping, and valves) are inservice inspected according to ASME, B&PVC, Section XI, Subsection IWD, with appropriate addendum(s).

6.6.2 Accessibility

The design and arrangement of the ASME Class 2 and ASME Class 3 piping, pump, and valve components have been made acce ssible for inspection and examination as follows: Pipe and Equipment Welds Location and clearance envelopes have been established for inspection and examination. Co ntours and surface finish are acceptable for inspection and examination.

Insulation Removal Piping or components to be inspected according to the Section XI code which are insulated, have been designed with removable numbered insulation panels.

Shielding Piping or components to be inspected according to the Section XI code and are radiologically shielded have been designed with removable or accessible shields.

6.6.3 Examination Techniques and Procedures Inservice inspection will be in acco rdance with ASME, B&PV Section XI.

6.6.4 Inspection Intervals The initial 10-year inspection program for LaSalle units 1 and 2 was submitted to the NRC on July 13, 1982 and December 21, 1982, respectively. The inservice LSCS-UFSAR 6.6-2 REV. 17 APRIL 2008 inspection program for both units 1 and 2 are based on the requirements of the ASME, Section XI 1980 edition including addenda through winter 1980. The inservice examinations conducted during the second 120 month Inspection Interval will comply with the 1989 Edition of ASME Section XI, except in cases where relief has been granted by the NRC. The inservice examinations conducted during the third 120 month Inspection Interval will comply with the 2001 Edition through the 2003 addenda, including the December of 2003 Erratum of ASME Section XI, except in cases where relief has been granted by the NRC.

6.6.5 Examination Categories and Requirements

The inservice inspection categories and requirements for Class 2, and Class 3 components are in agreement with ASME Section XI.

Specific written requests for relief from ASME code requirements determined to be impractical were contained in the initial in service inspection program. Relief from those requirements was granted by the NRC, detailed evaluation is included in Appendix C of NUREG-0519, Supplement No. 5, Safety Evaluation Report related to the operation of LaSalle County Station, Units 1 and 2.

6.6.6 Evaluation of Examination Results

The evaluation of Class 2 components ex amination results will comply with the requirements of Section XI.

The repair procedures for Class 2 and 3 components comply with the requirements of Section XI.

6.6.7 System Pressure Tests

All Class 2 system pressure testing complies with the criteria of Code Section XI, Article IWC-5000. All Class 3 system pres sure tests comply with the criteria of Article IWD-5000.

6.6.8 Augmented Inservice Inspection to Protect Against Postulated Piping Failures This inspection has been adequately cove red by the requirements of Section XI already adhered to previously.

LSCS-UFSAR 6.7-1 REV. 13 6.7 MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEM (MSIV-LCS)

Unit 2 deleted, Unit 1 abandoned in place The Main Steam Isolation Valve Leakage Control System provided originally has been deleted. The valve leakag es are processed by the Isolated Condenser Leakage Treatment Method as discussed in Section 6.8.

LSCS-UFSAR 6.8-1 REV. 13 6.8 Main Steam Isolation Valve - Isolated Condenser Leakage Treatment Method The Main Steam Isolation Valve - Isolated Condenser Leakage Treatment Method (MSIV - ICLTM) (Also called the MSIV Al ternate Treatments Leakage Paths) controls and minimizes the release of fiss ion products which could leak through the closed main steam isolation valves (MSIV's) after a LOCA. The system provides this control by processing valve leakage through the main steamlines, main steamline drains, and the main condenser.

6.8.1 Design Bases

6.8.1.1 Safety Criteria The following general and specific design criteria represent system design, safety, and performance requirements imposed upon the MSIV-ICLTM:

a. The safety function of the main steamlines and main steamline drains are described in LSCS-UFSAR Section 10.3.

b. The safety function of the main condenser is described in LSCS-UFSAR Section 10.4.1.

6.8.1.2 Regulatory Acceptance Criteria

The classification of the components and piping of the main steam supply system is listed in Table 3.2-1. All components and piping for the main steam supply system are designed in accordance with the code s and standards listed in Table 3.2-2 for the applicable classification.

The classification of the main condenser is described in LSCS-UFSAR Section 10.4.1.3.

6.8.1.3 Leakage Rate Requirements The MSIV-ICLTM has been incorporated as an integral part of the BWR plant design. The design features employed with this systems are established to reduce the leakage rate of radioactive materials to the environment during a postulated LOCA. Leakage control requirements are imposed upon the MSIV-ICLTM in order to:

a. eliminate the possibility of secondary containment bypass leakage of accident induced radioactive releases, b. allow for higher MSIV leakage limits, and LSCS-UFSAR 6.8-2 REV. 15, APRIL 2004

c. assure reasonable leakage verification test frequencies (once per fuel cycle).

The design and operational requirements imposed upon the MSIV-ICLTM relative to the foregoing criteria are established to:

a. allow MSIV leakage rates up to a total of 400 scfh for all MSIV valves, b. allow a MSIV leakage rate verification testing frequency compatible with the requirements of plant operating technical specifications, and

c. assure and restrict total plant dose impacts below 10 CFR 100 guidelines.

6.8.2 System Description

6.8.2.1 General Description The system provides this control by pr ocessing valve leakage through the main steamlines, main steamline drains, and the main condenser.

6.8.2.2 System Operation (U2 MSIV LCS delete, U1 Abandon-in-place)

With the deletion of the MSIV-LCS, MSIV leakage will pass from the outboard MSIV, through the main steamlines, main steamline drains and into the condenser. The large wetted volume in the main cond enser plates out inorganic iodine and holds up other fission products that esca pe through the MSIVs, limiting release to the environment. This alternate pathwa y is more reliable than the MSIV-LCS since less equipment is employed. The alternate pathway also has a much higher capacity for processing leakage than does the MSIV-LCS, with a capacity of only 100 scfh. In addition, the MSIV-LCS will on ly operate at less than 35 psig reactor vessel steam dome pressure, whereas the alternate pathway is independent of reactor pressure.

To properly align the pathway, in addition to closing the MSIVs and the containment isolation valves, operators will close valves to isolate the leakage pathway from the auxiliary steam supplies. The operating drains will also be closed and the shutdown drains will be opened. All of the remote manually operated valves that need to be moved are powered from Class 1E power supplies. Although these valves and their power supplies (with the exception of the MSIVs) are not maintained as safety-related, design contro l for all of these valves is maintained with respect to their importance to sa fety. Appropriate changes to station LSCS-UFSAR 6.8-3 REV. 13 procedures have been made to reflect deletion of the MSIV-LCS and use of the alternate leakage treatment method.

6.8.2.3 Equipment Required

The following equipment components are pr ovided to facilitate system operation:

a. piping - process piping is carbon steel throughout;

b. valves - motor-operated, standard closing speeds;

c. main condenser

6.8.3 System Evaluation

An evaluation of the capability of the MSIV-ICLTM to prevent or control the release of radioactivity from the main steamlin es during and following a LOCA has been

conducted. The results of this evaluation are presented in LaSalle County Nuclear Power Stations Units 1 and 2 Applicatio n for Amendment of Facility Operating Licenses NPF-11 and NPF-18, Appendix A, Technical Specifications, and Exemption to Appendix J of 10CFR50 Regarding Elimin ation of MSIV Leakage Control System and Increased MSIV Leakage Limits , NRC Docket Nos. 50-373 and 50-374. Additionally, Sargent & Lundy performed an evaluation on the piping, condenser and turbine building, to assure they would remain functional during a seismic event to mitigate the radiologically consequenc es of MSIV leakage (Reference Sargent & Lundy Calculation 068078 (EMD), Rev. 2, dated 8/9/95 for Unit 1 and 067927 (EMD), Rev. 2 dated 8/10/95 for Unit 2).

See Section 15.6.5.5 for more informat ion in the dose analysis and dose consequences.

6.8.4 Instrumentation Requirements The instrumentation necessary for contro l and status indication of the MSIV-ICLTM is designed to function under Seis mic Category I and environmental loading conditions appropriate to its installation with the control circuits designed to satisfy separation criteria. MSIV closed indication is powered from Class 1E power and is maintained as safety-related.

6.8.5 Inspection and Testing Preoperational tests for the main steamlines, main steamline drains, and the main condenser are discussed in LSCS-UFSAR Sections 10.3.4 and 10.4.1.4. No additional testing is required to support this operating mode.

LSCS-UFSAR TABLE 6.8-1 REV. 13 TABLE 6.8-1 DOSE CONSEQUENCES OF MSIV LEAKAGE LEAKAGE 30 DAYS FO LLOWING LOCA-UNIT 1 (100 SCFH per line)

WHOLE BODY DOSE (rem) THYROID DOSE (rem)

Exclusion Area (509 meters) 1.451E-3 3.14E-2 Low Population Zone (6400 meters) 3.3E-2 10.47 LSCS-UFSAR REV. 13

ATTACHMENT 6.A ANNULUS PRESSURIZATION

LSCS-UFSAR 6.A-i REV. 13 ATTACHMENT 6.A TABLE OF CONTENTS PAGE 6.A ANNULUS PRESSURIZATION 6.A-1 6.A.1 INTRODUCTION 6.A-1 6.A.2 SHORT-TERM MASS ENERGY RELEASE 6.A-2 6.A.2.1 Instantaneous Guillotine Break 6.A-3 6.A.2.2 Break Opening Flow Rate 6.A-4 6.A.2.3 Combined Break Flow 6.A-5 6.A.2.4 Determination of the Mass Flux, G 6.A-5 6.A.2.5 Biological Shield Wall 6.A-5 6.A.2.6 Comparison of the GE Model with the Henry/Fauske Correlation 6.A-6

6.A.3 LOAD DETERMINATION 6.A-11 6.A.3.1 Acoustic Loads 6.A-11 6.A.3.2 Pressure Loads 6.A-11 6.A.3.3 Jet Loads 6.A-11 6.A.3.4 Dynamic and Seismic Analysis (DYSEA) Computer Program 6.A-12

6.A.4 PRESSURE TO FORCE CONVERSION 6.A-15

6.A.5 SACRIFICIAL SHIELD ANNULUS PRESSURIZATION AND RPV LOADING DATA 6.A-17 6.A.6 JET LOAD FORCES 6.A-20

6.A.7 RECIRCULATION AND FEEDWATER LINE BREAK FORCING FUNCTION 6.A-21

LSCS-UFSAR 6.A-ii REV. 13 ATTACHMENT 6.A LIST OF TABLES

NUMBER TITLE 6.A-1 Time History for Postulated Recirculation Suction Pipe Rupture 6.A-2 Acoustic Loading on Reactor Pressure Vessel Shroud 6.A-3 RPV Coordinates of Nodal Points 6.A-4 Maximum Member Forces Due to Annulus Pressurization 6.A-5 Maximum Acceleration Due to Annulus Pressurization 6.A-6 RELAP4 Input Data, Recirculation Line Outlet Break 6.A-7 RELAP4 Input Data, Feedwater Line Break 6.A-8 Force Constants and Load Centers For Recirculation Line Outlet Break 6.A-9 Force Constants and Load Centers For Feedwater Line Break 6.A-10 DYSEA01 Program Input For Jet Load Forces 6.A-11 Time Force Histories - Recirculation Line Break 6.A-12 Time Force Histories - Feedwater Line Break LSCS-UFSAR 6.A-iii REV. 13 ATTACHMENT 6.A LIST OF FIGURES

NUMBER TITLE 6.A-1 Safe End Break Location 6.A-2 Break Flow Vs. Time - Feedwater Line Break 6.A-3 Geometry 6.A-4 Wave Speed 6.A-5 Mass Flux, Moody Steady Slip Flow 6.A-6 Break Flow Vs. Time 6.A-7 Nomenclature for Time History Computer Printout 6.A-8 Feedwater Line System Nodalization - Leg EA 6.A-9 Feedwater Line System Nodalization - Leg EB 6.A-10 Recirculation Line System Nodalization 6.A-11 Comparison of the GE and RELAP4/MOD5 Methods -

Feedwater Line Break, Leg EA 6.A-12 Comparison of the GE and RELAP4/MOD5 Methods -

Feedwater Line Break, Leg EB 6.A-13 Comparison of the GE and RELAP4/MOD5 Methods -

Recirculation Line Break, Finite Opening Time 6.A-14 Horizontal Model for Annulus Pressurization 6.A-15 Annulus Pressurization Loading Description 6.A-16 Annular Space Nodalization For Recirculation Line Break 6.A-17 Annular Space Nodalization For Feedwater Line Break

LSCS-UFSAR 6.A-1 REV. 13 6.A ANNULUS PRESSURIZATION 6.A.1 INTRODUCTION

Annulus pressurization refers to the load ing on the shield wall and reactor vessel caused by a postulated pipe rupture at the reactor pressure vessel nozzle safe-end to pipe weld. The pipe break assumed is an instantaneous guillotine rupture which allows mass/energy release into the drywell and annular region between the biological shield wall and the reactor pressure vessel (RPV).

The mass and energy released during the postulated pipe rupture cause:

a. A rapid asymmetric decomp ression acoustic loading of the annular region between the vessel and shroud from the pipe break at or beyond the vessel nozzle safe-end weld.

b. A transient asymmetric differential pressure within the annular region between the biological shield wall and the reactor

pressure vessel (annulus pressurization).

c. A jet-stream release of the reactor pressure vessel inventory and the impact of the ruptured pipe against the whip restraint

attached to the biological shield wall.

The results of the mass and energy release evaluation are then used to produce a dynamic structural analysis (force-time history) of the RPV and shield wall. The force time history output from the dyna mic analysis is subsequently used to compute loads on the reactor components. The following is a more detailed description of the annulus pressurization calculation performed for the LaSalle County Station.

6.A.2 SHORT-TERM MASS ENERGY RELEASE

The postulated pipe rupture at the weld between recirculation or feedwater piping and the reactor nozzle safe end leads to a high rate of water and steam mixture into the annulus between the RPV and the shie ld wall. Figure 6.A-1 illustrates the location of this break. Calculation of the mass/energy release is performed using the generic method for short-term mass releases. This method and a sample calculation are described below. Figure 6.A-2 illustrates a typical mass flux vs. time for a feedwater line break.

The purpose of this procedure is to document the method by which short-term mass release rates are calculated. The flow ra tes which could be produced by a primary system line break for the first 5 seconds include the effects of inventory and subcooling. Optionally, credit may be taken for a finite break opening time.

LSCS-UFSAR 6.A-2 REV. 13 ASSUMPTIONS

The assumptions are as follows:

a. The initial velocity of the fluid in the pipe is zero. When considering both sides of the break, the effects of initial velocities would tend to cancel out.

b. Constant reservoir pressure.

c. Initial fluid conditions inside the pipe on both sides of the break are similar.

d. Wall thickness of the pipe is small compared to the diameter.

e. Subcompartment pressure

~ 0. f. Mass flux is calculated using the Moody steady slip flow model with subcooling.

g. For steamline breaks, level swell occurs at 1 second after the break with a quality of 7%.

NOMENCLATURE (See Figure 6.A-3)

A BR - Break area.

A L - Minimum cross-sectional area between the vessel and the break. This can be the sum of the areas of parallel flow paths.

C - Sonic velocity (see Figure 6.A-4).

D - Pipe inside diameter at the break location.

F I - Inventory flow multiplier.

F I = 0.75 for saturated steam.

FI = 0.50 for liquid and saturated steam-liquid mixtures.

g c - Proportionality constant (=32.17 2 lbm-ft/lbf-sec 2).

G - Mass flux.

LSCS-UFSAR 6.A-3 REV. 13 G C - Maximum mass flux (see Figure 6.A-5).

h O - Reservoir or vessel enthalpy.

h P - Initial enthalpy of the fluid in the pipe.

h 7 - Enthalpy at P O and a quality of 7%.

L I - Inventory length. The distance between the break and the nearest area increase of A L whichever is less.

M - Mass flow rate.

I M - Mass flow rate during the inventory period.

P O - Reservoir or vessel pressure.

PSAT - Saturation pressure for liquid with an enthalpy of h P.

t - Time.

t I - Length of the inventory period.

v - Specific volume of the fluid initially in the pipe.

V I - Volume of the pipe between the break and A L .

X - Separation distance of the break.

6.A.2.1 Instantaneous Guillotine Break The following method should be applied to each side of the break and the results summed to determine the total flow.

LSCS-UFSAR 6.A-4 REV. 14, APRIL 2002 cL2 tFAA If I I' I BR L=> vFG A V tFAA If I BR I I' I BR L=<Inventory Period Prior to a pipe break, the fluid in the pipe is moving at a relatively low velocity. After the break occurs, a finite time is required to accelerate the fluid to steady-state velocities. The length of this time period is conservatively estimated as follows: a. (6.A-1) b. (6.A-2) where G is calculated as shown in Subsecti on 6.A.2.4 for a large separation distance and t < t I.

During this time period, the mass flow rate is calculated as Steady-State Period

Following the inventory period, the flow is assumed to be choked at the limiting cross-sectional flow area. For t I < t < 5.0 seconds, (6.A-4) 6.A.2.2 Break Opening Flow Rate

See Table 6.A-1 for the pipe displacement time history for postulated recirculation suction pipe rupture and Figure 6.A-7 for the nomenclature used.

Inventory Period

The inventory period is determined as des cribed in Subsection 6.A.2.1. The flow rate as a function of pipe separation distance is given by where G is obtained by using the methods of Subsection 6.A.2.4 (a or b).

IF BRAG I M= G LA M=XDG M=(6.A-3) (6.A-5)

LSCS-UFSAR 6.A-5 REV. 13 Determining Flow Rate

Following the inventory period, equation 6.A-5 is used to deter mine the flow rate where the mass flux, G, is determined from Subsection 6.A.2.4 (a, c, or d).

6.A.2.3 Combined Break Flow

To determine the total flow rate released from the break, the results of Subsections 6.A.2.1 and 6.A.2.2 are compared and whichever produces the smallest flow rate at any time is used (see Figure 6.A-6). Both methods produce maximum flow rates based on different limiting areas. The transfer from one curve to the other represents a change in the point where the flow is choked.

6.A.2.4 Determination of the Mass Flux, G

Depending on the time period, fluid conditions, and break separation distance, the mass flux is determined as follows:

a. If X < X B ,

b. If X > X B and t < t I G = G c (P o , h p) from Figure 6.A-5

c. If X > X B and t > t I G = G c (P o , h o) from Figure 6.A-5

d. If the break is a steamline and T > 1.0, level swell occurs.

G = G c (P o , h 7) from Figure 6.A-5

Note that for complete break separation (Subsection 6.A.2.1), X is always greater than X B, and for saturated water, X B is equal to zero.

6.A.2.5 Biological Shield Wall

For the purpose of analyzing the biological shield wall pressurization, credit may be taken for flow which escapes through the wall penetration. If the initial break location is in the annulus region between the wall and the vessel, no flow is assumed to escape through the penetration. If, however, it is located within the penetration itself, some of the flow may be assumed to escape. It is recommended

()()2D o P SATP1 B X= vo P c2gG= (6.A-6)

LSCS-UFSAR 6.A-6 REV. 13 that the fraction of the flow which escapes be calculated based on the ratio of the minimum annular flow area between the pe netration and pipe surface and between the penetration and pipe surface and between the penetration and the safe-end nozzle. 6.A.2.6 Comparison of the GE model with the Henry/Fauske Correlation The GE methodology for calculating the mass energy release from a recirculation line break which results in an annulus pressurization event was provided the NRC's Mr. Denwood F. Ross, Assistant Director for Reactor Safety, via GE letter dated May 2, 1978, from Mr. E. D. Fuller of BWR Licensing. This methodology was used in the adequacy assessment made for LSCS.

The definition of the annulus pressurization is given in the introduction (Subsection 6.A.1). A description of the time aspect s of the calculated mass and energy flow rates followed by a description of the modeling for the feedwater line and separately for the recirculation line is provided below. A comparison is then made between GE's analytical method and the method used in RELAP4/MOD5. Finally, both

graphical and numerical results of this comp arison are provided to substantiate the conclusion that the resulting break flows using the GE methods are much more conservative than those predicted by the use of RELAP for the LaSalle plant.

Timing Aspects of Mass and Energy Flow Rates

The GE method for calculating the short-term mass/energy release assumes that the overall time for mass release may be divided into two periods, the inventory period and the quasi-steady period. The inventory period is defined as the time required to accelerate the pipe fluid to steady-state velocities, at which time the flow is assumed to choke at minimum flow cross sections. During this time, the mass flux is based on initial thermodynamic conditions exis ting within the pipe. In the quasi-steady period, during which the flow is choked, the mass flux is based on thermodynamic conditions upstream from th e choke points. For both time periods the mass flux is determined from a graph of critical mass flux versus enthalpy, as calculated by the Moody Slip Flow Method. Each side of the break is analyzed separately and the results summed to give the total mass release rate.

Method for Feedwater Line Modeling

The feedwater system for LaSalle County St ation consists of the pumps, heaters, valves, and piping necessary for the tran smission of hotwell condensate to the reactor vessel as part of the closed cycl e cooling loop. LSCS has three feedwater pumps, two steam- driven and one electric-driven. During normal operation, the electric pump is in standby. The flow passes through a complex series of pipes and components from the feedwater pumps to the reactor vessel.

LSCS-UFSAR 6.A-7 REV. 13 The break location for the feedwater line break is the safe-end to the pipe weld housed within the vessel to shield wa ll subcompartment. For the feedwater line break, instantaneous break opening is assumed. Flow for the vessel side passes through the feedwater nozzles of the broken line and out the break. Flow from the system side passes through the feedwater piping network and out the break.

The nodalization of the feedwater system is shown in Figures 6.A-8 and 6.A-9. A series of 24 modes was selected after sensitivity studies were completed which demonstrated that a 24-node model was conservative relative to higher-noded systems.

The broken feedwater leg to be analyzed was chosen by multiple RELAP runs to determine the limiting break location. The critical assumptions in the analysis are as follows:

a. The feedwater pumps are simulated as (constant) mass flow sources. b. The reactor pressure vessel (RPV) is an infinite reservoir at constant temperature and pressure.

c. The temperature of the pump-side hydraulic network remains constant.

d. Appropriate sections of the hydraulic network are combined by means of "Ohm's Law" expressions for series and parallel circuits, assuming constant fanning friction actions.

e. The RPV thermodynamics stat e is subcooled at the prevailing temperature in the lower plenum (532

° F).

The break is modeled as an instantaneous guillotine pipe break with complete pipe offset. Before the break occurs, a fully open valve connects, Volumes 18 and 19. Closed valves connect those volumes to Volume 1, an infinite sink at constant pressure and temperature (atm ospheric conditions). The break is initiated at time zero by closure of the valve between Volumes 18 and 19 simultaneous opening of the valves to Volume 1.

Method of Recirculation Line Modeling

The recirculation system for LaSalle County Station is similar to the recirculation system of other BWR's. Flow is taken from the lower jet pump diffuser region, passed through 21-inch lines to a constant-speed pump, and then through a flow control valve to a header which feeds flow to five risers which provide flow to two jet pump nozzles each.

LSCS-UFSAR 6.A-8 REV. 13 The nodalization for the recirculation line leak is shown in Fi gure 6.A-10. The system has been modeled using 21 nodes. The break is located at the vessel nozzle safe-end to pipe weld on the recirculation pump suction side. The type of break considered here has a finite break opening time. For this case the break opening is complete after 30 milliseconds, at which time the pipe offset longitudinal distance is 5.8 inches. The break area is modeled as the surface area of an imaginary volume having a length of 5.8 inches and a diameter equal to that of the recirculation pipe ID. This volume (#18) is connected by a valve (Type 3) to an infinite reservoir (volume #19), and also by valves (Type 2) to the vessel side volume (#27) and pump side volume (#21). Both valves (Type 1) also connect Volumes 17 and 21. It is normally open before the break, and at the initiation of the break, closes at the same rate as the other valves open. The sum of the areas of the Type 2 valves equals the pipe area.

This network of valves best represents the break with finite opening time. Valves of Type 2 are opened at the same rate as Type 3 to ensure that choking occurs at Junctions 21 and/or 22. Junction 23 (having valve Type 3) is in reality a fluid surface, and choking cannot physically occur there. Choking must at least occur at Junctions 21 and/or 22, where the fluid is constrained by the pipe diameter.

Other assumptions in th e analysis include:

a. Negligible effects of core reactor kinetics on rated heat transfer to the core volume (Volume 2).

b. Constant flow of steam from the steam dome (Vol ume 5) at rated conditions.

c. Constant flow of feedwater at rated conditions.

d. Recirculation pumps trip at the time zero and are modeled via pump characteristic curves for coastdown.

e. Jet pump hydraulics were modeled as one equivalent pump per recirculation loop.

Comparison of General Electr ic Analysis to RELAP4/MOD5

For the annulus pressurization event, th e NRC has questioned General Electric's method for computing mass and energy flow rates following a postulated LOCA from long lines containing subcooled fluid. A program was developed to expedite the licensing of the LaSalle County Station to perform RELAP analyses using appropriate assumptions and to compare the results with those obtained using General Electric's method.

LSCS-UFSAR 6.A-9 REV. 13 RELAP4/MOD5 is a general computer prog ram which can be used to analyze the thermal hydraulic transient behavior of a water- cooled nuclear reactor subjected to postulated accidents such as loss-of-coolant accidents. The program simultaneously solves the fluid flow, heat transfer, the reactor kinetics equations describing the behavior of the reactor.

Numerical input data is utilized to describe the initial conditions and geometry of the system being analyzed. This data includes fluid volume, geometry, pump characteristics, power generation, heat exchanger properties, and nodalization of fluid flow paths. Once the system has been described with initial flow, pressure, temperature, and power level boundary co nditions, transients such as loss-of-coolant accident can be simulated by control action inputs. RELAP then computes

fluid conditions such as flow, pressure, mass inventory an d quality as a function of time. For the brief transients considered here (t 0.5 seconds), appreciable simplification of the overall thermal-hydraulic system, including the reactor pressure vessel, is justified owing to the relatively longer time constants which apply for heat transfer.

Brief summaries of the modeling approaches for feedwater and recirculation line breaks are given below.

The assumptions applied to th ese analyses are as follows:

a. Feedwater line:

1. LaSalle RELAP deck as basis.

2. Henry-Fauske-Moody flow model is used.

3. Instant break opening.

4. Mass flux terms between ve ssel and break (short side) are eliminated.

b. Recirculation line:

1. LaSalle RELAP deck as basis.

2. Finite break opening time is allowed for.

3. Henry-Fauske-Moody flow model is used.

4. Momentum flux terms in RELAP between vessel and break (short side) are eliminated.

Results of the Analysis

LSCS-UFSAR 6.A-10 REV. 13 The resulting break flows using the GE methods are much more conservative than those obtained by the use of RELAP. This is indicated graphically in Figures 6.A-11 through 6.A-13.

Conclusions The mass release result for the GE mass energy release method and the RELAP4/Mod 5 calculations are compared in Figures 6.A-11 through 6.A-13 for the postulated feedwater line break and reci rculation line break respectively. The analyses show that the GE method is conservative relative to RELAP 4/Mod 5 for both cases. The ration (r) of the GE method flow rates to those from RELAP/MOD5 is as follows:

Break Location r(t = 0.1 sec) r(t = 0.5 sec)

Feedwater (Leg EA) 2.300 1.70 Feedwater (Leg EB) 2.200 1.60 Recirculation Line 1.065 1.21 6.A.3 LOAD DETERMINATION 6.A.3.1 Acoustic Loads

Because the boiling water reactor (BWR) is a two-phase system that operates at or close to saturation pressure (1000 psi), th e differential pressure across the reactor shroud is of short duration, and the BWR system is not subjected to a significant shock-type load with respect to structural supports. This short- duration acoustic load is confined to a bending moment and shear force on the reactor pressure vessel and reactor shroud support. Typical results of the integrated force acting on the reactor pressure vessel shroud are given in Table 6.A-2.

6.A.3.2 Pressure Loads The pressure responses of the RPV-shield wall annulus for a recirculation suction line and a feedwater line were investigated using the RELAP4 computer code. An asymmetric model using several nodes and flow paths was developed for the analysis of the recirculation and feedwater line breaks. Further description of these analytical models and detailed discussion of the analyses may be found in Section 6.2.

The pressure histories generated by the RELAP4 code were in turn used to calculate the loads on the sacrificial wall and the reactor pressure vessel. The

annulus was divided into seven zones and an eighth-order Fourier fit to the output LSCS-UFSAR 6.A-11 REV. 14, APRIL 2002 pressure histories made for each zone to produce the Fourier coefficients required for the structural analysis of the shield wall. The specific loading data consisted of the time-pressure (psia) hist ories for each node within the annulus. Time-force histories representing the resultant loads on the RPV for each node through its geometric center were generated by taking the product of the node pressure and its "effective" surface area.

A sample pressure-to-force ca lculation is shown in Subsection 6.A.4. Subsection 6.A.5 shows the nodalization schemes and pressure areas used in this calculation. The time-force histories (forcing functions) calculated at each nodal point for both a recirculation and a feedwater line break are shown in Subsection 6.A.7. The nodal points are illustrated in Figure 6.A-14.

6.A.3.3 Jet Loads To address structural loads on the vessel and internals completely, jet thrust, jet impingement, and pipe whip restraint loads must be considered in conjunction with the above mentioned pressure loads. Jet thrust refers to the vessel reaction force with results as the jet stream of liquid is released from the break. Jet impingement refers to the jet stream force which leaves the broken pipe and impacts the vessel.

The pipe whip restraint load is the force which results when the energy-absorbing pipe whip restraint restricts the pipe separation to less than one full pipe diameter. This restricted separation is accounted for as a finite break opening time in the mass/energy release calculation. These je t loads are calculated as described in

ANSI 176 (draft), "Design Ba sis For Protection Of Nuclear Power Plants Against Effects Of Postulated Pipe Ruptures", January 1977.

The jet load forces used in this analysis are shown in Subsection 6.A.6. Although these values have been calculated for a re circulation line break only, they are also conservatively used for the feedwater load evaluation. This is conservative because the calculation of these jet effects depends largely on the area of the break, and the recirculation line is about 2.5 times larger in area. Figure 6.A-15 illustrates the location of the pressure loads and jet loads with respect to the RPV and shield wall.

The pressure loads and jet loads describe d above are then combined to perform a structural dynamic analysis. Both of these loads are appropriately distributed

along a horizontal beam model, which is shown in Figure 6.A-14. The vessel coordinates of these nodal points are described in Table 6.A-3.

The force time histories are then applied to a composite lumped- mass model of the pedestal, shield wall, and a detailed repres entation of the reactor pressure vessel

and internals. The DYSEA01 computer program is used for this analysis. This computer program is described in Subsec tion 6.A.3.4. The analysis produces acceleration time histories at all nodes for use in evaluating the reactor pressure vessel and internal components. Response spectra at all nodes are also computed.

LSCS-UFSAR 6.A-12 REV. 13 The peak loading on the major components used to establish the adequacy of the component design is shown in Tables 6.A-4 and 6.A-5.

6.A.3.4 Dynamic and Seismic Anal ysis (DYSEA) Computer Program

The DYSEA (Dynamic and Seismic Analysis) program is a GE proprietary program developed specifically for seismic and dynamic analysis of RPV and internals/building systems. It calculates the dynamic response of linear structural systems by either temporal modal superposition or response spectrum method. Fluid- structure interaction effect in the RPV is taken into account by way of hydrodynamic mass.

The DYSEA program was based on the program SAP-IV with added capability to

handle the hydrodynamic mass effect. St ructural stiffness and mass matrices are formulated similar to SAP-IV. Solution is obtained in the time domain by calculating the dynamic response mode by mode. Time integration is performed by using Newmark's -method. Response spectrum solution is also available as an option.

Program Version and Computer The DYSEA version now operating on the Honeywell 6000 computer of GE, Nuclear Energy Systems Division, was developed at GE by modifying the SAP-IV program.

Capability was added to handle the hydrodyn amic mass effect due to fluid-structure interaction in the reactor. The progra m can handle three-dimensional dynamic problems with beam, trusses, and springs. Both acceleration time histories and response spectra may be used as input.

History of Use

The DYSEA program was developed in the su mmer of 1976. It has been adopted as a standard production program since 1977 an d it has been used extensively in all dynamic and seismic analysis of the RPV and internals/building systems.

Extent of Application The current version of DYSEA has been used in all dynamic and seismic analysis since its development. Results from test problems were found to be in close agreement with those obtained from either verified programs or analytic solutions.

LSCS-UFSAR 6.A-13 REV. 13 Test Problems Problem 1:

The first test problem involves finding the eigenvalues and eigenvectors from the following characteristic equation:

(2 [M]-[K]) {x} = 0 where is the circular frequency, x is th e eigenvector, and [K] and [M] are the stiffness and the mass matrices given by:

(6.A-8) The analytical solution and the solution from DYSEA are:

a) Eigenvalues i: i DYSEA SOLUTION ANALYTIC SOLUTION 1 5.7835 5.7837 2 30.4889 30.4878 3 75.0493 75.0751 []

=2 25 4-1 Symmetric 2 4 2 q 4 1 2 q 4 2 4 2 4 1 M[]

+++=4 2 251 Symmetric15 4 2 g1 q 5 3 4 2 1 K (6.A-7)

LSCS-UFSAR 6.A-14 REV. 13 b) Eigenvectors i: 1.

DYSEA SOLUTION ANALYTIC SOLUTION

0.0319

0271.20666.00072.02105.15536.10319.0000.1000.1000.1 027.20666.00072.0211.1554.10319.0000.1000.1000.1 Problem 2: The second test problem compares the dynamic responses of the reactor pressure vessel, internals and reactor building subjected to earthquake ground motion.

The mathematical model of the reactor pressure vessel, internals and reactor building is given in Figure B-1. The inputs in the form of ground spectra are applied at the basement level. Response spectr um analysis was used in the analysis.

Natural frequencies of the system and the maximum responses at key locations have been calculated by both DYSEA and SAMIS. Result comparison are given in

B-1 and B-2. It can be seen that the results calculated by DYSEA agree closely with those obtained by SAMIS.

6.A.4 PRESSURE TO FORCE CONVERSION The RELAP4 pressure distribution output is converted to equivalent forces which are input into the DYSEA01 computer progra

m. Each pressure is represented by a force acting normal to the RPV or shield wall at the center of the given pressure surface area. These forces are then converte d into resulting forces (x component) acting on the respective DYSEA01 RPV beam nodal points. Mathematically, this is described as:

F R = PA cos where: F R = resultant force (lb), P = RELAP4 node pressure (psia), A = RELAP4 node surface area (in 2 ), and = Component angle.

LSCS-UFSAR 6.A-15 REV. 14, APRIL 2002 The results of these calculations are summarized in Table 6.A-4.

As an example, the pressure to force conv ersion at DYSEA01 node points 31 and 32 is shown below:

Time = 0.0800 seconds NODE ELEV (inches) PRESSURE (lb/in 2) AREA* (in 2) ANGLE (degrees) FORCE (lb) 6 1089.14 43.61 5828.44 15 245516 7 1089.14 35.34 5828.44 45 145660 8 1089.14 39.24 5828.44 75 59188 9 1089.14 41.40 8617.79 112.5

-136539 10 1089.14 39.99 8617.79 157.5

-318367

- 4543

• See Table 6.A-8 For 360°, the resultant force is 2 times 4543 lb or an inward (positive) force of 9086 lb.

Since DYSEA nodal points 31 and 32 are at Elevations 1065.2 inches and 1125.7 inches respectively, the RELAP4 pressure

/force at Elevation 1089.14 inches is distributed accordingly.

Consequently:

F 31 = 1125.7 - 1089.14 (9086) = 5491 lb, and 1125.7 - 1065.2 F 32 = 1065.2 - 1089.14 (9086) = 3595 lb.

1065.2 - 1125.7 These values can be compared to the co mputer-calculated DYSEA01 results, which are 5832.6 lb and 3252.7 lb respectively.

In the matrix displacement method of stru ctural analysis, externally applied nodal forces and moments are required to produce nodal displacements equivalent to

those that would be produced by forces or pressures applied between nodes. GE LSCS-UFSAR 6.A-16 REV. 13 considers the external moment effects for La Salle AP to be negligible because of the close nodal spacing of the LaSalle RPV mathematical model.

6.A.5 SACRIFICIAL SHIELD, ANNULUS PRESSURIZATION, AND RPV LOADING DATA

This subsection provides a brief descri ption of the analyses performed and the nodalization schemes, force constants, and load centers for the recirculation and feedwater line breaks. These data are used as input to the pressure to force conversion calculation.

The pressure responses of the RPV-sacrificial shield wall annulus to postulated pipe breaks at the RPV nozzle safe-end to pipe weld in a recirculation outlet line and a feedwater line were investigated using th e RELAP4 computer code. Throughout the analyses the following assumptions were made:

a. RPV thermal insulation displaces to the shield wall while retaining its original volume and leaving its support structure in

place. b. Insulation above the shield wall yields to elevated pressures and blows out into the drywell allowing venting of annulus at the

summit of the shield wall.

c. sacrificial shield penetration doors remain closed, allowing for limited venting of the annulus through all nozzle penetrations.

The nodalization schemes for both studies remain consistent with the guidelines cited above, with the exception of the region directly above the break, where it was

anticipated that a finer mesh would be necessary to properly account for the highly localized pressure gradients expected there (see Figures 6.A-16 and 6.A-17). The final nodalization was determined on the ba sis of available sensitivity studies for similar analyses.

The mass and energy release rates were derived with the methods outlined in Subsection 6.A.2. The blowdown rates for the recirculation outlet line break analysis account for actual pipe displa cement, while those for the feedwater line reflect an assumption of instantaneous pipe displacement (see RELAP4 input listings, Tables 6.A-6 and 6.A-7).

The specific loading data compiled for th e NSSS adequacy evaluation for postulated pipe breaks within the annulus consists of the time-pressure history (psia) and two time-force (lbf) histories for each node within the annulus. The latter two histories represent integrated forces acting through the center of each node on the RPV and the sacrificial shield wall respectively. The time-force histories were generated by LSCS-UFSAR 6.A-17 REV. 13 taking the product of the node pressure and a predetermined constant, or ss, which accounts for the curved surface of the RPV and the sacrificial shield respectively (see Tables 6.A-8 and 6.A-9). The two loadin g histories, one for the RPV and one for the shield wall, are defined below.

(6.A-9) = P i v Where: F v i nodal resultant force on RPV (lbf), P i node absolute pressure (psia), i node height (inches), R v RPV radius (inches), azimuthal width of node (degrees), and D p j pipe OD (in.). 4 2 j p D i P j - 0d cos 2 2 vR i iP i v F+=4 2 j p D j () iP - 2sin vR i2 iP =

LSCS-UFSAR 6.A-18 REV. 14, APRIL 2002 (6.A-10) = P i ss Where: F ss i nodal resultant force on shield wall (lbf), P i node absolute pressure (psia), i node height (inches), R ss shield wall inner radius (inches), azimuthal width of node (degrees), D ss j penetration ID (inches), and proportionality factor

6.A.6 JET LOAD FORCES This subsection provides the jet load forces which result from pipe separation during the postulated accident. The pipe whip schematic is shown in Figure 6.A-7, and the resulting loads are listed in Table 6.A-1.

These loads are applied to the appropriate nodal points for input to the DYSEA01 computer program. The DYSEA01 progra m input is provided in Table 6.A-10.

4 j ss 2 D P - d cos R P i j 2 2ssii i s F=+ s () u iP - 2sin ssR i2 iP = 4 j ss 2D j ()2 2 sin=

2 360 LSCS-UFSAR 6.A-19 REV. 13 6.A.7 RECIRCULATION AND FEEDWATER LINE BREAK FORCING FUNCTION The time force histories provided in T ables 6.A-11 and 6.A-12 are those values converted from the time-pressure histories which were calculated with the RELAP4 computer program. These ti me forces histories are used as input to the DYSEA01 computer program.

LSCS-UFSAR TABLE 6.A-1 (SHEET 1 OF 5) TABLE 6.A-1 REV. 0 - APRIL 1984 TIME HISTORY FOR POSTULATED RECIRCULATION SUCTION PIPE RUPTURE*, ** Time (sec) Pipe Displ. At Restraint (in.) Pipe Velocity At Restraint (ft/sec) Pipe Acceler. At Restraint (ft/sec 2) Rel. Displ.

Of End (in.) Total Displ. Of End (in.) Restr. Load Comp. PD1 (lb) Restr. Load Comp. PD2 (lb) Blowdown Force (lb) 0.00153 4.147E-02 3.547E 00 1.679E 03 0. 4.648E-02 0. 0. 346919.

0.00233 8.294E-02 4.889E 00 1.655E 03 0. 9.295E-02 0. 0. 346919. 0.00297 1.244E-01 5.932E 00 1.645E 03 0. 1.394E-01 0. 0. 346919. 0.00351 1.659E-01 6.816E 00 1.640E 03 0. 1.859E-01 0. 0. 346919.

0.00398 2.074E-01 7.597E 00 1.635E 03 0. 2.324E-01 0. 0. 346919.

0.00441 2.488E-01 8.304E 00 1.632E 03 0. 2.789E-01 0. 0. 346919. 0.00481 2.903E-01 8.955E 00 1.630E 03 0. 3.253E-01 0. 0. 346919. 0.00519 3.318E-01 9.561E 00 1.628E 03 0. 3.718E-01 0. 0. 346919.

0.00554 3.732E-01 1.013E 01 1.626E 03 0. 4.183E-01 0. 0. 346919. 0.00587 4.147E-01 1.067E 01 1.624E 03 0. 4.648E-01 0. 0. 346919. 0.00687 5.427E-01 1.077E 01 3.194E 02 2.689E-02 6.351E-01 50588. 0. 346919.

0.00787 6.742E-01 1.117E 01 4.350E 02 9.147E-02 8.471E-01 108204. 0. 346919.

0.00887 8.108E-01 1.159E 01 3.863E 02 1.808E-01 1.089E 00 168037. 0. 346919. 0.00987 9.519E-01 1.190E 01 2.419E 02 2.875E-01 1.354E 00 229892. 0. 346919. 0.01087 1.096E 00 1.203E 01 3.532E 01 4.076E-01 1.636E 00 293042. 0. 346919.

0.01187 1.240E 00 1.194E 01 -2.099E 02 5.388E-01 1.928E 00 356421. 0. 346919.

• Output parameters are listed at the end of this table. ** Except for the restraint load components PD1 and PD2, all variables below are in a direction parallel to the blowdown force.

LSCS-UFSAR TABLE 6.A-1 (SHEET 2 OF 5) TABLE 6.A-1 REV. 0 - APRIL 1984

Time (sec)

Pipe Displ.

At Restraint (in.) Pipe Velocity At Restraint (ft/sec) Pipe Acceler. At Restraint (ft/sec 2) Rel. Displ.

Of End (in.)

Total Displ.

Of End (in.)

Restr. Load Comp. PD1 (lb) Restr. Load Comp. PD2 (lb) Blowdown Force (lb) 0.01287 1.381E 00 1.158E 01-4.744E 026.802E-01 2.228E 00418752. 0. 346919. 0.01387 1.517E 00 1.096E 01-7.414E 028.316E-01 2.531E 00478650. 0. 346919. 0.01487 1.643E 00 1.007E 01-1.027E 039.934E-01 2.835E 00538908. 0. 346919. 0.01587 1.757E 00 8.948E 00-1.197E 031.166E 00 3.136E 00581800. 0. 346919.

0.01687 1.857E 00 7.672E 00-1.335E 031.350E 00 3.431E 00618871. 0. 346919. 0.01787 1.941E 00 6.278E 00-1.438E 031.543E 00 3.719E 00649762. 0. 346919. 0.01887 2.008E 00 4.801E 00-1.504E 031.746E 00 3.996E 00674226. 0. 346919.

0.01987 2.056E 00 3.279E 00-1.531E 031.956E 00 4.261E 00692131. 0. 346919.

0.02087 2.086E 00 1.751E 00-1.519E 032.172E 00 4.510E 00703465. 0. 346919. 0.02187 2.098E 00 2.567E-01-1.469E 032.392E 00 4.744E 00708338. 0. 346919. 0.02222 2.098E 00 0. 0. 2.470E 00 4.822E 00708572. 0. 346919. 0.02242 2.098E 00 0. 0. 2.513E 00 4.865E 00708572. 0. 346919. 0.02262 2.098E 00 0. 0. 2.555E 00 4.907E 00708572. 0. 346919. 0.02283 2.098E 00 0. 0. 2.598E 00 4.950E 00708572. 0. 346919. 0.02304 2.098E 00 0. 0. 2.640E 00 4.992E 00708572. 0. 346919. 0.02325 2.098E 00 0. 0. 2.683E 00 5.035E 00708572. 0. 346919. 0.02347 2.098E 00 0. 0. 2.725E 00 5.077E 00708572. 0. 346919. 0.02370 2.098E 00 0. 0. 2.768E 00 5.120E 00708572. 0. 346919. 0.02393 2.098E 00 0. 0. 2.810E 00 5.162E 00708572. 0. 346919.

LSCS-UFSAR TABLE 6.A-1 (SHEET 3 OF 5) TABLE 6.A-1 REV. 0 - APRIL 1984

Time (sec)

Pipe Displ.

At Restraint (in.) Pipe Velocity At Restraint (ft/sec) Pipe Acceler. At Restraint (ft/sec 2) Rel. Displ.

Of End (in.)

Total Displ.

Of End (in.)

Restr. Load Comp. PD1 (lb) Restr. Load Comp. PD2 (lb) Blowdown Force (lb) 0.02417 2.098E 00 0. 0. 2.853E 00 5.2O5E 00 708572. 0. 346919. 0.02442 2.098E 00 0. 0. 2.895E 005.247E 00708572. 0. 346919. 0.02467 2.098E 00 0. 0. 2.938E 005.290E 00708572. 0. 346919. 0.02494 2.098E 00 0. 0. 2.980E 005.332E 00708572. 0. 346919. 0.02522 2.098E 00 0. 0. 3.023E 005.375E 00708572. 0. 346919. 0.02551 2.098E 00 0. 0. 3.065E 005.417E 00708572. 0. 346919. 0.02582 2.098E 00 0. 0. 3.108E 005.460E 00708572. 0. 346919. 0.02614 2.098E 00 0. 0. 3.150E 005.502E 00708572. 0. 3469l9. 0.02649 2.098E 00 0. 0. 3.193E 005.545E 00708572. 0. 346919. 0.02687 2.098E 00 0. 0. 3.235E 005.587E 00708572. 0. 346919. 0.02728 2.098E 00 0. 0. 3.278E 005.630E 00708572. 0. 346919. 0.02774 2.098E 00 0. 0. 3.320E 005.672E 00708572. 0. 346919. 0.02827 2.098E 00 0. 0. 3.363E 005.715E 00708572. 0. 346919. 0.02893 2.098E 00 0. 0. 3.405E 005.757E 00708572. 0. 346919. 0.02992 2.098E 00 0. 0. 3.448E 005.800E 00708572. 0. 346919.

LSCS-UFSAR TABLE 6.A-1 (SHEET 4 OF 5) TABLE 6.A-1 REV. 0 - APRIL 1984 Output Parameters Summary Effective clearance (inches) Length from restraint to break (ft)

Restraint loading direction 0.415 3.542 0 degrees Pipe bending strain limit (in/in) Pipe rotation stability limit (degr.)

Max. allowable bending moment (ft-lbs) 9.054E-02 7.7815 1417307 Impact Velocity = 10.67 ft/sec Impact Time = 0.0059 seconds Number of bars composing the restraint Defl. of struc. in direction of thrust (in.) Defl. of restr. in direction of thrust (in.) 2 0.7086 0.9754 Force on restr. in direction of thrust (lb) Force on struc. in direction of thrust (lb)

Time at peak dynamic load (seconds) 708572 708572 0.0221 Total energy absorbed by the restraint (ft-lb)

`Energy absorbed by the structure (ft-lb)

Energy absorbed by the bottom hinge (ft-lb) 30522 20920 1956 Energy absorbed by the top top hinge (ft-lb) Restraint load (peak) components (lb) PD1 PD2 Restraint load (static) components (lb) PS1 PS2 0. 708572 0. 138258 0.

LSCS-UFSAR TABLE 6.A-1 (SHEET 5 OF 5)

TABLE 6.A-1 REV. 0 - APRIL 1984 Relative defl. of pipe end in the direction of the thrust (in.)

Total defl. of the pipe end 3.4649 5.8168 Defl. time for pipe end (seconds after impact)

Total time of movement 0.0250 0 0309 Energy absorbed by the restraint hinge (ft-lb)

Total absorbed energy (ft-lb) 115445 168843 Pipe defl. at restraint components (in.) XR1 XR2 Pipe defl. at the break components (in.) XP1 XP2 2.0986 0. 5.8168 0.

LSCS-UFSAR TABLE 6.A-2 TABLE 6.A-2 REV. 0 - APRIL 1984 ACOUSTIC LOADING ON REACTOR PRESSURE VESSEL SHROUD TIME (msec) ACOUSTIC LOAD (kips) 0 0 1.2 0 1.6 150 2.0 320 2.5 650 2.8 250 3.0 100 3.2 0 LSCS-UFSAR TABLE 6.A-3 (SHEET 1 OF 2) TABLE 6.A-3 REV. 0 - APRIL 1984 RPV COORDINATES OF NODAL POINTS NODAL COORDINATES NODE NUMBER X-ORDINATE Y- ORDINATE Z-ORDINATE 1 -912.000 774.000 1563.000 2 -912.000 774.000 1556.000 3 -912.000 774.000 981.200 4 -912.000 774.000 740.000 5 -912.000 774.000 1356.000 6 -912.000 774.000 1316.000 7 -912.000 774.000 1279.200 8 -912.000 774.000 1240.400 9 -912.000 774.000 1201.600 10 -912.000 774.000 1163.600 11 -912.000 774.000 1141.700 12 -912.000 774.000 1125.700 13 -912.000 774.000 1065.200 14 -912.000 774.000 1035.200 15 -912.000 774.000 1021.300 16 -912.000 774.000 994.200 17 -912.000 774.000 1601.700 18 -912.000 774.000 1559.700 19 -912.000 774.000 1499.700 20 -912.000 774.000 1436.900 21 -912.000 774.000 1398.500 22 -912.000 774.000 1318.000 23 -912.000 774.000 1279.200 24 -912.000 774.000 1240.400 25 -912.000 774.000 1201.600 26 -912.000 774.000 1163.600 27 -912.000 774.000 1141.700 28 -912.000 774.000 1125.700 29 -912.000 774.000 1021.300 30 -912.000 774.000 1035.200 31 -912.000 774.000 1065.200 32 -912.000 774.000 1125.700 33 -912.000 774.000 1141.700 34 -912.000 774.000 1163.600 35 -912.000 774.000 1201.600 36 -912.000 774.000 1240.400 37 -912.000 774.000 1279.200 38 -912.000 774.000 1318.000 39 -912.000 774.000 1356.600 40 -912.000 774.000 1398.500 41 -912.000 774.000 1436.900 42 -912.000 774.000 1499.700 43 -912.000 774.000 1559.700 44 -912.000 774.000 1563.600 LSCS-UFSAR TABLE 6.A-3 (SHEET 2 OF 2)

TABLE 6.A-3 REV. 0 - APRIL 1984 NODAL COORDINATES NODE NUMBER X-ORDINATE Y- ORDINATE Z-ORDINATE 45 -912.000 774.000 1601.700 46 -912.000 774.000 1619.800 47 -912.000 774.000 1724.200 48 -912.000 774.000 1743.600 49 -912.000 774.000 1768.200 50 -912.000 774.000 1817.100 51 -912.000 774.000 1866.000 52 -912.000 774.000 1563.000 53 300.000 774.000 886.000 54 -912.000 774.000 446.000 55 -912.000 774.000 318.000 56 -912.000 774.000 0. 57 -912.000 774.000 740.000

LSCS-UFSAR TABLE 6.A-4 TABLE 6.A-4 REV. 0 - APRIL 1984 MAXIMUM MEMBER FORCES DUE TO ANNULUS PRESSURIZATION COMPONENT DESCRIPTION ELEMENT NUMBER FEEDWATER RECIRC. JET REACTION Top guide (L)

• 4 22.20 38.00 29.0 Core plate (L) 7 20.80 42.00 30.0 Fuel support (L) 8 19.00 69.00 74.0 CRD housing (L) 9.10 22.00 70.0 CRD housing (M)

.24 .56 1.9 Shroud head (L) 19 59.80 78.00 133.0 Shroud head (M) 19 6.40 5.90 6.1 Shroud support (L) 26 184.00 296.00 246.0 Shroud support (M) 26 19.80 40.00 22.0 Vessel skirt (L) 50 1220.00 3204.00 1858.0 Vessel skirt (M) 50 216.00 221.00 130.0 Pedestal cont. (L) 3 486.00 2325.00 859.0 Pedestal cont. (M) 3 326.00 680.00 206.0 Stabilizer (L)

III 1722.00 1949.00 746.0 CRD support beam (L) 4.50 27.00 50.0

• *(L) Load - 10 3 x lb (M) Moment - 10 6 x in. x lb All loads incorporate appropriate fact or to account for shell behavior

LSCS-UFSAR TABLE 6.A-5 TABLE 6.A-5 REV. 0 - APRIL 1984 MAXIMUM ACCELERATION

• DUE TO ANNULUS PRESSURIZATION (in./sec 2) COMPONENT DESCRIPTION NODE NUMBER FEEDWATER RECIRC. LINE BREAK JET LOAD P line 9 80 283 675 CRD guide tube 11 86 298 309 Separators 17 155 306 342 Head spray 51 178 416 898 Steam dryer 46 118 200 451 Feedwater sparger 43 109 157 538 Jet pump 38 133 362 406 RPV 30 62 253 514 RPV (bottom) 16 61 254 598 Shield wall 2 282 398 229 Top of shield wall 1 190 326 254 Fuel 5 74 198 394 Fuel 7 27 51 77 Fuel 9 80 283 675

• *All accelerations incorpor ate a factor to accoun t for shell behavior.

LSCS-UFSAR TABLE 6.A-6 REV.0 - APRIL 1984 TABLE 6.A-6 (SHEET 1 OF 3)

RELAP 4 INPUT DATA, RECIRCULATION LINE OUTLET BREAK 1 = LASALLE RPV-SHIELD ANNULUS PRESSURIZATION STUDY - NSLD CALC NO 3C7-0976-001 2

• PROJECT NO 4266-00 R.M. HOGAN - D.L. ROBINSON - NUCLEAR ANALYSTS 3

• RECIRCULATION OUTLET LINE BREAK 4

• 5

• CASE "A" BASE LISTING 12/27/76 6

• 7 *2345678901234567890123457890123457890123457890123457890123457890123457890 8

• PROBLEM DIMENSIONS 9

• CARD LDMP-NEDI-NTC-NTR P-NVOL-NBUB-NTDV-NJUN-NONE-NFLL-NONE 10 010001 -2 0 3 6 38 0 0 86 0 4 0 1 0 0 0 0 0 11

• 12 *PROBLEM CONSTANTS 13 010002 0.0 1.0 14

• 15

• TIME STEPS 16 030010 1 1 10 0 0.0001 1E-06 0.025 17 030020 1 1 5 0 0.001 1E-06 0.2 18 030030 1 1 1 0 0.01 1E-06 1.0 19

• 20

• TRIP CONTROLS 21 040010 1 1 0 0 0.2 0.0 *END PROBLEM ON ELAPSED TIME 22 040020 2 1 0 0 0.0 0.0

• ACTION #2 ON ELAPSED TIME (FILL) 23 040030 3 4 30 36 3.0 0.0

• ACTION #3 ON DP (OPEN VALVE) 24 040040 4 4 31 36 3.0 0.0

• ACTION #4 ON DP (OPEN VALVE) 25 040050 5 4 32 36 3.0 0.0

• ACTION #5 ON DP (OPEN VALVE) 26 040060 6 4 33 36 3.0 0.0

• ACTION #6 ON DP (OPEN VALVE) 27

• 28

• BEGIN VOLUME DATA 29

• 2345678901234567890123456789012345678901234567890123456789012345678901234567890 30 VOLUME B R PRESS TEMP QUAL VOLUME MT MIX TP FLOWA DIAMV ELEV 31 050011 0 0 15.45 -1. 0.946 100.6 5.07 5.07 0 18.40 0.0 755.29 32 050021 0 0 15.45 -1. 0.946 100.6 5.07 5.07 0 18.40 0.0 755.29 33 050031 0 0 15.45 -1. 0.946 100.6 5.07 5.07 0 18.40 0.0 755.29 34 050041 0 0 15.45 -1. 0.946 150.9 5.07 5.07 0 23.36 0.0 755.29 35 050051 0 0 15.45 -1. 0.946 150.9 5.07 5.07 0 23.36 0.0 755.29 36 050061 0 0 15.45 -1. 0.946 121.0 7.47 7.47 0 20.98 0.0 760.36 37 050071 0 0 15.45 -1. 0.946 121.0 7.47 7.47 0 20.98 0.0 760.36 38 050081 0 0 15.45 -1. 0.946 121.0 7.47 7.47 0 20.98 0.0 760.36 39 050091 0 0 15.45 -1. 0.946 181.5 7.47 7.47 0 25.64 0.0 760.36 40 050101 0 0 15.45 -1. 0.946 181.5 7.47 7.47 0 25.64 0.0 760.36 41 050111 0 0 15.45 -1. 0.946 39.87 6.92 6.92 0 10.02 0.0 767.83 42 050121 0 0 15.45 -1. 0.946 54.28 4.90 4.90 0 10.50 0.0 767.83 43 050131 0 0 15.45 -1. 0.946 61.94 4.90 4.90 0 10.50 0.0 767.83 44 050141 0 0 15.45 -1. 0.946 81.43 4.90 4.90 0 13.47 0.0 767.83 45 050151 0 0 15.45 -1. 0.946 80.54 4.90 4.90 0 13.47 0.0 767.83 46 050161 0 0 15.45 -1. 0.946 26.77 2.67 2.67 0 8.43 0.0 774.75 47 050171 0 0 15.45 -1. 0.946 52.18 4.69 4.69 0 10.30 0.0 772.73 48 050181 0 0 15.45 -1. 0.946 52.18 4.69 4.69 0 10.30 0.0 772.73 49 050191 0 0 15.45 -1. 0.946 78.28 4.69 4.69 0 13.27 0.0 772.73 50 050201 0 0 15.45 -1. 0.946 77.39 4.69 4.69 0 13.27 0.0 773.73 51 050211 0 0 15.45 -1. 0.946 67.48 6.41 6.41 0 12.44 0.0 777.42 52 050221 0 0 15.45 -1. 0.946 67.48 6.41 6.41 0 12.44 0.0 777.42 53 050231 0 0 15.45 -1. 0.946 67.48 6.41 6.41 0 12.44 0.0 777.42 54 050241 0 0 15.45 -1. 0.946 101.2 6.41 6.41 0 15.52 0.0 777.42 55 050251 0 0 15.45 -1. 0.946 101.2 6.41 6.41 0 15.52 0.0 777.42 56 050261 0 0 15.45 -1. 0.946 171.1 9.59 9.59 0 18.61 0.0 783.83 57 050271 0 0 15.45 -1. 0.946 155.8 9.59 9.59 0 18.61 0.0 783.83 58 050281 0 0 15.45 -1. 0.946 155.8 9.59 9.59 0 18.61 0.0 783.83 59 050291 0 0 15.45 -1. 0.946 171.1 9.59 9.59 0 18.61 0.0 783.83 60 050301 0 0 15.45 -1. 0.946 155.8 8.81 8.81 0 17.86 0.0 793.42 61 050311 0 0 15.45 -1. 0.946 153.4 8.81 8.81 0 17.86 0.0 793.42 62 050321 0 0 15.45 -1. 0.946 143.9 8.81 8.81 0 17.86 0.0 793.42 63 050331 0 0 15.45 -1. 0.946 164.1 8.81 8.81 0 17.86 0.0 793.42 64 050341 0 0 15.45 -1. 0.946 19.76 6.92 6.92 0 10.02 0.0 767.83 65 050351 0 0 15.45 -1. 0.946 19.52 4.92 4.92 0 7.04 0.0 769.56 66 050361 0 0 15.45 -1. 0.557 16315. 41.0 41.0 0 400. 0.0 793.42 67 050371 0 0 15.45 -1. 0.557 11665. 12.1 12.1 0 965. 0.0 781.32 68 050381 0 0 15.45 -1. 0.557 82775. 44.7 44.7 0 1850. 0.0 736.62 69 VOLUME B R PRESS TEMP QUAL VOLUME MT MIX TP FLOWA DIAMV ELEV 70

• 2345678901234567890123456789012345678901234567890123456789012345678901234567890 71

• END VOLUME DATA 72

• 73

• BEGIN HORIZONTAL FLOW PATHS WITHIN S.S. ANNULUS 74

• 2345678901234567890123456789012345678901234567890123456789012345678901234567890

LSCS-UFSAR TABLE 6.A-6 REV.0 - APRIL 1984 TABLE 6.A-6 (SHEET 2 OF 3)

RELAP 4 INPUT DATA, RECIRCULATION LINE OUTLET BREAK 75 JUNCT IN OT P V FLO AJUN ZJUN IN FJUF FJUR V C 1 EQ DM CC C E 76 080011 1 2 0 0 0.0 14.86 757.82 0.40 0.24 0.00 0 0 0 0 0.0 0.6 1 0 77 080021 2 3 0 0 0.0 14.86 757.82 0.40 0.24 0.00 0 0 0 0 0.0 0.6 1 0 78 080031 3 4 0 0 0.0 14.86 757.82 0.50 0.40 0.00 0 0 0 0 0.0 0.6 1 0 79 080041 4 5 0 0 0.0 14.86 757.82 0.60 0.42 0.00 0 0 0 0 0.0 0.6 1 0 80 080051 6 7 0 0 0.0 20.19 764.10 0.30 0.22 0.00 0 0 0 0 0.0 0.6 1 0 81 080061 7 8 0 0 0.0 20.19 764.10 0.30 0.22 0.00 0 0 0 0 0.0 0.6 1 0 82 080071 8 9 0 0 0.0 20.19 764.10 0.38 0.39 0.00 0 0 0 0 0.0 0.6 1 0 83 080081 9 10 0 0 0.0 20.19 764.10 0.45 0.41 0.00 0 0 0 3 0.0 0.6 1 0 84 080091 35 34 0 0 0.0 7.04 772.02 0.30 0.85 0.00 0 0 0 0 0.0 0.6 1 0 85 080101 34 11 0 0 0.0 10.02 771.29 0.32 0.35 0.00 0 0 0 0 0.0 0.6 1 0 86 080111 11 12 0 0 0.0 7.47 770.28 0.64 0.56 0.00 0 0 0 3 0.0 0.6 1 0 87 080121 12 13 0 0 0.0 7.09 770.28 0.90 0.84 0.00 0 0 0 0 0.0 0.6 1 0 88 080131 13 14 0 0 0.0 7.09 770.28 1.13 0.85 0.00 0 0 0 3 0.0 0.6 1 0 89 080141 14 15 0 0 0.0 7.09 770.28 1.35 1.64 0.00 0 0 0 3 0.0 0.6 1 0 90 080151 11 17 0 0 0.0 2.11 773.74 2.26 0.05 0.00 0 0 0 0 0.0 0.6 1 0 91 080161 16 17 0 0 0.0 3.87 776.09 1.46 0.38 0.00 0 0 0 3 0.0 0.6 1 0 92 080171 17 18 0 0 0.0 6.79 775.07 0.94 0.83 0.00 0 0 0 3 0.0 0.6 1 0 93 080181 18 19 0 0 0.0 6.79 775.07 1.17 0.85 0.00 0 0 0 3 0.0 0.6 1 0 94 080191 19 20 0 0 0.0 6.79 775.07 1.41 1.63 0.00 0 0 0 3 0.0 0.6 1 0 95 080201 21 22 0 0 0.0 9.83 780.62 0.65 0.36 0.00 0 0 0 3 0.0 0.6 1 0 96 080211 22 23 0 0 0.0 9.83 780.62 0.65 0.36 0.00 0 0 0 3 0.0 0.6 1 0 97 080221 23 24 0 0 0.0 9.83 780.62 0.81 0.67 0.00 0 0 0 3 0.0 0.6 1 0 98 080231 24 25 0 0 0.0 9.83 780.62 0.97 0.68 0.00 0 0 0 3 0.0 0.6 1 0 99 080241 26 27 0 0 0.0 14.68 788.62 0.65 1.28 0.00 0 0 0 3 0.0 0.6 1 0 100 080251 27 28 0 0 0.0 14.68 788.62 0.65 0.68 0.00 0 0 0 3 0.0 0.6 1 0 101 080261 28 29 0 0 0.0 14.68 788.62 0.65 1.28 0.00 0 0 0 3 0.0 0.6 1 0 102 080271 30 31 0 0 0.0 13.49 797.83 0.71 1.27 0.00 0 0 0 3 0.0 0.6 1 0 103 080281 31 32 0 0 0.0 13.49 797.83 0.71 1.13 0.00 0 0 0 3 0.0 0.6 1 0 104 080291 32 33 0 0 0.0 13.49 797.83 0.71 1.27 0.00 0 0 0 3 0.0 0.6 1 0 105 JUNCT IN OT P V FLO AJUN ZJUN IN FJUF FJUR V C 1 EQ DM CC C E 106

• 2345678901234567890123456789012345678901234567890123456789012345678901234567890 107

• END HORIZONTAL FLOW PATHS WITHIN 5.5. ANNULUS 108

• 109

• BEGIN VERTICAL FLOW PATHS WITHIN S.S. ANNULUS 110

• 2345678901234567890123456789012345678901234567890123456789012345678901234567890 111 JUNCT IN OT P V FLO AJUN ZJUN IN FJUF FJUR V C I EQ DM CC C E 112 080301 6 1 0 0 0.0 18.40 760.36 0.33 0.03 0.03 1 0 0 3 0.0 0.6 1 0 113 080311 7 2 0 0 0.0 18.40 760.36 0.33 0.03 0.03 1 0 0 3 0.0 0.6 1 0 114 080321 8 3 0 0 0.0 18.40 760.36 0.33 0.03 0.03 1 0 0 3 0.0 0.6 1 0 115 080331 9 4 0 0 0.0 23.36 760.36 0.22 0.03 0.03 1 0 0 3 0.0 0.6 1 0 116 080341 10 5 0 0 0.0 23.36 760.36 0.22 0.03 0.03 1 0 0 0 0.0 0.6 1 0 117 080351 34 6 0 0 0.0 3.61 767.83 1.40 1.13 0.90 1 0 0 3 0.0 0.6 1 0 118 080361 11 6 0 0 0.0 3.61 767.83 1.40 1.13 0.90 1 0 0 3 0.0 0.6 1 0 119 080371 12 7 0 0 0.0 7.22 767.83 0.62 1.13 0.90 1 0 0 3 0.0 0.6 1 0 120 080381 13 8 0 0 0.0 7.22 767.83 0.62 1.40 1.17 1 0 0 3 0.0 0.6 1 0 121 080391 14 9 0 0 0.0 10.84 767.83 0.41 1.13 0.90 1 0 0 0 0.0 0.6 1 0 122 080401 15 10 0 0 0.0 10.84 767.83 0.41 1.13 0.90 1 0 0 0 0.0 0.6 1 0 123 080411 12 17 0 0 0.0 8.56 772.73 0.56 0.46 0.00 1 0 0 3 0.0 0.6 1 0 124 080421 13 18 0 0 0.0 8.56 772.73 0.56 0.46 0.00 1 0 0 0 0.0 0.6 1 0 125 080431 14 19 0 0 0.0 14.50 772.73 0.33 0.59 0.00 1 0 0 0 0.0 0.6 1 0 126 080441 15 20 0 0 0.0 14.50 772.73 0.33 0.68 0.00 1 0 0 0 0.0 0.6 1 0 127 080451 34 16 0 0 0.0 5.94 774.75 0.94 0.03 0.00 1 0 0 3 0.0 0.6 1 0 128 080461 11 16 0 0 0.0 5.94 774.75 0.94 0.88 0.00 1 0 0 3 0.0 0.6 1 0 129 080471 16 21 0 0 0.0 7.72 777.42 0.44 0.67 0.00 1 0 0 0 0.0 0.6 1 0 130 080481 17 22 0 0 0.0 7.72 777.42 0.59 0.68 0.00 1 0 0 0 0.0 0.6 1 0 131 080491 18 23 0 0 0.0 7.72 777.42 0.59 0.68 0.00 1 0 0 0 0.0 0.6 1 0 132 080501 19 24 0 0 0.0 11.57 777.42 0.40 0.68 0.00 1 0 0 0 0.0 0.6 1 0 133 080511 20 25 0 0 0.0 11.57 777.42 0.40 0.68 0.00 1 0 0 0 0.0 0.6 1 0 134 080521 21 26 0 0 0.0 7.72 783.83 0.80 0.96 0.00 1 0 0 0 0.0 0.6 1 0 135 080531 22 26 0 0 0.0 3.86 783.83 1.60 1.04 0.00 1 0 0 3 0.0 0.6 1 0 136 080541 22 27 0 0 0.0 3.86 783.83 1.60 1.04 0.00 1 0 0 0 0.0 0.6 1 0 137 080551 23 27 0 0 0.0 7.72 783.83 0.80 0.69 0.00 1 0 0 3 0.0 0.6 1 0 138 080561 24 28 0 0 0.0 11.57 783.83 0.54 0.96 0.00 1 0 0 0 0.0 0.6 1 0 139 080571 25 29 0 0 0.0 11.57 783.83 0.54 0.97 0.00 1 0 0 0 0.0 0.6 1 0 140 080581 26 30 0 0 0.0 11.57 793.42 0.60 1.00 0.00 1 0 0 0 0.0 0.6 1 0 141 080591 27 31 0 0 0.0 11.57 793.42 0.60 1.04 0.00 1 0 0 0 0.0 0.6 1 0 142 080601 28 32 0 0 0.0 11.57 793.42 0.60 0.97 0.00 1 0 0 0 0.0 0.6 1 0 143 080611 29 33 0 0 0.0 11.57 793.42 0.60 1.00 0.00 1 0 0 0 0.0 0.6 1 0 144 JUNCT IN OT P V FLO AJUN ZJUN IN FJUF FJUR V C I EQ DM CC C E 145 *2345678901234567890123456789012345678901234567890123456789012345678901234567890 146 *END VERTICAL FLOW PATHS WITHIN S.S. ANNULUS 147

• 148

• BEGIN FLOW PATHS TO CONTAINMENT - PENETRATIONS WITH SHIELDING DOORS 149 *2345678901234567890123456789012345678901234567890123456789012345678901234567890 150 JUNCT IN OT P V FLO AJUN ZJUN IN FJUF FJUR V C I EQ DM CC C E 151 080621 30 36 0 1 0.0 9.27 797.83 1.05 0.75 0.00 0 0 0 0 0.0 0.6 1 0 152 080631 31 36 0 2 0.0 13.90 797.83 0.70 1.69 0.00 0 0 0 0 0.0 0.6 1 0 LSCS-UFSAR TABLE 6.A-6 REV.0 - APRIL 1984 TABLE 6.A-6 (SHEET 3 OF 3)

RELAP 4 INPUT DATA, RECIRCULATION LINE OUTLET BREAK 153 080641 32 36 0 3 0.0 13.90 797.83 0.70 1.69 0.00 0 0 0 0 0.0 0.6 1 0 154 080651 33 36 0 4 0.0 9.27 797.83 1.05 0.75 0.00 0 0 0 0 0.0 0.6 1 0 155 080661 33 36 0 0 0.0 2.04 797.83 1.05 1.72 0.00 0 0 0 3 0.0 0.6 1 0 156 080671 32 36 0 0 0.0 0.68 797.83 3.39 1.71 0.00 0 0 0 3 0.0 0.6 1 0 157 080681 31 36 0 0 0.0 2.10 797.83 1.11 1.71 0.00 0 0 0 3 0.0 0.6 1 0 158 080691 30 36 0 0 0.0 1.77 797.83 1.25 1.72 0.00 0 0 0 3 0.0 0.6 1 0 159 080701 36 37 0 0 0.0 400. 793.42 0.06 0.05 0.00 1 0 0 3 0.0 0.6 1 0 160 080711 29 37 0 0 0.0 1.39 788.62 1.50 1.73 0.00 0 0 0 3 0.0 0.6 1 0 161 080721 28 37 0 0 0.0 0.71 788.62 3.30 1.71 0.00 0 0 0 3 0.0 0.6 1 0 162 080731 27 37 0 0 0.0 0.71 788.62 3.30 1.71 0.00 0 0 0 3 0.0 0.6 1 0 163 080741 26 37 0 0 0.0 1.39 788.62 1.50 1.71 0.00 0 0 0 3 0.0 0.6 1 0 164 080751 37 38 0 0 0.0 965. 781.32 0.03 0.05 0.00 1 0 0 3 0.0 0.6 1 0 165 080761 20 38 0 0 0.0 1.25 775.07 1.97 1.71 0.00 0 0 0 3 0.0 0.6 1 0 166 080771 19 38 0 0 0.0 1.07 775.07 2.20 1.71 0.00 0 0 0 3 0.0 0.6 1 0 167 080781 18 38 0 0 0.0 0.71 775.07 3.30 1.71 0.00 0 0 0 3 0.0 0.6 1 0 168 080791 17 38 0 0 0.0 0.71 775.07 3.30 1.71 0.00 0 0 0 3 0.0 0.6 1 0 169 080801 15 38 0 0 0.0 1.25 770.28 1.97 1.71 0.00 0 0 0 3 0.0 0.6 1 0 170 080811 14 38 0 0 0.0 1.07 770.28 2.20 1.71 0.00 0 0 0 3 0.0 0.6 1 0 171 080821 13 38 0 0 0.0 1.47 770.28 1.50 1.71 0.00 0 0 0 3 0.0 0.6 1 0 172 080831 12 38 0 0 0.0 0.71 770.28 3.30 1.71 0.00 0 0 0 3 0.0 0.6 1 0 173 080841 11 38 0 0 0.0 0.71 772.02 3.30 1.71 0.00 0 0 0 3 0.0 0.6 1 0 174 080851 35 38 0 0 0.0 1.08 772.02 2.43 1.71 0.00 0 0 0 0 0.0 0.6 1 0 175 JUNCT IN OT P V FLO AJUN ZJUN IN FJUF FJUR V C I EQ DM CC C E 176 *2345678901234567890123456789012345678901234567890123456789012345678901234567890 177 *END FLOW PATHS TO CONTAINMENT - PENETRATIONS WITH SHIELDING DOORS 178

• 179 *BEGIN FILL PATH 180 *2345678901234567890123456789012345678901234567890123456789012345678901234567890 181 JUNCT IN OT P V FLO AJUN ZJUN IN FJUF FJUR V C I EQ DM CC C E 182 080861 0 35 1 0 0.0 1.00 772.02 0.00 0.00 0.00 0 0 0 3 0.0 1.0 1 0 183 JUNCT IN OT P V FLO AJUN ZJUN IN FJUF FJUR V C I EQ DM CC C E 184 *2345678901234567890123456789012345678901234567890123456789012345678901234567890 185

• END FILL PATH 186

• 187

• VALVE DATA CARDS 188 110010 -3 0.0 0.0 0.0 189 110020 -4 0.0 0.0 0.0 190 110030 -5 0.0 0.0 0.0 191 110040 -6 0.0 0.0 0.0 192

• 193

• FILL TABLE DATA CARDS 194

• FILL CONTROL 195 130100 16 2 0 0 1060. 533.

196

• CARD TIME FLOW TIME FLOW TIME FLOW 197 130101 0.0 0.0 0.002 371. 0.004 1194. 198 130102 0.006 2476. 0.008 4463. 0.010 7081.

199 130103 0.0173 18092. 0.019395 18092. 0.019405 9162.

200 130104 0.022 10573. 0.024 11445. 0.026 12147.

201 130105 0.028 12611. 0.030 12865. 0.031 12885.

202 130106 5.0 12885.

203

• 204 *2345678901234567890123456789012345678901234567890123456789012345678901234567890 205 *******************************************************************************

206

• MODEL REVISIONS 207 *******************************************************************************

208 LSCS-UFSAR TABLE 6.A-7 REV.0 - APRIL 1984

• TABLE 6.A-7 (SHEET 1 OF 3)

RELAP 4 INPUT DATA, FEEDWATER LINE BREAK 29 = LASALLE RPV-SHIELD ANNULUS PRESSURIZATION STUDY - NSLD CALC NO 3C7-0976-001 30

• PROJECT NO 4266-00 R.M. HOGAN - D.L. ROBINSON - NUCLEAR ANALYSTS 31

• FEEDWATER LINE BREAK 32

• 33

• CASE "C" BASE LISTING 1/3/77 34

• 35 *2345678901234567890123457890123457890123457890123457890123457890123457890123457890 36

• PROBLEM DIMENSIONS 37

• CARD LDMP----NEDI---------NTS--------NTRP---------NVOL------NBUB--------NTDV-------NJUN-------NONE--------NFLL---------NO NE 38 010001 -2 0 3 8 32 0 0 70 060 1 00000 39

• 40 *PROBLEM CONSTANTS 41 010002 0.0 1.0 42

• 43

• TIME STEPS 44 030010 1 1 50 0 0.0001 1E-06 0.025 45 030020 1 1 25 0 0.001 1E-06 0.2 46 030030 1 1 1 0 0.01 1E-06 1.0 47

• 48

• TRIP CONTROLS 49 040010 1 1 0 0 0.2 0.0 *END PROBLEM ON ELAPSED TIME 50 040020 2 1 0 0 0.0 0.0

• ACTION #2 ON ELAPSED TIME (FILL) 51 040030 3 4 23 30 3.0 0.0

• ACTION #3 ON DP (OPEN VALVE) 52 040040 4 4 24 30 3.0 0.0

• ACTION #4 ON DP (OPEN VALVE) 53 040050 5 4 25 30 3.0 0.0

• ACTION #5 ON DP (OPEN VALVE) 54 040060 6 4 26 30 3.0 0.0

• ACTION #6 ON DP (OPEN VALVE) 27 040070 7 4 27 30 3.0 0.0 *ACTION #7 ON DP (OPEN VALVE) 28 040080 8 4 28 30 3.0 0.0

• ACTION #8 ON DP (OPEN VALVE) 29

• 30

• BEGIN VOLUME DATA 31

• 2345678901234567890123457890123457890123457890123457890123457890123457890123457890 32 VOLUME B R PRESS TEMP QUAL VOLUME HT MIX TP FLOWA DIAMV ELEV 33 050011 0 0 15.45 -1. 0.946 150.9 5.07 5.07 0 23.36 0.0 755.29 34 050021 0 0 15.45 -1. 0.946 150.9 5.07 5.07 0 23.36 0.0 755.29 35 050031 0 0 15.45 -1. 0.946 150.9 5.07 5.07 0 23.36 0.0 755.29 36 050041 0 0 15.45 -1. 0.946 150.9 5.07 5.07 0 23.36 0.0 755.29 37 050051 0 0 15.45 -1. 0.946 181.5 7.47 7.47 0 23.80 0.0 760.36 38 050061 0 0 15.45 -1. 0.946 181.5 7.47 7.47 0 23.80 0.0 760.36 39 050071 0 0 15.45 -1. 0.946 181.5 7.47 7.47 0 23.80 0.0 760.36 40 050081 0 0 15.45 -1. 0.946 181.5 7.47 7.47 0 23.80 0.0 760.36 41 050091 0 0 15.45 -1. 0.946 159.7 9.59 9.59 0 17.83 0.0 767.83 42 050101 0 0 15.45 -1. 0.946 157.9 9.59 9.59 0 17.83 0.0 767.83 43 050111 0 0 15.45 -1. 0.946 157.9 9.59 9.59 0 17.83 0.0 767.83 44 050121 0 0 15.45 -1. 0.946 167.4 9.59 9.59 0 17.83 0.0 767.83 45 050131 0 0 15.45 -1. 0.946 67.48 6.41 6.41 0 12.44 0.0 777.42 46 050141 0 0 15.45 -1. 0.946 67.48 6.41 6.41 0 12.44 0.0 777.42 47 050151 0 0 15.45 -1. 0.946 67.48 6.41 6.41 0 12.44 0.0 777.42 48 050161 0 0 15.45 -1. 0.946 101.2 6.41 6.41 0 15.79 0.0 777.42 49 050171 0 0 15.45 -1. 0.946 101.2 6.41 6.41 0 15.79 0.0 777.42 50 050181 0 0 15.45 -1. 0.946 100.8 9.59 9.59 0 15.52 0.0 783.83 51 050191 0 0 15.45 -1. 0.946 110.0 9.59 9.59 0 15.52 0.0 783.83 52 050201 0 0 15.45 -1. 0.946 116.1 9.59 9.59 0 15.52 0.0 783.83 53 050211 0 0 15.45 -1. 0.946 171.1 9.59 9.59 0 18.61 0.0 783.83 54 050221 0 0 15.45 -1. 0.946 155.8 9.59 9.59 0 18.61 0.0 783.83 55 050231 0 0 15.45 -1. 0.946 45.22 10.58 10.58 0 13.39 0.0 793.42 56 050241 0 0 15.45 -1. 0.946 55.63 10.58 10.58 0 13.39 0.0 793.42 57 050251 0 0 15.45 -1. 0.946 116.2 10.58 10.58 0 16.48 0.0 793.42 58 050261 0 0 15.45 -1. 0.946 131.5 10.58 10.58 0 16.48 0.0 793.42 59 050271 0 0 15.45 -1. 0.946 176.7 10.58 10.58 0 19.57 0.0 793.42 60 050281 0 0 15.45 -1. 0.946 171.8 10.58 10.58 0 19.57 0.0 793.42 61 050291 0 0 15.45 -1. 0.946 16.12 4.00 4.00 0 5.42 0.0 796.75 62 050301 0 0 15.45 -1. 0.557 16315. 41.00 41.00 0 400. 0.0 793.42 63 050311 0 0 15.45 -1. 0.557 11665. 12.10 12.10 0 965. 00 781.32 64 050321 0 0 15.45 -1. 0.557 82775. 44.70 44.70 0 1850. 00 736.62 65 VOLUME B R PRESS TEMP QUAL VOLUME HT MIX TP FLOWA DIAMV ELEV 65 *2345678901234567890123457890123457890123457890123457890123457890123457890123457890 66

• END VOLUME DATA 67

• 68

• BEGIN HORIZONTAL FLOW PATHS WITHIN S.S. ANNULUS 69

• 2345678901234567890123457890123457890123457890123457890123457890123457890123457890 70

• JUNCT----IN----------0T-----------P-------------V--------------FLO---------AJUN--------ZJUN---------IN-----------FJUF-------FJUR------V -C-I-EQ---DM----------CC------------C-E 72 080011 1 2 0 0 0.0 14.86 757.82 0.60 0.29 0.00 0 0 0 0 0.0 0.6 1 0 73 080021 2 3 0 0 0.0 14.86 757.82 0.60 0.43 0.00 0 0 0 0 0.0 0.6 1 0 74 080031 3 4 0 0 0.0 14.86 757.82 0.60 0.29 0.00 0 0 0 0 0.0 0.6 1 0 75 080041 5 6 0 0 0.0 20.19 764.10 0.45 0.25 0.00 0 0 0 0 0.0 0.6 1 0 LSCS-UFSAR TABLE 6.A-7 REV.0 - APRIL 1984 TABLE 6.A-7 (SHEET 2 OF 3)

RELAP 4 INPUT DATA, FEEDWATER LINE BREAK 76 080051 6 7 0 0 0.0 20.19 764.10 0.45 0.41 0.00 0 0 0 0 0.0 0.6 1 0 77 080061 7 8 0 0 0.0 20.19 764.10 0.45 0.25 0.00 0 0 0 0 0.0 0.6 1 0 78 080071 9 10 0 0 0.0 13.88 772.63 0.69 1.31 0.00 0 0 0 0 0.0 0.6 1 0 79 080081 10 11 0 0 0.0 13.88 772.63 0.69 1.27 0.00 0 0 0 3 0.0 0.6 1 0 80 080091 11 12 0 0 0.0 13.88 772.63 0.69 1.31 0.00 0 0 0 3 0.0 0.6 1 0 81 080101 13 14 0 0 0.0 9.83 780.62 0.65 0.51 0.00 0 0 0 0 0.0 0.6 1 0 82 080111 14 15 0 0 0.0 9.83 780.62 0.65 0.51 0.00 0 0 0 3 0.0 0.6 1 0 83 080121 15 16 0 0 0.0 9.83 780.62 0.81 0.38 0.00 0 0 0 3 0.0 0.6 1 0 84 080131 16 17 0 0 0.0 9.83 780.62 0.97 0.39 0.00 0 0 0 3 0.0 0.6 1 0 85 080141 18 19 0 0 0.0 14.68 788.62 0.44 0.79 0.00 0 0 0 3 0.0 0.6 1 0 86 080151 19 20 0 0 0.0 14.68 788.62 0.44 0.83 0.00 0 0 0 3 0.0 0.6 1 0 87 080161 20 21 0 0 0.0 14.68 788.62 0.54 0.51 0.00 0 0 0 3 0.0 0.6 1 0 88 080171 21 22 0 0 0.0 14.68 788.62 0.65 0.85 0.00 0 0 0 3 0.0 0.6 1 0 89 080181 29 23 0 0 0.0 5.42 798.75 0.40 0.85 0.00 0 0 0 0 0.0 0.6 1 0 90 080191 23 24 0 0 0.0 16.19 798.75 0.20 0.33 0.00 0 0 0 3 0.0 0.6 1 0 91 080201 24 25 0 0 0.0 16.19 798.75 0.30 0.05 0.00 0 0 0 3 0.0 0.6 1 0 92 080211 25 26 0 0 0.0 16.19 798.75 0.40 1.33 0.00 0 0 0 3 0.0 0.6 1 0 93 080221 26 27 0 0 0.0 16.19 798.75 0.50 1.34 0.00 0 0 0 3 0.0 0.6 1 0 94 080231 27 28 0 0 0.0 16.19 798.75 0.60 0.39 0.00 0 0 0 3 0.0 0.6 1 0 95

• JUNCT----IN----------0T-----------P-------------V--------------FLO---------AJUN--------ZJUN---------IN-----------FJUF-------FJUR------V C-I-EQ---DM----------CC----------C-E 96 *2345678901234567890123456789012345678900123456789012345678900123456789012345678901234567890 97

• END HORIZONTAL FLOW PATHS WITHIN 5*5* ANNULUS 98

• 99

• BEGIN VERTICAL FLOW PATHS WITHIN S*S* ANNULUS 100 *0123456789012345678901234567890123456789012345678901234567890123456789012345678901234567890 101

• JUNCT----IN----------0T-----------P-------------V--------------FLO---------AJUN--------ZJUN---------IN-----------FJUF-------FJUR------V -C-I-EQ---DM----------CC---------C-E 102 080241 5 1 0 0 0.0 23.80 760.36 0.26 0.03 0.00 1 0 0 3 0.0 0.6 1 0 103 080251 6 2 0 0 0.0 23.80 760.36 0.26 0.03 0.00 1 0 0 3 0.0 0.6 1 0 104 080261 7 3 0 0 0.0 23.80 760.36 0.26 0.03 0.00 1 0 0 3 0.0 0.6 1 0 105 080271 8 4 0 0 0.0 23.80 760.36 0.26 0.03 0.00 1 0 0 0 0.0 0.6 1 0 106 080281 9 5 0 0 0.0 10.84 767.83 0.54 1.13 1.28 1 0 0 0 0.0 0.6 1 0 107 080291 10 6 0 0 0.0 10.84 767.83 0.54 1.13 1.28 1 0 0 3 0.0 0.6 1 0 108 080301 11 7 0 0 0.0 10.84 767.83 0.54 1.13 1.28 1 0 0 0 0.0 0.6 1 0 109 080311 12 8 0 0 0.0 10.84 767.83 .054 1.13 1.28 1 0 0 0 0.0 0.6 1 0 110 080321 13 9 0 0 0.0 7.22 777.42 0.83 0.96 0.00 1 0 0 3 0.0 0.6 1 0 111 080331 14 9 0 0 0.0 3.61 777.42 1.66 0.96 0.00 1 0 0 3 0.0 0.6 1 0 112 080341 14 10 0 0 0.0 3.61 777.42 1.66 0.96 0.00 1 0 0 3 0.0 0.6 1 0 113 080351 15 10 0 0 0.0 7.22 777.42 0.83 0.96 0.00 1 0 0 0 0.0 0.6 1 0 114 080361 16 11 0 0 0.0 10.84 777.42 0.56 0.96 0.00 1 0 0 0 0.0 0.6 1 0 115 080371 17 12 0 0 0.0 10.84 777.42 0.56 1.01 0.00 1 0 0 0 0.0 0.6 1 0 116 080381 18 13 0 0 0.0 7.71 783.83 0.80 0.68 0.00 1 0 0 0 0.0 0.6 1 0 117 080391 19 14 0 0 0.0 7.71 783.83 0.80 1.03 0.00 1 0 0 0 0.0 0.6 1 0 118 080401 20 15 0 0 0.0 7.71 783.83 0.80 0.96 0.00 1 0 0 0 0.0 0.6 1 0 119 080411 21 16 0 0 0.0 11.57 783.83 0.54 0.97 0.00 1 0 0 0 0.0 0.6 1 0 120 080421 22 17 0 0 0.0 11.57 783.83 0.54 0.96 0.00 1 0 0 0 0.0 0.6 1 0 121 080431 23 18 0 0 0.0 3.86 793.42 1.94 0.70 0.00 1 0 0 3 0.0 0.6 1 0 122 080441 24 18 0 0 0.0 3.86 793.42 1.94 0.70 0.00 1 0 0 0 0.0 0.6 1 0 123 080451 25 19 0 0 0.0 7.71 793.42 0.97 0.98 0.00 1 0 0 0 0.0 0.6 1 0 124 080461 26 20 0 0 0.0 7.71 793.42 0.97 1.00 0.00 1 0 0 0 0.0 0.6 1 0 125 080471 27 21 0 0 0.0 11.57 793.42 0.65 0.99 0.00 1 0 0 0 0.0 0.6 1 0 126 080481 28 22 0 0 0.0 11.57 793.42 0.65 0.97 0.00 1 0 0 0 0.0 0.6 1 0 127

• JUNCT----IN----------0T-----------P-------------V--------------FLO---------AJUN--------ZJUN---------IN-----------FJUF-------FJUR------V -C-I-EQ---DM----------CC---------C-E 128 *23457789012345678901234567890123456789012345678901234567890123456789012345678901234567890 129

• END VERTICAL FLOW PATHS WITHIN S*S*ANNULUS 130

• 131

• BEGIN FLOW PATHS TO CONTAINMENT - PENETRATIONS WITH SHIELDING DOORS 132

• 23457789012345678901234567890123456789012345678901234567890123456789012345678901234567890 133

• JUNCT----IN----------0T-----------P-------------V--------------FLO---------AJUN--------ZJUN---------IN-----------FJUF-------FJUR------V -C-I-EQ---DM----------CC---------C-E 134 080491 23 30 0 1 0.0 1.54 798.75 3.60 1.61 0.00 0 0 0 0 0.0 0.6 1 0 135 080501 24 30 0 2 0.0 3.86 798.75 1.30 1.07 0.00 0 0 0 0 0.0 0.6 1 0 136 080511 25 30 0 3 0.0 7.71 798.75 1.06 1.99 0.00 0 0 0 0 0.0 0.6 1 0 137 080521 26 30 0 4 0.0 7.71 798.75 1.06 1.99 0.00 0 0 0 0 0.0 0.6 1. 0 138 080531 27 30 0 5 0.0 9.27 798.75 0.79 2.40 0.00 0 0 0 0 0.0 0.6 1 .0 139 080541 28 30 0 6 0.0 11.57 798.75 0.65 1.82 0.00 0 0 0 0 0.0 0.6 1 0 140 080551 29 30 0 0 0.0 0.68 798.75 3.96 1.71 0.00 0 0 0 0 0.0 0.6 1 0 141 080561 28 30 0 0 0.0 0.68 798.75 3.96 1.71 0.00 0 0 0 3 0.0 0.6 1 0 142 080571 27 30 0 0 0.0 1.36 798.75 1.98 1.71 0.00 0 0 0 3 0.0 0.6 1 0 143 080581 26 30 0 0 0.0 1.36 798.75 1.70 1.73 0.00 0 0 0 3 0.0 0.6 1 0 144 080591 25 30 0 0 0.0 0.68 798.75 3.96 1.71 0.00 0 0 0 3 0.0 0.6 1 0 145 080601 30 31 0 0 0.0 400. 793.42 0.06 0.05 0.00 1 0 0 3 0.0 0.6 1 0 146 080611 22 31 0 0 0.0 0.71 788.62 3.86 1.71 0.00 0 0 0 3 0.0 0.6 1 0 147 080621 21 31 0 0 0.0 1.39 788.62 1.70 1.73 0.00 0 0 0 3 0.0 0.6 1 0 148 080631 20 31 0 0 0.0 0.68 788.62 2.98 1.74 0.00 0 0 0 3 0.0 0.6 1 0 149 080641 19 31 0 0 0.0 1.42 788.62 1.93 1.71 0.00 0 0 0 3 0.0 0.6 1 0 150 080651 31 32 0 0 0.0 965. 781.32 0.03 0.05 0.00 1 0 0 3 0.0 0.6 1 0 151 080661 12 32 0 0 0.0 2.89 772.63 0.90 1.71 0.00 0 0 0 3 0.0 0.6 1 0 152 080671 11 32 0 0 0.0 2.50 772.63 1.17 1.71 0.00 0 0 0 3 0.0 0.6 1 0

LSCS-UFSAR TABLE 6.A-7 REV.0 - APRIL 1984 TABLE 6.A-7 (SHEET 3 OF 3)

RELAP 4 INPUT DATA, FEEDWATER LINE BREAK 153 080681 10 32 0 0 0.0 2.50 772.63 1.17 1.71 0.00 0 0 0 3 0.0 0.6 1 0 154 080691 9 32 0 0 0.0 2.14 772.63 1.29 1.71 0.00 0 0 0 3 0.0 0.6 1 0 155

• JUNCT----IN----------0T-----------P-------------V--------------FLO---------AJUN--------ZJUN---------IN-----------FJUF-------FJUR------V -C-I-EQ---DM----------CC-----

---C-E 156

• 23457789012345678901234567890123456789012345678901234567890123456789012345678901234567890 157

• END FLOW PATHS TO CONTAINMENT - PENETRATIONS WITH SHIELDING DOORS 158

• 159

• BEGIN FILL PATH 160

• 23457789012345678901234567890123456789012345678901234567890123456789012345678901234567890 161

• JUNCT----IN----------0T-----------P-------------V--------------FLO---------AJUN--------ZJUN---------IN-----------FJUF-------FJUR------V -C-I-EQ---DM----------CC----

---

C-E 162 080701 0 29 1 0 0.0 1.0 789.75 0.0 0.0 0.0 0 0 0 3 0.0 1.0 1 0 163

• JUNCT----IN----------0T-----------P-------------V--------------FLO---------AJUN--------ZJUN---------IN-----------FJUF-------FJUR------V -C-I-EQ---DM----------CC-----

---C-E 164

• 23457789012345678901234567890123456789012345678901234567890123456789012345678901234567890 165

• END FILL PATH 166

• 167

• VALVE DATA CARDS 168 110010 -3 0.0 0.0 0.0 0.0 169 110020 -4 0.0 0.0 0.0 0.0 170 110030 -5 0.0 0.0 0.0 0.0 171 110040 -6 0.0 0.0 0.0 0.0 172 110050 -7 0.0 0.0 0.0 0.0 173 110060 -8 0.0 0.0 0.0 0.0 174

• 175

• FILL TABLE DATA CARDS 176

• FILL CONTROL 177 130100 4 2 0 0 1045. 420.

178

• CARD TIME FLOW TIME FLOW 179 1030101 0.0 14200. 0.001050 14200.

180 1030102 0.001060 21600. 1.00 21600. 181

• 182

• 23457789012345678901234567890123456789012345678901234567890123456789012345678901234567890 183

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • M 184

• MODEL REVISIONS 185 130101 0.0 7100. 0.001050 7100. 186 130102 0.001060 10800. 1.00 10800. CARD ABOVE IS REPLACEMENT CARD.

187

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • M 188

• LSCS-UFSAR TABLE 6.A-8 (SHEET 1 OF 2) TABLE 6.A-8 REV. 0 - APRIL 1984 FORCE CONSTANTS AND LOAD CENTERS FOR RECIRCULATION LINE OUTLET BREAK NODE v ss ELEVATIO N 1 3696.03 4948.35 757.82 15.0°, 345.0° 2 3696.03 4948.35 757.825 45.0°, 315.0° 3 3696.03 4948.35 757.825 75.0°, 285.0° 4 5464.86 7316.51 757.825 112.5°, 247.5° 5 5464.86 7316.51 757.825 157.5°, 202.5° 6 5828.44 7290.77 764.095 15.0°, 345.0° 7 5828.44 7290.77 764.095 45.0°, 315.0° 8 5828.44 7290.77 764.095 75.0°, 285.0° 9 8617.79 10779.95 764.095 112.5°, 247.5° 10 8617.79 10779.95 764.095 157.5°, 202.5° 11 2857.42 2503.87 771.290 22.5°, 337.5° 12 4038.29 3887.97 770.280 45.0°, 315.0° 13 4022.57 2990.40 770.280 75.0°, 285.0° 14 5970.91 5748.63 770.280 112.5°, 247.5° 15 5891.80 5523.42 770.280 157.5°, 202.5° 16 2234.85 2605.94 776.085 15.0°, 345.0° 17 3862.52 3683.01 775.075 45.0°, 315.0° 18 3862.52 3683.01 775.075 75.0, 285.0 19 5711.02 5445.58 775.075 112.5°, 247.5° 20 5631.91 5220.37 775.075 157.5°, 202.5° 21 5325.49 6256.20 780.625 15.0°, 345.0° 22 5325.49 6256.20 780.625 45.0°, 315.0°

LSCS-UFSAR TABLE 6.A-8 (SHEET 2 OF 2)

TABLE 6.A-8 REV. 0 - APRIL 1984 NODE v ss ELEVATION 23 5325.49 6256.20 780.625 75.0°, 285.0° 24 7874.13 9250.27 780.625 112.5°, 247.5° 25 7874.13 9250.27 780.625 157.5°, 202.5° 26 11713.96 11338.09 788.625 22.5°, 337.5° 27 11713.28 12957.61 788.625 67.5°, 297.5° 28 11713.28 12957.61 788.625 112.5°, 247.5° 29 11713.96 11338.09 788.625 157.5°, 202.5° 30 12864.45 12694.81 798.710 22.5°, 337.5° 31 12809.98 12622.87 798.710 67.5°, 297.5° 32 12934.41 14386.28 798.710 112.5°, 247.5° 33 12867.88 11885.05 798.710 157.5°, 202.5° 34 1557.92 2042.96 771.290 7.5°, 352.5° 35 1140.80 0.00 772.020 0.0°, 360.0°

LSCS-UFSAR TABLE 6.A-9 (SHEET 1 OF 2) TABLE 6.A-9 REV. 0 - APRIL 1984 FORCE CONSTANTS AND LOAD CENTERS FOR FEEDWATER LINE BREAK NODE v ss ELEVATION 1 5464.86 7316.51 757.825 22.5°, 337.5° 2 5464.86 7316.51 757.825 67.5°, 292.5° 3 5464.86 7316.51 757.825 112.5°, 147.50 4 5464.86 7316.51 757.825 157.5°, 202.5° 5 8617.79 10779.95 764.095 22.5°, 337.5° 6 8617.79 10779.95 764.095 67.5°, 292.5° 7 8617.79 10779.95 764.095 112.5°, 247.5° 8 8617.79 10779.95 764.095 157.5°, 202.5° 9 11681.94 11194.20 772.625 22.5°, 337.5° 10 11523.72 10743.78 772.625 67.5°, 292.5° 11 11523.72 10743.78 772.625 112.5°, 247.5° 12 11666.44 10309.43 772.625 157.5, 202.5 13 5325.49 6256.20 780.625 15.0°, 345.0° 14 5325.49 6256.20 780.625 45.0°, 315.0° 15 5325.49 6256.20 780.625 75.0°, 285.0° 16 7874.13 9250.27 780.625 112.5°, 247.5° 17 7874.13 9250.27 780.625 157.5°, 202.5° 18 7967.46 9359.90 788.625 15.0°, 345.0° 19 7841.24 7570.97 788.625 45.0°, 315.0° 20 7963.08 7716.94 788.625 75.0°, 285.0° 21 11713.96 11338.09 788.625 112.5°, 247.5° 22 11718.28 12957.61 788.625 157.5°, 202.5° 23 3530.66 4305.38 798.710 7.5°, 352.5°

LSCS-UFSAR TABLE 6.A-9 (SHEET 2 OF 2)

TABLE 6.A-9 REV. 0 - APRIL 1984 NODE v ss ELEVATION 24 4432.90 5207.63 798.710 22.5°, 337.5° 25 8726.85 9431.69 798.710 45.0°, 315.0° 26 8722.47 7788.73 798.710 75.0°, 285.0° 27 12872.20 13504.58 798.710 112.5°, 247.5° 28 12934.41 14386.28 798.710 157.5°, 202.5° 29 840.94 0.00 798.710 0.0°, 360.0°

LSCS-UFSAR TABLE 6.A-10 (SHEET 1 OF 2) DYSEA01 PROGRAM INPUT FOR JET LOAD FORCES TIME FUNCTION NUMBER = ( 1)

FUNCTION DESCRIPTION = ( RESTRAINT LOAD AT NODE 2 )

NUMBER OF ABSCISSAE = ( 51) FUNCTION SCALE FACTOR = ( 3.8880E-01)

TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION 0.00163 0. 0.00233 0. 0.00297 0. 0.00351 0. 0.00398 0. 0.00441 0. 0.00481 0. 0.00519 0. 0.00554 0. 0.00667 0. 0.00887 5.0588E 04 0.00787 1.0820E 05 0.00887 1.6604E 05 0.00987 2.2989E 05 0.01087 2.9304E 05 0.01187 3.5842E 05 0.01287 4.1875E 05 0.01387 4.7365E 05 0.01487 5.3891E 05 0.01587 5.8180E 05 0.01687 6.1887E 05 0.01787 6.4976E 05 0.01887 6.7423E 05 0.01987 6.9213E 05 0.02087 7.0347E 05 0.02187 7.0834E 05 0.02222 7.0857E 05 0.02242 7.0857E 05 0.02262 7.0857E 05 0.02283 7.0857E 05 0.02304 7.0857E 05 0.02325 7.0857E 05 0.02347 7.0857E 05 0.02370 7.0857E 05 0.02393 7.0857E 05 0.02417 7.0857E 05 0.02442 7.0857E 05 0.02467 7.0857E 05 0.02494 7.0857E 05 0.02522 7.0857E 05 0.02551 7.0857E 05 0.02582 7.0857E 05 0.02614 7.0857E 05 0.02649 7.0857E 05 0.02687 7.0858E 05 0.02728 7.0857E 05 0.02774 7.0857E 05 0.02827 7.0857E 05 0.02893 7.0857E 05 0.02992 7.0858E 05 0.19740 7.0857E 05 TIME FUNCTION NUMBER = ( 2)

FUNCTION DESCRIPTION = ( RESTRAINT LOAD AT NODE 3 )

NUMBER OF ABSCISSAE = ( 51) FUNCTION SCALE FACTOR = ( 6.1120E-01)

TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION 0.00153 0. 0.00233 0. 0.00297 0. 0.00351 0. 0.00398 0. 0.00441 0. 0.00481 0. 0.00519 0. 0.00554 0. 0.00587 0.

0.00687 5.0588E 04 0.00787 1.0820E 05 0.00887 1.6604E 05 0.00987 2.2989E 05 0.01087 2.9304E 05 0.01187 3.5642E 05 0.01287 4.1875E 05 0.01387 4.7365E 05 0.01487 5.3891E 05 0.01587 5.8180E 05 0.01687 6.1887E 05 0.01787 8.4976E 05 0.01887 6.7423E 05 0.01987 6.9213E 05 0.02087 7.0347E 05 0.02187 7.0834E 05 0.02222 7.0857E 05 0.02242 7.0857E 05 0.02202 7.0857E 05 0.02283 7.0857E 05 0.02304 7.0857E 05 0.02325 7.0857E 05 0.02347 7.0857E 05 0.02370 7.0857E 05 0.02393 7.0857E 05 0.02417 7.0857E 05 0.02442 7.0857E 05 0.02467 7.0857E 05 0.02494 7.0857E 05 0.02522 7.0857E 05 0.02551 7.0857E 05 0.02582 7.0857E 05 0.02614 7.0857E 05 0.02649 7.0857E 05 0.02687 7.0858E 05 0.02728 7.0857E 05 0.02774 7.0857E 05 0.02827 7.0857E 05 0.02893 7.0857E 05 0.02992 7.0858E 05 0.19740 7.0857E 05 TABLE 6.A-10 REV. 0 - APRIL 1984 LSCS-UFSAR TABLE 6.A-10 (SHEET 2 OF 2) DYSEA01 PROGRAM INPUT FOR JET LOAD FORCES TIME FUNCTION NUMBER = ( 3)

FUNCTION DESCRIPTION = ( BLOWDOWN LOAD AT NODE 34 & JET LOAD )

NUMBER OF ABSCISSAE = ( 51) FUNCTION SCALE FACTOR = ( -2.4270E 00)

TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION 0.00153 2.8666E 05 0.00233 2.8666E 05 0.00297 2.8666E 05 0.00351 2.8666E 05 0.00398 2.8666E 05 0.00441 2.8666E 05 0.00481 2.8666E 05 0.00519 2.8666E 05 0.00554 2.8666E 05 0.00587 2.8666E 05 0.00887 2.8666E 05 0.00787 2.8666E 05 0.00887 2.8666E 05 0.00987 2.8666E 05 0.01087 2.8666E 05 0.01187 2.8666E 05 0.01287 2.8666E 05 0.01387 2.8666E 05 0.01487 2.8666E 05 0.01587 2.8666E 05 0.01687 2.8666E 05 0.01787 2.8666E 05 0.01887 2.8666E 05 0.01987 2.8666E 05 0.02087 2.8666E 05 0.02187 2.8666E 05 0.02222 2.8666E 05 0.02242 2.8666E 05 0.02262 2.8666E 05 0.02283 2.8666E 05 0.02304 2.8666E 05 0.02325 2.8666E 05 0.02347 2.8666E 05 0.02370 2.8666E 05 0.02390 2.8666E 05 0.02417 2.8666E 05 0.02442 2.8666E 05 0.02467 2.8666E 05 0.02494 2.8666E 05 0.02522 2.8666E 05 0.02551 2.8666E 05 0.02582 2.8666E 05 0.02614 2.8666E 05 0.02649 2.8666E 05 0.02687 2.8666E 05 0.02728 2.8666E 05 0.02774 2.8666E 05 0.02827 2.8666E 05 0.02893 2.8666E 05 0.02992 2.8666E 05 0.19740 2.8666E 05

TIME FUNCTION NUMBER = ( 4)

FUNCTION DESCRIPTION = ( BLOWDOWN LOAD AT NODE 35 & JET LOAD )

NUMBER OF ABSCISSAE = ( 51) FUNCTION SCALE FACTOR = ( -2.4270E 00)

TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION TIME VALUE FUNCTION 0.00163 6.0254E 04 0.00233 6.0254E 04 0.00297 6.0254E 04 0.00351 6.0254E 04 0.00398 6.0254E 04 0.00441 6.0254E 04 0.00481 6.0254E 04 0.00519 6.0254E 04 0.00554 6.0254E 04 0.00587 6.0254E 04 0.00887 6.0254E 04 0.00787 6.0254E 04 0.00887 6.0254E 04 0.00987 6.0254E 04 0.01087 6.0254E 04 0.01187 6.0254E 04 0.01287 6.0254E 04 0.01387 6.0254E 04 0.01487 6.0254E 04 0.01587 6.0254E 04 0.01687 6.0254E 04 0.01787 6.0254E 04 0.01887 6.0254E 04 0.01987 6.0254E 04 0.02087 6.0254E 04 0.02187 6.0254E 04 0.02222 6.0254E 04 0.02242 6.0254E 04 0.02262 6.0254E 04 0.02283 6.0254E 04 0.02304 6.0254E 04 0.02325 6.0254E 04 0.02347 6.0254E 04 0.02370 6.0254E 04 0.02390 6.0254E 04 0.02417 6.0254E 04 0.02442 6.0254E 04 0.02467 6.0254E 04 0.02494 6.0254E 04 0.02522 6.0254E 04 0.02551 6.0254E 04 0.02582 6.0254E 04 0.02614 6.0254E 04 0.02649 6.0254E 04 0.02687 6.0254E 04 0.02726 6.0254E 04 0.02774 6.0254E 04 0.02827 6.0254E 04 0.02893 6.0254E 04 0.02992 6.0254E 04 0.19740 6.0254E 04 TABLE 6.A-10 REV. 0 - APRIL 1984 b I t-' t-' U;, OD +P FUNCTION OESCRlPTlON NWWER OF ABSCISSAE FUNCTION SCILE FACTOR T 1 tiE VALUE FUNCTION 0. -7.0000E-01 0.00050 -7.OOOOE-01 0.00100 -7.0000E-01 0.00150 8.0000E-01 0.00200 -3.0000E-01 0.00250 1.COOOE-01 0.00300 8.0000E-OI 0.00350 -1.7000E 00 0.00400 - 3,5000E 00 0.00450 -7.4000E 00 0.00500 -1.SBOOE 01 0.00550 -3.1400E 01 0.00600 -5.6600E 01 0.00650 -9.3700E 01 0.00700 -1.4950E 02 , 0.00750 -2.2730E 02 0.00800 -3.331OE 02 0.00850 -4.6970E 02 0.00900 -6.462OE 02 0.00950 -8.6620E 02 0.01000 -1,1345E 03 0.01050 -1,4570E 03 0.01100 -1.8377E 03 0.01150~-2.2840E 03 0.01200 -2.7986E 03 0.01250 -3.3861E 03 0.01300 -4.0496E 03 0.01350 -4.7924E 03 0.01400 -5.6173E 03 0.01450 -6.5270E 03 0.01500 -7.5213E 03 0.01550 -8.6015E 03 0.01600 -9.7706E 03 0.01660 -).1026E 04 0.01700 -1.2370E 04 O.O,l750 -1.38016 04 0.01800 -1.5322E 04 0.01850 -1.6927E 04 0.01900 -1.8619E 04 0.01B50 '-2.0396E 04 0.02000 -2.2260E 04 0.02050 -2.4210E 04 0.02100 -2.62496 04 0.02180 -2.8375E 04 0.02200 -3.0576E 04 0.02250 -3.2842E 04 0.02300 -3.8170E 04 0.02380 -3.7M7E 04 0.02400 -3.09b4E 04 a02450 -4.2392E 04 0.02500 -4.4844E 04 0.03000 -6.665GE 04 0.03500 - 7.0605E 04 TABLE 6.A-11 (SHEET 1 OF 32) TIME FORCE HISTORIES - RECIRCULATION LINE BREAK = ( 1) t FORCING FLiNCTlOEl AT NODE 30 00>0iY 0. -0.7 ) TIHEVALUE FUNCTION 0.00010 -7.000OE-01 0.00060 -7.0000E-01 0.00110 8.0000E-01 0.00160 8.0000E-01 0.00210 -1.4000E 00 0.00260 1.0000E-01 0.00310 5.0000E-01 0.00360 -2.0000E-01 0.04410 -3.1000E 00 0.00460 - 9.3000E 00 0.00510 -1.9600E 01 0.00560 -3.5800E 01 0.00610 -6.1300E 01 0.00660 -1.O28OE 02 0.00710 -1.6370E 02 0.00760 -2.4630E 02 0.00810 -3.6830E 02 0.00860 -8.0310E 02 0.00910 -6.8660E 02 0.009t0 -9.1550E 02 0.01010 -1.1942E 03 0.01060 -1.5278E 03 0.01110 -1.9219E 03 0.01160 -2.3808E 03 0.01210 -2.9103E 03 0.01260 -3.5123E 03 0.01310 -4.1918E 03 0.01360 -4.9504E 03 0.08410 -5.7918E 03 0.01460 -6.7180E 03 0.01510 -7.7294E 03 0.01b60 -8.8283E 03 0.01610 -1.0014E 04 0.01660 -1.1288E 04 0.01710 -1.265OE 04 0.01760 -1.4099E 04 0.01810 -1.5634E 04 0.01860 -1.7259E 04 0.01910 -1.896qE 04 0.Q1960 -2.0762E 04 0.02010 -2.2643E 04 0.02060 -2.4612E 04 0.02110 -2.6668E 04 0.02160 -2.8811E 04 0.02210 -3.1023E 04 0.02260 -3.3302E .04 0.02310 -3.5642E 04 0.02360 -3. 8025E 04 0.02410 -4.0440E 04 0.02460 -4.2881E 04 0.02600 -4.9743E 04 0.03100 -6.9403E 04 0.03600 -6.63206 04 TIM VALUE FUWTION 0.00020 -7.0000E-01 0,00070 -7.0000E-01 0..00120 6.0000E-01 0.00170 6.0000E-01 0.00220 1.000GE-01 0.00270 1.00GOE-01 0.00020 -6.00uOE-01 0.00370 -1,330OE 00 0,00420 -4.2000E 00 0.00470 -1.00QOE 01 0.00520 -2.OEuOE 01 0.00570 -3.B4oOE 01 0.00620 -6.8200E 01 0.00670 -1.lZ3OE 02 0.04720 -1.7660E 02 0.00770 -2.6470E 02 0.00820 -3.8390E 02 0.00870 -8.3GSOE 02 0.00920 -7.P81OE 02 0.00970 -9.6600E 02 0.01020 -1,2557E 03 0.01070 -1.6023E 03 0.01120 -2.OOBDE 03 0.01170 -2.4817E 03 0.01220 -3.024CE 03 0.01270 -3.6~16E 03 0.01320 -4.3337E 03 0.01370 -8.1121E 03 0.01420 -5.97ilE 03 0.01470 -6.9131L 03 0.01820 -7.B418E 03 0.01870 -S.OS&lE 03 0.01620 -1.02636 04 0.01670 -1.l553E 04 0.01720 -1.2932E 04 0.01770 -1.4399E 04 0.01820 -1.E023E 04 0.01870 -l.7593E 04 0.01920 -1.032tE 04 0.01970 -2.113cE 04 0.02020 -2.3031E 04 0.02070 -2.50t7E 04 , 0.02120 -2.708SE 04 0.02170 -2.9247E 04 0.02220 -3.1476E 04 0.02270 -3.3764E 04 0.02320 -3.61166 04 0.02370 -3.8b~CE 04 0.02420 -4.0927E 04 0.02470 -4.33/2E 04 0.02700 -8.4525):

04 0.03200 -7.125OE 04 0.03700 -6.5130E 04 TIME VALUE FUNCTION 0.00030 r7.0000E-01 0.000613 - 7. t100OE - 1) t O.CO130 6.0000E-01 0.OQleO 8.00UOC-01 0.00230 1.00OOE-01 0.00280 1,OOUOE-01 0.00330 -6.OOOOE-01 0.00360 -1.bOQOE 00 0.004:.0 -5. WOOL 00 0.00480 -1.1400E 01 0.00530 -2.3800E 01 0.00360 -4.4900E 01 0.00C30 -7.6600E 01 0.00680 -1.24POE OY 0.00730 -1.0370E 02 0,007e0 -2. RCiY1X 03 0.00830 -4.t110E 0; 0.00580 -5.rl70E 02 0.00930 -7.72'OE 02 0.00960 -1.0201E 03 0.01030 -1.3202E 03 0.01080 -1.6786E 03 0.01130 -2.0970E 03 0.01180 -2.8841E 03 0.01230 -3.8411E 03 O.Otp8O -3.7747E 03 0.01330 -4.4PC7E 03 0.01380 -5.2771E 03 0.01430 -6.1522E 03 0.01480 -7.1120E 03 0.01630 -0.1593E 03 0.01580 -9.e9PIE 03 0.01630 -1.0513E 04 0.01680 -1.1822L 04 0.01730 -1.3218E 0:' 0.01780 -1.4703E 01 0.01830 -1.G273E 04 0.01880 -1.7932f.

04 0.01930 -1.9675E 01 0.01980 -2.1504E 04 0.02030 -2.3421E 04 0.02080 -2.5424E 04 0.02130 -2.7614E C1 0.02180 -2.968GE 04 0.02230 -3.1928E 04 0.02280 -3.4231E 04 0.02330 -3.65916 04 0.02380 -3.8989E 04 0.02430 -4.1413E 04 0.02480 -4.3863E 01 0.02800 -5.BOGOE 04 0.03300 -7.2097E 04 0.03800 -6.1191E 04 T 1 ErE VALUE FIJNC T 1 ON Cl.6*13040

-7 O-'oCC-01 O.VOO"0 -7.0000~-01 0.00140 C.0300t-01 O.UOI9O 6.0000E-01

0. C*OE.40 I. OOOOE -0 1 0. LJo290 -B. OWOL 00 0.00?40 -1.7000E 00 0.00390 -2.4000E 00 0. (to440 -6.4000E 00 0.00490 -1.3GOOE 01 0.005-10-2.7100EOt . 0. C*C6PO -4. sr00E D t 0.00610 -6. SWDC 01 0.00690 -' 3c.EtbI 02 @. 00740 -2. OWOE 02 0.00700 -3 WDOE 02 U. (10813 -4.3L160L 02 0.00820 -6.OC30Z 02 11.0>040 -(t.lblOE 02 O.OuSS3 - 1 .U7EOE 03 0.01040 -1.3877E 03 0.0;090 -1.7bGSE:

03 0.01 140 -2.11190E 03 0.0 t 1 90 -2.6899E 03 0.01240 -3.2614E 03 0.01290 -3.9105E 03 0.01340 -4.6377E 03 0.01390 -5.4457E 03 0,01440 -6.3382E 03 0.01490 -7.3150E 03 0 01540 -6.3768E 03 0.OlBBO -Q.5304E 03 O.Ol64O -.t.O76LE 04 0.01690 -i.2034E 04 0.01740 -1.3509E 04 @ 01790 -I.t*blOE 04 l>.01840 -3.6C47lSE 04 0.01690 -1.6274E 04 0.01940 -2.0334E 04 0.01990 -2.1861E 04 0.020<0 -2.3814E 04 0.02050 -2.583SE 04 0.02140 -2.7942E 04 0.021EO -3.0130E 04 0.02240 -3.2384E 04 0,0228U -3.47005 04 0.02340 -3.7069E 04 0.02390 -3.9471E 04 0.02440 -4.1903E 04 0.02490 -4.4353E 04 0.02900 -6.3150E 04 0.03400 -7.1M4E 04 0.03900 -5.6687E 04 TABLE 6.A-ll (SHEET 2 OF 32) 0.04100 -4.6802E 04 0.04600 -2.82518 04 . 0.06100 -1.4412E 04 0.05600 -9.7343E 03 0.OG100 -7.2478E 03 0.06600 -6.6609E 03 0.07100 -1.0139E 03 0.07600 6.4342E 03 0.08100 1.0700E 04 0.08600 0.67tDE 03 0.09100 7.8560E 03 0.0BG00 0.6945E 03 0.10100 1,48726 04

0. I OCOO I . 900OE 04 0.11100 1.0668E 04 0.11600 1.3356E 04 0.12100 9.7634E 03 0.12600 2.8282E 03 0.13100 3.2196E 03 0.13600 b.0901E 03 0.14100 E.2325E 03 0.14600 8.46168 03 0.15100 1.74111 03 0,15600 7.00868 03 0.16100 0.3275E 03 0.16800 1.0173E 04 0,17100 9.4993E 03 0.17600 7.9701E 03 0.iEl00 0.452IE 03 0.10600 8.2672E 03 0.19100 4.6763E 03 0.l9600 4:4279E 03 O.PO200 4.OE97E 03 0.2lZOO 6.0242E 03 0.22200 4.3353E 03 0.23200 6.6013E 03 0.24200 6.968OE 03 0.28200 4.9933E 03 0.26200 3.18866 03 0.27200 3.b48bE 03 0.26200 4.1483E 03 0.29200 3.6756E 03 0.30200 3.0479E 03 0.31200 2.9937E 03 0.32200 2. 6345E 03 0.33200 2.3840E 03 0.34200 2.420SE 03 0.35200 2.904SE 03 0.36200 3.1333E 03 0.37209 2.6547E 03 0.38200 P.1689E 03 0.39200 1.7348E 03 0.40200 1.1799E 03 0.41200 4.6l6OE 02 0.42200 -3.9630E 02 0.43200 -6.4080E 02 0.44200 -0.1600E OI 0.48200 4.07906 02 0.46200 4.9370E 02 0.47200 4.7140E 02 0.48200 1.0456E 03 0.49200 1.0636E 03 TABLE 6.A-11 (SHEET 3 OF 32) T i ME FUNCT l CrN NUMGER = ( 2) 0\ P I I-' I-' % T: t-' I-' U3 CO s. FUNCT 1014 OLSCRI PT 1 ON = ( FORC~~!JFUNCIIONATNOOE 31 00>:15 0. 8.6 ) N~IHF.CF~

OF ABSC I SSAE = a 577) FUNCT ION SCALE FACTOR

'

• t 1 .0000E ,00) TlnE VALUE FUNCTION 0. 1 .6000E 00 0.00050 1 .6000E 00 0.00 100 1 . GOCrOE 00 0.00150 4.0000E-01 0.0020Lt - 6.3CIOOE 00 0.00250 -2.1200E 01 0.00300 -5.9300E 01 0.00350 - 1 .3870E 02 0.00th0 -2.7~20~ 02 0.00450 -4.0990E 02 0.00500 -8.122OE 02 0.00550 -1.224tE 03 0.00600 -1.7347E 03 0.00650 -2.3470E 03 0.00700 -3.0601E 03 0.00750 - 3,8777E 03 0.008oo -4.e035~ 03 0.00850 -5.8415E 03 0.00900 -7.0018E 03 0.00930 -8.2921E 03 0.01000 -0.726GE 03 0.01050 -1.131GE 04 0.01B00 -1.3077E 04 0.01150 -I.SOlPE 04 0.01200 -1.7128E 04 0.01250 -1.9422E 04 0.01300 -2. lSlOE 04 0.01350 -2.4612E 04 0.01400 -2.7541E 04 0.01450 -3.0709E 04 0.01900 -3.41326 04 0.01:*50 -3.7823E 04 0.01600 -4.1795E 04 0.01650 -4.60ClE 04 0.01700 -5.0637E 04 0.01750 -5.3332E 04 0.01800 -6.0755E 04 0.01850 -6.6312E 04 0.01900 -7.2204E 04 0.01950 -7.8419E 04 0.02000 -8.4937E 04 0.02030 - 0.1708E 04 0.02100 -9.84966 04 0.02150 -1.05016 05 0.02200 - 1 .0704E 05 0.02250 -1.12546 03. 0.02300 -1.1759E 05 0.02350 -1.22166 05 0.02400 - 1 .26266 05 0.02450 -1.2973E 05 0.02500 -1.32536 05 0.03000 -1.339lE 05 0.03500 -1.OP59E 05 TI6E VALUE FUNCTl OW 0.00010 1.6000E 00 0.00060 1.6000E 00 0.00110 4.0G00E-01 0.00160 -2.5000E 00 0.0021 0 - 7.5000E 00 0.00260 -2,6100E 01 0.00310 -7.2100E 01 O.OOtG0 -1.6110E 02 0 00410 -3.1510E 02 0.004GO -5.8SlOE 02 0.00510 -8.8650E 02 0.00560 -t.3180E 03 0.00610 -1.8892E 03 O.OO6GO -2.48166 03 0.00710 -3.2160E 03 0.00760 -4.05446 03 0.00810 -5.0015E 03 0.00860 -6,0639E 03 0.00010 -7.2489E 03 0.009GO -8. tiC.73E 03 0.01010 -1.0032E 04 O.OlOG0 -1.1654E 04 0.01110 -1.3453E 04 0.01160 -1.5426E 04 0.01210 -1.7574E 04 0.01260 -1.9903E 04 0.01310 -2.2433E 04 0.01360 -2.5178E 04 0.01410 -2.8154E 04 0.01460 -3.1373E 04 0.01510 -3.4648E 04 0.01560 -3,8594E 04 0.01610 -4.2G24E 04 0.01660 -4.GS51E 04 0,01710 -6.l569E 04 0.01760 -5.G552E 04 0,01810 -6.1640E 04 '0,01860 -6.74G4E OA 0.01010 -7.3424E 04 0.019GO -7.96986 04 0.02010 -8.6274E 04 0.02060 -9.3070E 04 0.02110 -0.0764E 04 0.02160 -1.0235E 00 0.02210 -1.O817E 05 0.02260 -1.1350E 05 0.02310 -1.1054E 05 0.02360 -1.2302E 05 0.02410 -1.2701E 05 0.024GO -1.3035E 05 0.02600 -1.3671E 05 0.03100 -1.2963E 05 0.03600 -I.O508E 05 TIME VALUE FUNCTIOld TI 0.00020 l . t000E 00 0.00070 1 .G000E 00 0:00120 4.00GOE-01 0.00170 -2.500OE 00 0.00220 -1.0000E 01 0.00270 -3.2300E Ot 0.00320 -6.3700E 01 0.0037U -1.8GaOE 02 0.90420 -3.S740E 02 0.00470 -6.13906 02 0,00520 -9.65POE 02 0.00570 -1.4174E 03 0.00L20 -1.EC*6E 03 0.00670 -2.6185E 03 0.00720 -3.3748E 03 0.00770 -4.23b4E 03 0.00820 -5.2038E 03 0.00870 -6.2398E 03 0.00920 -78 6020E 03 0.00@70 -8,8472E 03 0.01020 -1.0313E 04 0.01070 -1.199YE 04 0.01120 -1.3633E 04 0.01170 -1.58stF 04 0.01 220 - \ .80:.4". 04 0.01270 -2.0393E 04 0.01320 -2.296bE 04 0.01370 -2.57*1* 04 0.01420 -2.87796 04 0.01470 -3.204EE 04 0.01520 -3.5377E 04 0.01570 -3.8377E 04 0.01620 -4.3466E 04 0.016i0 -4.78C4E 04 0.01720 -5.25h4E 04 0.01770 -b.7beZE 04 0.01820 -6.2937E 04 0.01870 -6.8G28E 04 0.01920 -7.4651E 04 0.01970 -8.ODGOE 04 0.02020 -8.7F2lE 04 0.02070 -B.4425E 04 0.02120 -1.G111E 05 0.02170 -1.03Y4E 05 0,02220 - 1.09; 9E 05 0.02270 -1.1462E 05 0.02320 -1.194SE 05 0.02370 - 1.23ebE 05 0.02420 -1.2773E 05 0.02470 -1.30b3E 05 0.02700 - 1.386rE 05 0 03200 -1.2476E 05 0.03700 -1.0102E 05 IME VALUE FUrICT 11. rl f 1 EZ VFLlE FJNCT I ON 0.00039 1 .60SOE 00 0.00040 1 .6000E 00 0. OOC~PO t . eooo~ oo 0. [10(1~'r0 1 .~OOOE 00 0.00130 4.0000E->1 , 0.00140 4.0000E-01 0.00180 -3.000OE 00 0.001ZO -5.OOOOE 00 0. CICl230 - 1. T70nE OI 0.00240 - 1.7400E 01 0.00280 -3. E8CtOE 01 0. O'>iQO -4.8400E 01 0,00330 - 8 . OOGOE 02 0.00340 -1.l89OE 02 fi.00360 -2.1410E 09 P.00393 -2.4470E 02 Lo.00-430

-4.0170E 02 :. , CtCi,i..:

3 -4. 4C-lOE 02 0.00400 -6.7620E 02 .O. 00490 -7.4260E 02 . 0.09530 -1.04765 03 0,03540 -1.1339E 03 0.00504 -\.61H2E 03 O. UO5: 0 -1 . G257t 02 ' 0.00630 -2,0!!112E 03 0.0U640 -2.2161E 03 0. OOGGO -.?.7G::3E 03 G .00600 -2.6093E 05 0.00730 -3.F371E 03 0.00740

-3.70Y3E 03 ' 0.007t;O -4. rPll2E 03 O.UCl700 -4.6101E 03 0.00630 -5.4121E 03 CI . 00040 -6. G2!:,2E 03 0.00800 -6.5221E 03 0.OOWO -6.75b2E 03 O.O@93i1 -7.76116 03 0.00L40 -8. OLZOt 03 0.00980 -0.1341E 03 . 0.00990 -0.4277E 03 0.01030 -1.06596 04 G.01040 -1.09E4E 04 0.01000 -1.23636 04 0.01090 -1.2710E 04 0.01130 -1.4220E 04 0.01140 -1.4616E 04 0.01180 -1.E264E 0" O.OtlE0 -1'.6GS4E 04 0.01230 -1.C4ClE 0.: 0.01240 -I. e948E 04 0.01200 -2. C.608E 04 0.01290 -2.t397E 04 0.01330 -3.3SCl4E 04 0.01340 -2.4354E 04 0.01360 -2.63COE 04 0.01390 -2.6E36E 04 0.01430 -2.9412E 04 0.01440 -3.0056E 04 0.0l4h0 -3.Z732E 01 0.01450 -3.3427E 04 a0.01530 -3.C314E 04 0.01540 -3.7063E 04 0.01560 --.!.0171E 04 0.01090 -4.OY76E 04 0.01650 -4.43185 04 0.01640 -4.5165E 04 0.016(*0 -4.e>b9~ w 0.01690 -4.S5SCE 04 0.01730 -5.ZR3SE 04 0.01740 -5.4517E 04 0.017bO -5.8t27E 01 0.017YO -5.9G63C 02 0.01830 -6.404PE 04 0.01#~?0 -6.5173E 04 0.018~0 -~.CBII~E 0.1 0.01~90 -7.0~98~ 04 0.01930 -7.5900E 0.4 0.019'0 -7.7158E 04 0.01900 -8.2266E 04 0.019EO -8.3bl)E 04 0.02030 -8.8979E 04 0.02040 -9.0346E 04 0.02080 -9.5777E 04 0.02090 -9.7118E 04 0.02130 -1.0242E 05 0.02140

-1.0372E 05 0.02180 -1.0422E 05 0.02190 -1,058SE 05 0.02230 -1.1039E 05 0.02240 -1.1147E 05 . 0.02280 - 1 .16(.26 05 0.02290-1.166lE05 0.02330 -1.20396 05 0.02340 -1.2129E 05 0.02360 - 1.2468E 05 0.02390 -1.2549E 05 0.02430 - 1.2642E 05 0.02440

-1.2909E 05 0.02480 -1.31496 05 0.02490 -1.3203E Ob 0.02800 -1.3909E 05 0.02900 - 1.3727E 03 0.03300 -1.19626 05 0.03400 -1.tA45E 05 0.03800 -9.7482E 04 0.03900 -9.4409E 04 t-' Cn 0 Cn 1 C 'TI !J3 (SHEET 4 OF 32) 0.04200 -8.LG04E 04 0.043C10 -0.2284E 04 0.04700 -6.6040E 04 0.049012 -6.0160E 0.1 0.05200 -3.9849E 04 0.06300 -3.5013E 04 0.05700 -1 .E-:CBE 04 0. 0560i) -I . dt-'~?E -11 0.06200 -1.4060E 02 0.06300 2.7473E 03 0.06700 1.1216E 04 0.06800 1.2336E 0.1 0.07200 1.20e8E 04 0.07300 1 . 1050E 04 0.07700 6.3317E 03 0.07800 5.C937E 03 0.06200 2.6611E 03 0.08300 0.22'/OE 03 0.08700 1.9684E 04 0.08800 2.27;:.?1500.1 0.09200 3.2746E 04 . 0.095C10 5.447C1E 09 0.09700 3.7220E 04 . 0.09@0o ::.Fe16E 04 ' 0.10200 3.2691E 04 0. 103llrJ G. 1OI.iE 01 0.10700 2.4912E 04 0. 1 OBOO 2.37tIZE 04 0.11200 2.1228E 04 0.11300 2.1171E 01 0.11700 2.21elE 04 0.1 180il 2.25'77E 04 0.12200 2.320CE 04 0. 12311+-, 2. ::+ctni: 01 0.12700 2.OSEUE 04 0.12800 l.QC27E 01 0.13200 1.5311E 04 0.13300 1.42665 ?)4 0.13700 1 . O86BE 04 0. 1 :I800 1 . 028 .;7E 0.1 0.14200 9.3567E 03 O.1430o Si.C1>::71'.

133 0.14700 1.2628E 04 0. 14< on 1.9777~ I):! 0.15200 1.8370E 04 0.1 530CI 1 ,626GE 01 0.15700 2.0808E 04 0. 1 Ua<IO 2. C16','4E I) 0.16200 1.86el~ 04 0. 1 6300 I . DO l OE 0.1! 0.16700 1

.6627E 04 0.16800 1.5220E 04 0.17200 1.4322E 04 0.17300 1 .$2<15~ 04 0.17700 1.38G"E 04 0. 17800 1 .36U1 E 01 0.18200 1.3016E 04 0.18300 1.26/2E 04 0.18700 1.2223E 04 0.18800 1.207BE 04 0.19200 1.1450E 04 0.19300 1.1258E 04 0.19700 1.0223E 04 0.18800 9.8B71E 03 0.20400 7.7301E 03 0.20600 7.20!.* 1 E 03 0.21400 7.5759E 03 0.21600 8.24046 03 0.22400 1.1125E 04 0.22600 1.1499E 01 0.23400 1.0845E 04 0.23600 1 .035.;trC 0;: 0.24400 8.8367E 03 0,24600 8.7041:tE 03 0.25400 8.5264E 03 0. 256fln 8. 37;.7E 03 0.26400 7.1166E 03 0,266ilO 6.7!?97E 03 0.27400 b.7960E 03 0.2i6(10 5. C.3ClDE 03 0.2E400 5.6611E 03 0, E~'6DO 5. tr711.~1E 03 0.29400 5.44E2E 03 0.296i10 f..3G-.GE 03 0.30400 4.8238E 03 0.3CI6C10 4.642t E 03 0.31400 3.CE.12E 03 0.3161*,3

a. 7012F 03 0.32400 3.1980E 03 0.3260Ck 3.14..1E.

03 0.33300 3.2030E 03 0.33600 3.26:IUE 113 0.34400 3.5274E 03 0.31600 3.5712E 03 0.35400 3.S911E 03 0.35600 3.5Y90E 03 0.36400 3.2715E 03 0.36600 3.2419E 09 0.37400 2.6359E 03 0.37600 2.9171E 03 0.38400 1.7712E 03 0.38600 1.3832E 03 0.39400 5.41DOE 02 0.39600 4.6180E 02 0.40400 -1.7140E 02 0.40600 -5.7320E 02 0.41400 -1.9639E 03 0.41 600 .r2.22Ci5E 03 0.42400 -3.1568E 03 0.42600 -3.2164E 03 0.43400 -3.0540E 03 0.43600 -2.9403E 03 0.44400 -2.C885E 03 0.44600 -2.6729E 03 0.45400 -2.59CZE 03 0.45600 -2.5054E 03 0.46400 -1.7780E 03 0.46600 -1.blt6E 03 0.47400 -Q.OOEOE 02 0.47600 - 0.78ilOE 02 0.48400 -1.2094E 03 0.48600 -1.31C;lE 03 0.49400 -1.4910E 03 0.4TG00 - 1 .47 /%li 03 0.0"*00 -7.8471E 04 U . Ol:;C.lfl

-5, 52CbVE 06 0 Of. fnr? -3.0.15,'E 04 (I. ,.,9. 8-LO - 1 . 4 t.r!;c 08 0. 0'.--00 6. 3:> l9E 03 0. OC:'CCI 1 .2617:i 04 0. 0?-400 9. P,*52E 03 0.079r10 5.5~103~ 03 O.OE"(IQ l.lbO65 0.1 0.01300 2.6519E 04 D . 1:lS.,iOO

3. P61.15t: Or: 0. OS".f*O 3. 63 I 1C 0.4 C).)(..?CI(I k.C?.?:'f f,1+ O.lOP*,O 2.2713t: 04 0. ll,!c10 2. 126.11- 04 0. 1 1 O*'*fl 2. 290':E 04 0. 1 z-fl!~ 2, 2, 23, (1.3 0. 12903 1 .8:~9;:L 04 0. I 2400 1.32e::~ 0.q 0.1, ;it9 9.7~~81E 05 O.lr?.,'~.*O l.oil!-.:'-

114 0. IT,: <*<f 1 ,~"L!..L 04' 0. t t..?no I . 9na59r 04 (-6. 1 .>< :I(* 2. o:?!>l!c.

04 O.lF?l*+ 1.7311E 04 0. 1 r.x91:~(1 1 .4QC:9E 04 0. 17-2l.lF 1 .4 1 @6E 0.4 0.17Di.13 1.3540E 04 (I. 1 B.?l'*(l t ,2700E 04 0.189ircl 1.19%9E 04 0.19400 1.1039E 04 0.19900 9.5314E 03 0. PORO 6.303dE 03 0.21800 9.0164E 03 0.22r *r~ 1.164OE 04 0.23E .#I* 9.061 OE 03 0. 2.A800 6.6502E 03 0.2: I-lit '8. lll0E: 03 0.2C.800 6.4364E 03 0.2 VC.--O 5. G2ec 03 0.2b. 5.571bt 03 (1. 2?bilC1 5. 2:1UOL 03 (1.30f *3 4.4527E 00 (I. :.L 1 (.. I;* :. , C : C3L 0:* 0.32eno 3.1228~ 03 o ri3'.+.10

3. 3331 F: 03 0.31311n S.+lieE 03 0.3: 900 3.513.4E 03 o.~...~c*o 3.2023~ 03 0.37800 2.5401E 03 0.3C.8130 1.0451E 03 0.39800 4.1730E 02 0.40800 -0.1Q90E 02 0.41800 -2.5996E 03 0.42300 -3.2570E 03 0.43800 -2.8511E 03 0.44300 -2.6675E 03 0.458fi0 '2: 360OE 03 0.4G600 -1.5199E 03 0.47 900 -9.13POE 02 ' 0.4590' - 1 .401,:?E 03 0.4P'Fll:~

-1.4614E 03 TABLE 6.A-11 (SHEET 5 OF 32) TlHE FUNCTION NUHBER 8 ( 3) FUNCTION DESCRIPTION

• t FORCING FUNCTION AT NODE 32 00#02I 0,. 0.9 8 NJtIDEC Of ABSC I SSAE 8 ( 677) FUNCTION SCALE FACTOR 8 6 1.0000E 00) HE VALUE FUNCTION 0. 9. OOOOE -01 0.00050 9.0000E-01 0.00100 9.0000E-01 0.00150 2.0000E-01 0.00200 -3.5000E 00 0.00250 -1.18OOE 01 0.00300 -3.3100E 01 0.00350 -7.7400E 01 0.00400 -1.652OE 02 0.00450 -2.7880E 02 0.00500 -4.5290E 02 0.00550 -6,8290E 02 0.00600 -9.6740E 02 0.00650 -1.3089E 03 0.00700 -1.7066E 03 0.00750 -2.1625E 03 0.00800 -2.6788E 03 0.00850 -3.2577E 03 0.00900 -3.9047E 03 0.00950 -4.6243E 03 0.01000 -5.4243E 03 0.01050 -6.3104E 03 0.01100 -7,2927E 03 0.01150 -8.3746E 03 0.01200 -9.5GlSE 03 0.01250 -1.0831E 04 0.01300 -1.2219E 04 0.01350 - 1.3726E 04 0.01400 -1.5359E 04 0.01450 -1.712GE 04' 0.01500 - 1.90356 04 0.01550 -2.1093E 04 0.01600 -2.3308E 04 0.01 650 - 8.5687E 04 0.01700 -2,8239E 04 0.01750 -3.096UE 04 0.01800 -3.3881E 04 0.01850 -3.G980E 04 0.01900 -4.026GE 04 0.01950 -4.37326 04 0.02000 -4.7367E 04 0.02050 -5.1143E 04 0.02100 -5.4906E 04 0.02150 -6.8564E 04 0.02200 -6,9694E 04 0.02250 -6.2760E 04 0.02300 -6.557tiE 04 0.02350 -6.81281:

04 0.02400 -7.04tlE 04 0.02450 -7.2349E 04 0.02500 - 7.3909E 04 0.03000 -7.4676E 04 0.03500 -6.11t6E 04 TIME VALUE FUNCTION TI 0.00010 9.0000E-01 0.00060 S.0000E -01 0.00110 2.0000E-01 0.00160 -1.4000E 00 0.00210 -4.2000E 00 0.00260 -1.4600E 01 0.00310 -4.0200E 01 0.00360 -8.98006 01 0.00410 -1.7570E 02 0.00460 -3.09505 02 0.00510 -4.9440E 02 0.00560 -7.3bOOE 02 0.00610 -1.03125 03 0.00660 -1.3839E 03 0.00710 -1.7S35E 03 0.00760 -2.2610E 03 0.00810 -2.7892E 03 0.00860 -3.3817E 03 0.00910 -4.0425E 03 0.00960 -4,7777E 03 0.01010 -6.5943E 03 0.01060 -6.4992E 03 0.01110 -7.5022E 03 0.01160 -8,6028E 03 0.01210 -9.8004E 03 0.01260 -1.1099E 04 0.01310 -1.2511E 04 0.01360 -1.4041E 04 0.01410 -1.6701E 04 0.01460 -1.7496E 04 0.01510 -1.94346 04 0.01560 -2.1523E 04 0.01610 -2.3770E 01 0.01660 -2.6183E 04 0.01710 -2.87706 04 0.01760 -3.1537E 04 ' 0.01810 -3.44865 04 0.01860 -3.7623E 04 0.01010 -4.0947E 04 0.01960 -4.4446E 04 0.02010 -4.8113E 04 0.02060 -5.1903E 04 0.02110 -5.5647E 01 0.02160 -5.7077E 04 0.02210 -6.0325E 04 0.02260 -6.3344E 04 0.02310 -6.6106E 04 0.02360 -6.8602E 04

• 0.024 10 -7.0829E 04 0.02460 -7.2692E 04 0.02600 -7.6240E 04 0.03100 -7,229tE 04 0.03600 -6.86005 04 T I ME VALUE FUFlCT l ON TI IIFT VALIF FUNCT I OM o.ooo:.o ~.~OIICE-01

0. UC*~?O e. ~000~~ -(I t 0.00060 0. UCt00E - 0 1 O.OOOI1" R . 001lOE - 0 1 0.00130 icl.U~:~.OE-l)1 (I. 00lr.!r 2.Or~oUt-01 O.OOIE0 -Y.l13rli)E Oi) ' 0,UUlCICt

-2.C301)f:

00 0, 60230 - 7 . Lc.oi~rrE OG ti. il*)~'~ -9. ?r%jaE oo 0.00280 -2.22210E 0 1 0. 002Gt :b -2. 70-.10E 0 1' 0.00330 -5.E OOE 01 0.003t:O -6.6::It ,E 01 0.063t.0 -1 .19.!OE 02 0 . ft0:rgn - 1 , 3ti!-.l~h' 02 O.OC~b:*O

-2.&! LIE 02 0. C*~.I...*~~

-2. i.1~6~1: 0% 0.00480 -3.7710E 1'2 0.00403 -4.1410E 02 0.00550 -5.8-1; ~JE 0 " 0. 0~1EEO -6.3230E 02 0.03560 -8.472ClE 02 0.0ot.90 - S.OG60E 02' 0.00630 -1.1GS75 03 0.006r0 - I. 235SE 03 0.00680 -1.6-!lOE 03 0.0U60t -1.62226 03 ' 0.00730 -1.9726E 03 0.00740 -2.0663E 03 0.00780 -2.46tOE 03 0.00790 -2.5708E 03 0.00830 -3.0102E 03 0.00840 -3.1370E 03 0.00880 -3.6374E 03 O.Cl.lc30

-3.7694E 03 0.009:*0 -4.3361E 03 n.0~940 -4.4741E 03 0.00980 -5.0939E 03 0.00990 -5.2576E 03 0.01030 -5.0441E 03 0.01040 -6.1256E 03 0.01060 -6.0887E 03 0.0 IOIIO -7.0882E 03 0.01130 -7.S301E 03 0.01140 -8.1507E 03 0.01 1eO -9.0697E 03 0.01 I90 -9.5099E 03 0.01230 - l .~3tl6E 04 11.01240 -1.Cl967E 04 0.01280 -1.1649E 04 0.01290 -1.i932E 04 " 0.01330 -1.3107E 04 0.01340 -1i3414E 04 0.01360 -1.4G89E 04 0.013rr3 -1.6021E 04 0.01430 -1.6403E 04 0.01440 -1,S762E 04 0.01460 -8.t254E 04 0.01490 -1.0641E 04 0.01630 -2.02a1E 04 0.01540 -2.06G9E 04 0.01580 -2.2403E 01 0.OlbVO -2.2882E 04 0.01630 -2.47l5E 04 0.01640 -2.519EE 04 0.01680 -2.7197E 01 0.01690 -2.77145 04 0.01730 -2.EESSE 01 6.01740 -3.0408E 04 0.01560 -3.2685E 04 0.01790 -3.3285E 04 0.01630 -3.5718E 04 0.01840 -3.6345E 04 0.01880 &3.8930E O? 0.01890 -3.9594E 04 0.01930 -4.2327E 04 0.01940 -4.3029E 04 0.01980 -4.5694E 04 O.Ol990 -4.662BE 04 0.02030 -4.9G21E 04 0.02040 -6.0383E 04 0.02080 -6.3412E 04 0.02090 -5.4160E 04 0.02130 -S.7116E 04 0.02140 -5.7843E 04 . 0.02180 -5.8402E 04 0.02190 -6.9033E 04 0.02230 -6.156 3E 04 0.02240 -6.2166E 04 0.02280 -6.4480E 04 0,02290 -6.5033E 04 0.02330 -6.7139E 04 0.02340 -6.7G38E 04 0.02360 -6.9533E 04 0.02390 .-6.9B82E 04 0.02430 -7.16lOE 04 0,02440 -7.1001E 04 0.02480 -7.3331E 01 0.02490 -7.3628E 04 0.02800 -7.7563E 04 0.02900 -7.6552E 04 0.03300 -6.6711E 04 0.03400 -6.3025E 04 0.03800 -5.4363E 04 0.03900 -1.964SE 04 TABLE 6.A-ll (SHEET 6 OF 32)

TABLE 6.A-11 (SHEET 7 OF 32) TIME FUNCTION NWER m ( 4) o'\ P 1 I-' t-' 'dm C 0 I % TI t' t-' U3 a rP FUNSTION DESCRIPTION

= t FORCINO FUNCTION AT NODE 33 OO>O2I 0. 113623.8 ) NUMBER OF ABSCISSAE t 577) FUNCT I (>I+ SCALE FACTOR f 1 .000OE 00) HE VALUE 0. 0,00050 0.00100 0.00150 0.00200 0.00250 0.00300 0.00350 0.00400 0.00450 0.00500 0.00550 0.00600 0.00650 0.00700 0.00750 0.00800 0.00850 0.00900 0.00950 0.01000 0.01050 0.01 100 0.01 150. 0.01200 0. 0 1'250 0.01300 0.01350 0.01400 0.01450 0.01500 0.01550 0.01600 0.01650 0.01700 0.01730 , 0.01800 0.01850 0.01900 0.01950 0.02000 0.02050 0.02100 0.02150 0.02200 0.02250 0.02300 0.02350 0.02400 0.02450 0.02500 0.03000 0.03500 FUNCTION T 1 1 .3624E 04 1.3624E 04 1.3624E 01 1.3624E 04 1.3624E 04 1.3624E 04 1.36236 04 1.3621): 04 1.3617E 04 1 .3608E 04 1.3591E 04 1.35638 04 1.35lliE 04 1.3456E 04 1.3372E 04 1,3268E 04 1.3146E 04 1 .3008E 04 1.2860E 04 1,2705E 04 1.255OE 04 1 .2398E 04 t .2251E 04 1.2109E 04 1.1973E 04 1 . 1 838E 04 1.1694E 04 1 . l534E 04 1.1351E 04 1.l137E 04 1 .08E8E 04 l.0599E 04 1.0268E 04 9.8932E 03 9.4738E 03 9.0103E 03 e. 5038E 03 7.9559E 03 . 7.3688): 03 6.7369E 03 6.1140E 03 . 8.4550E 03 4.7629E 03 4.1014E 03 3.4155E 03 2.7630E 03 2.1B87E 03 1 ,62226 03 1.9875E 03 1.6852E 03 1 .4758E 03 3.69606 02 1.1462E 03 HE VALUE 0.00010 0.00060 0.001 10 0.00160 0.00210 0. 002GO 0.00310 0.00360 0.00410 0.004GO 0.00610 0.00560 0.00610 0.00660 0.00710 0.00760 0.008tO 0.00860 0.00910 0.00960 0.01010 0.01060 0.01110 0.01 160 0.01210 0.01260 0.01310 0.01 360 0.01410 0.01460 0.01510 0.01560 0.01610 0.01660 0.01710 0.01760 0.01610 0.01860 0.01910 0.01960 0.02010 0.02060 0.021 10 0.02160 0.02210 0.02260 0.02310 0.02360 0.02410 0.02460 0.02600 0.03100 0.03600 FUNCIION 1 1 .3624E 04 1.36245: 04 1.3424E 04 1 .3621E 04 1.3624E 0-4 1.3624E 04 1.3623E 04 1.3621E 04 1.3616E 04 1.36QBE 04 1.3587E 04 1.3555E 04 1.3608E 04 1.3441E 04 1.3353E 04 1.3245E 04 1.3119E 04 1,2979E 04 I. 2829E 04 1.2674E 04 1.2619E 04 1.2368E 04 1.2222E 04 . 1.2C181E 04 1.1946E 04 I. lOlOE 04 1 .1664E 04 I. 1 5OOE 04 1.l311E 04 1 .1060E 04 1.0833E 04 1.0536E 04 1.0197E 04 9.8186E 03 9.3844E 03 8.9122E 03 8.3975E 03 7.8416E 03 7.24CBE 03 6.629bE 03 5. Q836E 03 b.3216E 03 4.6471E 03 3.9645E 03 3.2815E 03 2.6382E 03 2.0457E 03 2.2991E .03 1.9l95E 03 1.6361E 03 1 .0969E 03 3.3840E 02 1.6426E 03 'IHE VALUE 0.00020 0.00070 0,00120 6.001 70 0.00220 0. ~10270 0. Up320 0.00370 0.00420 0. GO470 0.00520 0.00570 0.00C20 0.00670 0.00720 0.00770 0.00820 0.00870 0.00920 0.00970 0,010.?0 0.01070 0.01120 0.01170 0.01220 0.01270 0.01 320 0.01 370 0.01420 0.01470 0.01626 0.01570 0.01620 0.01670 0.01720 0.01770 0,01820 0.01870 0,01920 0.01970 0.02020 0.02070 0.02120 0.02170 0.02220 0.02270 0.02320 0.02370 0.02420 0.02470 0.02700 0.03200 0.03700 FUNCTION TIME VALUE 1 3CS4E 04 0 0003f1 1. SGIbE 04 0.00080 1.36i4E 04 0.00139 1 3694E 04 0.031CO 1.3C.24E 04 O.Od230 1.5' P?E 04 0. rju2ur; 1.3622E 04 0. 0035.J I . rtC20E 04 0.003f.O I. 3614E 04 0 00430 1.56G2E 04 0.OQ48b 1.3581E 04 0,00530 1.3547E 04 0.OG5EO 1.3496E 04 0.00630 1.34252: 04 0.00660 I. 33338 04 0.00750 1.3221E 04 0.00780 1.3082E 04 0.00830 1.2950E 04 0.00880 1.0798E 04 0.00930 1 ,264 38 04 0. 00960 1. L?e9E 04 0.01030 1 . 2c38E 04 0.01080 1.2193E 04 0.01130 1 . Yt134E 04 0,OtlbO 1 1919E 04 0.012?0 1.17e2~ 04 O.OIZ~C~ 1.1633E 04 0.01330 1.14645: 04 0.01310 1.1269E 04 0.01430 1 . l042E 04 0.01480 I. C777E 04 0.01530 1 .C472E 04 0.01580 1.C123E 04 0.01630 9.73G7E 03 0.016eO 9. k035E 03 0.01730 8.e126~ 03 0.017tO 8,:trC.E 03 0.01630 7.7256E 03 0.01880 7.1234E 03 0.01930 6.t016E 03 0.01960 5.8651E 03 0.02030 5.1873E 03 0.02083 4.t113E 03 0.02120 3.6271E 03 0.02160 3.1492E 03 0.02230 2.51BlE 03 0.02280 1. Q35OE 03 0,02330 2.2153E 03 0.02380 1 .L(553E 03 0.02430 I . tPO5E 03 0,02480 6.17705: 02 0.02800 3.tceo~ 02 0.03300 2 ZE95E 03 0.0380~ FUNL I I ON 1 .~cL"I~. n: 1 3h24E 04 1 , 36 24L Cl-? 1.3GZ4C 04 1.3-2 ' 03 1 . 31823~ 03 1.3CPZL 94 1.3618E 01 8 3EliV 04 b .35I*lul.

04 1 . 35(.rC 04 1 . 351'%.Em 04 I .3170E 04 1.3390C 0.4 1.329lE 01 1.3172E 04 1,3037tr 64 1.28991. 04 1.27365: 04 1.25PIC 04 t.2428k 04 1.22EOE 04 1.L137E 0.1 1.200nE 0.1 I.tt1C.tK 01 t. liZ.'?t 04 I . i6t;t.r 0.1 1.13P I 04 1.11 6SL 1.14 t.OC?.41E 01 1 .ow '*L 01 1.033,E 04 9. S'Ilbt. 03 9.561 IE 03 9.1OG4E 03 P.COOGE n3 E.Ot;BiE 03 7.489?E 03 .6.672YE 03 6.2440E 03 8. 58BOE 03 4.8181E 03 6.2364E 03 3.51514E 03 2.8900E 03 2.2742E 03 1 .723CC 03 2.OJ83E 03 1.7381E 03 . 1.5105E 03 4.62COE 02 7.7840E 02 3.79685: 03 TABLE 6.A-11 (SHEET 8 OF 32) I). 7OZIJE 03 1.4045E 04 I. ISGEE 04 2.2313E 0.1 2.42-' :f 0e 2. SSrl9E 01 2.70elE 13.) 3.0151E 0.: 3.lt'IlE 04 3.22'.1L 04 ?.2??.1E C14 5. 41: 'E U.$ 3. t l'* +.-*L cwl . tlUE 0.1 3.0S67E 04 . O.OP,7L'E 04 4.2183E 04 4.3b5OE 04 4.423GE 04 - 4.3ef 9E 04 4.5-133E 04 ' 4.5S93E 03 4.5328E 04 4.5604E 04 4.738St 04 4.OU23E 04 4.91.dZE 04 5.039*L' 04 S.tr7GCC 04 b.OS18E 5.09)6E 04 04 6. 1003E 04 6.215.1E 04 5.29;rlE Ibrl 5.3307E 061 5.3.irr)E 04 5.4165E (14 b.4139E 04 5. 5.':G@E 04 &.5.:97E 04 5 5587E 04 Z.5239E 03 5.6569E 04 5.6591E 04 5.61HGE 04 6.754GE 04 6.7724E 04 '5.7607E 04 5.766;JE 04 6.8072E 04 5.8959E 04 8.9781E 04 0.83DlE 04 b. 6590E 04 6.89tJE 04 b.6338E 04 5.6972E 04 6. Q302E 04 S.9570E 04 5.9508E 04 8.93ClE 04 5.B737E 04 TABLE 6.A-11 (SHEET 9 OF 32) TIME FUNCTlON NUneER 8 6 6) FUNCTION DESCRIPTION

= t FORCING FUNCTION AT NODE 34 OO>uTb 0. -10080.4 ) NlIMOER OF ABSCISSAE

= 4 077) FUNCTION SCALE FACTOR 8 4 1.0000E 00) IME VALUE FU1.ICT ION TI 0. - 1 .0080E 04 0.00050 -1.0351E 04 0.00100 -1.107SE 04 0.00150 -1.2218E 04 0.00200 -1.4004E 04 0.00250 -1.6585E 04 0.00300 -2.0338E 04 0.00350 -2.6259E 04 0.00400 -3.1261E 04 0.00450 -3.7923E 04 0.00500 -4.5597E 04 0.00550 -6.4284E 04 0.00600 -6.4023E 04 0.00650 -7.4895E 04 0.00700 -8.7506E 04 0.00750 -1.0177E 05 0.00800 -1.1765E 05 0.00850 -1.3b22E 05 0.00900 - 1. G486E 05 0.00950 - t.7647E 05 0.01000 -1.9990E 05 0.01030 -2.2602E 00 0.01100 -2.819SE 0s 0.01150 -2.7465E 05 0.01200 -3.0323E 05 0.01250 -3.3389E 05 C 01300 -3.6587E 05 0.01350 -3.9878E 05 0.01400 -4.3251E 05 0.01450 -4.6696E 05 0.01500 -6.0200E 05 0.01550 -5.3787E 05 0.01600 -6.7449E 05 0.01630 -6.1169E 05 0.01700 -6.4937E 05 0.01750 -6.8769E 05 0.01800 -$.2520~ 05 ' 0.01'850 -7.6136E 05 0.01900 -7.9606E 05 0.01950 -8.1789E 05 0.02000 -8.3330E 05 0.02050 -8.4614E 05 . 0.02 100 -8.5687E' 05 0.02150 -8.6483E 05 0.02200 -8.7365E 05 0.02250 -8.8099E 08 0.02300 -8.8765E 05 0.02350 -6.4374E 05 0.02400 -8.S302E 05 0.02450 -8.9764E 05 0.02500 -9.0174E 05 0.03000 -S.4609E 05 ' 0.03500 -0.7767E 05 ME VALUE FUNCT 1018 0.00010 -1.0099E 04 0.000GO -1.0461E 04 0.00t10 -1.1228E 04 0.00160 -1.2126E 04 0.00210 -1.44308 04 0.00260 -1.7241E 04 0.00310 -2.1232E 04 0.00360 -2.6379E 04 0.00410 -3.2506E 04 0.00460 -3.938aE 04 0.00510 -4.7252E 04 0.00560 -5.6148E 04 0.00610 -6.6OSSE 04 0.00660 -7.7281E 04 0.00710 -0.0229E 04 0.00760 -1.0481E OS 0.00810 -1.2098E 05 0.00860 -1.3698E 05 0.00910 -1.6903E 0s 0.00960 -1.8102E 05 0,01010 -2.0477E 05 0.01060 -2.3031E 06 0.01110 -2.6757E 05 0.01160 -2.8016E 05 0.01210 -3.0926E 05 0.01260 -3.4017E 05 0.01310 -3.7237E 05 0.01360 -4.0545E 03 0.01410 -4.39346 05 0.01460 -4.739tE 05 0.01610 -5.0911E 08 0.01560 -5.4513E 65 0.01610 -8.8187E 05 0.01660 -6.19196 05 0.01710 -6.669bE 05 0.01 760 -6.9527E 05 0.01810 -7.3256E 05 0.01860 -7.6843E 05 0.01910 -8.O28tE 08 0.01960 -8.21176 05 0.02010 -8.3607E 05 0.02060 -8.4841E 05 0.02110 -8.888lE 05 0.02160 -8.6744E 05 0.02210 -8.7515E 05 0.02260 -8.8239E 05 0.02310 -B.8892E 05 0.02360 -8.8893E 05 0.02410 -8.9399E 05 0.02460 -8.9830E 05 0.02600 -9.1046E 05 0.03100 -9.5461E 05 0.03600 -9.8OO5E 06 TlME VALUE FlC.rCT ION 0.00020 -t.O13dE 04 0.00070 -1.C6ETE 04 0.00120 -1.144ZE 04 0.p0170 -1.2827E 04 0.00220 -1.4901E 04 0.00270 -1.7Q4GE 04 0.90320 -2.217r~ 04 0.00970 -2.7beCE 04 0.00420 -3.3bi:lE 04 0.00470 -4. G874E 04 0.00520 -4.89bOE 04 0.00570 -8. 8O5bE 04 0.00G20 -6.61SDE 04 0.00670 -7.0335E 04 0.00720 -9. SOEOE 04 0.00770 -1 .O?YPE Ob 0.00820 -1,244lE 05 0.00878 -1.42C3E 08 0.00920 -1.632BE 05 0.00970 -1.8SCJE 05 0.01020 -2.OC72E 05 0.01070 -2.3563E 06 0.01120 -2.5633E 05 0.01170 -2.8572E 05 0.01220 -3.ltaJE 05 0.01270 -3.46tlE 05 0.01320 -3.78C~E 05 0.01370 -4.1218E 06 0.01420 -4.46t2E 05 0.01470 -4,8067E 05 0.01520 -1.1623E 05 0.01570 -5.5243E 05 0.01620 -5.8Y:SE 05 0.01670 -6.PG71E 05 0,01720 -6.643CE 05 0.01770 -7.0264 05 0.01820 -7.39L3C 05 0.01870 -7.7b4t~E 05 0.01920 -8.093bE 05 0.01970 -8.2476E 05 0.02020 -8.3biOE 05 0.02070 -b.bDbtE 05 0.02120 -8 C069E 05 0.02170 -8.685~E 05 0.02220 -6.7605E 08 0.02270 -8.6377E 05 0.02320 -8.90tci 05 0,02370 -6.899PE 05 0.02420 -8.S4i;E 05 0.02470 -8.99'aCE 05 0.02700 -9.1622E 05 0.03200 -B 6PtRC 05 0.03700 -9.81CJE 05 TlME VALUE F1J14CT I Oti TI 0.00030 -1 .Olli8E 04 0.00C150 - 1 .0?3 iE 04 0.00130 -1.1677E 04 O.OO1RO -1.3214E 04 0,00230 -1.5117E 04. 0. OOkF~O -1.86:18C 04 0.00930 -2.3156E 04 0.00380 -2.87GQE 04 0.00430 -3.6139E 04 0.00400 -4.2SOSE 04 0.00530 -5.0GH2E 04 0.00580 -6.0003E 04 0.00630 -7.0335E Q4 . 0.06660 -8.2256E 04 0.00730 -9. 5871E 04 ' 0.00780 -1.lllOE 05 0.008$0 - 1.27R2E 06 0.00880 -1.40Y5E 06 O.OC930 -1.67616 05 0.00980 -I.P032E 05 0.01030 -2.1475E 05 0.01080 -2,4102E 05 0.01130 -2.G373E 05 0.01180 -2.9135E 05 0.61230 -3.2147E 08 0.01260 -3.5293E OG 0.01330 -3.8551E 05 0.01380 -4.1892E 05 0.01430 -4.5310E 05 0.01460 -4.8789t 05 0.01530 -5.2342E 05 0.01580 -5.6975E 05 0.01630 -5.SC74E 05 0.01680 -6.342% 05 0.01730 -6.7233E 05 0.01780 -7.1036E 05 0.01830 -7.4707E 05 0.01&80 -7.8240E 05 0.01930 -8.1603E 05 0.01980 -8.2746E 05 0.02030 -8.4125E 05 0.02080 -8.S2OIE 05 0.02130 -8.6243E 05 0.02180 -6.706lE 05 0.02230 -8.3813E 03. 0.02280 -8.85lSE 05 0.02330 -8.9138E 05 0.02380 -6.S103E 05 , 0.02430 -8.9587E 05 0.02480 -9.0017E 05 0.02800 -9.2827E 05 0.03300 -9.68706 05 0.03800 -9.8078E 05 PiE VALUE FUNCTION 0.00040 -3.0262E 04 6. 0LJD:'O - 1 .(:387E 04 0.00140 -1.8936E 04 0.00180 -1.350!E 04 0. ('ll.?40 - 3 . 5979E 04 O.COZR0 -l.E-?PIE 04 0.00340 -2.4185E 04 0.00390 -3.0003E 04 0,00140 -).6517E 04 0.00.90 -4.38626 04 0.03540 -5.24E4E 04 0.00590 -6.1992E 04 0.00640 -7.2581E 04 0.00690 -8.404EE 04 0.00740 -9.9788E 04 - 0.00790 -1.1434E 05 o.oos.ao -1.3183~ 05 t*. OObOO -1 .&077E 0s O.OOS~3U -1.7201E 05 0.009Cl0 - 1.9507E 05 .0.01040 -2.19G1E 05 0.01090 -2.4617E 05 0.0b140 -2.6917E 06 0.01190 -2.9726E 05 0.C11240 -3.27GSE OS 0. tri?'.*O -3.5642E 05 0.01340 -3.921 tE 05 0.01390 -4.25G9E 05 0.01440 -4.6003E 05 0.01480 -4.9492E 08 0.01540 -f .3063E 05 0.01590 -6.6710E 05 0,01640 -6.0420E 05 0.01690 -6.4160E OS 0.01740 -6.8003E 05 0.01790 -7.1779E 05 0.01840 -7 J426E 05 0.01890 -7.8925E 05 0,.01040 -8.2464E 05 0.01890 -6 3043E 05 0.02040 -8.4373E 05 0.02090 18 54866 06 0.02140 -8.6416E 05 0.02190 -6.7209E 00 0.02240 -8.7968E 05 0.02260 -8.8637E 05 0.02340 -8.9257E .05 0.02300 -8.9204E 06 0.02440 -8.S676E 05 0.02490 -0.009JE 05 0.02900 -9.3723E 05 0.03400 -9.Y38W 05 0.03900 ?S39E 00 TABL 9 6.A-11 (SHEET 10 OF 32) cn Y t-' w TABLE 6.A-ll TIME FUNCTION NUFIBER - t 6) (SHEET 11 OF 32) FUNCTION OESCRIPTION 8 t FORCItiO FUNCTION AT NODE 35 00>02S 0. -2867.6 B NUNBEfi OF ABSCISSAE ( U77) ' FUNCTION SCALE FACTOR = ( 1.0000E 00) t-' U) 03 IP ME VALUE FUNCTION 0. -2.85768 03 0.00050 -3.1065E 03 0.00100 -3.7499E 03 0.00150 -4.6570E 03 0.00200 -5.8763E 03 0.00260 -7.4579E 03 0.00300 -9.6849E 03 0.00350 -1.2154E 04 0.00400 -1.6264E 04 0.00450 -1.90926 04 0.00500 -2.3921% 04 0.00550 -2.96436 04 0.00600 -3.62028 04 0.00650 -4.3603E 04 0.00700 -5.2313E 04 0.00750 -6.2179E 04 0.00800 -7.3060E 04 0.00850 -8.4948E 04 0.00900 -9.8178E 04 0,00550 -1.1263E 05 0.01000 -1,2817E 05 0.01050 -t.4475E 05 0.01100 -1.6250E 05 0.01150 -1.7924E 05 0.01200 -1.9841E Ob 0.01230 -2.1864E 05 0.01300 -2.3968E 06 0.01350 -2.6130E 05 0.01400 -2.8344E 05 0.01450 -3.0605E 05 0.01500 -3.2898E 05 0.01650 -3.6221E 05 0.01600 -3.7572E 05 0.01650 -3.9816E 05 0.01700 -4.2337E 05 0.01750 -4.4739E 03 0.'01800 -4.7104E 05 O.O\85O -4.9334E 05 0.01900 -5,1459E 05 0.01950 -6.322tE 05 0.02000 -6.3740E 05 . 0.02050 -5.4209E 06 0.02100 -5.4650E 05 0.02150 -6.5097c 05 0.02200 -6.6551E 05 0.02250 -5.6029E 05 0.02300 -5.6506E 08 0.02350 -8.6696E 05 0.02400 -3.7160E 05 0.02450 -6.7630E 05 0.02500 -5.@lOOE 05 0.03000 -6.2583E 05 0.03500 -6.40706 Ob TIME VALUE FUNCTION 0.00010 -2.8750E 03 0.00060 -3.2052E 07 0.00110 -3.8743E 03 0.00160 -4.8814E 03 0,00210 -6,. 1439E 03 0.00260 -7.8428E 03 0.00310 -1.0068E 04 0.00360 -1.2713E 04 0.00410 -1.b940E 04 0.00460 -1,9983E 04 0.00510 -2.4996E 04 0.00560 -3.0089E 04 0.00610 -3.7870E 04 0.00G60 -4.6246E 04 0.00710 -5.4199E 04 0.00760 -6.4278E 04 0.00810 -7.5317E 04 0.00860 -8.7491E 04 0.00910 -1.0097E 05 0.00960 -1.ib66E 05 0,01010 -1.3139E 05 0.01060 -1.4824C 06 0.01110 -1.6617E 05 0.01160 -1.8298E OS 0.01210 -2.0239E 05 0.01260 -2,2278E 05 0.01310 -2.439SE 05 0.01360 -2.65676 05 0.01410 -P.B792E 06 0.01460 -3.1062E 05 0.01510 -3.336tE 05 0.01560 -3.5689E 05 0.01610 -3.8045E 05 0.016G0 -4.04236 05 0.01710 -4.2016E 05 0.01760 -4.1236E 05 0.01810 -4.7561E 05 0.01860 -4.9768E 05 0.01910 -5.1871E 05 0.01960 -5.3331E 05 0.02010 -5,38376 06 0.02060 -6.4298E 05 0.02110 -6.4747E 05 0.02160 -6.b185E 05 0.02210 -5.16476 05 0.02260 -6.6123E 05 0.02310 -6,6331E 05 0.02360 -8.6787E 05 0.024t0 -5.7255E 05 0.02460 -6.7724E 05 0.02600 -6.9053E 05 0.03100 -6.2930E 05 0.03600 -6,4204E 05 1 IHE VALUE FUNC 6 ION T InC VALUE FUNCT l ON 1 0.00020 -2.907BE OJ 0.00030 - 2. 9677L 03 0.00070 -3.5208E OS 0.0008r) -3.411L45E 03 0.00120 -4.0521E 03 0.00.130 -4.2421E 03 0.00170 -6.1152E 03 0.00180 -6.36976 03 0.00220 -6.43G4E 03 0.05230 -6.7533E 03 0.00270 -8.2502E 03 0.00280 -8.6759E 03 0.OP320 -1.0567E 04 0.00330 -1.1OBlE 04 0.00370 -1.3304E 04 0.00380 -1.39cOE 04 0.00420 -1.6666E 04 0.00430 -1,7437E 04 0.00470 -2.O913E 04 0,00480 -2.1880E 04 0,00520 -2.6107E 04 0.00530 -2.7247E 04 0.00570 -3.2171E 04 0.00680 -3.3403E 04 0,00620 -3.8996E 04 0.00630 -4.0498E 04 0.00670 -4.ES38E 04 0.00689 -4.8600E 04 0.00720 -6.6132E 04 0.00730 -5.CIlO8E 04 0.00770 -6.6416E 04 0.00780 -6.6530k 04 0.00620 -7.7639F 04 0.00630 -8.0010~ 01 0.00870 - 9.0087E 04 0.00680 -0.27243E 6.4 0.00920 -1.0382E 05 0.00930 -1.0671E 05 0.00970 -1.1872E 05 0.00960 -1.2163E 0'5 0.01020 -1.3466C 05 0.01030 -1.S7U7E 0:b 0,01070 -1.5174E 05 0.01080 - 1.5520E 0:i 0.01120 -136823~ 05 0.01130 -t.718GE 05 0.01170 -1.86746 05 0.01180

-1.905BE 06 0.01220 -2.0641E 05 0.01230 -2.t046E 05 0.01270 -2. P696E 05 0.01280 -2.3110E 08 0.01320 -2.4827E 05 0.08330 -2.5259E Ob 0.01370 -2.7010E 05 0.01380 -2.7461E 05 0.01420 -2.9244E 05 0.01430 -2.96f15E 05 0.01470 -3.1617E 05 0.01480 -3.1977E 05 0.01520 -3.3824E 05 0.01630 -3.4289E 05 0.01670 -3.6lb9E 65 0.01580 -3.G62SE 05 0.01620 -3.e51BE 03 0,01630 -3.8994E 05 0.01670 -4.G801E 05 0.01680 -4.1379E 05 0.01720 -4.3306E 05 0.01730 -4.3784E 05 0.01770 -4.S710E 05 0.01780 -4.6181E 05 0.01820 -4.COOSE 05 0.01830 -4.6156E 05 0.01870 -6.OlD8E vS 0.01880 -5.n624E 06 0.01920 -5.2255E 09 0.01930 -8.2GPOE 0% 0.01970 -6.3437E 05 0.01980 -5.3540E 05 0.02020 -5.3S31E 05 0.02030 -5.4024E 05 0.02070 -5.4389E G5 0.02060 -5.4400E 05 0.02120 -5.4835E 05 0.02130 -6.oR21E 05 0.02170 -6.5272E 05 0.02160

-5.b3GlE 05 0.02220 -6.8741E 06 0.02230 -6.5636E 05 0.02270 -8.6219E 05 0.02280 -5.6318E 05 0.02360 -5.6421E US 0.02330 -6.6512E 05 0.02370 -6.6680E 05 0.02380 -b.6974E 05 0.02420 -8.7349E 05 0.02430 -5.7443E 05 0.02470 -6.78lSt 03 0.02480 -5.7913E 05 0.02700 -6.99P6E OY 0.02800 -6.0919E 05 0.03200 -6.337tE 6s 0.03300 -6.3662E 05 0.03700 -6.4298E ed 0.03800 -6.4359E 05 TABLE 6. (SHEET 12 0.04200 -G.4i..*C 05 0.04700 -6.3562E 05 0.05200 -6.25tiE 05 0.05700 -6.12 7E 06 0.06200 -6, C2 1 $16 05 0.06700 -6.98SSE 08 0.07200 -6.S85bE 05 0.07700 -6.9OtdE 05 0.06200 -6.96sSE 05 0.OE700 -5.9013E 05 0.09200 -6.8076E 06 0.09700 -5.74CSE 05 0.10200 -5.69ECE 05 0.10700 -5.G407E 05 0.11200 -6.6620E 06 G.11100 -6.4714E 05 0,12200 -5.4113E 05 0,12700 -5.3870E 05 0,13200 -6.33C9E 06 0.13700 -6.PfB3E 00 0.14200 -.5.3626E 05 0.14700 -5.48bGE 05 0.15200 -5.40EBE 05 0.15700 -5.4Ct7E 06 0.16200 -6.4012E 05 0.16700 -6.4564' 05 0,17200 -6.3742E 05 0.17700 -6,3073E 05 0.18200 -5.P701E 05 0.18700 -5.2719E 05 0.19200 -5.2672E OS 0.19700 -5.POBlE 05 0.20400 -5.1712E 05 0.2t400 -5.103iE 05 0.22400 -5.ll79C 05 0.23400 -6.1236E 05 0.24400 -5.0963E 05 0.25400 -5.0109E 05 0.26400 -4.9664E 05 0.27400 -5.OObL-E 05 0.26400 -4.9951E 05 0,29400 -4.9523E 05 0.30400 -4.6507E 05 0.31400 -4.9578C 06 0.32400 -4,91406 05 0.33400 -4. e6t4E 05 0.34400 -4.3703E 05 0.35400 -4.e909~ 05 0.36400 -4. PO886 05 0.37400 -4.85LSE 05 0.36400 -4.8005E 05 0.39400 -4.82:,1E 05 0.40400 -4.85G4E 05 0.41400 -4,8300E 06 0,42400 -4.8112f 05 0.43400 -4.818UE 05 0.44400 -4.83CbE 05 0.45400 -4.7971C 05 0.46400 -4.7870E 05 0.47400 -4.7926E 05 0.48400 -4.7860E 05 0.49400 -4.776bE 06 t-' (n 0 V1 i C . r $ Y TABLE 6.A-ll TIME FUNCTION NUMBER d 7) (SHEET 13 OF 32) ? I-' I-' I-' Ui a3 tP FUNCTION DESCRIPTION ( FORCING FUNCTION AT NODE 36 00>02% 0. -515.9 9 NUMBER OF ABSCISSAE . 877) FUNCTION SCALE FACTOR w 6 1.OOOOE 00) TlflE VALUE FUNCTION 0. -6.1590E 02 0.00050 -5.1580E 02 0.00100 -5.1730E 02 0.00150 -5.2100E 02 0.00200 -5.3500E 02 0.00250 -5.815OE 02 0.00300 -6.89806 02 0.00350 -0.1140E 02 0.00400 -t.2956E 03 0.00450 -1.8798E 03 0.00500 -2.6718E 03 0.00350 -3.6549E 03 0.00600 -4,7944E 03 0.00650 -6.0537E 03 0.00700 -7.3987E 03 0.00750 -8.8089E 03 0.00800 -1.0280E 04 0.00850 -1.1821E 04 0.00900 -1.3460E 04 0.00930 - 1 .6243E 04 0.01000 -1.7226E 04 0.01060 -1.9472E 04 0.01100 -2.2050E 04 0.01 150 -2.4982E 04 0.01200 -2.8284E 04 0.01260 -3.1989E 04 0.01300 -3.6t41E 04 0.01350 -4.0773E 04 0.01400 -4.5BSGE 04 0.01450 -5.1523E 04, 0.01500 -5.7641E 04 0.01530 -6.4210E 04 0.01600 -7,120lE 04 0.01650 -7.8600E 04 0.01700 -8.63866 04 0.01750 -'#.4563E 04 0.01800 -1.0311E 08 o.dteao -1.1202E 05 0.01900 -1.2l26E 05 0.01950 -1.3066E 05 0.02000 -1.4037E 05 0.02050 -1.5006E 05 0.02100 -1.5880E 06 0.02150 -1.6886E 05 0.02200 -1.60d6~ 05 0.02250 -I. 8985E 05 0.02300 -1.9888E 05 0.02350 -1.9821E 01 0.02400 -2.0478E 05 0.02450 -2.IOSPE 05 0.02500 -2.1571E 05 0.03000 -2.4627E 05 0.03500 -2.5018E 05 TlflEVALUE FUNCTION 0.00010 -5.1690E 02 0.00060 -5.1590E 02 0.00110 -5.17306 02 0.00160 -5.2250E 02 0.00210 -8.40hOE 02 0.00260 -5.9600E 02 0.00310 -7.2460E 02 0.00360 -9.7360E 02 0.00410 -1.3969E 03 0.00460 -2.0218E 03 0.005to -2.e53e~ 03 0.00560 -3.8709E 03 0.00610 -5.0373E 03 0.00660 -6.3169E 03 0.00710 -7.6755E 03 0.00760 -9.1005E 03 0.00810 -~,O581E 04 0.00860 -1.2139E 04 0.00910 -1.3803E 04 0.00960 -1.5622E 04 0.01010 -1.7653E 04 0.01060 -1.9961E 04 0.01110 -2.2612E 04 0.01160 -2.5612E 04 0,01210 -2.8991E 04 0.01260 -3.2783E 04 0.01310 -3.7029E 04 0.01360 -4.1758E 04 0.01410 -4.6982E 04 0,01460 -6.2708E 04 0.01510 -5.892OE 04 0.01560 -6.5575E 04 0.01610 -7,2649E 04 0.01660 -8.0127E 04 0.01710 -8,7990E 04 0.01760 -9.6242E 04 0.01810 -1.0486E 05 0.01860 -1.1364E 05 0.01910 -1.2315E 05 0.01960 -1.3260E 05 0.02010 -1.4232E 05 0.02060 -1.5201E 05 0.02110 -1.618QE 05 0.02160 -1.7187E 00 0.02210 -1.82OBE 05 0.02260 -1.9170E 05 0.02310 -1.9236E 05 0.02360 -1.9858E 05 0.02410 -2.0601E 05 0.02460 -2.1167E 05 0.02600 -2.P412E 05 0.03100 -9.4708E 05 0.03600 -2.5026L 05 T l ME VALUE FUtICT I ON 0.00020 -5.lS3CE 02 0.00070 -5.1735E 02 0.00120 -5.17SOE 02 0.00170 -6.24LOE 02 0.00220 -6.4blDE 02 0.00270 -6. I LavOE 02 0.00320 -7.62tOE 02 0.00?70 -1.0436E 03 0.00420 -l.5045E 03 0.00470 -2.1758E 03 0.00620 -3,0442E 03 0.00570 -4.O93fE 03 0.00620 -6.264dE 03 0.00670 -6.S819E 03 0.00720 -7.9566E 03 0.00770 -9.3917E 03 0.00820 -1.OledE 04 0.00870 -1.P4E2E 04 0.00920 -1.41S2E 04 0.00970 -1.EOOOE 04 0.01020 -1.COEOE 04 0.01070 -2.04CIE 04 0.01120 -2.3176E 04 0.01170 -2.62F7E 04 0.01220 -2.9714E 04 0.01270 -3.3364E 04 0.01320 -3.7936E 04 0.01370 -4.P764E 04 0.01420 -4.eCCBE 04 0.01470 -5.3611E 04 0.01520 -6.0216E 04 0.08570 -6.tbE7E 01 0.01620 -7.4113E 04 0.01670 -8. lbrSE 04 0.01720 -8.E614E 04 0.01770 -0.7938E 04 0.01820 -1.GG63E 05 0.01870 -1.1807E 05 0.01920 -1.25i4E 05 0.01870 -1.3405E 05 0.02020 -1.49ZSE 05 0.02070 -1.83P8E 05 0.02120 -1.6388E 05 0.02170 -1.7388E 06 0.02220 -1.64OGE 05 0,02270 -1.9354E 05 0.02320 -1.9387E 05 0.02370 -2.0003E 05 0.02420 -2.0720E 05 0.02470 -2.1272E 05 0.02700 -2.3698E 05 0.03200 -2.4672E 05 0.03700 -2tbtPOE 05 7 1 HE VALUE FUIIGT 1014 0.00030 -5.16POE 02 0.00080 -6.1730E 02 0.09130 -5. I6COE 02 0,00180 -5.26"nE 02 0.00230 -6.66C.OC b2 0.00280 -6.35506 02 0.00330 -6,nGYOE 02 0.00380 -1.1205E 03 0.00430 -1.622PE 03 0.00480 -2.3313E 03 0.00530 -3.240GE 03 0,00563 -4.321 7E 03 0.00630 -5.6376E 03 0.00680 -6.8524E 03 0.00730 -E.23U6E 03 0.00780 -S.E947C 03 0.0003*3 -1.llt25E 04 0.006; .0 - 1 .27I;91 04 0.00830 -1.4bo9E 01 0.00E80 -1.6401E 01 0.01030 -I.BJS~E 04 0.01080 -2.097SE 0-1 0.01130 -2.37G3E 04 0.01180 -2.6918E 04 0.01230 -3.0455E 04 0.01280 -3.4425E 04 0.01330 -3.8861E 04 0,01380 -4.3708E 04 0,01430 -4.9213E 04 0.01480 -6.5136E 04 0.01530 -6.1531E 04 0.01580 -6.835GE 04 0.01630 -7.5G03E 04 0.01680 -8.3226; 04 0.01730 -9.1248E 04 0.017CO -9.9648E 04 0.01830 -).0842E 05 0.01800 -1.17bZE 05 0.01930 -1.2b96E 05 0.01680 -1.361RE 0: 0.02030 -1.4t17E 05 0.020~0 -1.6~05~ ns 0.02130 -1.6587E 05 0 02183 -1.7592E 05 0.02230 -1.8602E 05 0.02260 -1,95426 05 0.02330 -1.9536E 05 0,02380 -2.0224E 05 0.02430 -2.0836E 05 0.02480 -2.1374E 05 0.02800 -2.3692E 05 0.03300 ,-2.4944E 05 0.03800 -2.5001E 05 'ul P !a TABLE 6.A-11 (SHEET 14 OF 32) r t-'

TABLE 6.A-ll (SHEET 15 OF 32) T 1 WE. FUNCT I Ott HU13CER = t 0) FUNCTION OESCR I PTI ON 8 ( EORCING FUNCTIOll AT WDE 37 00~0i.3 0. -1.2 ) NWElEFi 01: ABSCISSAE 8 ( 677) FUNCT l6lJ SCALE FACTOR 8 4 1 .00OC*E 00) 'IME VALUE FUNCTION 11 0. -1.2000E 00 0.00050 -l,POOOE 00 0.00100 -1.2000E 00 0.00150 -1.2000E 00 0.00200 -1.2000E 00 0.00250 -2. BOOOE 00 0.00300 -6.1000E 00 0.00350 -l.OlOOE 01 0.00400 -5.0100E 01 0.0C1450 -I. lG'.inE 02 0.00500 -2.4420E 02 O.OOb50 -4.F520E 02 O.OOGO0 -8.1020E 02 0.00650 -1.3412E 03 0.00700 - 2.0722E 03 0.00750 -3.0418E 03 0.00000 -4.2693E 03 0.00050 -5.7674E 03 0.00900 -7.6308E 03 0.00950 -9. t497E 03 0.01000 -1.1801E 04 0.01030 -1.426IE 04. 0.01100 -1.69126 04 0.01150 -1.973GE 01 0.01200 -2.2729E 04 0.01250 -2.6390E 04 0.01300 -2.B224E 04 0.01350 -3.2747E 04 0.01400 -3.6478E 04 0.01450 -4.0447E 04 0.01500 -4.4GC7E 04 0. OiSPO -4.9240E 04 0.01600 -5.4142E 04 . 0.01650 -5.US2QE 04 0.01700 -6.5130E 04 O.Ol750 -7.1274E 04 0.01800 -7.7878E 04 0.01830 -8.4955E 04 0.01900 -0.2F13E 04 0.01050 -1.0056E 05 0.02000 -1.OSt3E 05 0.02050 -1.1816E 03 0.02100 -1.274,4*

Ob 0.02150 -1.3694E 05 0.02200 -1.4G67E Ob 0.02250 -1.865RE 05 0.02300 -1.66596 05 0.02350 -1.7628E 06 0.02400 -1.86136 05 0.02450 -1.95876 05 0.02500 -2.0543E 03 0.03000 -2.6797E 05 0.03500 -3.0947E 05 HE VALUE FUNCTION TI 0.00010 -1.20C~0E 00 0.000G0 -1.2000E 00 0.001l0 -1.2000E 00 0.001G0 -I .20C,OE 00 0.00210 -1.2000E 00 0.00260 - 2.80ChOE 00 0.00310 -9.3000E 00 0.00:GO -2.4000E 01 0.00410 -6.1100E 01 0.00460 - 1.3740E 02 0.00810 -2.8130E 02 0.00560 -6.2630E 02 0.00610 -9.07GOE 02 O.OOEG0 -1.4691E 03 0.00710 -2.P474E 03 0.00760 -3.2661E 03 0.00t 10 -4.5481E 03 0.00660 -6.OQ97E 03 0.00910 -7.Cl49E 03 0.00900 -9.9816E 03 0.01010 -1.2276E 04 0.01060 -1.4776E 01 0.01110 -1.7462E 04 0.01160 -2.0321E 04 0.01210 -2.3348E 04 0.01250 -2.604tE 04 0.01310 -2.SSt3E 04 0.013GO -3.347GE 04 0.01410 -3.7ZBZE 04 0.01460 -4.1270E 04 0.01610 -4.b~70E 01 0.018G0 -6.Ol9tE 01 0.01510 -6.5167E 01 0.01663 -6.053bE 04 0.01710 -6.6323E 04 0,01760 -7.2558E 04 0.01810 -7.9256E 04 0.01800 -3.6428E 04 0.01910 -9.4086E 04 0.01960 -1.0223E 05 0.02010 -1.1091~ 05 0.02060 -1.2000E 05 0.02110 -1.2932E 05 0.02160 -1.38876 05 0.02210 -1.486SE 05 0.02260 - 1.565QE 05 0,02310 -1.6859E 05 0.02360 -$.782GE 05 0.02410 -t.BeoeE 05 0.02460 -1.078QE 05 0.02600 -2.2391E 05 0.03100 -0.76GlE 08 0.03600 -3.1424E 06 InE VALUE FUNCTION 0.00020 - 1 ,0000E 00 0.06070 -).ZOG>E 00 0.00120 -1.200ctE 00 0.00170 -1,POOOE 00 0.00220 - 1 ~OIJ~E 00 0.002: 9 -4.4OcOE 00 o.oa320 -1.10uOE 01 0.00370 -2.8aLWE 01 0.0042C1 -7.2FUOE 01 0.00470 -I.EOiJE 02 0.00520 -3. PC* I CsF 02 .O.OLS~O -t.esto~ 02 0.00620 - 1 . OOOJE 03 0.0067u -1.C0~3E 03 0.00720 -2.4298E 03 0.00770 -3.6002E 03 0.00030 -4. &.*LGC 03 0.00870 -6.4411E 03 0.00~20 -8.3nw~ 03 0.00970 -I.O?i2E 01 0.01020 -1.t7beE 01 0.01070 -1.62YOE 04 0.01120 -1.8['21E 04 0.01170 -2.Ou13C 04 O.bI22P -2.3971E 04 0.01270 -2.72b:iE 04 0.01310 -3.061 ?E 04 0.013?0 -3.421JE 04 0.01420 -3. ?03.iE 04 0.01470 -4 21C.E 04 f~.OIt126

-4.C;.?IOE 04 0.01570 -5.1 lt3E 04 0.01520 -6.62OCE 04 0.01670 -6.1657E 04 0.01720 -6.76JZE 04 0.01770 -7.3tiJE 04 0.01020 -6.t6CIE 04 0 01670 -8.79taE 04 0.01Q20 -9.Ef77i 04 0.01970 -1.03S.5 03 0.02OZU -1. 127tE OS 0.02070 -1.2165f 05 0.02120 -1.31ilf Ob 0.02170 -1.4GduE 05 0.02220 -1.bOcLE 05 0.02270 -1.60CSE 05 0.02320 -1.703wE 05 0.02370 -1 .8OZJE 06 0.02420 - 1 .900!X 05 0.02470 -1.09726 05 0.02700 -2.4097E 05 0.03200 -2.8YVuE 05 0.03700 -3.17*YE 05 . T IHE VALUE FUlJCf told TI 0.00030 - 1 . 20r*0E 00 0.0~360 - 1 .2L .iOE 00 . 0.0Cf130 -1.2OllOE 00 O.Oc)tCO -1.2Oc1OE 00 . 0.00230 - 1 .2000E 00 0.0020G -d04F90L PO 0.00330 - I .2G(BOE 0 1 0.00300 -3.64UOE 01 0, 00$? *3 - (? . GF;l*t>F 01 0.06-1t'J - 1 .CIS.* IL 02 0.0i853t4

-3. (iY'fOt 02 O.OO?~CO -tr.nn;':nI-l or 0.0065 *1 - 1 , I 6*'53E 07 0.0(16(:6

.. 1 .7! 9 .$E I'?: o.Ols'/r.>

-2. c;n;.zr. + t.3 0. 00700 -3. 7;!l ;L.::- i17 0.f83Yb0 -G. Ir\I:1E k:j 0. OO@!lrl -6 7?;.?CE 1):' O.OC*>:O -0.7i;'KE U'I 0.6S'JfrO - 1 . OU;";L 04 0.01030 -1.3262; 04 0.01880 -1.8830E 04 0,01130 -1.8SOGE 04 O.Oll@O -2.1til3E 04 0.0!230 -2.4605E 04 0.01913 -2.7869E 34 0.01330 -0.1314E 04 0.013f.0 -3.49t13E 0.1 0.014?0 -%.C027E U3 0.014I~O -4.3953

04 0.016;10 -4.79LSI.

04 0.01500 -'..2IDGE 04 O.ZIIC:<O

-6.721.75 04 0.016PO -6. i!'.'W,E 04 0.01730 -6.971.tE U4 0. 01700 -7.ci:~~t:

nl 0.01c3li -fi.Z..l!':

E Od 0.018t:O -(Z. f',' ICY Oz? ' 0.0165iO -P.723.?E 04 0.01 9T.9 - I . Ot.G.?E OIj 0,02030 - 1.84GIE 0:. ' 0.02080. - 1 .23'10E 05 0.02130 -1.3312E CfiS 0.02160 -1.4295E 05 . 0.02230 -1.6261E 05 0.02269 -1,625SE 05 0.02330 -1.72376 05 ' 0.02360 -1.6220E 06 0.02430 -1.0199E 05 0.02480 -2.OIC3E 05 0.02tiOO -2.41676 05 0.03300 -2.9511E 05. 0.03600 -3.1924E 05 TABLE 6.A-11 (SHEET 16 OF 32) -3. I39-at: 05 0.0.1~~ -3. I n::~ CC. -2.6012T 05 0. W~CI~J -2. ;.;. ..rc CI;> -2.673.ei 05 O.65Soi) -p. 5-51 ?.1E 1.15 -2.296tii 05 0. ObEW -L. S'ijOtE it3 -2.101 :IE 05 O.Ot:*r10

-2. 1 I IGC' ..I5 -2.Olt;E 0'3 0.06~iOO -;c.vii,IL uxi - I . StSfrC 05 0.07:tOQ - 1 . g17'4E 05 -1.834of 05 0.07600 -1.C03PE 09 -1,66316 08 0.08300 -1.62218 09 - 1 ,47906 03 0, 0.q~t.1')

-) ..-,4itoe ti;j -1.8064E 08 0.061~00 - Il: ZhUJE US -I. 1041E 05 , 0.09000 -1 .Ot'/GE 05 -9.54tSE 04 0. IO~OCI -P.~.c~E i14 -0.C41~C 04 0. 10001l 71 ;"3:: 04 -8.634SE 04 0.11300

-0.4 JAOL i4 -8.274GE 04 ' 0,11600 -8.LiOt;E la*$ -8.1334E 04 0. 12301:1 -8. ltk?3rr" .il -0.3997E 04 O.~YBCI~I -P..?.l/t.'E 111 -9.169M 04 0.1;53rw -~.lt;c.t~:

-0.9729~ 04 O. t~$trt*o -l.r.l.:nE OP -0.7001*f:

04 0, 143~10 -u.r.#.I'i::C

.. t . -0.3031E 04 Q, ftj,.)<11'1

-~I.:;..'::+.I:

tt.:' -0.I13SE 04 0. te31.1n.1

-+ae :,:q* ,:.;: 41 ' -0.13btiC 04 0.16000 -l~.l'litriE u! -0. I93RE 0.9 0, 163C)(_ -~.cflt."'r:

ft? -3.443rE 04 0. 160[.*> - 0. L:.. 19 v'i -8.8186C 04 0.17350 -9.OICi'E 31 -9.3004E 04 0. 17600 -Q.E!',i7E 111 -t?.I657E 04 0. l839Ci -i'. [.R4Z O? -6.0443E 04 0, I6000 -7,8822E 0.1 -7.4?24E 04 0.19300 -7.IG4IE 04 -7.00366 04 0.10800 -6. G003E 01 -6.4073E 04 0.20600 -6.243fiE 04 -7.01879 04 0.21riOO -C.G'7CgC 04 -6.51648 04 O.226W -6.6317E M -6.245FE 04 0.23600 -6.19?6E I*? -6.0613E 01 0. ?1C*or) -6.OZ-4E 0d -6.9724E 04 0.2*:bbO -7. IUil;; 04 -7.023:!E 04 0.26iCI.0 -C.C'l; 1E 01 -6.0234C 0.1 0.27600 -6.lIB. : I L 8.1.4 -6.C3dlL 04 0.29COO -6.1237C 114 -6.90t4C 04 0.291:-~0

-~.III* 'I:? (I t -7.006iE 04 0. 30CGt10 -6. Gl t:I:E 0-1 -6.7233E 04 0. 3 1600 -L . .?';;.Gi.

U:! -5,rigt 1f 04 0,32r,i10

-6.Q. "'?? 11: -6.406zr 04 0. 3:;6Cl0 -G. c ;'.,:LL 111 -6.49'1 ti 64 0.34f30 -6.C..:l<;i ibi -8.8341E 04 0. 3fiC.03 -t.S?a:ZE (1.1 -4.672oE 04 0.3Gt*20 -1.SlGlE 15-1 -5.llbit 04 0.376110 -6.3'171E U.1 -6.4502E 04 0,39600 -G.6469l2 04 -6.294tE 04 0.39EO0 -6.Ot:IDE . -6.blCVE 04 0.40606 -6.@X<lE 04 . -6.1645E 04 0.41600 -6.11Y7E 04 -8.4381E 04 0.42600 -G.45106 W -6.132r)E 04 0.43600 -5.7f409E 04 -6. llblE 04 0.4460 -6.2057E 04 -8.6497E 04 0.46600 -6.7942E 04 -6.9217E 04 0.46600 -6. BCt.3E 04 -6.69::lE:

04 0.47610 -8.6382E 04 -8.681'rlC:

04 0.48600 -f. 68*?4E U1 -8.6897E 04 0.49GDO -8.R025E 04 3. nf.?@O -11. o!,;ti: 'T: 05 tl. {I.' I.*U -k. >/L~;li OL> ti.8.1: -2.5kl31:

05 (I. Cl!.t.k~ll . 2.2 1 1 3E 05 ,-I. .'. ' 1 . ::* , pyi v.ut *I:~S -2.u11it~

111) Q.9itI.tu . #.91t.:.6 01, (I. <I;'. .uU - I . i;'ll...C 06 0. cZC-~.t+t'& - 1 . BihkbE (I. 0:. 'I* 'I . 1 . ?%I! tI,C 05 0.09.?L*O

-1 .ZlnlF. Ol.5 a. (I$*< 1.~0 - 8 . l'v:a;,l:

OP . t*, l t*..!C*(! - s . l ;'(lUT 04 (1, 1 * :t'lF) C, f $:\I: 1.14 0. 1 IC.l"O -3, ril.i2rlii 04 e1.1)300 -C).i:lO~;!;

04 ta. 1 ::. ?it0 -0 *' 1 .sUGI: 04 11, I.'tI"I -.C. .",?:I&..

It1 IS, 1 L...:~IC~ -a. .t A-~C ~i: I 1.3 11. \:~~n'I<* "I .b;tl<,~~I:

llli 0, i'!.:i : .*.. 4:t; 1:: WI f-t.1. 1.0 -~../811~~~4 . ,.. i; !+*'t ..$,. $i:rr,T frt . (I. 1 I>:'.*.II*I -C.. i"':: C 614 n. is', . :, , .',.I:;. 04 kt. 1 <. -4L.O , t:. #..'.a! !I: 0.q 0. tI:!l'.u -ZI. 114:':. 04 0. I i -*1#*3 -@. 4**l?L a1 0. 3 :-:re -0. prtl ,:P o.+ Q.~C.~.I.II>

o.! f1.306 -?.r~lt.~~

-7. tfi191 04 (~4 0 . & $:!~:l>l~f

-6 . Ol.<Y&L 04 0.2td.30 '-6.44 161: M F.kl.S*~*c

-7,C,6*.?*il:

04 0. ;iiC:.&>O -G. bWi...L 04 ' n. 2;:000 -ci. c" c [. a~ 04 (1. ::.:<.131*:*

-6. 2.136..!:

0.4 0.2b8W -7. l lah2E 04 !.. . :>* .r1-11-: - o.. . 6 ,;$ .:c 0.4 61.2. :I:I(I -G.,l'a?..i Il! b). Ci;:tjll - 6. :-I :. ..tE it.? , f#.c:..l.,.l

-*/. ;;*t. .!. I>.+ 0. 3.1C.01' -ti. $l'ZCIli 01 (1.31:1...

-6.4;fjS'i 04 I,, '.....' 0 . :;. - .:: .I: 81' k}. ti.. .;:1.t+: -1.. ;:>:?I: 4.1. f1.::,:..0:1 -G.: .IG.iC 04 (> . i ; ..,.>Ce;*

' - 5. I;? 1 04 .O.:+.').l'&;

-1.4.1C7Z 04 u.bidOO +l..i/4HE-04, 0. :':3800 -6 ; 421.C E: 04 ~.;:.:<.iti, -b.591 1E 04 0. .i0006 -S., Mi!YE 04 ' '0.4tb0O -6.4174E 04 0.42800 .-6. a2368 04 0.13600 -6.1141E 04 0.45600 .-PL. 3Q7SC. 04. 0.45800 -8.C-iti9E 04 0.4G80G b.OlOBEM ' 0.47000 -6.66252c 04 ... , 0.40800 -6.769BE.

04 P.43600 -0.9023fjO4.

' . . . . .. .

TABLE 6.A-ll (SHEET 17 OF 32) FUNCTION DEbCfiIPTION

= ( FbRCINQ FUNC'I'IBW AT NUDE 38 00,02S 0, -0.2 b HI.:. li.ER OF APSC l SS4E c t (i77) TUtISi l liN SCALE FAClOCg I .OOOOE 00) T l HE VALUE FUtJCT l ON 0. -2.OOOOE-03 0.00050 -2.0060E-01 O.OOIOO -2.000n~-ot 0.00160 -0,r'OCl I*-01 0.00200 -2.C.00GE-01 0.00250 -6.0600E-01 0.00300 -1.1000E 00 0.00360 -3.6000L 00 0. 00400 -9.200c~2 00 0.00450 -2.l4OClE 01 0.00500 -4,490OE 01 0.00600 -8.5400E 01 0.00600 - 1 ,604CIE 02 0.00650 -2.4630C 02 0.00700 -3.OCGOE 02 0.00750 -5,566OE 02 0.00800 -7.8410E 02 0.00650 -1.0692E 03 0.00900 -1.3830E 03 0.00950 -1.753PE 03 0.0t000 -2,167QE 03 0.01050 -2.GlSIE 03 0.01100 -3.lObRE 03 0.01150 -3.6245E 03 0.01200 -4.1741E 03 0.01250 -4,7647E 03 0.01300 -5.3670E 03 0.01350 -6.013CE 03 0.01400 -6.G9BlE 03 0.01450 -7.4ZeOE 03 0.01600 -8.20G7E 03 0.01550 -9.0419E 03 0.01600 -9.6431E 03 0.01650 -1.0014E 01 0.01700 -1.1961E 04 0 :0 1 750 - 1 .3086E 04 0.01800 -1.4302E 04 0.01850 -1.b602E 04 0.01900 -1,6990E 04 0.01960 -1.6468E 04 0.02000 -2.0042E 04 0.02050 -2,170OE 04 0.02100 -2.3404E 04 0.02150 -2.6149E 04 0.02200 -2.6936E 04 0.02260 -2.8750E 04 0.05300 -3,0584E 04 0.02350 -3.2374E 04 0.02400 -3.4163E 04 0.02430 -3.5971E 04 0.02500 -3.7728E 04 0.03000 -4.9213E 04 0.03500 -5.6638E 04 TIWE VALUE FUNCTtblJ 0.00020 -P.COGOE-01 0.00070 -2.0000E-01 0.00120 -2.OUOitE-01 0.00170 -2.oi~OilE-01 o.ooaen -?.oaoo~-i11 0.00270 -6.0000E-01

0. do320 -2. OOOtlE 00 0.00370 -6.30006:

00 0.00420 -1.930OL 01 0.00470 -2.94GOE 01 0.00520 -b.Bt\DDE 01 0.00570 - 1.0020E 02 0.00520 - \ BCCCkE 02 0.00570 -2. Pt~006 02 0.00720 -4.4620E 02 0.00770 -6.12ROE 02 0.00820 -0.6BPOE 02 0.00870 -1.162BE 03 0.00920 -1.8261E 03 0.00970 -1.b140E 03 0.0102n -e.a.m2~ 03 0,01070 -2.aOOGE 03 0.01120 -3.30f16E 03 0.01170 -3.8406E 03 0.01226 -4. 40Z8E 03 0.01270 -4.? 367E 03 0.01320 -5.GZ13E 03 0.01370 -fi.P&31E 03 0.01430 -6. B0.18E 03 0.01470 -7.7C2CE 03 . 0.01520 -8.5341E 03 0.01570 -U.3982E 03 0.01620 -I CISZPE 04 0.0\670 -1 t383L 01 0.01720 -t.k402E 04 0.01770 -1.355aE b4 0.01620 -1.4EllE 04 0.01870 -1.CI46E 04 0.01B20 -1.7670E 04 0.01970 -1.BOOSE 04 0.02Q20 -2.OC.SBE 04 0.02'570 -2.Y377E 04 0.02120 -2.4097E 04 0.02170 -2.Eb59E 04 0.02220 -2.7662E 04 0.02270 -2.34a3E 04 0.02320 -3.1292E 04 0.02370 -3.3090E 04 0.02420 -3.4602E 04 0.02470 -3.6678E 04 0.02700 -4.4254E 04 0.03200 -5.2RBGE 04 0.03700 -6.8306E 04 TIRE VALUE FlWtCT 1 ht? 0. GO.,bU - 2. 0i;!*CtS .U I O.OOOBO - e . .:S'II'IOE - o I 0.06130 -2.n*e~lDE-Vl 0.0018II -Y.c*1t*~-0t 0.0il200 -2. ,r#iaQ!. . 0i 0.00260 -G.ui***'.i-08 0.0~380 -2.31.~ <IE 0. I 0.00380 -6. 501 gt+E C*LI 0.09130 -1. Si.: 16 $61 0.00.180 -3.41(1fblr:

(11 0. @*ti30 -G. C'. *I*JE 111 0. fi.)SOO -8 . 2 l;:ir~:.ct:r 0.ib363*.1

-2,(-:'!<*1!

(12 0. (10: :'3 -3. ',;"II~IE

$\I 0.00740 -rt.&l(:DE O? 0. 607L.D -6. ;";lic;C ic;. 0.WGSO -B.49zoE 02 0.06E60 -1.2.?90E 03 0.00930 - t .CO<nE 03 O.O(rDciO - 1. CUb7E OS 0.01030 -2.483110 03 0.0106d -2.PI172E 63 0.01130 -3.4132E 03 0.01100 -3.@508E 63 0.012361 -4.5107E 03 0.01260 -6,110OE 03 0.01330 -5.7511RE 63 0.013CO -C..419!lE 0:- 0.0i430 -7.l3llrjir 03 0.01460 -: . C 1;;;2E ti3 0.01630 -i.. 7i.t 2E 03 O.Ot5t;O -9.ti73lE 03 O.(tlG30 -t.(iF172 134 O.li\FBO -l.l*-.::;E 01 0.01 Xi3 - t ,262CE 0.4 0.017hO -1 .3.:iilY:-

e? 0.01630 -I.SClilE U.4 0.01350 -1.6$22E 04 0.0192.0 -1.78ti6E 04 0.01900 -\.@4C*lE 01 ' 0.01030 -a. 103lE 04 0.02040 -2.271BE 04 0.02130 -2.4147E 04 0.02100 -2.6216E 04 0.01230 -2.6027E 04 0.02280 -2. LR6OE 04 0.02330 -3.I056E 04 O.Ct2380 -3.3461E 04 0.02430 -3.62tSE 04 0.02480 -3.70~62 04 0.02C00 -4.5301E M 0.03300 -5.4197E 04 0.03600 -6.OCOCE 04 - B. 0Cll?i,*:e

-2.oOourr-01 o, (K*143 -2.ita'ttX-01 o,i*irt~n -O.OWJ.~C-~)~

41 &I.'*. 'ri;:* -6, cl;dji.C-Ol

0. i~meo -I .t wmt on It. ttc)3.? .I -2. BOc.W.lE 00 0 rc(v83o -7. /tt(ttli~

no n, rho.? lo - t . c ?oo~ 01 t@,(VrltG

-8.52(WE 01 0. tl(*f*30 -7:. ir(UJE Ul 0. (li~'i.':*O - I . Wiic*bE I12 I>.#.-I~.

Q -2.k4;',\K 0;: 1) , I'll 1l:'LIO - 3, 5lebDt 0% 0. tlCl'.*rlO

-6. 1310E 02 U.Co*,Elir

-7.&bo.*i.

$12 $:I. t~1.1 i(,r* - 1 .it* WE 0s 0.00..00 -1 .E14tE 03 0.OtlC.lL1

-1 . W60E 03 4). c*oL<.il -2. 68t i E 03 0.01MC1 -2.PI8.3E 03 n.0tc.e~ OWE 03 0.01110 -3.b183t.

03 0.@1190 -4.0018E 03 1,. 01 9..0 -4. 6:tr.nE 03 0.01 290 -6.2421.1E 03 W:OlslO' -6. ' .?tlit 03 0,1*13':b3

-6.6b89E 03 0.011.10 .-7.2783E 03 '0.I114tirJ

-8.0 tGGC 03 O. ~~IWO --8. tt707~ rt3 LD. r*lfi9Q '-9.7e76F:

03 0.61610 -1.071Ci 04 0.01C.LQ -1.1745L:

04 ib.01540 -1.2057C M C~.LI\?LJO

-1 ,405:C 04 0.91640 -1.639PL 04 9.01E;BD -1.67UUE 04 O.U!940 "1,8166E 04 0.01S90 -t.SflBE 04 0.02040 -0.8364E.M O.02ORO -2.3060E 04 0.02140 -.2,4797E 04 '0.02100 -P.b578E 04 0.02240 -2.8392E 04 0.02290 -3.0227E 04 0.02340 -3.2017E 04 0.02300 -0.3025E 04 0.02440 -3.6616E 04 0.02490 -3.2379E 04 0. 02BOO -3,7409E 04 0,03100 -6.6666E 04 0.03909 -@.659SE~04 . . .. . .

TABLE 6.A-11 (SHEET 18 OF 32) 0. fbl4ClO -6. C170E 04 0. <l.l900 -6.0W31E 04 0. C1.4C10 -4.6 'ME 04 0.1 *tl:'C*CJ

-4.01199E 04 0.05400 -3.77blE 04 0.062100 -3.6776E 04 0.0:'?C*9

-3.5lE2tC 03 0. USP:.*.3 -3. 2I2qE 04 $3. (44219 ..2.8070E 04 0.0:1900 -2.5017E 04 ti. n94*:*n -2.2370E 04 (1.1 :I~C*J - 1. 8U7bk Orf 0. 1 :'ul@c) -1.6B?rtt 04 0, 109*:10 - I .Ckt3E 04 '0. 1 1400 - 1 .5b2:iE 04 r a. I 1 JCID . .I . t.c*;.~ w U. tl'cdflo -1.4032E 04 0. Ilk*..) -1.61b8E fW 0. t *4**b0 - 1 .71?r? ;E 04 fa. !- C.10 -1.0.:ibE awl . O.I?lOO-1.740?L04 . 0. 1 .I! -fItl - 1 .702!tit 64 0. 1Blfi:b - 1.61:30E 04 ' (8. IS+:I'I -1.C'J)YF 04 0. 1li !I. I -1 .t718E 04 0.1 >@00 -1.6WlE 04 CI, I 741i -1 .6;sr~ 04 0.179C10 -1.72WE 04 0.1W00 -1.61ME 04 0. 188cl3 -1.3037E 04 0.i04QO -1.3153E 04 0.13900 -1.OTiOE 04 0. ;?iWl*O - 1 .1 WOE 04 O.trC.00 -1.28E3E 04 O.ZBL*CIO - 1 .2057E 04 0.2.; *10 -1. 1.IElE 03 0.3.1L'1:10 - 1 .14S7E 04 0. 2 S(.Cl:, -'I .3 l SfiE 04 0.236il'l

-1.P2114 04 0. L7bt1O . f .0782L. 04 0. 2 171.1C. - 1 . 1 723E (14 0 . 28:?1:** I - 1 .3D3 7E C44 n. :xi.. 1.16 - t . i W~FF 04 11. C 1 Xlv -8. B74 It. O9 0.22:..00 . I .0931E CM 0. i3L.iI?l - 1 . e477li 011 G. :?390 -1 . t.28Z.f 0-? 0. :.,bE.I)O

-21.7373E U3 0:sa..:06 -B. iKi16E 03 0.9f6u0 -t'.O[iOBE U1) 0.311:410 - 1.1 b02C 04 0.30d00 -1,OQME 04 0.4tIC00 .L1.017SE 04 0.41COC -l.,178GE 04 0;428CtO -1.1617E 04. 0.43C00 -1.1229E 04 0.44500 -9.7472E 03 0.45003 -1.0020E 04 0.45C00 7l.OG71E 04 0.47640 -1.omnE 04 0.48606 - 1. OfigGE 04 0. *DO06 - 1 ..0839E 04 . .

TABLE 6.A-11 (SHEET 19 OF 32) TIME FUNCTION NWER r ( 10) ' FUNCTION OESCHIPTtON r 4 FORCINQ FUNCT 1Okt C.T NUtll4ZR OF AUSCI SSAE 8 6 177) FUNCTlON SCALE FACTOR

• I 1.0000E 00) flME VALUE FUNCTION 0. 3.0000E-01 0.00050 3.0000E-01 0.0C1100 3.6CM3OE-01 0.00150 9.0000E-01 0.00200 3.0000E-01 0.00250 3.00~~0~-01 0.00300 3.OGO(lE-01 0.00350 3.0000E-01 0.00400 3.0000E-01 0.00480 -5.0000E-01 0.00500 -1.3000E 00 0.00550 -2.8000E 00 0.00600 -6.OOOOE 00 0.00650 -1.3100E 01 0.00700 -2.4300E 01 O.C@075O -4,3600E 01 0.00800 -7.6100E 01 0.00150 -1.2130E 02 0.00S00 -1.0900E 02 0.00060 -2.8440E 02 0.01000 -4.1270E 02 0.01060 -6.82ZOE 02 0.01100 -7,9930E 02 0.01150 -I.O717E 03 0.01200 -1.405SE 03 0.01250 -1.80G7E 03 0.01300 -2.2794E 03 0.01350 -2.8272E 03 0.01400 -3.4540E 03 0.01450 -4.1601E 03 0.01600 -4.9461E 03 0.01550 -5.PIOBE 03 0.01600 -6.7522E 03 0.01660 -7.7677E 03 0.01700 -8.854BE 03 0.01750 -1.0010E 04 0.01800 -1.122QE 04 0.0'18L3 -1,PtiIOE 04 0.0l900 -1.3646E 04 0.01950 -1.6234E 04 0,02000 -1.6674E 04 0.02050 -1.8162E 04 0.02100 -1.9692E 04 0.02160 -2.12666 04 0.02200 -2. ze75~ 04 0.02250 -2.4519E 04 0.02300 -2.6185E 04 0.02350 -2.7698E 04 0.02400 -2 Q620E 04 0.02450 -3.1361E 04 0.02300 -3.3116E 04 0.03000 -8.1022E 04 0.03500 -6.2475E 04 TIHE VALUE FUNCTION 0.00010 3.OQOOE-01 0.00000 3.0000E - 0 1 0.00110 3.0000E-01 0.00160 3.0000E-01 0.00280 3.0(,00E-01 0.00260 3.0000E-01

@.00310 3.FOODE-01 0.00360 3.0000E -0 I 0.00110 3.0000E-01 0.00460 -8.0000E-01 0.00510 -1.3000E 00 0.00560 -3.6000E 00 0.06610 -7.400OE 00 0.006G0 -1.4GOOE 01 0.00710 -2.iDOOE 01 0.00760 -4.8000E 01 0.0081b -6.2800E 01 0.00060 -1.3310E 02 0.00910 -2.O610E 02 0.00960 -3.0680E 02 0.01010 -4.4310E 02 0.01050 -6,2170E 02 0.01110 -8.5040E 02 0.01160 -1.1340E 03 0.01210 -1.4006E 03 0.01260 -1.8046E 03 0.01310 -2,332GE 03 0.01360 -2,9461E 03 0.01410 -3.5680E 03 0.01460 -4.3114E 03 0.01810 -8.l132E 03 0.01860 -5.9024E 03 0.01610 -6.9494E 03 0.01600 -7.9GbOE 03 0.017lO -9.0797E 03 '0.01760 -1.0249E 04 0.01810 -1.1481E 04 0.01860 -1.27i2E 04 0.01910 -1.4120E 04 O.Ol9GO -1.8tlSE 04 0.02010 -1.6Zl69E 04 0.020G0 -1.8461E 04 0.02110 -2.OOOflE 04 0.02160 -2.168SE 04 0.0221U -2.3201E 04 0.02260 -2.4852E 04 0.02310 -2.6534E 04 0.02360 -2,624tE 04 0.02410 -2.B96EE 04 0.02460 -3.l7llE 04 0.02600 -3.6634E 04 0.03100 -1.4003E 04 0.03600 -6.3943E 04 llNE VALUE CUN3T1513 0.00020 3.0000E-01 0.0007~1 ~.O~OOE-OI 0.00120 3.OCOOE-01 0,a00170 3.0000E-01 0.00220 3.0u0C1E-0 I 0.0027(* 3.001:1f1E-(rt O.Oil3ZCl 3.000vE -0 1 0.OW70 3.600QF-01 0.00420 -0.0006E;OI 0.00470 -0. onow-01 0.OOSPu -2.0oOOL 03 0.00370 -3.60UOE 00 0.00620 -8.1000E 00 0.00870 -1.G600E 01 0.00720 -3.10UOE 01 0.00770 -5.4300E 01 0.00820 -9.lBOOE 01 0.00870 - 1 . J5E.OE 03 0.00920 -2.2360E 02 0.00970 -3.3160E 02 0.01020 -4.iS>08E 02 0,01070 -8.G310E 02 0.01 120 '-9.0230E 02 0,01170 -l,lF78E 03 0,01220 -1.bB72E 03 0.0127U '-1. bP63E 02 0.01320 -2.4E52E 03 0.OtfiSO -3.Ob8,SC 03 0.01420 -:;.7263E 03 0.01470 -4.4615E 03 0.01620 -6.2824E 03 0.01670 -6.176C.Lr:

03 . 0.01620 -7.1491E 03 0.01670 -8.lC4GE 03 0.01720 -9.3069E 03 0.01770 -1.0441E 0.1 0.01820 -1.1754E 04 0.01870 -1.3057E 04 0.01920 -1.43LSE 04 0.01870 -1.5804E 04 0.02020 - 1 .7264E 04 0.02070 -1.87CCE 04 0:02120 -2.0318E 04 0.02170 -2.19045 04 0.02220 -2.3S29E 04 0.02270 -O.bI&GE 04 0.02320 -P.6875E 01 0.02370 -4.IIS8CE 04 0.024110 -3.0314E 04 0.02470 -3.PO62E 04 0.02700 -4.OOBtE 04 0.03200 -6.6h02E 04 0.03700 -6.b177E 04 T 1 NE \'Z.LUE Ftll'lC~f'l

• I! 0. OO0f 0 0.6*0L*C*E

-01 O.CICV:*; (I ?.'.I%I'~~-.II~

0. ;-XI 3. i*.ba MIE-kt I O.CI?I~(;O R.I~.~II.~:~-I:~I O.Oia::;':c> .i.iN*.a'tE-U)

0. @f,g[:*:~

3. #:ete.#r q-:-1.11 0. oU'r;:<~)

3. C* ~i*.;E-i#l O.CKt:itO

,Mi>tbE-i~l O.UW

.:O -ti.+*:lLIOE-Ol O,OrJ.:';G

-!;. jqacj!:-&:$I 0.0~6: I) -2. i1e.11.~~

8 Ct . OC"X3 -4 . &0?li*Z 00 O.ObC.:40 -O. i@>@C 0. OmM - I . *i.nOE 8: 0,00730 -3.-?:. .OE 01 0.007(K' -G.lbt*OL 01 0.096Slj - 1 , 00i OE 02 0. @:*"0G - I . f.::iiDE 02 0.009C9 -2.43dOE 02 0.00080 -3. WPOE 02 0. 0 1030 - ti, crc.(ic*i:

CIi? 0.OIOCIO -7.Uf~"*OE 02 0.011S.U -S.*'iUcr?:

02 0.011&0 -I .LD?i\E 03 0.01 230 - 1 .6:**jL'E 03 0. O! PC12 -2, 1:~f.Il~t::

6):: 0.01353 -i:.6W;&t:

08 O.OtaP0 -3. IFIDE r):* 0.0i.f: IJ -::.~:i:;i';

.: *a 0.01400 -4.6?1.?E 04 0. 01 S~-.I? -[i. 41.p.2': 63 0. Ol titi0 -6.3' /L*L 03 6.01630 -7,3+>EBE OR 0.01600 -(;.l?llGE 0.01730 -9.b'J'at.E 09 0.01780 -1.07::SE 64 0.01C30 -t.l3D1E 04 O.OlC*@O -i ,J3(@5E 04 0.01930 -1.4673E 04 0.01900 -1.6093E 04 0.02030 -1.7661E 04 0.020120 -1.9078E 04 0.02130 -2.0631E 04 O.OZlh0 -2.2227E 04 0.02230 -2.58b7E 0? 0.02200 -2.5820E 04 0.02330 -2.7214E 04 0.023CO -2.6026E 04 0.02430 -3.0663E 04 0.024CO -3.2413E 04 0.02800 -4.37ti5E 04 0.0~36*0 -6. b&'Z 1 E 04 0.03(:00 -6.616GE 04 TABLE 6.A-11 (SHEET 20 OF 32) t-, I-' 0.04300 - t .5(;9?E 0.1 0.0rG0a -r..b130L 01 0.OE31.13

-3. tsS4'tbE 01 O.,Ct'X~GO

-2.3b'lF-6 04 0. OG~~D -9.1 =txY 0 i o . or..nr)..~

.. 8 . P I ! cz lr .. 0.07400 - I . -19L7i: 6.1 0.07EW -I.DIPtE 04 0.08LStt0 -G,U,;IBE 01 0, P>CI.I - I . s?T,:,~; 0% 0.03;:lt.b C.: 3ti.L: ul: 0.0~Cf.fl 1.,4.!:.?F rrs O,$O:~l'!.

>l.t\(t:.?!:

a)::: 0. 1 <1t:ct*-c 1 . ll?'.i:; 0.3 6 . 1 1 31.1.1 ; . 4 0.1.' :a;: I 0. 1 1 C;(n:# O. uto XI: 1.1.. 0. 12&IK1 G.,lt*-61E 115 0.1PCl.~-*

P.,'Cl:'?.:

I)? 0. 1?;(.0 -1 :..c .,>L ..If 0, 16:;I.il -G.L.I61E it3 0.142.CiO

-1 .OI'/YL 04 0. lhG:.b..~ - 1 .4:'<1?~5 0.1 0.15311.1

-1.t.: t5E *'bl 0. ltiO(W -2. f'(.*t'C k3.1 0.16If10 -2.4?<-67 0.1 0. 16fil.14l -e.41:,"..E 04 0.1780~1 -:'!.2?:0E M 0. 178CI0 -Z?.kli!OE 04 0. 163(.i* -l.Pltli!?

0.1 0. lBbl*U ..l.(ri*btE 04 0. I SrjOO - l .2;'9'clE 04 0.19Ec@* -S.Cl~..iZ 03 0. :!'.l;clo

-7. 2t:s/g 0:.i 0.2 I $.fin -r.. r.n:.Pr I>*< 0.22C.ilil

-.Z...t<lE 0.1 0. : ::601, -2. ns. .;4E 0.1 0. 2.fC.0l.l

-" F!tcl::: n4 ' 0. lf*C4li.* - 1 . .:(ri V2 04 0. 2kj6r*l.l

-7. C.$* tiiE 4.::) 0. Z./i*tiu -9. iO;+i &I: 0.20G00 - 1 .7203E 04 ' 0.23600 -l.C'!&?Z 04 0,80(.0~ - 1 .I B;i&C U4 0.31f.00 -to. S&?OE ('3 0.32'.'.00

-9.0926E 03 0.33t@0 -1 . I Lt.GL 04 O.34C00 -). 40.?6E Or: ' 0.3FG00 -1.43t6E 03 . 0.3FCOO -).E:l42E

04. 0.376LlO -1.20f43E 01 0.38G00 -1.1843E 04 ' O.3Uf.00 . 1 .33RGE 04 0.406C10 - 1.30b7E 04 0.41600 -1.1690E.

04 ' 0.42600 -1.1007E 04 0.43600 -:.~B~z!x 04 0.44600 -1.63SOE 04 0.45600 -1.22COE M 0.46.C.W -I. lBlOE 021 0.47F00 -1.E4CQE 04 0,48600 - 1.23t 4E 04 0.49600 -1.2173E LM 0. F~.IL~ -6. 4509r. w (1. ~'.?f*:~:*

-6. 21 16t 04 is. ~f..tCiG -3. 21-1:'E 04 0. C*.i;MlO -2.21193E 04 0.9;iPO -2.1201C 04 ct k-..: I ail - 1 , CFbGb &i <I.+ il.c~'~ilG

-1 .43tlE 04 (I. ~l?ll(@G -U. YB i3C 03 0.0>,103 -5.2~ilOC 03 t ,.$ -s~.-er~ - n . 1 1 GQ~ CQ CI. <*s.jil-, 3. 41:<'.,.,E (8s n . .*-I..I~~

6 . t: :17.,: . CI:.~ 11. u.11. 0. n;'t,C :: I .; : it. ? t.lL:k@.@*.b I . 1 15 I E ti4 11 r I p!l.~l I . gr :9:ctr S,L~ It. 1 l<~lx' 7,4~< $1: 03 I$. I L'RibCJ 4. G:>i:PE <13 I,. v:5c@w'b 1. 79iR1E 03 0. I ..' <I - 1 . 21.:LIW iv t1,t3..*Y@ -U.4tiiwlE 0s - tr. te.!nf* -8 .09cis1: 05 0. 14. C*tl - 1 . rlDt'!.iE 04 ' (#. \ !.i,?t:10 - 1 . (l!.i;.'E 0.3 0. 1,j:~iO -2. I44W D4 0.1C.100 -2.3166E 04 0. tCs::.If~, -2.45291:

04 0.1 71GCl -2.2.124E 04 t' Cn . 0.1 Vt100 - 1.94448 04 C] .O. LC. :0.3 -l . ti772E 0.1 Cn (I. t &tH)J - 1 .6;*13E 04 1 (1. 'i WOO - 1 .1872E 04 C 6, I 9*1Clil - 0. 2: W!E 03 "3' 0,2O:'O?)

r6.6?4OE 03 fa. a I ;.;.no - I . osuo!: 04 0, L.:l;O(i - 2.ti057E 04 !a 6.2:';;1CtO - 2.6 jEOE 04 0. ::. \0ti -2.3f12GE 04 (t.yioiw~

-1.2160~ 04 0. %i*(iUU -7.62ME 03 (8. i('.~U1:3

-1 .2Z61E 09 (1.23300 -1.7n61E 04 n.a:l?:oG

-1.6t~se n4 0.80@00 -1.0009E 04 0.3ib00 -7.694GE 03 (I. 32800 - 1 .0036E 04 0.38Y00 - 1 .97706 64 0.34800 ' - 8.4843E 04 ,P. 3b890. - 1 .3669E 04 0.36800 -6.3416E 04 0.37EOO -).1293E M P.38000' -1.2963E 04 0.88300 -1.3730E 04 U. 40600 . -,l . kS99E 04 " 0.41000 -1;O(MGE04 0.42Bi10 -1 .P647E 04 0;4086@ -1.3107E 04 0.44000 -1.4crt7E 04 0.4r.900 -I . 2OWE 04 0,46300 - 1 .2067E 04 0147000 .-I. 2G28E 04 -.0.481)0~.-1.010ZE 04 . O.49Bc.U)

-1.. 90778 04 . . , . .,. . , . . .

TABLE 6.A-11 (SHEET 21 OF 32) FUNCTION DESCRIPTION

• ( FORCINO FUI~CTIOtI AT WOE 40 O0,OZt 0. 0.4 8 NW4DEH OF AbSCISSAE

• ( 877) FUIICT 1 Otl SCALE FACTOR 8 ( 1.000OE 00) 1-3 iz t' M rn b 1 F I--' G 0 I % T: t' I-' ID 03 A TIME VALUE FUNCTION 0. 4.0000E-01 0.00050 4.0000E-01 0.00100 4.0000E-01 0.00150 4.0000E-01 0.00200 4.0000E-01 0.00250 4. C*000E-01 0.00300 4.0030E-01 0.00350 4.0000E-01 0.00400 4.0000E-01 0.00450 -7. (IClOOE-01 0.00500 -1,8000E 00 0.00550 -4.0000E 00 0.00000 -8.3000E 00 0.00660 -1.D5OOE 01 0.00700 -3.6200E 01 0.00760 -6.lG00E 01 0.00800 -1.0620E 02 0.00050 -1.7140E 02 0.00000 -2.6710E 02 0.00950 -4.0190E 02 0.01000 -b.U830E 02 0.0t050 -8.228OE 02 0.01100 -1.lZQGE 03 0.0i150 -1.V146E 03 0,01200 -1.OB63E 03 0.012SO -2.55326 03 0.01360 -3.2312E 03 0.01360 -3.9954E 03 0.01400 -4.8812E 03 0.01450 -6.0791E 03 0.0l500 -6. (1890E 03 0.01850 -8.2119E 03 0.01600 -9.5422E 03 0.016b0 -1.0977E 04 0.01700 -1.2514E 04 0.01760 -1.4146E 04 0 .'01800 - I . JESSE 04 0.01650 -1.7679E 04 0.01900 -1.95G7E 04 0.01050 -2.1528E 04 0.02000 -2.3564E 04 0.02050 -2.SG66E 04 0.02100 -2.7829E 04 0.02150 -3,0053E 04 0.02200 -3.2326E a1 0.02250 -3.465OE 04 0.02300 -3.7019E 04 0.02350 -3.e425E 04 0.02400 -4.l658E 04 0.02450 -4.4320E 04 0.02600 -4.660OE 04 0.03000 -7.2104E 04 0.03500 -0.62SOE 04 TlHE VALUE FUNCT: ON 0.00020 4.000OE-01 0.00070 4.0060E 0 1 0.00120 4.0000E-01 0,00170 4.0000E 01 0,00220 4.0000E-01 0.00270 4.0000E -0 1 0.00320 4.0000E-01

0. ob370 4. OOO~E-01 0,00420 -Y.OOOOE-01 0,00470 -7.00GOE-01 0.00620 -P.@OOOE 00 0.00570 -6.0006E 00 0.00620 -1.1BOCE 01 0.00670 -2.3000E 0) 0.00720 -4.3600E 01 0.00770 -7.6700E 01 0.00820 -1.2940E 02 0.00970 -2.05bOt 02 0.00320 -3.1590E 02 0.001170 -4. G340E 0.2 0.01020 -6.7l30E 02 0.01070 -P,b?QOE 02 0.01120 -1.E75lE '3 0.01 170 -t .6RZBF 'b3 0.01220 -2.2006E 33 0.01270 -2.OL:tt 03 0.01330 -3.6177E 03 0,01370 -4,3250E 03 0. o I 420 -5. er.BaE 03 0.01470 -6.3093E 03 O.OI6iiO -7.465OE 03 0.01b70 -8,7318E 03 0.01620 -1.0103C 04 0.01670 -1.l581E 04 0.01720 -1.3165E 04 0.01770 -1.4826E 04 0.01820 -1.65ESC 04 0.01870 -1.8424t 04 0.0 1C120 -2.0343E 04 '0.01970 -2,2335E 04 0.02020 -2.4387E 04 0.02070 -2.6b26E 04 0.02120 -2.87I3E 04 0.01170 -3.OSlJE 04 0,02220 -3.3251E 04 0.02270 -3.5502E 04 0.02320 -3.7977E 04 0.02370 -4.0395E 04 0.02420 -4.2839E 04 0.02470 -4.0310E 04 0,02700 -6.6662E 04 0,03200 -7.9961E 04 0.03700 -@.2106E 04 TI I.tE VALUE FUt 15.1 I olJ 0.0r3030 4.0O'lOE-01 O.FCtOc6 s.t.*iltl .l-01 0.00130 4.0?11:1uE-01 0 . OCD I OCI 4 . 1% II:I~IE - I I i 0. OOi'SO 6. @~.~.~liC-Ol

0. CI~I&&~ e. e ....w-I~ er: -1:1 0.00330 4 . r.iK.OK-*.i 1 0.00380 4.00~OC-9) 0.00#!36 -7. Ot~~iOE-Cll 0.06160 -7.1?*:1ilE-01 0.00530 -2, S WOE 00 0.00600 -8. 1 3l.lOE. 00 0.00630 -~.Z~CIOE.,OI 0 ; O0tOO - P .7til.l*ZE 0 1 0.00730 -4. $20-*I Cll 0.00700 -6.GaO00 01 0.03630 - 1 .42:tCtE 02 0.0068iI -2.2470F 612 0.C13U30 -3.40CWL:

02 0. OCIPGO -t:. <l?l-?OE 112 0.01030 -i.Z(ICW%

0:: 0,01080 -E,GB.OE 02 0.01130 -1.351uE 113 0.01100 -1,737kIE 03 0.01230 -2. 3Ir.2E 03 0.082iXS -2.5 '!*..& &>'> 0.0133*:1

-3.r-71 )E 113 0.013Ff* -4.f.lr!ilC Oi 0.0~43~ -5.4cr.'sE CK$ 0.01480 -6.C3t06 03 O.C@1(13(1

-7.7b9.2E

@3 O.Olt6ic -O.S27GE.

03 0.01630 -i.i*w~r 01 0.01 GOO - 1 . 18C;:'E 0.1 0.01730 -1.34PlE 0.1 0.017EO -1.617OE 04 0.61830 - 1 .GQ.?t,E 01 O.OlEG0 -1.L.8OPE 04 0.01930 -2.07::6E 04 O.OlB0O -2.2743L U4 0.02030 -2.4817E:

04 0. C12000 -2.69:iGE 04 0.02130 -*?.Ql:i6E 04 0.021(;0 -3.141 1E 04 0.02230 -3.3714E 01 0.02200 -3. c na;e.r ui 0.02330 -3.C4t;PE 04 0.02380 -4.OC61E 04 0.02430 -4.3'3::ZE 04 0.02480 -4.6t.Oi;E OQ 0.02000 -C. 1 P8.;ilE 04 0.03300 -8.3125E 04 0.03000 -9.

&5(16E 04 TABLE 6.A-11 (SHEET 22 OF 32) t3 t' M .m , 'P I-' I-' Ei C 0 I % Ttt t? I-' w m tP 0. QWO@ -9.1163f 04 0. 1 .!Z;*>D -7. W'J~K 04 0.Q :I10 -4.iF.l'sr*E O? CI.*~:,:'QIJ

-3.L:.W4E 0.4 1.1 . .,. .. r*, - 2. :, 8 1: re il. 1 .... . . :. ,,;..*:i.

(,.,: o.**,.::.1 -g.lF' %:L 134 t.>.~rii11~.1

-1 .51(:E il . * .t'.r!l'qj - 7. & (16: 11.1. .,.I1 -I.. .ti, t1. 11. r*$J!"':\.U

3. ti. * ,%I. *.#a (8. I. *.-j $1. q.,;&-,.*&

$7: t>. ! e. ..I'I@ 1 . ;- '.. .. I.:.: 8.1 7 ts. sW.0 1 .4';1,',2t~

(1: t*.114t0 1..$j&71I'll?

' . 1 I PibO 1 . C. WiE 04 (1. 1::*11:'0 6.4tS?C OCI 9. t 2:. .'dl 2. 5: .i.T: 09 11, tiZ..~*il

-1.7314E $13 $8. I .r9c*Ct - t , 1"l.IZL' (v( U.il.~!t~l l.rl.a2ii c14 f r1..1 1:?*o . 2,C:.:?Ci O.? 11. I ./. CI .~2.5-1U 0,: . 0. I l.if..1..3

-3. W..(tE:E 04 n . . c. lilt> . 3.2:.:e.l.:

ne 1 1. ! :. 8'. . $ . s:..r. c' .:,.* (I. t 78. .-#.. . :a . 4 - ,rs'b~ 04 4,. 17:*C1+.1

• ,2. 7,!'/'6 0.1 11. .. *a'$. .P.lr .:.*: 3.) rc. l,.:;r' -:,;..4t~::

G,? tl. 6 $I*:*. &.I - 1 . 1. ?.",'p: &j! I I 1.2.t%,:~'

e.? ##.. ,:.I,~I .. II,;..Q;.~.

&l,:,b, ***-t ..#.'.....*

1.1: 81.i:;: :u *4.zg81%-

UJ; f1.2:' I 0'-3.?li L 04 1 . (I? ti,i..~!\*O

-1.78kE 05 0, 31. 't.613 - 1 ;Q?31q- O<' i*..2..l:-*bU

-1.7i<.i:

04 0.2":C49 -2.SL4IL:

04 1 I. 2 .*,::I .I." - 2. if. r*,1: I>4 0. ::;Or 00 .- I .425@E 04 U.:.lC+00

-1.115'1E 04 0.32:Y1 -I.dla?E 04 0 ,;32.*:: - 1 . &i,?GL 01, 0. ,:4t'OC* -2,0376E 04 0,'.':CDO

-1.03tFi 04 0. 31:C.OCl - 1 . (' -,:l.&E 04 6. aierir! . - t . tih.96 04 0. .~~WCI r f . BJI BE 04 {I, ,'?I- ;.0 "1 ,820: :E 0.4 , 0. riUC.C f .1)37Ut; 0.4 0.4t800 -1 .ZCICUE 134 ' 0.42300 -1.7iii2E 04 0.433&0 .-l;(lki~~

09 , . , 0. .14300:.?\ ,U#;S1: 04 o;~sP!'~o -I ,X%3BL;: '0.1 0.4GF.00 -1.7039E 04 0.47800 .-.1,7a46E 04 ' . . 0.469rQ .-1;71tOE 04 O.~SC... - 8 .?WN p4 . ..

TABLE 6.A-11 (SHEET 23 OF 32) FUNCTION DESCRIPTION

= ( F0T:C I Eli3 FUNCT 1 ON AT WDE 42 OL)OOOOO~:~O~~~~~~~@O~~CY~~~G~~OO~)

N(W1RK9 OF ARCiCI SLhE r I 877) FUIK I I O:.( SCALE FACTOR

= ( 1 .0000E 00) TlflE VALUE FUNCTION TI 0. I . 5D3EE 03 0.00050 1. I;O3SE 03 0.00100 1.6032E 03 0.00160 1.6032E 03 0.002 ?0 1 .5032E 03 0.002S0 1,602YE 03 0.00300 1.6035E 03 0.00350 1,6035f 03 0.00403 1.5028E 03 0.00150 1.5021E 03 0.0@500 I.b(b31E 03 0.00553 1.5013E 03 0.00600 1.6006E 03 0.00350 1 .4083E 03 O.O0/00 1.4991E 03 0.00760 1 .4084E 03 0.00800 1 .4959E 03 0.00850 1.4041E 03 0.00800 1.4CB8E03 t' 0.005i50 1. 4828E 03 ? &I 0.01000 1.4714E 03 cn 0.01030 1.45GSE 03 0.01100 l.4335E 03 0.01150 1.3f177E 03 0.Ol2W 1.34PSE 03 F 0.01250 1.283tE 03 I-' 0.01300 1.IQ35E 03 0.01350 1.0737E 03 0.01400 9. IQOOE 02 0.01450 7.2180E 02 0.01600 4.7100E 02 0.01650 1.5040E 02 0.01600 -2.2570E 02 0.01660 -6.0760E 02 a M 0.01700-1,2406E03 4 0.01750 -t.C1934E 03 0.01800 -2.6684E 03 0 0.01650 -3.5416E 03 0.01900 -4.6553E 03 I 0.01950 -1.706SE 03 0.02000 -7.0016E 03 0.02050 -8,449GE 03 % 0.02100-I.0050E04 T: 0.02160 -1.lBlOE 04 t' 0.02200 - 1.3734E 04 0.02250 -1.5814E 04 0.02300-\,80G1E04 0.02350 -2.0465E 04 U3 0.02400 -2.3007E 04 a2 IP 0.02160 -2.66QOE 04 0.02500 -2.bSO6E 04 0.03000 -6.1493E 04 0.03500 -0.8387E 04 HE VALUE FUNCTION TlHE VALUE FUt;CTION 71 0.00010 l.bO39E 03 0.00325 1.5073E 03 0.00060 1.603kE 03 6.00470 1 . b030C 03 0.00110 1.5032E 03 .0,00\20 t.8032E 03 0.00160 l.5032E 03 0.00170 1,SOtZE 03 0.00210 1.R032E 03 0.0<1220 1 ,5021F 03 0,00260 1.5043E 03 0.00270 1,5013'; 03 0.00310 1.6036E 03 0100320 1.8036E 03 0.00360 1.6026C 03 0.00370 i,b028E 03 0.00410 1.502EE 03 0.00120 1.E021E 03 0.00460 1,6021E 03 0.00470 l.bul3E 03 0.00510 1.603tE 03 0.00120 l.6031E 03 0.00660 1.6013E 03 0.006 /0 1 . SOOLE 03 0.006t0 1.4999E 03 0.00020 1.4BSDE 03 O.OOGC0 1.4991E 03 0.00570 1.4391E 03 0.00710 1.4991E 03 0.00720 1,4991E 03 0.00760 1.4304E 03 0.00770 1.4CtCE 03 0.00810 1.4358E 03 0.09820 1.49dBE 03 0.00D60 1.494tE 03 0.00970 1.40.11E 03 0.00910 1.4081E 03 0.00920 1.4853E. 03 0.00!160 . 1 .4a2CE 0'3 0.00970 t .4elOE 03 0.01010 1.46QGE 03 0.01020 l.46ClE 03 0.01060 1.4530E 03 0.01070 1.4477L 03 0.01110 1.4267E 03 0.011R0 1.421.3E 03 0.011G0 1.3906E 03 0.01170 1.3blPE 03 0.0l~la 1.338~~ 03 0.0 12%0 1 ,325i;C 03 0.01ZbO 1.2687E 03 0.01270 .l.ZbZOE 03 0.01310 1.172GE 03 0.01 320 1,1606E 03 0.01360 1.0474E 03 0.01370 1.0160E 03 0.01410 8.0320E 02 0.01420 8.4740E 02 0.01460 6.7670E 02 0.01470 6.27BOE 02 0.01510 4.1420E 02 0.01520 3.6SSkrE 02 0.01560 8.7000E 01 . 0.01670 1.30C10E 01 0.01610 -3.1090E 02 0.01620 -3.QDEOE 02 0.01660 -7.0990E 02 0.01670 -8.8620E 02 0.01710 -1.3620E 03 0.01720 -1.48C2E 03 0.01760 -2.0370E 03 0.01720 -2.1C73E 03 0.01&10 -2.0250C 03 0.0t820 -2. SC.70E 03 0.01860 -3.733BE 03 0.01870 -3.8296E 03 0.01910 -4.7760E 03 0.01920 -4. QL297E 03 0.01600 -5.B52PE 03 0.01970 -G.k@84E 03 0.02010 -7.27736 03 0.02020

-7.SGlBE 03 0.02060 -8.7652E 03 0.02070 -9.0710E 03 0.02110 -1.0390E 04 0.02120 -1.0736C 04 0.02160 -1.2182E 04 0.02170 -1.P561E 04 0.02210 -1.4135E 04 0.02220 -1.45.76E 0.1 0.02260 -1.6252E 04 0.02270 -1.6tti4E 04 0.02310 -1.8831E 04 0.02320 -1.900SE 04 0.02360 -2.OSGOE 04 0,02370 -2.1463E 04 0.02410 -&?.3533E 04 0.02420 -2.4053E 04 0.02460 -2.6244E 04 0.02470 -2.6P02r 04 0,02600 -3.45OOE 04 0.02700 -4.0C$SE 04 0.03100 -6.8565E.04 0.03200 -7.5423E 04 0.03600 -8.OlOOE 04 0.03700 -0.77396 04 HE VALUE FUI1L;'t:Oi.!

T O.on(~'.g

~.[;C'I?:E c~i o.Oi1e60 r . t.05:CL 03 0. (101 30 % .&I r: ';'E O.3 0.001 fir3 1 . F432:I 09 . o.OC126n 1 . c.~sF::,~-:

re.- O. 0(12tiri r . sc1.5 ;2~ 03 0.00331% I. W lCsE 03 0.00380 . 1 . hic26E 03 O.'0f@35 1 . RfG!lE 0% 0.00~1&0 1,bOli'E 0b 0.00630 1 . t;O:: 1 E 03 0. 00G60' I .SOCIC.E 03 , 0.00E30 1 .d91'CE 03 0. DOCCG 1 ..A901E 03 0.0073U I .46:4 IE 03 0.007eo 1 .49*:~~ i!3 0, 00C.30 1 .4:.*. C E 03 0, ooeec~ r . ,. r-.:::.:.'r 03 O.00(130 t .4t*; 0% 0, 0.0011t)l i $ . ri7! ?ti 0% 0.01030 1.46.73E 03 0.01060 1,4?::.1E 03 0.01130 1.4126E 03 0.011c)o l.;.:'e2E 0s 0.01239 ).31:'!;':

0? O.UI280 1. L.iBL;E Ob 0.01 33'5 1 . 1262ii 0'3 0.013E0 P.P840E 0:: 0.01430 8.073UE 02 , 0.01460 6.7710E 02 0. OltiSil 2. ~L~?CIE 02 O.Ol880 -6.2SOOE 01 0.01630 -4.02.10E 02 0.016eG -1.0076E 03 0.01730 -1.GlDQE 03 0.017CO -2.33CCiE 03 0.01630 -9.173ilE 09 0,01680 -1.135OE 03 0.01930 -7,2279F-03 O.OlbBO -C..l(:*OE 03 0.02030 -7.HUlaE 0:s 0.02080 -9.8WlE 03 0.02130 -I.%CJi;7E 04 0.02160 -1,284flL 04 . 0.02230 -1.4963E 04 ' 0.02260 -1.71NE 01 0.02330 -1.9406E 04 0.02C80 -2.1973E 04 0.02430 -2.4GOlE 04 0.02480 -2.7365E 04 0.02900 -4.7561E 04 0.03300 -6.1206E 04 0.03800 -0.44ie~ M TABLE 6.A-11 (SHEET 24 OF 32) w I I-' I-' w a2 &.

TABLE 6 .A-ll (SHEET 25 OF 32) TIHE I-'UNCTIOH NWER 1 13) > FU14CT I ON 9EYCFr l PT I ON 8 1 FORCING FUNCT1O)I Af NCiDE 43 00*0';:, 0. 66.6 ) NLWK'CII OF ABSCISSAE ( 877) FUlJb 1101' SCALE r'C.CTGl: ( 1 .0<*00E 00) TitlE VALUE FUflCTlON TI 0. 6.68OOE 01 0.00050 6.GSOOE 01 .0.00100 6. 6400E 01 0.00160 6.6400E 01 0.00200 6.610C*E 01 o.OoE-30 6. OdOOE 01 0.00300 6.6600E 01 0.00350 6.6600E 01 , 0.00100 6.6400E01 0.00~50 G.C~O~E 01 0.00500 6.6490E 01 0.00fit50 6.6400E 01 0.00600 6.63JOE 01 0.00650 6.6SOOE 01 0.00700 6.630OE 01 0.00750 6.0200E 01 0.01l800 6.6100E 01 t-' 0.00850 6.6000E 01 m 0.00900 6.bUOOE 01 0.00950 6.65hOE 01 ' o\ 0.01000 6,BOUOE 01 0.0)050 6.4400E 01 > 0.01100 6.3100t 01 I 0.01150 6.1800E 01 I-' 0.01200 5,@700E at P 0.01290. b.6700E 01 0.01200 5.2600E 01 0.01350 4.7500E 01 0.01400 4.OEOOE 01 0.01460 3.1800L OI 0.01500 2.080CrE 01 0.01&60 7.01.0OE 00 0.01600 -1.0000E 01 O.Ol650 .3,04OOE 01 0.01700 -5.4eOOE 01 0.01750 -6.3700E 01 0. bl8OO -1.1750E .02 0.01850 -1.5GGOE 02 0.01B00 -2.OlIOE 02 0.01960 -2.5230E 02 0.02000 -3.ODSOE 02 0.02050 -3.73tbOE 02 0.02100 -4.4420p 02 0.02150 -5.2210E 02 0.02200 -6.0710E 02 0.02250 -6.0910E 02 0.02300 -7.0840E 02 0.02350 -0.0470E 02 0.02400 -1.017OE 03 0.02450 -1.1356E 03 0.02500 -1.2601E 03 0.03000 -2.7163E 03 0.03500 -3.007tE 03 HE VALUE FUi ICT I Obi T 1 HE V.?LUE f UfIC I I OH T 1 0.000l0 6.6ti00E 01 0.000?n ti 6fiOOF 01 0.000.3(1 S.6500E 01 0.0C@01o 6. GbDvE 01 0.00110 6.6400E 01 0.00120 6.G40OE 01 O.00l60 6.640OE 01 0:06170 6.li400li 61 0.00210 6.6100E 01 0.00220 6.6,lOOE 01 0.00260 6.6600E 01 0.00271t O.GLOL.E 01 0.00310 G.6500E 01 0.09320 6. bb01.1E 01 0.M1360 6.640OE 01 0.00370 6.frlbOE 01 0.00410 6.6400E 01 0.00.110 6.6400E 01 0.00460 6.C40OE 01 0.00470 6.G40CbE 01 0.00510 6.6400E 01 0.00520 6.G40uE 01 O.OOW9 0.6400E 01 O.OCIB70 6.63i:1'1E 01 O.OO610 0.6300E 01 0 s 00826 6. G3 *)UL 01 0.00?60 6.6300E 01 0,00G7(1 6. C:?IOOE 01 0.00710 6.630OE 01 0.01172~ 6, ut:30fi 01 O.OO7G6 6.62OOE 01 O,0077i* 6.66.5W)or 01 U.CIOOIO 6.6100~ 01 O. onezn P za,ltn~~ 0 I 0.0OOGO 6.601.*OE 0 1 0.011970 6.600UE 01 O.OOOtO 6.5G00E 01 0. OC*O20 6, F300E 01 O.OC3GO 0.6500E 01 0.00970 6. Bt.O@E 01 0.01010 6.b000E 01 (1.01000 6..ibCIVE 01 0.01060 6.4200& 01 0.01070 B..~I-~DDE 01 O.Olll0 6.3100E 01 ,O.O112O 6.VUO6E 01 0.01160 6.1600E 01 O.Oll7O 0.1100E 01 O.OlklO 5.e200E 01 0.01220 6.CGOOE 01 0.01260 5.6100E 01 0.01270 5.5300E 01 0.01310 5.16OOE 01 0.01320 6.0900E 01 0.01360 4.6300E 01 0.01370 4,5000E 01 0.01410 3.0000f 01 0,01420 3.7b00E 01 0.01460 9.0000E OI 0.01470 @.7( OOE 01 0.01510 1.6300E 01 0.01020 l.13t3rE 01 0.01660 3.800OE 00 0.01S70 6.00UOE-01 0.01610 -1.3700~ 01 0.01620 -1.7G00E 01 0.016G0 -3.4903E 01 0,01670 -3.QSODE 01 0.01710 -6.0200E 01 0,01720 -6.6000E 01 0.01760 -0.0000E 01 0,0177~ -9.6700E 01 0.0181U -1.240OE 02 0.01 820 - 1 . Y;?60E 02 0.018GO -1.6610E 02 0.01876 -1,737OE 02 0.01810 -2.1110E 02 0.0192Cf -2.2100E 02 0.01960 -9.6310E FZ 0.01970 -2.7430E 02 0.02010 -3.21706 uZ 0.02020 -3.3fi30t 02 0.02060 -3.8700E 02 0,02070 -4.0100E 02 0.02110 -4.B930E 02 0.02120 -4.7460E 02 0.02160 -6.3b0GE OE 0.02170 -6 tu20E 02 0.02210 -6.0480E 02 0.02220 -6.4P00E 02 0.02260 -7.1840E 02 0.OP270 -7.5650E 02 0.02310 -8.lOPOE 02 0.02320

-8.4010E 02 0.02360 -0.06EOE 02 0.02370 -9.4880E 02 0.02410 -1.01036 03 0.02420 -1.0637E 03 0,02460 -#.l6OlE 03 0.02470 - # . l b4BE 03 0.02600 -$.52blE 03 0.02700 -1.6072E 03 0.03100 -3.030SE 03 0.03200 -3.1440E 03 0.03600 -3.038CE 03 0.03700 -3.6786E 03 IW VE.LlMi FUJI.. .Y I I 8; I 7 1 i V . E TUCI. : 8 i riH 0.000?8" a. -6'8f%(: *-*,@,.+ !I., G.C!51, .: 01 0. 0011L .3 C,. c ..J*.*I.'

U I w. I.. -. I* ty. :?..rI.-~k' CJI 0.~1:*0 ,:.c$rn>~

tit . , ,. , 1 : C. c.~~'..)!

' I .I O.OOtC0 G.'.'~'I?QL.

111 i#.na I,~*.I U.b..:6,.tk 01 0,002:'0 6. 64'2*.*E 11 l @.I*..*)" I;.Q~.?(I,~c 01 rt. Ucli?li!l

6. i.',~.. .i:: I. t II.I..*::I'I ki., rbI.~E.t#I 0.003?0 L; . BCE~IDB r) l tb.,~::.:~

6.1. r.3t 01 0.003L:O S.Li.??-frlI Oi 0. CI.'*":,:.,~

6, 6.109;: 01 0.0~30 G.G?CI..W nt C1.'%.3'"O

9. @f?W3. 01 o.o(-?.;u ti.cqitoi;:

01 n.:e.a.:. .I G.G?~I)L 111 0.066W C .6.+?OE 0 1 (r.t!'il t(. 6.6.!1'.~

01 0. On'Jbi. C. O:*:r'**>.l 01 cb. -,*I :W 6.liru.r.':

all, O.OUG3O O,F'?O~..'.ul

&I. <I*'#. '&.'I '~,BYU(IL.'

fbl 0. (@f;:-.> 6. :, c I t ..: , **11 {b. ** f: , .'. ij.c.::#*.b:'

0. my53 l,, ?.'>l'lI.l?. . I) *. tb.. !t* 6

• c: '.l!I. .. ttl 6. ilr)7@& ti. <~!LI*:;L LI l t*.@~.*?:.lt~t 6.66dUb" (11 (r. 0Oij$1:1 8 .. c,:: *an ' ill 8 ,I' .:.?,.@ 6.('1.** L; I,# v.CIOL.~Q 1~.6~".*;9 01 4' : I*. . :::Q ti. tis~..~ u 1 0. CM~LI<@ 6. b?llk?I: i (1.1.*11.70

6. (.'X**JI.

t$l 0. truS~< <I G . b-!*:f*t 1 1 6.,. O' .!I 6.t.3llC UI 0.01 u::3 G.~:*.R.:<

UI 1:. (1 1.: .:..! 5. rlCL*~..r; 01 0. 010t'(~ 1[;. 3n~i. (11 ft.tl~*.+:n.)

6. htQ>T 01 0.61160 i.Srd!OOIE it1 0.111 1 '!0 .6.Lf00E 08 O.OllU0 6.0','::0t 01 C1.10 Ib.3 6. C12WE 01 o.oiz:*o ii.eo.tir~r 01 o.(I~?.:u h.i06QE 01 0. 01CP:t fi.d'.:Ot)E 01 . CI . I: : : ': ti .':a f @:bL [I 1 0.013P.Q 4.0800E01 . 0.015!21 4.87UOEOI 0.013t;O 4.3600E Ot ' O.Oi?S+u' 4.010O5 01 0.01i30 3.670UE 01 u.Ot4.~~~

3.eJWIE 01 0.014*+0 0.66*-*0E 01 0.01490 .P.320rJE 01 o.ot~+o 1,29~li*t 01 O.~-~I:~ICI i.ooo*.: ar 0. o t 5co -r . atsooE ccn . o . o I ticair - s.aocr.,~

no 0.01~30 -2. I~~IOT: 01, O.O1*>10 L2.tOWE Ot 0.016GO -4.45i10E 01 f&.Ol.SSG

-4.W00E 01 0.0lf5c. -7. tti(8iC 01 6.8 17.10 -7;70f Dl! 111 0.01 706 -I . o.4. $1.1~ i*& 0.01530 -1.102:0~

c12 0. 01 @$IJ - 1 , 41.1'. LI~: #.b . f1.t.11 '4:-1 .43:*liC 02 0.162 'O.L~lOt~:-i.~I@t~Co'Y 0.01@:.::1

-2.31 Iik 02 ~1.01040 '-g.416OE 02 0.6tsi0 -r.8$to~ cb2 . n.r;!rtbo 4!.~760f 02 0.02(*20 -5.4710E 02 '. 11.0Pr?O.-3.6310f 112 0. 02~1f.0 -4. lfrlc~ 02 0.0201*2*.

-1. I 46~~ .02 0.Ci~lZo -L.BCI~CIE re .n. nPl.Zi4 - F; tltiI:n3E 03 11. u, I HO . -6.89bt1E 1)2 0.62lBri -b.it;'uL-I 02 O.OPi!JO -C;. 61 CIOE 02 .' . 0.02240 -6.6020E.

4.12 0.0221;0 -7. Bl/tt~E 02 0.02290 .-7:7800E 02 0.0233O -8.6140E 02 .. 0.OY5<0 -8.b2WE 02 0.02380 -9.7130E 0.2 . 0.02340 -0.94lOE 02 0.02430 -1.087t;:

(13 '.U.0114Q

-1.1116E03 0.024EO -I ,4007E 03 4.02490 - 1. P346E 03 0.02800 -2,lOZOE 03 0.02900~-~.4070E 03 0.03300 -3.6697E 03 0.03-100 4.7844E 03 0.03800 -3.7317E 03 . 0.03900'.~3..PWOE 03 . .. . ,

TABLE 6.A-11 (SHEET 26 OF 32) 3' I I-' I-' 0. f*+400 - 1.8006E 03 0. O.iS:..IO

-7.0G,.!.JE 02 t@.s-*: rib: -2.CG:OE UX P.:? .:i*: -l.Q3r**bE 01 11.051O9 8. slf.OZ 0:: is. : ; nc. 1 .3 3:?7~ 113 11. .I . b$;.'E c!.: o . 0 .,9;3 I . c:anL'E 03 le. *>-!*+0 1 .6.103E 0s U.QJL:O 1.P913C 0::t 30 !.:a 1 .(I# n$E 03 a. tb;.tlOil

7. tr2iici 02 0.11V@0 b.fli9.1E 02 t,, 6. t li*.?rir) 1.tOO ti. 4.2000E 47WE 02 OY 0.Il49CtO 5.9790E'OL

' 0.12400 6.2660E 02 t.129.W 4.6600E 02 o.i3.-wn P.+~BDE 02 11. I srS(i3 - 1 .62u3E 02 0 ..I4400 1 .7920E 02 ' - o. i laoti -2.76b:lE (12

• ra. I ::.!ta -2. nwiti na 0. I ~&'JO - 1 .4!361.bE 1'2 . It. 16406 -3. Q3in)E 1.1 I 0 1C"OO 8.47110F 01 t' (1. 1?.:00 @.1':30C 01 ul fi. 1 761L(O 1 .09413E 01: a 11. 1 f3+?00 1 .004M 02 cn lb. b.s:~CICb

-1 .t*lbdE 0.2 I C 1.. 1:.?.?03 -4.07W.E 02 , tq 0. i: . 56 -9.0330E 02 '41. i??*$F*> -7,762Uf 112 0. ZIEGO .-9. 29CtOE 01 F ' % o . 1~:?a00 1 . O~PZE on 0. k>.iOO . 1 ,2275E 03 O.lbe00 6.2280' 02 0.5';-300

-8,401OE 02 0.26000 -6.6209E 02 0.27900 - 1 ; 68SOE 02 0.2 .idO(r 8.6BOOE 02 0.23800 6,2700E 02 Ca. '. 08011 -2.9n70E 02 0.31&00 -6.6390E 02 0.328ufl - 3.682OE 02 0.23900 7.32006 01 0.34900 7.eJOOE 01 n. ?!.:3oct -a, aoenr 02 U. QGdUtJ -4 ..?I JOE 02 0.376L30 -3.0120E 02 6. BBiQ6. -3 72001; 01 0.39C100 -3,3400E Ot 0.40b00 '2.021OE 02 .O.41800 -p.O81OE 02 0.42&00 -1.4160E 02 ' 0.43830 -1.3300E 02 0.44600 -2.0710E 02 , .' 0.45800.-P,3mO~

02 0.46300 -1 id620E 02 0.47600 -1.40MK 02 . '. 0.40600 -1..035OE 02 ,. .'. 0.49800 ,,:I ;2820L 02 . . '. .:*... .. . I.. .,, . : . . . .-, ,% .. .

TABLE 6.A-ll (SHEET 27 OF 32) 0 b I I-' t-' I-' w 03 rP Fur1..: I 1 \el I Or SIX l l't l ON

• i t ORCI FIG FllNCTlON AT WOE 3 00>02S 0. -7300.0 J El~'lll'4'l:

or .rC?CI SSPE = f 877) rub: I I u,! t:..a~.t ~E.L tor; = 4 1 . OOOOE 00) TIHE VALUE FUNCTI0I.I TI 0.00010 -7.3006E 03 0.000GO -7.2ODSE 03 0.001 10 -7, 2inx 03 0.00160 -G,0127E 03 0.002lO -6.1530E 03 0.002C.0 -4.7YObE 03 0.C10510 -2.6030E 03 0.003GO 8.1260E 02 0.00410 4.7371E 03 , O.OO4tV 8.6C15E 03 O.OOSl0 1.2731E 04 O.OOC6O 1.7142E 04 0.00010 2.203bE 0.1 0. OOGCO 2.55YOC 04 0.007 10 3.30GOE 0.1 0.06750 4.1032E 01 0.00i10 4.E?DOE 01 0. C@(ZOCiO 6. 6(13**E 0 1 0.01)6 1 0 C .9OBC.E 04 ('.UljC)GO C 2110E 04 0.alOiO b.6T.RCE 0.1 CI.01Q80 1.1002E 05 0.01110 1.2CO3E 09 0.01160 1.80PC.E 06 0.01210 l.bL.95E 05 0.01 2c;o 1 . ?6Ib?E 06 0.01310 l .<d9SE 05 O. 01 :-GO 2.1:;:42~;

05 0.01410 2.39C"E 05 O.Ot4~D P.B.?LCE 06 0.01810 P.C3231' t'.i t1.OlCBO 3.ltt2Li 03 0.0lOln 3.ci:..lr;.E 0!i 0. Ulii@;lI 3. 2264E 0:li O.U171(' 4.0.':4CE Od O.(t17S!i 4.1".'-.?K 05 0. 610'10 .?. i :'I. BE 05 0.OIC.Gt.

t ChlhSi: WG O.OIfil(~

t*. 3';6'/E 05 tl.0131.*

• .6212L.

05 0.02010 6. T'.'lftClE 115 0.O21:jt;O G . 23S':E 03 0,021IO 6.G*S*oE Cod 0.u21CO 6.7o60E 05 0,62310 7.0059E 05 O.OY2GO 7.24319 05 0.023tO 7.4211E 05 0.02360 7.6071E 00 0.02410 7.7611E 03 0.02166 7.9332E 05 0.02500 8.3202.E 06 0.03100 b.P:GBE 05 0.03090 8.t311SE OS W VALUE O.OC~C'20 0.06670 .o, 001 20 0,00170 0.00220 0.00270 Or00320 0.00370 0.004?C* 0.00470 O.OG320 0.00570 0.00520 O.Oc?i70 0.00720 0.00770 0.00320

0. OOE 70 0.0062..

0.00976 0.01020 0,01070 0,oi teo 0.01170 n.01220 O.Ot2iir 0.01320 0.01370 0.01.020 0,01470 0.0tf2'ir) 0.01570 0.211627 0,016i0 0.01 720 0.01Y70 0.01610 0.01 9'70 0.01920 6.0i370 0.02020 0.62:170 0.02120 0.02170 0.022iib 0.02270 0.02320 0. v2370 0.024 20 0.02470 0.02700 0.03200 0.03700 TltE VALUE PUWCTIC)i1 0.OC*U30 -7. 2C:'rsSE 03 0.0UObO -7.2724E 03 0.00180 -7.13S.1E 03 0.001BO -G.G7l7E 03 0.00290 -8.BP34E 03 0.00260 -3,5356E Uii 0.00330 -1.3083E 03 0.00380 e.3892E 03 0.60430 6.3fillE 03 0.00480 I .62SUE r)4 0. C.3530. 1 .4 1 17E 04 0.005LO t.F.:!'c 04 .0.00630 2.41b3E a1 0. W661'1 i? . "Y'/ZE 04 0.00760 :+. :;C+l.lE C*4 0. Oi.7Cli ..!. ..4: :14E 01 O.O(:PSl*

U. ;.,I Z?E U.1 0.000f't)

G. :::'lGE 0-1 0. UOL*?O 1-45 t HI: n4 @.000:.1 c..~~:'trf:

13.1 O.OI049 1 .Ol<!BE O.OlC*PJ l,lT.:0ECi,.r 0.01150 1.P9s7E 03 0. 01 180 1 .46c2E @:19 0.01 2.:'9 1 . 6C.i 7E 0 > U. 61 290 1. tJ2f-1E OU 0.0 1380 2.031CE 05 0. Ol bFO :. .2!.:2t:E 05 0.01430, P.48-lSE 05 0.016C;O r:. 72GOE O!f 0.01 SsO 2. il:..'?iE WO 0.01 C~90 3.2Cr13F tlli O.Oii.313 R:'.."fPl.

0': 0. 01 f.01.8 3. l:?(SiilI (6l.i O.flt7?l>

4.b':it:-:I 4..,.: O.e.17!.]

4.6 '27. .t'i 0. (11 GPO 4. : l lr-1: rj!7 0.01~GO 5. l!i?li.- .::I; 0. 01?"1l~ ti..?'? e1L r..; 0.cIb:~';bl I.. ; .:..t11: .*, 0. UdlILSu ti. C*;;<* I cv'i 0.0s cisn 6. ::.,::.LC 11t.i 0. 02 1 .'+it 6, ti 3 ',-ie tb'> 0.021(:0 c.(;i,.;f.C.it--..

0.62233 7.10rCE 00 0.022&0 7.23-2E 0'; 0.02330 7.P030E 03 0.02380 7.WEOE 03 0.00430 7.8370E 06 0.62480 7.Q.'.9E 05 0. u2:i.00 B.70YrlE 05 0.033011 8.S318E 00 0.03600 F.6DIf5E 06 FUNCTION -7.2879C 03 -7.25b8E 63 -7.079SE 03 -5.6226E 03 -6.4149E 03 - 3. wo~e '03 -6.4440E 02 3.1735E 03 7.0853E 03 1.1072E 04 1,6329E 04 - 2.OOlOE 6( 2.52cJ'JE 04 . 3.12Z@E.04 3.0077E 04 4.63CDE 04 . 6.6ulSE (4 .6. ti?t-:,i 04 '/ ..7'.B'z 5 E 04 9. (':I$ 1 t 64 1.0447E 05 1 . P0;:riE 05 I. 32S7E 05 .1.46SOE 06 ' 1 .6 )O'..E 05 1.86666 05 P,OIC(IE 05 z.2sao~ 08 2.63kYE 00 0.7793E 05 3.W13E 06 3.3176E 01 3.L*J7lL 05 3.50G'K 08 .4, 223::, Of,. .I . t: a: I ,r; 4. bi311.. OG !:i.2.h*.

L CIS 0. 5L. 1. .l: hi5 U (12:~.&$,..:

itG 6.1349; 05 6.425'.*1 tlb 6. iaanv.. $15 ' 6.B;k.N-,US 7.'1%1lt 05 7.3760; t'5 ?.tir(>lE 06 7.6906E 05 '1.8698E 05 0.02d2E 05 e.qjwlc oo 'I). w70E 08 . . , 0'.4774E 05 . . C r 'cn TABLE 6.A-11 (SHEET 28 OF 32) I-' I-' Q 03 1P 8.3416E 06 7.641*16 06 6.GOCOE 05 '-1.0 113E 06 b.413GE OG 5.0 8 /:IE 05 4.7042E 06 d .G2GGE 05 4.4?;2lE 06 4.2077E 08 3.01<,:7E O!! Q.63li7E 05 L'..1641E 06 3.9(5(111L 06 3. I'J40E Oli 3.2347E 06 3. It1548 08 3.lAlRE 06, '; ,200 1E 05 3.21616 00 9. :Z;+OOE 06 3 . SOfi%E CIS 3. 1 t;itiE Ob 3. 103311 Oti 3.0ti37E 05 2. !'ll>EL oli G.!l2f3E 0s 3 . Li?lbltL 015 T . on2r.r: ns 5. .?'jl.L'I:

OG f ,C100E 06 L . ?3r, ;E. 05 2. ~QG!~E 06 3.741UE 06 2. ,.:/,:1:1 0:; :!,G1421E (16 2, !)I! 1 tiL 011 2. ,.*'/<!I,L O!b 2. G(;..Iq>E Oti .? , ? i:,fl:,E O!

• 2. 1.. I;IE Ob 2. ' .2>.u1 0s 2. ;. I -ft,r. nti 2 , t .,I: i',L' 1IL; %!..:'Ii:kE (lb * . , ..+.i;+E 1.1! . . .? . e-*:? i E oli i..: 1 'J7t 06 '!.l),?iltl-05 3 , :. .'I '.,I- OG :!. :y$C 0li 2 L3'JGI:

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0. 3;:.31*4 t (1. !.. : .! g'> 0, ;,*.'00 0.9b;?400 Ci. 96490 0.3; -40'1 0. 36-?03 (1. 3L-100 0. .I0400 0.41400 0.42400 0.4SfIOO 0.44400 0.45.100 0.46.tOO 0.47100 G.4C!30 0.39?00 8. 04OCE 66 7. IP12E 05 6.366% 08 5.7L1O.IE 06 6. 2th14C 06 4. $15'5?L 05 4.7167L 05 4. bc^BSE P'> 4 . 420.i:: 08 rt. 092CtE 05 3.3077E 05 3.6722E 06 3.4121E CB 3.3272E PY 3. PC 04E 05 3.12135E G5 3. ICfUL' 05 3.16786 06 3 21.2fiE Lj'3 3.?PO%E 05 3. Pi.43E 05 3.1SUlE 06 9. ) :.'olrf O!;. 3,IM ILL it6 3. IY.'~ hE 05 7. t5..:?l;k (It, ~.BIO.IF O6 2. .,clt)PC 2.6773E 06 2.i;L.i3lE 05 2. P407E 00, P. 6%0:3E 00 2.739oE 05 2.7143F 05 2. f./!?Ok 10% , 2.60436 05 Z.C?PQE. 08 5. br:67f G5 5. bt572E (8% 2. !j?c;c.ic 06 3. ti5 3(li? 09 2. aC1i.L 05 2 . :r I 6C.L O!J 2 . -? C '; 1 1 05 2.4'131i:

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0,44bOO 0.49900 0.46900 0.47UOO '0:40$00 .' o. 4~e~s 7. TOME 08 6. t>?QC 05 6.Ut)EQL 06 8.488iK 08 8. (1'33r)E 06 rt.622fW OOb 4.667* 06 4.49376 05 4. k633E 08 3.8772E 06 0.70rtE 06 3.4946E 08 3.3719E 06 3.3032E 06 3.LY4S7c 06 . 3.1@76(. 06 3.182M 06 D.lOc,'E OP, 3. L'IG.lt' CIS 3.231% 06 3.PIP;YI 05, 3.1:iF:lr' 05 3.1l:."C 06 3,64~:'.:c (85 3.111 12E 05 Y. e!.i.V+i 06 2.ausroa , 2.0009E 06 2.87t3E 06 2.0'4G1; 06 9. 83eq!E 06 k.61MI-: 06 2. -/UP.* OD 2.687NE 05 2. C2C2E 06 2.695SE 06 P .B73iE 01) P.OGlM O!. 2.IIj31E OPi 2.! ?5Lk 06 &;ti5llE 06 9 . v-',: na /. ritrl Lt Ub 2.4 41: ($6 .2.r(t;Oit'

2. .let. .z 116 3.4 JP4C 05 2.~11. 'qi 06 4. 38 I*..; 05 0,3'/7)t 05 , 2. 'J759E 06 2.4062E 06 , 2.44C'I.E 06 0;18M 06 P.47bM 06 0.47IK 06 4.4332K 06 2.4191E OS 2.S647E 06 0.3702E 06 2.3693E 05 . g.3702t 05 ' . . .

TABLE 6.A-11 (SHEET 29 OF 32) FUNCTION OESCRI PT ;ON ( FORCINO FUNCTION AT NODE P LI;*O2S 0. NUIWER OF ABSC I SSAE = l 577) . FUNCTION SCALE FACTOR = ( 1.0000E 00) 1 t3 t' M cn ;; I-' I-' E G 0 I % !a H t? I-' Lo CO 1P FUNCTION TI -4.6990E 02 -4.6410E 02 -4. 08GOE 02 -1.63BOE 02 5.218OE 02 1.8154E 03 3.9928E 03 7.257SE 03 1.1382E 04 1.5581E 04 1 ,9954E 04 2.4709E 04 3.0018E 04 3.6035E 04 4,2950E 04 5.0913E 04 6.0133E 04 7.0748E 04 8.283OE 04 0.6506E 04 1.1175C 05 1. 2857E 05 1.4704E 05 1.6173E 05 1.6120E 05 2.028OE 05 2.2616E 05 2.5112E 08 2.7767E 05 3.0578E 05 3.35tilE 05 3.6713E 05 4.005lE 05 4.3553E 05 4.7213E 05 6.1039E 05 6.4993E 05 5.90606 05 6.3212E 05 6.66038 05 7.0739E 05 7.4661E 05 7.8411E 06 8.2011E 06 8. 5295E 06 6.8701E 05 O. 1 957E 08 9.4110E 05 9.6628E 05 0. Q289E 06 1.0170E 06 1.1052E 06 1.2777E 06 HE VALUE 0.00010 0.00060 0.001 10 0.00160 0.00210 0.00260 0.00310 0.00360 0.00410 0.00460 0.00510 0,00560 0.00610 0.00660 0.00710 0.00760 0.00810 0.00860 0.00910 0.00960 0.01010 O.Ot060 0.01110 0.01 160 0.01ZIO 0.01260 0.01310 0.01360 0.01410 0.01460 0.01610 0.01560 0.01610 0.01660 0.01710 0.01760 0.01810 0.01860 0.01910 0.01960 0.02010 0.02060 0.021 10 0.02160 0.02210 0.02260 0.02310 0,02360 0.02410 0.02460 0.02600 .0.03100 0.03600 FUNCT l ON -4.69906 02 -4.ClOOE 02 -3.'$15OE 02 4.8700E 01 9.5320E 02 I .5662E 03 b.1616E 03 8.8817E 03 1 .3053E 04 i. 7297E 04 2,1798E 04 P. 6766E 04 3.2330E 04 3.8605E 04 4,6001E 04 6.4438E 04 6.4206E 04 7. 5405E 04 8.8125E 04 1.0241E 06 1.1628E 05 1.3577E 05 1.5054E 06 1.6925E 05 1 .8S63E 05 2 1 'I 92E 06 2.3595E 05 2.6155E 05 2.8873E 05 3.1744E 05 3.47948 05 3. 8027E 05 4.1432E 05 4.4998E 05 4.8735E 06 (J 1605E 05 I. 0608E 05 6.0713E 05 6 4882E 05 6, 8266E 05 7 23314 05 7,6179E 05 7.Y871E 05 8.3193E 06 8 6674E 05 0.OC23E 06 9.3131E 06 9.4953E 05 S.7713E 05 1.0031E 06 1.1066E 06 1 2359E 06 1.2070E 06 TIHE VALUE 0.00030 O.OOa~0 0.00130 0.001eo 0.00230 0.00280 0.00330 0.00380 0.00430 0.00480 0.00530 0.00580 0,00630 0.00680 0.00730 0.00780 0.00830 0.00880 0.00930 0.00980 0.01030 0.01080 0.01 130 0.01 160 0.01230 0.01280 0.01330 0.01380 0.01430 0.01480 ' 0.01530 0.01680 0.01630 0.01680 0.01730 0.01780 0.01830 0.01680 0.01930 0.01 980 0.02030 0.02080 0.02130 0.02180 0.02230 0.02280 0.02330 0.02360 0.02430 0.02460 0.02800 0.03300 0.03800 ME VALUE 0.00040 0.00090 0.00140 a,00190 0,00240 0.00290 0.00340 0.00390 0.00440 0.00490 0. 00640 0.00590 0.00640 0.00690 0.00740 0.00790 0.00840 0.00690 0.00940 0.00990 0.01040 0.01090 0.01 140 0.01190 0.01240 0.01290 0.01340 0.01390 0.01440 '0.01490 o.otS40 0.01590 0.01640 0.01690 0,01740 0.01790 0.01640 0.01soo 0.01940 0.01990 0.02040 0. 02090 0.02140 0.021*0 0.02240 0.02290 0.02340 0.02390 0.02440 0.02490 0.02900 0.03400 0.03900 FUNCT 1 ON -4.6690E 02 -4.9670E Ot -2.4210E a2 3.4080E 02 1.4956E 03 3.474bE 03 6.5114E 03 $ . OUS2E 04 1.4736E 04 1.9052E 04 2.3718E04 - U.0904E 04 ' 3.4766E 04 4.1487E 04 4.92308 04 5.0179E 04 6.6513E 04 0.0305E 04 0.3650E 04 1 .O857E 05 1.P506E 06 1.4322E 05 1.5799E 05 1.7709E 05 1.0834E 05 U.Pl37E 08 2.4600E 05 2.7224E 05 3.0004E 05 3.2941E 06 3.6066E 05 3.0369E 08 4.2040E 05 4.6469E 05 1.0265E 05 5.4193E 08 5.8239E 05 6.2377E 05 6.6569E 05 6.0930E 06 7.368BE 08 7.7674E 06 0.1302E 05 0.4596E 08 0.0031E 05 9.1323E 05 B.382QE 00 0.6079E 05 0.1)772* 05 1.0t30E 06 I.167OE 06 1 ,26735: 06 I. t8Z6E 06 I-' \O a3 lb 0.04000 0.04500 0.05000 0.05500 0.06000 0.06500 0.07000 0.07500 0.08000

0. 08500 0.09000 0.09500 0. foooo 0.10500 0. I1000 0.11500 0.l2000 0.12500 0.13000 0. I3500 0.14000 0.14500 0.15000 0.15500 0.16000 0. 1 C.500 0.17000 0.17500 0, 18000 0.10~00 0.19000

0. I P500 0.20000 0.21000 0.2%000 0.23000 0.21000 0.25000 0.26000 0.27000 0.28000 0.29000 0.300CtO 0.31000 0.32000 0. $3000 0.34000 0.35000 0.36000 0,37000 0.38000 0.39000 0.40000 0.41000 0.42000 0.43000 0.44000 0.45000 0.46000 0.47000 0.48000 0.49000 TABLE 6.A-11 (SHEET 30 OF 3 0.04200 1.2534E 06 0.04700 1.t595E 06 0.08200 1.0343E 06 0.05700 9 3449E 05 0.06200 0.7173E 05 0.06700 8.3cse~ 05 0.07200 6.GI9SE 05 0.07700 7.7568E 05 0.08200 7.1974E 05 0.08700 6.*llSE 05 0.09200 6.4775E 06 0.09700 6.0709E 05 0: 10200 6.74666 05 0.10700 5.6162E 05 0.11200 ~.3SOOE 05 0.11700 6.2278E 05 0.12200 6,160tE 08 0.12700 5.1932E 05 0.13200 5.3018E 05 0.13700 5.6I29E 05 0,14200 5.6067E 05 0.14700 5.7130E 08 O.lb200 5.6564E 05 0.15700 6.6387E 08 0.16200 6.76276 08 0.16700 6.6517E 08 0.17200 6.5733E 05 0.17700 5.4628E 05 0.18200 6.3977E 05 0.18700 5.2403E 05 0.19200 5.464E 06 0.19700 4.9427E 00 0.20400 4.8010E 05 0.21400 4.7231E 05 0.22400 4.8188E 05 0,23400 4,9052E 05 0.24400 4.8655E 08 0.25400 4.6976E 05 0.26400 4.8595E 08 0.27400 4.4597E 05 0.28400 4. b463E 05 0,29400 4.5840E 05 0.30400 4. -000E 05 0.31400 4,3258E 05 0.32400 4.2703E 05 0.33400 4.3184E 06 0.34400 4,3525E 05 0.35403 4.3132E 08 0.36400 4.3094E 08 0,37400 4,lBbOE 05 0.38400 4.2169E 08 0.39400 4.2420E 05 0.40400 4.2234E 06 0.41400 4.2434E 05 0.42400 4.2594E 05 0.43400 4.2442E 05 '0.44400 4.i85lE 05 0.45400 4.t678E 05 0.46400 4.1462E 05 0.47400 4. I286E 05 0.48400 4.1189E 05 0.49400 4.12WE 05 TABLE 6.A-11 (SHEET 31 OF 32) FUNCTlOti DESCRIPTION

• ( FORCINQ FUNCTION AT WOE 1 (rJ>O2S 0. 1588.4 ) NUMBER OF ABSCISSAE , ( 577) FUNCTION SCALE FACTOR 1 1.00OOE 00) TIME VALUE 0. 0.00050 0.00100 0.00160 0.00200 0.00250 0.00300 0.00350 0.00400 0.00450 0.00500 0.00550 0.00600 . 0.00650 0,00700 0.00760 0.00800 0.00850 0.00900 0.60950 0.01000 0.01050 0.01 100 0.01 160 0.01200 0.01250.

0.01300 0.01350 0.01400 0.01460 0.01500 0.01650 0.01600 0.01650 0.01700 0.01 760 0.01800 0.01860 0.01900 0.01950 0.02000 0.02060 0.02100 0.02160 0.02200 0.02250 0,02300 0.02350 0.02400 0.02460 0.02500 0.03000 0.03500 FUNCTION TIME VALUE 1.6884E 03 0.00010 1.6884E 03 0.00060 1.6884E 03 . 0.001 10 1. 5889E 03 0.00160 1. 68OSE 03 0.00210 l.6894E 03 0.00260 1.5881E 03 0.00310 1.8887E 03 0.00360 1.6892E 03 0.00410 1,690OE 03 0.00460 1.5907E 03 0.00610 1.6904E 03 0.00660 1.6918E 03 0.00610 1.6953E 03 0.00660 1. 6989E 03 0.00710 1.6047E 03 0.00760 1,6134E 03 0.00810 l.6283E 03 0.00860 1.6493E 03 0.00910 1.6802E 03 0.00960 1,7209E 03 0.01010 1.7782E 03 0.01060 1.8525E 03 0.01110 1.9493E 03 0.01 160 2.0734E 03 0.01210 2.2287E 03 0.012G0 2,4189E 03 0.01310 2.6526E 03 0.01 360 2.931GE 03 0.01410 3.2648E 03 0.01460 3.6671E 03 0.01810 4.1132): 03 0,01860 4,64368 03 0.01610 6.2616E 03 O.Ql660 5.9446E 03 0.01710 6.7288E 03 0.01760 7.6094E 03 0.01810 0.6931E 03 0.01860 9.6868E 03 0.01010 1.0893E 04 0.01960 1,2217E 04 ' 0.02010 1.3662E 04 0.02060 1 ,6232.E 04 0.021lO 1.6924E 04 0.02160 1.0742E 04 0.02210 2.0685E 04 0.02260 2.2750E 04 0.02310 2.49338 04 0.02360 2.7220E 04 0.02410 2.9GlOE04 . 0.02460 3.2096E 04 0.02600 6.034SE 04 0.03100 8.2348E 04 0.03600 FUNCTION 1 I. 6884E 03 1.5884E 03 1.6684E 03 1.6889E 03 1.5889E 03 1.6894E 03 1.5087E 03 I, 6887E 03 1.5892E 03 1.5905E 03 l.5913E 03 ' 1,69046: 03 1. b92OE 03 1. 5958E 03 1.6003E 03 1 .6062E' 03 1.6156E 03 1.6313E 03 1.6637E 03 1.6858E 03 l.7303E 03 1.7911E 03 1.0710E 03 1.97226 03 2.1012E 03 2.2631E 03 P.4626E 03 2.7036E 03 9.9941E 03 3.3386E 03 3.7426E 03 4.2138E 03 ' 4.7691E 03 6.3844E 03 6.0935E 03 6.8966E 03 7.79765: 03 8.8046E 03 9.91908 03 1.1148E 04 1.24979 04 1.3967% 04 1,6560E 04 1.7277E 04 1.9121~ 04 2.1087E 04 2,3178E 04 2.5360E 04 2.7689E 04 3.01OOE 04 3.7322): 04 6.6148E 04 8.3199E 04 'IME VALUE 0.00020 0,00070 0.00120 0:0017o 0.00220 0.00270 0.00320 0.00370 0.00420 0.00470 0.00620 o.oos70 0.00620 0.00670 0.00720 0.00770 0,00820 0.00870 0.00920 0.00970 0.01020 0.01070 0.01120 0.01 I70 0.01220 0.01270 0.01320 0.01370 0.01420 0.01470 0.01520 0.01570 0.01620 0.01670 0.01720 0.01770 0.01820 0.01870 0.01920

• 0,01970 0.02020 0.02070 0.02120 0.02170 0.02220 0.02270 0.02320 0.02370 0,02420 0.02470 0.02700 0.03200 0.03700 FUNCTION TIME VALUE 1 .5884E 03 0.00030 l.5884E 03 0.00080 l.6884E 03 0.00130 1. SOBSE 03 0.00180 I. 6889E 03 0.00230 1.5894E 03 0.00280 1.6887E 03 0.00330 1 6892E 03 0.00380 I. 59OOE 03 0.00430 1 . 89OSE 03 0.00480 1.5900E 03 0.00630 I. 6906E 03 0.00580 1.6926E 03 0.00630 1.5969E 03 O.00680 1 .6008E 03 0.00730 1 .6075E 03 0.00780 l.6190E 03 0.00830 I. 6347E 03 0.00880 1.6699E 03 0.00930 1.69405: 03 0.00980 1.7404E 03 0.01030 I. 80b3E 03 0.01080 1.8887E 03 0.01 130 1.99445: 03 0.01180 2.1313E 03 0.01230 2 2996E 03 0.01280 2.6071E 03 0.01330 2.7686E 03 0.01380 3.0678E 03 0.01430 3 4130E 03 0.01480 3.8309E .03 0.01630 4.31716 03 0.01680 4.C774E 03 0.01630 6.3104E 03 0.01680 6.2467E 03 0.01 730 7.0688E 03 0,01780 7.99136 03 0.01830 9.0185E 03 0.018~0 1.0156E 04 0.01930 1 .1407E 04 0.01980 1.2780E 04 0.02030 1.42768 04 0.02080 1.5893E 04 0.02130 t. 7637E 04 0,02180 1.9603E 04 0.02230 Y.1496E 04 0.02280 3609E 04 0.02330 2.8834E 04 0.02380 2.61632: 04 0.02430 3.0692E 04 0.02480 4.2799E 04 0.02800 7.1685E 04 0.03300 b.2610E 04 0.03800 FUNCTION T 1.6884E 03 1.5884E 03 1.6884E 03 1. 6889E 03 . l.6887E 03 1,688lE 03 1. 6887E 03 1.6892E 03 1.6900E 03 1,6905E 03 1.5902E 03 I. 8908E 03 1. 6932E 03 1. 6976E 03 1.6016E 03 1.6093E 03 1.6216E 03 1.6401E 03 1.6665E 03 1.7021E 03 1.76368 03 1.8194E 03 1. 9077E 03 2.0206E 03 2.1628E 03 2.3377E 03 2.5529E 03 2.6139E 03 3.1238E 03 3.4914E 03 3.9234E 03 4.4227E 03 4.99896 03 5,655SE 03 6.4031E 03 7.2448E 03 0.1873E 03 9.23696 03 1.0397E 04 1.1673E 04 1 .3069E 04 1 .4688E 04 1.6230E 04 1.7S99E 04 , 1.9893E 04 2.1900E 04 2.4047E 04 2.6291E 04 2.8641E 04 3.1090E 04 4.8521E 04 7.6359E 04 1.0371E 04 TABLE 6.A-ll (SHEET 32 OF 32) 0.04200 6.tv6QE 04 0,04300 0.04700 3,692SE 04 0.04800 0.05200 2.0U22E 04 0.06300 0.05700 1.2877E 04 0.05800 0.06200 4,5991E 03

• 0:06300 0.06700 -8.3501E 03 ' 0.06800 0.07200 -1.7005E 04 0.07300 0.07700 -2.0820E 04 0.07800 0.08200 -2.0714E 04 0.08300 0.08700 -1.8165E 04 0.08800 0.09200 - 1 .4477E 04 0.09300 0.09700 -1.0753E 04 0.09800 '0.10200 -8.0822E 03 0.10300 0,10700 -6.8929E 03 0.10800 0.11200 -7.0285E 03 0.11300 0,11700 -7.6930E 03 0.1'1800 0.12200 -6.OG34E 03 0.12300 0.12700 -3.4112E 03 0.12800 0.13200 1.0507E 03 0.13300 0.13700 4.8382E 03 0.13800 0.14200 9.71368 03 0.14300 0.14700 1.Z792E 04 0.84800 0.15200 l.6072E 04 0.16300 0.15700 1.4664E 04 0.15800 0.16200 l.2407E 04 0, 16300 0.16700 1.0628E 04 0.16800 0.17200 1,0153E 04 0.17300 0.17700 0.0714E 03 0.17800 0.18200 8.9036E 03 0.18300 0,18700 t.0106E 04 0.18800 0.19200 1.3729E 04 0.19300 0.19700 1.8189E 04 0.8 9800 0.20400 2.3021E 04 0. PO600 0.21400 1.7258E 04 0.21600 0.22400 6.P070E 02 0.22600 0.23400 -3.1965E 03 0.23600 0.24400 -1.5C56E 03 0.24600 0.25400 8 .272QE 04 0.25600 0.26400 2.6143E 04 0.26600 0,27400 1.9070E 04 '0.27600 0.28400 6.4261E 03 0.28600 0.29400 -2.4680E 02 0.29600 0.30400 9.1571E 03 0.30600 0.31400 1.6992E 04 0.31600 0.32400 1.7937E 04 0.32600 0.33400 1.1278E 04 0,33600 0.34400 8.4614E 03 0.34600 0,30400 1.2178E 04 0.35600 6.36400 1 .6908E 04 0.36600 0.37400 1.6362E 04 0.37600 0.38400 l.2lOOE 04 0.38600 0.39400 1.0544E 04 0.39600 0.40400 1.2890E 04 0.40600 0.41400 1,3887E 04 0.41600 0.42400 1.3040E 04 0.42600 0,43400 1.2473E 04 0.43600 0.44400 1.3769E 04 0.44600 0.45400 1.4343E 04 0.48600 0.46400 1 .3708E 04 0.46600 0.47400 1.2008E 04 0.47600 0.48400 1.2432E 04 0.48600 0.49400 1.2343E 04 0.49600 TABLE 6 .A-12 (SHEET 1 OF 28) TIHE FUNCTION NUMBER TIME FORCE HISTORIES - FEEDWATER LINE BREAK = t 1) - fA5AlLC FUNCTION DESCRlPTlON

= ( FORCINQ FUNCTION AT NODE 30 00,023 - 0. NllfiRER OF ABSCISSAE 6771 FUNCTION SCALE FACTOR f 1.0000E 00) TlHE VALUE FUNCTION 0. 2.0000E-01 0.00050 2.OOOOE-01 0.00100 2.0000E-01 0.00150 2.0000E-01 0.00200 2.0000E -01 0.00250 2.0000E-01 0.00300 2.0000E-01 0.00350 2.0000E-01 0.00400 2.0000E-01 0.00450 2.0000E-01 0.00500 2.0000E-01 0.00550 2.0000E-01 0.00600 2.0000E-01 0.00G50 2.0000E-01 0.00700 2.0000E-01 0.00750 2.0000E-01 0.00~00 2.OOOOE-01 0.00850 2.0000E-01 0.00900 2.0000E-01 0.00950 2.OOOOE-01 0.01000 2.0000E-01 0.01050 2.0000E-01 0.01100 2.0000E-01 0.01150 -9.0000E-01 0.01100 2.0000E-01 0.01250 -9.0000E-01 0.01300 -1.3000E 00 0.01350 -1.9000E 00 0.01400 -1.9000E 00 0.01450 -1.9000E 00 0.01500 -4.4000E 00 0.01350 -5.5OOOE 00 0.01600 -7.5000E 00 0.01650 -1.0100E 01 0.01700 -1.3200E 01 0.01 750 - 1.7800E 01 0.01800 -2.24006 01 0.01850 -2.9500E 01 0.01900 -3.9700E 01 0.01950 -5.1400E 01 0.02000 -6.5200E 01 0.02050 -8.1900E 01 0.02100 -1.0330E 02 0.02150 -1.2710E 02 0.02200 -1.56GOE 02 0.02250 -1.92YOE 02 0.02300 -2.3460E 02 0.02350 -2.8310E 02 0.02400 -3.3870E 02 0.02450 -4.0440E 02 0.02500 -4.7960E 02 0.03000 -1.8888E 03 0.03500 -4.2740E 03 TINE VALUE FUNCTION 0.00020 2.OOOOE-01 0.00070 2.0000E-01 9.00120 2.0000E-01 0.00170 2.0000E-01 0.00220 2.000OE - 0 1 0.00270 2.0000E-01 0.00320 2.0000E-01 0.00370 2.OOOOE-01 0.00420 2.0000E-01 0.00470 2.0000E-01 0.00520 2.0000E-01 0.00570 2.0000E-01

' 0.00620 2.0000E-01 0.00670 2.0000E-01 0.00720 2.000OE-01 0.00770 2.0000E-01 0.00820 2.0000E-01 0.00870 2.0000E-01 0.00920 2.0000E-01 0.00970 2.0000E-01 0.01020 2.0000E-0t 0.01070 -9.0000E-01 0.01120 2.0000E-01 0.01 170 -9.0000E-01 0.01220 2.0000E-01 0.01270 -9.0000E-01 0.01320 -9.0000E-01 0.01370 -1.9000E 00 0.01420 -3.0000E 00 0.01470 -1.9000E 00 0.01520 -5.5000E 00 0.01570 -6.5000E 00 0.01620 -8.6OOOE 00 0.01670 -1.llOOE 01 0.01720 -1.4200E 01 0.01770 -1.88OOE 01 0.01820 -2.5900E 01 0.01870 -3.3000E 01 0.01920 -4.4300E 01 0.01970 -5.60OOE 01 0.02020 -7.1300E 01 0.02070 -9.0100E 01 0.02120 -1.1240E 02 0.02170 -1.3770E 02 0.02220 -1.6970E 02 0.02270 - 2.0780E 02 0.02320 -2.6280E 02 0,02370 -3.0490E 02 0.02420 -3.6410E 02 0.02470 -4.3420E 02 0.02700 -8.9090E 02 0.03200 -2.78blE 03 0.03700 -5.1219E 03 TIHE VALUE FUNCTION 0.00030 2.0000E-01 0.00080 2.0000E-01 0.00130 2.0000E-01 0.00180 2.0000E-01 0.00230 2.OOOOE-01 0.00280 2.0000E-01 0.00330 2.0000E-01 0.00380 2.0000E-01 0.00430 2.0000E-01 0.00480 2.0000E-01 0.00530 2.0000E-01 0.00580 2.0000E-01 0.00630 2.0000E-01 0.00680 2.0000E-01 0.00730 2.0000E-01 0.00780 2.0000E-01 0,00830 2.0000E-01 0.00880 2.0000E-01 0.00930 2.0000E-01 0.00980 2.0000E-01 0.01030 2.0000E-01 0.01080 2.0000E-01 0.01130 2.0000E-01 0.01180 -9.0000E-01 0.01230 2.0000E-01 0.01280 -9.OOOOE-01 0.01330 -1.9000E 00 0.01380 -1.9OOOE 00 0.01430 -3.0000E 00 0.01480 -3.0000E 00 0.01530 -5.5000E 00 0.01580 -6.5OOOE 00 0.01630 -8.6000E 00 0.01680 -1.2100E 01 0.01730 -1.5700E 01 0.01780 -2.0300E 01 0.01830 -2.7OOOE 01 0.01880 -3.5100E 01 0.01930 -4.6800E 01 0.01980 -6.85OOE 01 0.02030 -7.4800E 01 0.02080 -9.3600E 01 0.02130 -1. t700E 02 0.02180 -1.4380E 02 0.02230 -1.7730E 02 0.02280 -2.1680E 02 0.02330 -2.6240E 02 0.02380 -3.1660E 02 0.02430 -3.7770E 02 0.02480 -4.4840E 02 0.02800 -1,1706E 03 0.03300 -3.2822E 03 0.03800 -6.4261E 03 TIHE VALUE FUNCTION 0.00040 2.0000E-01 0.00090 2.0000E-01 0.00140 2.0000E-01 0.00190 2.0000E-01 0.00240 2.0000E-01 0.00290 2.0000E-01 0.00340 2.OOOOE-01 0.00390 2.0000E-01 0.00440 2.OOOOE-01 0.00490 2.0000E-01 0.00540 2.0000E 0.00590 2.0000E-01, 0.00640 2.0000E-01 0.00690 2.0000E-01 0.00740 2.000OE-01 0.00790 2.0000E-01 0.00640 2.0000E-01 0.00890 2.OOOOE-01 0.00940 2.0000E-01 0.00990 -9.0000E-01 0.01040 2:OOOOE-01 0.01090 2.0000E-01 0.01 140 -9.0000E-01 0.01190 2.0000E-01 0.01240 -9.0000E-01 0.01290 -1.3000E 00 0.01340 -1.9000E 00 0.01390 -1,900OE 00 0.01440 ,-1.9000E 00 0.01490 -4.0000E 00 0.01540 -5.5000E 00 0,01590 -7.6000E 00 0.01640 -9.6000E 00 0.01690 -1.2100E 01 0.01740 -1.6700E 01 0.01790 -2.1300E 01 0.01840 -2.800OE 01 0.01890 -3.6600E 01 '0.01940 -4.89OOE 01 0.01990 -6.1600E 01 0.02040 -7.8400E 01 0.02090 -9.8200E 01 0.02140 -1.2210E 02 0.02190 -1.505OE 02 0.02240 -1 .@WOE 02 0.02290 -2.2600E 02 0.02340 -2.7240E 02 0.02390 -3.2660E 02 0.02440 -3.9020E 02 0.02490 -4.6300E 02 0.02900 -1.504lE 03 0.03400 -3.78366 03 0.03900 -5.61796 03 TABLE 6. A- (SHEET 2 OF TABLE 6.A-12 (SHEET 3 OF 28) T IHE FUNCl ION NUtiDER t 2) FUNCTION DESCRIPTION

= f FORCING FUNCTION AT NODE 31 00~021 0. 0:3 NUHUCK or ABSCISSAE

= ( 577) FUI.!Cl ION SCALE FACTOR f 1.0000E. 00) TIHEVALUE FUNCTION 0.00010 3.0000E-01 0.00060 3.0000E-01 0.00110 3.0000E-01 0.00lG0 3.0000E-01 0.00210 3.0000E-01 0.00260 3.0000E-01 0.00310 3.0000E-01 0.00360 3.0000E-01 0.00410 3.0000E-01 0.00460 3.0000E-01 0.00510 3.0000E-01 0.00560 3.0000E-01 0.00610 -9.0000E-01 0.00660 3.0000E -01 0.00710 3.0000E-01 0.00760 3.0000E-01 0.00810 -9.OOOOE-01 0.00860 - 1.4000E 00 0.00910 -2.1000E 00 . 0.00960 -9.0000E-01 0.01010 -2.1000E 00 0.01060 -4.4000E 00 0.01110 -4.4000E 00 0.01160 -7.3000E 00 0.01210 -1.0900E 01 0.01260 -1.4900E 01 0.01310 -1.Q70OE 01 0.01360 -2.6100E 01 0.01410 -3.4900E 01 0.01460 -4.7700E 01 0.01510 -6.3400E 01 0.01560 -8.2700E 01 0.01610 -1.0720E 02 0.01660 -1.3790E 02 0.01710 -1.758OE 02 0.01760 -2.2060E 02 0.01810 -2.7390E 02 0.01860 -3.3700E 02 0.01910 -4.128OE 02 0.01960 -8.0000E 02 0.02010 -6.0270E 02 0.02080 -7.1970E 02 0.02110 -8.5160E 02 0.02160 -1.0018E 03 0.02210 -1.1678E 03 0.02260 -1.3551E 03 0.02310 -1.5598E 03 0.02360 -1.783BE 03 0.02410 -2.0290E 03 0.02460 -2.2922E 03 0.02600 -3.1309E 03 0.03100 -6.6967E 03 0.03600 -8.8111E 03 7 Jt%E VALUE FUNCTION 0.00020 3.0000E-01 0.00070 3.OOOOE-01 0.00120 3.0000E-01 0.00170 3.0000E-01 0.00220 3.0000E-01 0.00270 3.0000E-01 0.00320 3.0000E-01 0.00370 3.0000E-01 0.00420 3.0000E-01 0.00470 3.0000E-01 0.00520 3.0000E-01 0.00570 3.0000E-01 0.00620 -9.0000E-01 0.00670 3.0000E-01 0.00720 3.0000E-01 0.00770 -0.0000E-01 0.00820 -9.0000E-01 0.00870 -9.0000E-01 0.00920 -2.1000E 00 0.00970 -9.0000E-01 0.01020 -2.1000E 00 0.01070 -3.8000E 00 0.01120 -6.6OOOE 00 0.01170 -6.8000E 00 0.01220 -1.1400E 01 0.01270 -1.4POOE 01 0.01320 -2.1300E 01 0.01370 -2.8900E 01 0.01420 -3.7700E 01 0.01470 -S.O600E 01 0.01520 -6.7000E 01 0.01570 -8.E700E 01 0.01620 -1.129OE 02 0.01670 -1.4390E 02 0.01720 -1.6270E 02 0.01770 -2.29lOE 02 0.01820 -2.8650E 02 0.01870 -3.5200E 02 0.01920 -4.2940E 02 Q.01970 -6.l900E 02 0.02020 -6.2460E 02 0.02070 -7.4500E 02 0.02120 -8.8040E 02 0.02170 -1.0328E 03 0.02220 -1.2046E 03 0.02270 -1.3940E 03 0.02320 -1.604OE 03 0.02370 -1.8321E 03 0.02420 -2.0805E 03 0.02470 -2.3472E 03 0.02700 -3.80236 03 0.03200 -7.3394E 03 0.03700 -8.842OE 03 TIME VALUE FUNCTION 0.00030 3.0000E-01 0.00080 3.0000E-01 0.00130 3.0000E-01 0.00180 3.0000E-01 0.00230 3.0000E-01 0.00280 3.0000E-01 0.00330 3.0000E-01

'0.00380 3.0000E-01 0.00430 3.0000E-01 0.00480 3.0000E-01 0.00530 3.0000E-01 0.00580 -9,000OE-01 0.00630 -9.0000E-01 0.00680 3.0000E-01 0.00730 3.OOOOE-01 0.00760 -9.OOOOE-01 0.00830 -9.OOOOE-01 0,00880 -Q.OOOOE-01 0.00930 -2.1000E 00 0.00980 -2.1000E 00 0.01030 -3.3000E 00 0.01080 -3.POOOE 00 0.01130 -6.1000E 00 0.01160 -8.SOOOE 00 0.01230 -1.0900E 01 0.01280 -1.6100E 01 0.01330 -2.2500E 01 0.01380 -3.0100E 01 0.01430 -4.0100E 01 0.01480 -5.4100E 01 0.01530 -7.1000E 01 0.01580 -9.2000E 01 0.01630 -1.l940E 02 0.01680 -1.52OOE 02 0.01730 -1.915OE 02 0.01780 -2,3960E 02 0.01830 -2.9820E 02 0.01880 -3.6650E 02 0.01930 -4.46ROE 02 0.01980 -5.3930E 02 0.02030 -6.4GLOE 02 0.02080 -7.6960E 02 0.02130 -9.O860E 02 0.02180 -1.0655E 03 0.02230 -1.2408E 03 0.02280 -1.4348E 03 0.02330 -1.64696 03 0.02380 - 1 .8796E 03 0.02430 -2.1326E 03 0.02480 -2.4037E 03 0.02800 -4.5195E 03 0.03300 -7.89lOE 03 0.03800 -0.7389E 03 TINE VALUE FUNCTION 0.00040 3.0000E-01 0.00090 3.0000E-01 0.00140 3.0000E-01 0.00190 3.0000E-01 0.00240 3.0000E-01 0.00290 3.0000E-01 0.00340 3.0000E-01 0.00390 3.0000E-01 0.00440 3.0000E-01 0.00490 3.0000E-01 0.00540 3.0000E-01 0.00590 -S.OOOOE-01 0.00640 3.0000E-01 0.00690 3.0000E-01 0.00740 3.0000E-01 0.00790 -9.OOOOE-01 0.00840 -1.4000E 00 0.00890 -9.0000E-01 0.00940 -2.1000E 00 0.00990 -2.6000E 00 0.01040 -3.3000E 00 0.01090 -4.QOOQE 00 0.01140 -7.3000E 00 0.01190 -8.5OOOE 00 0.01240 -1.2100E 01 0.01290 -1.610OE 01 0.01340 -2.3200E 01 0.01390 -3.2000E 01 0.01440 -4.2500E 01 0.01490 -5.6500E 91 0.01540 -7.5100E 01 0.01690 -9.7200E 01 0.01640 -1.25306 02 0.01690 -1.6010E 02 0.01740 -2.0080E 02 0.01790 -2.5120E 02 0,01840 -3.1080E 02 0.01890 -3.82OOE 02 0.01940 -4.6470E 02 0.01990 -8.6000E 02 0.02040 -6.7020E 02 0.02090 -7.9580E 02 0.02140 -9.3910E 02 0.02190 -1.0989E 03 0.02240 -%;2781E 03 0.02290 -8.4744E 03 0.02340 -1.6927E 03 0.02390 -1.S283E 03 0.02440 -2.1841E 03 0.02490 -2.4581E 03 0.02900 -5.2585E 03 0.03400 -8.3296E 03 0.03900 -0.5123E 03 TABLE 6. A- (SHEET 4 OF b I t-' 0.04000 -8.1784E 03 0.04500 -5.3401E 03 0.05000 -9.9890E 02 0.05500 4 2838E 03 0.06000 8.2254E 03 0.06500 8.1167E 03 0.07000 4.3684E 03 0.07500 -8.2100E 02 0.08000 -6.3315E 03 0.08400 -1.1265E 04 0.09000 -1.4117E 04 0.09500 -1.3980E 04 0.10000 -1.1405E 04 0.10500 -7.1867E 03 0. 11000 -2.2434E 03 0.11500 2.0010E 03 0.12000 4.P732E 03 0.12500 5.2893E 03 0.13000 4.4644E 03 0.13500 2.6250E 03 0.141300 3.3870E 02 0.14500 -1.8931E 03 0.15000 -3.6677E 03 0.15500 -4.7200E 03 0.16000 -4.9917E 03 0.16500 -4.6347E 03 0.17000 -3.9983E 03 0.17500 -3.0548E 03 0.18000 -2.2115E 03 0.18500 -1.R50GE 03 0.19000 -6.2280E 02 0.19500 -3.6410E 02 0.20C00 -1.129OE 02 0.21000 -2.7150E 02 0.22000 -1.2274E 03 0.23000 -2.2445E 03 0.24000 -2.595GE 03 0.20000 -2.0283E 03 O.PC.000 -1.29GCC 03 0.27000 -7.0010E 02 0.260C10 -2 CGOOE 02 0.29000 2.2970E 02 0.30600 5.1260E 02 0.31C100 -1 O42OE 02 0.321'100 - 1 ,7883E 03 0.33000 -2.3388E 03 0.34000 -1.5740E 03 0.35000 - 3.7200E 02 0.36000 8.1560E 02 0.37000 1.8169E 03 0.38000 2.2572E 03 0.39000 1.7970E 03 ' 0.40000 7.0530E 02 0.41000 -2.5490E 02 0.42000 -6.1260E 02 0.43000 -5.1060E 02 0.44000 -2.3G60E 02 0.45000 -3.9600E 01 0.46000 2.8090E 02 0.47000 7.4340E 02 0.48000 1.0947E 03 0.49000 1.2585E 03 TABLE TIME FUNCTION NUEIBER * ( 31 ( SHEET FUNCTI ON 3ESCRI PTI ON 8 ( FORCING FUNCTION AT NODE 32 00>02S 0. 0.2 f NUNPER OF ABSCISSAE 8.1 577) FUNCTlON SCALE FACTOR 8 t 1.0000E 001 TlME VALUE FUNCTlOM 0. 2.0000E-01 0.00050 2.0000E-01 0.00100 2.0000E-01 0.00150 2.0000E-01 0.00200 Z.OOOOE-01 0.00250 2.0000E-01.

0.00300 2.0000E-01 0.00350 2.0000E-01 0.00400 -5.0000E-01 0.00450 2.OOOOE-01 0.00500 2.0000E-01 0.00580 2.0000E-01 0.00600 -5.0000E-01 0.00650 2.0000E-01 0.00700 2.0000E-01 0.00750 2.0000E-01 0.00Ck00 -5.000OE-01 0.00350 -8.0000E-01 0.00900 -5.0000E-01 0.00950 -5.OOOOE-01 0.01000 -1.4000E 00 0.01050 -1.8000E 00 0.01100 -2.SOOOE 00 0.01150 -4.1000E 00 0.01200 -5.4000E 00 0.01230 -7.4000E 00 0,01300 -9.6000E 00 o.r~i350 -1.3900~ 01 0.~1400 -1.8100C 01 0.01450 -2.4600E 01 0.01500 -5.3100E 01 0.01550 -4.r1100E 01 0.OlGOO -5.6900E 01 0.01650 -7.3iOOE 01 0.01 700 *8,3500E 01 0.0)730 -1.1740E 02 0.01800 -1.4620E 02 0.01850 -1.8050E 02. 0.01900 -2.2110E 02 0.01950 -2.6890E 02 0.02000 -3.2390E.

02 O.OP.j50 -3.8730E 02 0.02100 -4.5950E 02 0.02150 -5.4070E 02 0.02200 -6.3210E 02 0. Q.2250 -7,3390E 02 0.02300 -8.4590E 02 0.02350 -9.6930E 02 0.02400 -1.1034E 03 0.02450 -1.2480E 03 0.02500 -1.4033E 03 0.03090 -3.3433E 03 0.03500 -4.8182E 03 TIME VALUE FUNCTION 0.00010 2.0000E-01 0.00060 2.0000E-01 0.00110 2.0000E-01 0.00160 2.0000E-01 0.00210 2.0000E-01 0.00260 2.0000E-01 0.00310 2.0000E-01 0.00360 2.0000E-01 0.00410 2.0000E-01 0.00460 2.0000E-01 0.00510 2.OOOOE-01 0.00560 2.0000E-01 0.00610 -6.0000E-01 0.00660 2.OOOOE-01 0.00710 2.0000E-01 0.00760 2.00OOE-01 0.00810 -5.0000E-01 0.00860 -8.OOOOE-01 0.00910 -1.2000E 00 0.00960 -5.0000E-01 0.01010 -1.2000E 00 O.OlOG0 -2.5000E 00 0.01110 -2.5000E 00 0.01160 -4.1000E 00 0.01210 -6.1000E 00 0.01260 -8.3000E 00 0.01310 -1.1000E 01 0,01360 -1.45OOE 01 0.01410 -1.9500E 01 0.01460 -2.6600E 01 0.01510 -3.5400E 01 0.01560 -4.6100E 01 0.01610 -5.9800E 01 0.01660 -7.6900E 01 0.01710 -9.80OOE 01 0.01760 -1.2300E a2 0.01810 -1.5270E 02 0.01860 -1.879OE 02 0.01910 -2.3020E 02 0.01960 -2.78BOE 02 0,02010 -3.3610E 02 0.02060 -4.0140E 02 0.02110 -4.7490E 02 0.02160 -5.5870E 02 0.02210 -6.5130E 02 0.02260 -7.6570E 02 0.02310 -8.6990E 02 0.02360 -9.9480E 02 0.02410 -1.1316E 03 0.02460 -1.2783E 03 0.02600 -1.7460E 03 0.03100 -3.7346E 03 0.03600 -4.9137E 03 TIflEVALUE FUNCTION 0.00020 2.0000E-01 0.00070 2.0000E-01 0.00120 2.0000E-01 0.00170 2.00GOE-01 0.00220 2.0000E-01 0.00P70 2.OOOOE-01 0.00320 2.0000E-01 0.00370 2.0000E-01 0.00420 2.0000E-01 0.00470 2.0000E-01 0.00520 2.0000E-01 0.00570 2.0000~-0i 0,00620 -5.0000E-01 0.00670 2.0000E-01 0.00720 2.0000E-01 0.00770 -5.0000E-01 0,00820 -5.0000E-01 0.00870 -5.0000E-OI 0:00920 -1.2000E 00 0.00970 -5.OOOOE-01 0.01020 -1.2000E 00 0.01070 -2.1000E 00 a.01120 -3.1000E 00 0.01170 -3.e000E 00 0.01220 -6.3000E 00 0.01270 -8.3000E 00 0.01320 -1.19OOE 01 0.01370 -1.6100E 01 0.01420 -2.1000E 01 0.01470 -2.8200E 01 0.01620 -3.7400E 01 0.01570 -4.8400E 01 0.01620 -6.3000E 01 0.01670 -8.0200E 01 0.01720 -1.0140E 02 0.01770 -1.2780E 02 0.01820 -1.e980E 02 0.01870 -1.Q630E 02 0.01920 -2.39SOE 02 0.01970 -2.6950E 02 0.02020 -3.4830E 02 0.02070 -4.155OE 02 0.02120 -4.9100E 02 0.02170 -5.7600E 02 0.02220 -6.7180E 02 0.02270 -7.7740E 02 0.02320 -8.94SOE 02 0.02370 -1.0217E 03 0.02420 -1.1603E 03 0.02470 -1.3069E 03 0.02700 -2.1204E 03 0,03200 -4.0930E 03 0.03700 -4.9309E 03 TIME VALUE FUNCTION 0.00030 2.0000E-01 0.00080 2.0000E -01 0.00130 2.OOOOE-01 0.00180 2.0000E-01 0.00230 2.0000E-01 0.00280 2.0000E-01 0.00330 2.0000E-01 0.00380 2.0000E-01 0.00430 2.OOOOE-01 0.00480 2.0000E-01 0.00530 2.OOOOE-01 0.00580 -6.0000E-01 0.00630 -5.0000E-01 0.00680 2.0000E-01 0.00730 2.0000E - D 1 0.00780 -5.0000E-01 0.00830 -5.OOOOE-01 0.00880 -5.0000E-01 0.00930 -1.2000E 00 0.00980 -1,200OE 00 0.01030 -1.8000E 00 0.01080 -2.1000E 00 0.0113Q -3.4000E 00 0.01180 -4.7000E 00 0.01230 -6.1000E 00 0.01280 -9.0000E 00 0.01330 -1.2600E 01 0.01380 -1.6800E 01 0.01430 -2.240OE 01 0.01480 -3.0200E 01 0.01530 -3.9600E 01 0.01580 -6.1300E 01 0.01630 -6.6600E 01 0.01680 -8.4700E 01 0.01730 -1.0680E 02 0.01780 -1.3360E 02 0.01830 -1.6630E 02 0.01880 -2.0410E 02 0.01930 -2.4920E 02 0.01980 -3.0070E 02 0.02030 -3.6060E 02 0.02080 -4.2930E 02 0.02130 -5.0670E 02 0.02180 -5.9420E 02 0.02230 -6.920OE 02 0.02280 -8.OOlOE 02 0.02330 -9. I840E 02 0.02380 -1.0482E 03 0.02430 -1.1893E 03 0.02480 -1.340SE 03 0.02800 -2.5204E 03 0.03300 -4.4006E 03 0.03800 -4.8734E 03 TIME VALUE FUNCTION 0.00040 2.OOOOE-01 0.00090 2.0000E-01 0.00140 2.0000E-01 0.00190 2.0000E-01 0.00240 2.0000E-01 0.00290 2.0000E-01 0.00340 2.0000E-01 0.00390 2.0000E-01 0.00440 2.OOOOE-01 0.00490 2.0000E-01 0.00540 2.0000E-01 , 0.00590 -5.0000E-01 0.00640 2.000OE-01.

0.00690 2.0000E-01 0.00740 2.0000E-01 0.00790 -5.0000E-01 0.00840 -0.0000E-01 0.00890 -5.0000E-01 0.00940 -1.2000E 00 0.00990 -1.4000E 00 0.01040 -1.8000E 00 0.01090 -2.8000E 00 0.01 140 -4,1000E 00 0.01190 -4.7000E 00 0.01240 -6.7000E 00 0.01290 -9.0000E 00 0.01340 -1.2900E 01 0.01390 -1.7900E 01 0.01440 -2.3700E 01 0.01490 -3.1500E 01 0.01540 -4.19OOE 01 0.01590 -5.4200E 01 0.01640 -6.99OOE 01 0.01690 -6.B300E 01 0.01740 -1.1200E 02 0.01790 -1.4010E 02 0.01840 -1.7330E 02 0.01890 -2.1300E 02 0.01940 -2.6920E 02 0.01990'-3.1230E 02 0.02040 -5.7370E 02 0.02090 -4.4380E 02 0.02140 -5.2370E 02 0.02190 -6.1280E 02 0.02240 -7.1270E 02 0.02290 -8.2220E 02 0.02340 -9.4400E 02 0.02390 -1.0753E 03 0.02440 -1.2160E 03 0.02490 -1.3706E 03 0.02900 -2. 9325E 03 0.03400 -4.6452E 03 0.03900 -4.7471E 03 0.04000 -4.5609E 03 0.04500 -2,9780E 03 0.05000 -5.5710E 02 0.05500 2.3889E 03 0.06000 4.5871E 03 O.CG500 4.6264E 03 0.07000 2.4361E 03 0.07500 -4.5780E 02 0.08000 -3.5309E 03 o.ua~oo -6.2622~ 03 0.09000 -7.8727E 03 0.09500 -7.7961E 03 0.10000 -6.3604E 03 0 10500 -4.0078E 03 0.11000 -1.2511E 03 0.11500 1.1159E 03 0.12000 2.5503E 03 0.12!~00 2.9437E 03 0.13000 2.4897E 03 0.13500 1.4639E 03 0.14000 1.889OE 02 0.14500 -1.0557E 03 0.15000 -2.0454E 03 0.15500 -2.6322E 03 0.16000 -2.783?E 03 0.16500 -2.5846E 03 0.17000 -2.1796E 03 @I 0.17500 - 1.7036E 03 t' 0 1 bOOO - 1.2333E 03 bl 0.18500 -8.0900E 02 0.19000 -4.5880E 02 a\ 0,19500 -2.0310E 02 b 0.20000 -6.2900E 01 0.21000 -1.5140E 02 I I-' 0.22000 -6.8450E 02 N 0.23000 -1.2517E 03 0.24000 -1.4475E 03 0.25000 -1.1311E 03 0 260C 0 -7.1750E 02 0.27000 -3.9040E 02 0.28000 -1.4880E 02 0 29000 1.2810E 02 0.30000 2.8580E 02 0.31000 -5.8100E 01 0.32000 -9.9730E 02 0.33000 -1.3043E 03 0.34000 -8.7780E 02 0.35000 -9.075OE 02 0.36000 4.5480E 02 0.37000 1.0132E 03 0.30000 1.2568E 03 0.39000 1.0022E 03 0.40000 3.9330E 02 0.41000 -1.4210E 02 0.42000 -3.4160E 02 0.43000 -2.8480E 02 0.44000 -1.3190E 02 0.45000 -2.2100E 01 0.46000 1.5660E 02 0.47000 4.1460E 02 0.48000 6.1050E 02 0.49500 7.0180E 02 I-' Ui CO IP TABLE 6.A-12 (SHEET 6 OF 28) 0.04200 -4.0421E 03 0.04700 -2.1022E 03

• 0.05200 6.0470E 02 0.05700 3.4692E 03 0.06200 4. e649~ 03 0.06700 3.8604E 03 0.07200 1.3243E 03 0.07700 -1.6835E 03 0.08200 -4.7142E 03 0.08700 -7.1018E 03 0.09200 -0.0302E 03 0.09700 -7.3619E 03 0.10200 -5.3072E 03 0.10700 -2.9071E 03 0.11200 -2.1720E 02 0.11700'1.8169E 03 0.12200 2.8297E 03 0.12700 2.8550E 03 0.13200 2.1295E 03 0.13700 9.6660E 02 0.14200 -3.P740E 02 0.14700 -1.4933E 03 0.15200 -2.3325E 03 0.15700 -2.7417E 03 0.16200 -2.7385E 03 0.18700 -2.4396E 03 0.17200 -1.9916E 03 0.17700 -1.512OE 03 0.18200 -).0573E 03 0.18700 -6.6840E 02 0.19200 -3.4290E 02 0.19700 -1.3270E 02 0.20400 -3.750OE 01 0.21400 -3.2500E 02 0.22400 -9.359OE 02 0.23400 -1.3974E 03 0.24400 -1.3693E 03 0.26400 -9.6800E 02 0.26400 -5.6860E 02 0.27400 -2.95POE 02 0.20400 -4.9100E 01 0.29400 2.3530E 02 0.30400 2.4990E 02 0.31400 -4.4110E 02 0.32400 -1.9297E 03 0.33400 -1.1913E 03 0.34400 -6.1620E 02 0.35400 6.2700E 01 '0.36400 6.9970E 02 0.37400 1.1637E 03 0.38400 1.218OE 03 0.39400 7.7930E 02 0.40400 1.4640E 02 0.41400 -2.581OE 02 0.42400 -3.4530E 02 0.43400 -2.lS9OE 02 0.44400 -8.6200E 01 0.45400 3.5300E 01 0.46400 2.5790E 02 0.47400 6.0620E 02 0.48400 6.5620E 02 0.49400 7.2780E 02 TABLE 6.A-12 (SHEET 7. OF 28) TIME FUNCTION NUMBER s ( 4) FUNCTION DESCRIPTION ( FORCING FUNCTION AT NODE 34 00>02S 0. NUMBER OF ABSCISSAE s t 1771 FUNCTION SCALE FACTOR

= ( 1.0000E 001 TlME VALUE FUNCTION 0. -1.4370E 02 0.00050 -1.4430E 02 0.00100 -1.4430E 02 0.00150 -1.4430E 02 0.00200 -1.443OE 02 0.00250 -1.4490E 02 0.00300 -1.4490E 02 0.00350 -1.4510E 02 0.00400 -1.4490E 02 0.00450 -1.4570E 02 0.00500 -1.4550E 02 0.00550 - 1 .4670E 02 0.00600 -1.4730E 02 0.00650 -1.4790E 02 0.00700 -1.4930E 02 0.00750 -1.5230E 02 0.00800 -1.5480E 02 0.00850 - 1 .5I 8OE 02 td 0.00900 -1,661OE 02 P 0.00950 -1.7540E 02 M 0.01000 -1.8850E 02 0.01050 -2.0610E 02 Cn 0.01100 -2.288OE 02 0.01150 -2.5880E 02 Y 0.01200 -2.S720E 02 0.01250 -3.4620E 02 I-' 0.01300 -4.0700E 02 N 0.01350 -4.8050E 02 0.01400 -5.6990E 02 0.01450 -6.7710E 02 0.01500 -8.0270E 02 0.01550 -9.5030E 02 0.01600 -1.8108E 03 0.01660 -1.3098E 03 0.017.00 -1.5261E 03 0.01750 -1.7679E 03 0.01800 -2.0341E 03 0.01850 -2.3264E 03 0.01900 -2.6428E 03 0.01950 -2.9839E 03 0.02000 - 3.3482E 03 0.02050 - 3.7329E. 03 0.02100 -4.1366E 03 0.02150 -4.5582E 03 0.02200 -4.9924E 03 0.02250 -5.4372E 03 0.02300 -5.8884E 03 0.02350 -6.3436E 03 0.02400 -6.7975E 03 0.02450 -7.2464E 03 0.02500 -7.6853E 03 0.03000 -1.0716E 04 0.03500 -1.0336E 04 TlflE VALUE FUNCTf ON 0.00010 -1.4370E 02 0.00060 -1.4430E 02 0.00110 -1.4430E 02 0.00160 -1.4430E 02 0.00210 -1.4430E 02 0.00260 -1.4430E 02 0.00310 -1.4430E 02 0.00360 -1.4490E 02 0.00410 -1.4550E 02 0.00460 -1.4550E 02 0.00510 -1.4610E 02 0.00560 -1.4690E 02 0.00610 -1.475OE 02 0.00660 -1.4850E 02 0.00710 -1.4990E 02 0.00760 -1.5230E 02 0.00810 -1.5560E 02 0.00860 -1.6100E 02 0.00910 -1.6850E 02 0.00960 -1.7840E 02 0.01010 -1.918OE 02 0.01060 -2.1000E 02 0.01110 -2.3380E 02 0.01160 -2.6630E 02 0.01110 -3.0670E 02 0.01260 -3.5720E 02 0.01310 -4.2000E 02 0.01360 -4.9710E 02 0.01410 -6.9030E 02 0.01460 -7.0100E 02 0.01510 -8.3100E 02 0.01560 -9.8210E 02 0.01610 -1.1546E 03 0.01660 -1.3514E 03 0.01710 -1.5723E 03 ' 0.01760 -1.8187E 03 . 0.01810 -2.0904E 03 0.01860 -2.3872E 03 0.01910 -2.7091E 03 0.01960 -3.0555E 03 0.02010 -3.4234E 03 0.02060 -3.8127g 03 0.02110 -4.2203E 03 0.02160 -4.64376 03 0.02210 -5.0798E 03 0.02260 -5.5268E 03 0.02310 -5.9800.E 03 0.02360 -6.43476 03 0.02410 -6.8878E 03 0.02460 -7.3344E 03 0.02600 -8.5218E 03 0.03100 -1.O884E 04 0.03600 -1.0058E 04 TlME VALUE FUNCTION 0.00020 -1.4370E 02 0.00070 -1.4430E 02 0.00120 -1.4430E 02 0.00170 -1.4430E 02 0.00220 -1.4430E 02 0.00270 -1.4450E 02 0.00320 -1.44906 02 0.00370 -1.44906 02 0.00420 -1.4550E 02 0.00470 -1.455OE 02 0.00520 -1.4610E 02 0.00570 -1.4670E 02 0.00620 -1.4790E 02 0.00670 -1.4870E 02 0.00720 - 1 . 505OE 02 0.00770 -1.5290E 02 0.00820 -1.5680E 02 0.00870 -1.6220E 02 0.00920 -1.6990E 02 0.00970 -1.8020E 02 0.01020 -1.9480E 02 0.01070 -2.1420E 02 0.01120 -2.4040E 02 0.01170 -2.7370E 02 0.01220 -3.1560E 02 0.01270 -3.687OE 02 0.01320 -4.3400E 02 0.01370 -6.148OE 02 0.01420 -6.1090E 02 0.01470 -7.25ZOE 02 0.01520 -8.5930E 02 0.01570 -1.0141E 03 0.01620 -1.1922E 03 0.01670 -1.3934E 03 0.01720 -1.6197E 03 0.01770 -1.8713E 03 0.01820 -2.1478E 03 0.01870 -2.4506E 03 0.01920 -2.7770E 03 0.01970 -3.1271E 03 0.02020 -3.49956 03 0.02070 -3.8924E 03 0.02120 -4.3034E 03 0.02170 -4.7302E 03 0.02220 -5.169OE 03 0.02270 -5.6170E 03 0.02320 -6.0707E 03 0.02370 -6.5256E 03 0.02420 -6.9775E 03 0.02470 -7.4227E 03 0.02700 -9.2684E 03 0.03200 -1.0898E 04 0.03700 -9.7889E 03 TJME VALUE FUNCTION 0.00030 -1.4370E 02 0.00080 -1.4430E 02 0.00130 -1.4370E 02 0.00180 -1.4430E 02 0.00230 -1.4430E 02 0.00280 -1.4430E 02 0.00330 -1.4490E 02 0.00380 -1.455OE 02 0.00430 -1.4550E 02 0.00480 -1.4630E 02 0.00530 -1,4670E 02 0.00580 -1.4670E 02 0.00630 -1.4730E 02 0.00680 -1.4850E 02 0.00730 -1.509OE 02 0.00780 -1.535OE 02 0.00830 -1.5720E 02 0.00880 -1.6280E 02 0.00930 -1.7090E 02 0.00980 -1.8310E 02 0.01030 -1.9860E 02 0.01080 -2.1950E 02 0.01130 -2.4600E 02 0.01180 -2.8110E 02 0.01230 -3.'2540~

02 0.01280 -3.809OE 02 0.01330 -4.4910E 02 0.01380 -5.3230E 02 0.01430 -6.3220E 02 0.01480 -7.1060E 02 0.01630 -8.8880E 02 0.01580 -1.0479E 03 0.01630 -1.2306E 03 0.01680 -1.4364E 03 0.01730 -1.6674E 03 0.01780 -1.9244E 03 0.01830 -2.2060E 03 0.01880 -2.5135E 03 0.01930 -2.6450E 03 0.01980 -3.1996E 03 0.02030 -3.5768E 03 0.02080 -3.9735E 03 0.02130 -4.3075E 03 0.02180 -4.8166E 03 0.02230 -5.2584E 03 0.02280 -5.7075E 03 0.02330 -6.1620E 03 0.02380 -6.6166E 03 0.02430 -7.0672E 03 0.024b0 -7.8107E 03 0.02800 -9.8974E 03 0.03300 -1.0789E 04 0.03800 -9.6540E 03 1 IflE VALUE FUNCTl ON 0.00040 -1.4370E 02 0.00090 -1.4430E 02 0.00140 -1.4370E 02 0.00190 -1.4430E 02 0.00240 -1.4430E 02 0.00290 -1.4490E 02 0.00340 -1.4490E 02 0.00390 -1.4490E 02 0.00440 -1.4550E 02 0.00490 -1.4610E 02 0.00540 -1.4610E 02 , 0.00590 -1.4690E 02 0.00640 -1.4810E 02 0.00690 -1.4930E 02 0.00740 -1.5llOE 02 0.00790 -1.5420E 02 0.00840 -1.5860E 02 0.00890 -1.649OE 02 0.00940 -1.7330E 02 0.00990 -1.8610E 02 0.01040 -2.OI9OE 02 0.01090 -2.2400E 02 0.01140 -2.5250E 02 0.01190 -2.8890E 02 0.01240 -3.3550E 02 0.01290 -3.9400E 02 0.01340 -4.6420E 02 0.01390 -5.5070E 02 0.01440 -6.0440E 02 0.01490 -7.7650E 02 0.01540 -9.lQlOE 02 0.01590 -1.0826E 03 0.01640 -1.2705E 03 0.01690 -1.4809E 03 0.01740 -1.7173E 03 0.01790 -1.9787E 03 0.01840 -2.2658E 03 0.01890 -2.5774E 03 0.01940 -2.9143E 03 0.01990 -3.2732E 03 0.02040 -3.6546E 03 0.02090 -4.0547E 03 0.02140 -4.4723E 03 0.02190 -4.9043E 03 0.02240 -6.3475E 03 0.02290 -6.7986E 03 0.02340 -6.2523E 03 0.02390 -6.7069E 03 0.02440 -7.16676 03 0.02490 -7.5978E 03 0.02900 -1.0385E 04 0.03400 -1.O591E 04 0.05900 -9.3748E 03 TABLE 6.A-12 (SHEET 8 OF 28) 0.04400 -9.2314E 03 0.04900 -3.19736 03 0.05400 -7.8699E 03 0.05900 -5.3537E 03 0.06400 -3.06S6E 03 0.06900 -3.OIOZE 03 0.07400 -6.2368E 03 0.07900 -1.1339E 04 0.08400 -1.6385E 04 0.08900 -1.9JDOE 04 0.09400 -2.0151E 04 0.09900 -1.OW6E 04 0.10400 -1.68336 04 0.10900 -1.4847E 04 0.11400 -1.293JE 04 0.11900 -1.IW8E 04 0.12400 -9.Si86E 03 0.12900 -8.4999E 03 0.13400 -7.9734E 03. 0.13900 -7.9285E 03 0.14400 -8.4663E 03 . 0.14900 -9.324OE 03 0.15400 -1.0313E 04 0.15900 -1.1241E 04 0.16400 -1.1893E 04 0.16900 -1.2Z52E 04 0.17400 -1.223OE 04 0.17900 -1.1&16E 04 0.18400 -1.1306E 04 0.18900 - 1. OBZOE 04 0.19400 -1.OWOE 04 0.19900 -1.6413E 04 0.20800 - 1 .1137E 04 0.21600 -1.2125E 04 0.22800 -1.2912E 04 0.23800 -1.3WOE 04 0.24800 -1.3412E 04 0.25800 -1.3913E 04 0.26800 -1.4402E 04 0.27600 - 1.4653E 04 0.28800 -1.4691E 04 0.29800 - 1.4669E 04 0,30800 - 1.4884E 04 0.31800 -1.3B06E 04 0.32800 -1.23SOE 04 0.33800 -1.1302E 04 0.34800 -1.W14E 04 0.35800 -9.4734E 03 0.36800 *8.8053E 03 0.37800 -@.7485E 03 0.38800 -9.2235E 03 0.39800 -*.WOE 03 0.40800 -9.4814E 03 0.41800 -D.S36SE 03 0.42800 -9.7W8E 03 0.43800 -1.0073E 04 0.44800 -9.SS49E 03 fl 0.46800 -9.4794E 03 0.46800 -8.WSBE 03 0.47800 -8.UI8E 03 0.48800 -7.U71E 03 0.49800 -7.4134E 03 TABLE 6.A-12 (SHEET 9 OF 28) T IME FUIICT I ON NUMBER r f 5) FUNCTION DESCRIPTION 0 ( FORCINQ FUNCTION AT NODE 35 00>02J 0. -297.3 E NllML.CR OF AUSC I SSAE s t 577) FUNCTION SCALE FACTOR 8 ( 1.OOOOE 00) ME VALUE FUNCTION 0. -2.9730E 02 0.00050 -2.9860E 02 0.00100 -2,986OE 02 0.0~J150 -2.9860E 02 0.00200 - 2. W60E 02 0.00250 -2.9980E 02 0.00300 -2.9980E 02 0.0 350 -3.0030E 02 0.00400 -2.9980E 02 0.00450 -3.0160E 02 0.00500 -3.0110E 02 0.00550 -3.0350E 02 0.00600 -3.0180E 02 0.00650 -3.0600E 02 0.00700 -3.0900E 02 0.00750 -3.1530E 02 0.00800 -3.2030E 02 0.00850 -3.3070E 02 0.00500 -3.4370E 02 0.00950 -3.6290E 02 0.01000 -3.bOlOE 02 0.01050 -4.2650E 02 0.01100 -4.7310E 02 0,OtlSO -5.35GOE 02 0.01200 -6.1500E 02 0.01250 -7.1640E 02 0.01300 -8.4230E 02 0.01350 -9.9440E 02 0.01400 -1.1793E 03 0.01450 -1.4012E 03 0.01500 -1.6612E 03 0.01550 -1.9GGGE 03 0.01F00 -2.3153E 03 0.01650 -2.7106E 03 0.017(t0 -3.1581E 03 O.pl750 -3.6506E 03 0.01800 -4.2096E 03 O.OlP50 -4.8144E 03 O.01EOO -5.4682E 03 0.01950 -6.1751E 03 0.02000 -6.9289E 03 0.02050 -7.7251E.03 0.02100 -8.5606E 03 0.02150 -9.4330E 03 0.02200 -1.0332E 04 0: 02250 - 1 .1252E 04 0.02300 -1.2186E 04 0.02350 -1.3128E 04 0.02400 - 1 .4067E 04 0.02450 -1.499GE 04 0.02500 -1.5901E 04 0.03000 -2.2177E 04 0.03500 -2.1390E 04 TIME VALUE FUNCTlON 0.00020 -2.9730E 02 0.00070 -2.6860E 02 0.00120 -2.9860E 02 0.00170 -2.986OE 02 0.00220 -2. 9860E 02 0.00270'-2.9910E 02 0.00320 -2.9980E 02 0.00370 -2.egeo~ 02 0.00420 -3.0110E 02 0.00470 -3.0110E 02 0.005iO -3.0230E 02 0.00570 -3.0350E 02 0.00620 -3.0600E 02 0.00670 -3.0780E 02 0.00720 -3.11bOE 02 0.00770 -3.1650E 02 0.00820 -3.2450E 02 0.00870 -3.3570E 02 0.00920 -3.5170E 02 0.00970 -3.7290E 02 0.01020 -4.0310E 02 0.01070 -4.4320E 02 0.01120 -4.9740E 02 0.01 170 -5.6630E 02 0.01220 -6.5320E 02 0.01270 -7.6310E 02 0.01320 -8.9820E 02 0.01370 -I .0654E 03 0.01420 -1.2643E 03 0.01470 -1.5007E 03 0.01520 -1.7783E 03 0.01570 -2.0986E 03 0.01620 -2.4671E 03 0.01670 -2.8835E 03 0.01720 -3.3519E 03 0.01770 -3.8725E 03 0.01820 -4.4448E 03 0.01870 -5.0715E 03 0.01920 -8.7469E 03 0.01970 -6.4715E 03 .0.02020 -7.2422E 03 0.02070 -8.0552E 03 0.02120 -8.9057E 03 0.02170 -9.7890E 03 0.02220 -1.0697E 04 0.02270 -1.1624E 04 0.02320 -1.2563E 04 0.02370 -1.3505E 04 0.02420 -1.4440E 04 0.02470 -1.5361E 04 0,02700 -1.9181E 04 0.03200 -2.2554E 04 0.03700 -2.0258E 04 TINE VALUE FUNCTION 0.00030 -2.9730E 02 0.00080 -2.9860E 02 0.00130 -2.9730E 02 0.00180 -2.9860E 02 0.00230 -2.9860E 02 0.00280 -2.9860E 02 0.00330 -2.998OE 02 0.00380 -3.0110E 02 0.00430 -3.0110E 02 0.00480 -3.0280E 02 0.00530 -3.0350E 02 0.00580 -3.035OE 02 0.00630 -3.0480E 02 0.00680 -3.0730E 02 0.00730 -3.1230E 02 0.00780 -3.178OE 02 0.00830 -3.2520E 02 0.00880 -3.3700E 02 0.00930 -3.5370E 02 0.00980 -3.7890E 02 0.01030 -4.1110E 02 0.01080 -4.5420E 02 0.01130 -5.0920E 02 0.01180 -5.818OE 02 0.01230 -6.7340E 02 0.01280 -7.8830E 02 0.01330 -9.2940E 02 0.01380 -1.1016E 03 0,01430 -1.3083t 03 0.01480 -1.5532E 03 0.01530 -1.8393E 03 0.01580 -2.1687E 03 0.01630 -2.WC7E 03 0.01680 -2.S3727E 03 0.01730 -3.4505E 03 0.01780 -3.9624E 03 0.01830 -4.5653E 03 0.01880 -5.2015E 03 0.01930 -5.8876E 03 0.01980 -6.6213E 03 0.02030 -7.4020E 03 0.02080 -8.223tE 03 0.02130 -9.0799E 03 0.02180 -B.9682E 03 0.02230 -1.0882E 04 0.02280 -1.1812E 04 0.02330 -1.2752E 04 0.02380 -1.3693E 04 0.42430 -1.4t25E 04 0.02480 -1.5543E 04 0.02800 -2.0482E 04 0.03300 -2.2327E 04 0.03800 -1.9772E 04 TIME VALUE FUNCTION 0.00040 -2.9730E 02 0.00090 -2.98606 02 0.00140 -2.9730E 02 0.00199 -2.9860E 02 0.00240 -2.9860E 02 0.00290 -2.99806 02 0.00340 -2.9980E 02 0.00390 -2.9980E 02 0.00440 -3.0110E 02 0.00490 -3.0230E 02 0.00540 -3.0230E 02 0.00590 -3.0410E 02 0.00G40 -3.0660E 02 0. 30690 -3.0900E 02 0.00740 -3.1280E 02 0.00790 -3.t900E 02 0.00840 -3.2820E 02 0.00890 -3.4120E 02 0.00940 -3.5870E 02 0.00990 -3.851OE 02 0.01040 -4.1780E 02 0.01090 -4.6350E 02 0.01140 -5.2260E 02 0 01190 -5.9780E 02 0.01240 -6.9440E 02 0.01290 -8.1530E 02 0.01340 -9.6070E 02 0.01390 -1.13966 03 0.01440 -1.3542E 03 0.01490 -8.6070E 03 0.01540 -1.9021E 03 0.01590 -2.2405E 03 0.01640 -2.6293E 03 0.01690 -3.0648E 03 0.01740 -3.5540E 03 0.01790 -4.0949E 03 0.01840 -4.6891E 03 0.01890 -5.3339E 03 0.01940 -6.0310E 03 0.01990 -6.7737E 03 0.02040 -7.5630E 03 0.02090 -8.3910E 03 0.02140 -3.2553E 03 0.02190 -1.0149E 04 0.02240 -1.1067E 04 0.02290 -1.2000E 04 0.02340 -1.2939E 04 0.02390 -1.3880E 04 0.02440 -1.481tE 04 0.02490 -1.5724E 04 0.02900 -2.1492E 04 0.03400 -2.1917E 04 0.03900 -1.9COIE 04 TABLE 6.A-12 (SHEET 10 OF 28) 0.04200 -1.8995E 04 0.04300 - 1.9029E 04 0.04700 -1.9257E 04 0.04800 -1.9195E 04 0.05200 - 1.77946 04 0.05300 rl.7lO5E 04 0.05700 -1.32906 04 0 05800 -1.2199E 04 0.06200 -7.9043E 03 0.06300

-7.0414E 03 0.06700 -5.5746E 03 0.06800

-5.8366E 03 0.07200 -9.6038E 03 0.07300 -1.1158E 04 0.07700 -1.9005E 04 0.07800 -2.12216 04 0,08200 -3.0014E 04 0.08300 -3.2028E 04 0.08700 -3.8362E 04 0.08800 -3.9463E 04 0.09200 -4.1727E 04 0.09300 -4.1821E 04 0.09700 -4.0570E 04 0.09800 -3.9875E 04 0.10200 -3.6486E 04 . 0.10300 -3.5641E 04 0.10700 -3.2367E 04 0.10800

-3.1559E 04 0.11200 -2.8374E 04 0.11300 -2.7815E 04 0.11700 -2.4406E 04 0.11800 -2.3659E 04 0.12200 -2.0819E 04 0.12300 -2.0224E 04 0.12700 -1.8244E 04 0.12800 -1.7883E 04 0.13200 -1.6740E 04 0.13300 -1.6584E 04 0.13700 -1.6351E 04 0.13800 -1.6337E 04 0.14200 -1.6993E 04 0.14300 -1.7226E 04 0.14700 -1.6504E 04 0.14800 -1.8874E 04 0.15200 -2.0651E 04 0.15300 -2.103tE 04 0.15700 -2.2500E 04 0.15800 -2.2899E 04 0.16200 -2.4078E 04 0.16300 -2.4359E 04 0.16700 -2.5097E 04 0.16800 -2.5229E 04 0.17200 -2.5499E 04 0.17300 -2.5403E 04 0.17700 -2.4809E 04 0.17800 -2,4622E 04 0,18200 -2.3736E 04 0.18300 -2.3553E 04 0.18700 -2.2809E 04 0.18800 -2.2582E 04 0.19200 -2.2058E 04 0.19300 -2.2010E 04 0.19700 -2.1777E 04 0.19800 -2.1770E 04 0.20400 -2.2314E 04 0.20600 - 2.2839E 04 0.21400 -2.4332E 04 0.21600 -2.4913E 04 0.22400 -2.5941E 04 0.22600 -2.6320E 04 0.23400 -2.7030E 04 0.23600 -2.7019E 04 0.24400 -2.7183E 04 0.24600 -2.7505E 04 0.25400 -2.C711E 04 0.25600 -2.870BE 04 0.26400 -2.9361E 04 0.26600 -2.9598E 04 0.27400 -3.0211E 04 0.27600 -3.0265E 04 0.28400 -3.0382E 04 0.28600 -3.0057E 04 0.29400 -3.0073E 04 0.29600 -3.0316E 04 0.30400 -3.0922E 04 0.30600 -3.0839E 04 0.31400 -2.5746E 04 0.31600 -2.9185E 04 0.32400 -2.6669E 04 0.32600 -2.6095E 04 0.33400 -2.418OE 04 0.33600 -2.3775E 04 0.34400 -2.2290E 04 0.34600 -2.1929E 04 0.35400 -2.0375E 04 0.35600 -1.9978E 04 0.36400 - 1.8662E 04 0.36600 -1.8422E 04 0: 37400 -1.7978E 04 0.37600 -1.8014E 04 0.38400 - 1.8650E 04 0.38600 -1.8871E 04 0.39400 -1.9567E 04 0.39600 -1.9653E 04 0.40400 - 1.9708E 04 0.40600 - 1.96646 04 0.41400 -1.9629E 04 0.41600 -1.9686E 04 0.42400 - 1.9952E 04 0.42600 -2.0086E 04 0.43400 -2.06646 04 0.43600 -2.0768E 04 0.44400 -2.OB67E 04 0.44600 -k.0799~ 04 0.45400 -2.0113E 04 0.45600 -1.9862E 04 0.46400 -1.8779E 04 0.46600 -1.8499E 04 0.47400 -1.7533E 04 0.47600 -1.7331E 04 0.48400 -1.6641E 04 0.48600 -1.6486E 04 0.49400 -1.8747E 04 0.49600 -1.6516E 04 TABLE 6.2%-12 (SHEET 11 OF 28) t-' UJ CO " 1P FUNCTION OESCI? 1 f?T I ON Nl;tlBER OF ABSCISSAE FUNCT I ON SC.tLE FACTOR T l ME VALUE FUNCT 1 ON 0. - 1 .2000E 00 0.00030 - 1 .2000E 00 0.00100 -1.2000E 00 0.00150 -1.2000E 00 0.00200 -1.2000E 00 0.002 50 - 1 .2000E 00 0.00300 -2.8000E 00 0.003JO -4.49006 00 0.00.!00 -9.3000E 00 0.00450 -1.9100E 01 0.00500 -3.5000E 01 0.00550 -3.9500E 01 0. @@COO -9.9100E 01 0.00650 - 1.6030E 02 0.C0700 -2.4200E 02 0.00750 -3.5940E 02 0.00000 -5.159V5 02 0.00350 -7.2120E 02 0.00000. -9.8240E 02 0.009S0 -1.3071E 03 0.01000 -1.7062E 03 0.01050 -2.1940E 03 0.01 100 -2.7461E 03 0.01 i50 -3:4011E 03 0.01200 -4.

1462E 03 0.01250 -4.9926E 03 0.01300 -3.9374E 03 0.0 1350 -6.9816E 03 0.01400 -a. 1380~ 03 0.01450 -9.3904E 03 0.01500 -1.0741E 04 0.01550 -1.2184E 04 0.01600 -1.37166 04 0.01650 -1.53326 04 0.01700 -1.7027E 04 0.01750 -1.8795E 04 0.01800 -2.0622E 04 0.01850 -2.2508E 04 0.01900 -2.4456E 04 0.01950 -2.G456E 04 0.05000 -2. E497E 04 0.02050 -3.0567E 04 0.02 100 -3.2650E 04 0.02150 -3.4739E 04 0.02200 -3.68 18E 04 0.02250 -3.8879E 04 0.02300 -4.090GE 04 0.02350 -1.28SOE 04 0.02400 -4. ,381 9E 04 0.02450 -4.6664E 04 0.02500 -1.8180E 04 0.03000 -6.24t5E 04 0.03500 -7.7558E 04 = ( FORCING FUNCTION A TIME VALUE FUNCf ION 0.00010 -1.2000E 00 0.00060 -1.2000E 00 0.001t0 -1.2000E 00 0.00160 -1.2000E 00 0.00210 -1.2000E 00 0.00260 - 1 .2000E 00 0.00310 -2.8000E 00 0.00360 -6.lOOOE 00 0.00410 -1.1000E 01 0.00460 -2.0700E 01 0.00510 -3.9900E 01 0.00560 -6.7200E 01 0.00610 -1.1030E 02 0.00660 -1.7450E 02 0.00710 -2.6370E 02 0.00760 -3.8810E 02 0.00810 -5.5370E 02 0.00860 -7.6800E 02 0.00910 -1.0418E 03 0.00960 -1.3815E 03 0.01010 -1.7956E 03 0.01060 -2.29016 03 0.01 110 -2.8704E 03 0.01160 -3.5412E 03 0.01210 -4.3086E 03 0.01260 -5.1739E 03 0.01310 -6.13aOE 03 0.01360 -7.2070E 03 0.01410 -8.3809E 03 0.01460 -9.6535E 03 0.01510 -1.10226 .04 0.01560 -1.2484E 04 0.01610 -1.403% 04 0.01660 -1.5665E 04 0.01710 -1.7376E 04 0.01760 -1.9154E 04 0.01810 -2.0995E 04 0.01860 -2.28936 04 0.01910 -2.4851E 04 0.01960 -2.66636 04 0.02010 -2.8908E 04 0.02060 -3.0983E 04 0.02110 -3.3069E 04 0.02160 -3.51566 04 0.02210 -3.72326 04 0.02260 -3.9286E 04 0.02310 -4.1305E 04 0.02360 -4.32BOE 04 0.02410 -4.3197E 04 0.02160 -4.70SOE 04 0.02600 -5.18336 04 0,03100 -6.5735~ 04 0.03600 -7.9155E 04 TIME VALUE FUNCTION 0.00020 - 1 .2000E 00 0.00070 -1.2000E 00 0.00120 -1.2000E 00 0.00170 -1.2000E 00 0.00220 - 1 .2000E 00 0.00270 -1.2000E 00 0.00320 -2.9000E 00 0.00370 -6. 1000E 00 0.00420 -1.2600E 01 0.00470 -2.1000E 01 0.00920 -4.3200E 01 0.00570 -7.3700E 01 0.00620 -1.2130E 02 0.00670 -1.8920E 02 0.00720 -2.84906 02 0.00770 -4.1820E 02 0.00820 -5.9160E 02 0.00870 -8. 1810E 02 0.00920 -1.1046E 03 0.00970 -1.4564E 03 0.01020 -1.88t6E 03 0,01070 -2.3993E 03 0.01120 -2.9971E 03 0.01170 -3.6374E 03 0.01220 -4.4722E 03 0.01270 -5.3571E 03 0.01320 -6.3413E 03 0.01370 -7.4325E 03 0.01420 -8.6272E 03 0.01470 -9.9180E 03 0.01520 -1.1307E 04 0.01570 -1.278SE 04 0.01620 -1.4354E 04 0.01670 -1.6003E 04 0.01 720 -1.7726E 04 0.01770 -1.9518E 04 0.01820 -2.1370E 04 0.01870 -2.32836 04 0.01920 -2.52516 04 0,01970 -2.7268E 04 ' 0.02020 -2.9323E 04 0.02070 -3.1100E 04 0.02120 -3.3488E 04 0.02170 -3.5573E 04 0.02220 -3.7646E 04 0.02270 -3.9694E 04 0.02320 -4.1705E 04 0.02370 -4.3668E 04 0.02420 -4.5574E 04 0.02470 -4.7411E 04 0.02700 -5 1808E 04 0.03200 -6.9368E 04 0.03700 -8.

C2C8E 04 TIME VALUE FUNCTION TI 0.00030 - 1 .2000E 00 0.00090 - 1 .2000E 00 0.00130 -1.2000E 00 0.00180 -1.2000E 00 0.00230 -1.2000E 00 0.00280 -2.8000E 00 0.00330 -4.4000E 00 0.00380 - 7.7000E 00 0.00430 -1.4200E 01 0.00480 -2.7300E 01 0.00530 -4.8100E 01 0.00580 -8.1900E 01 0.00630 - 1 ,32706 02 0.00680 -2.0510E 02 0.00730 -3.0970E 02 0.00780 -4.48406 02 0.00830 -6.3230E 02 0.00880 -8.7 1706 02 0.00930 -1.1717E 03 0.00980 -1.5384E 03 0.01030 -1.9841E 03 0.01080 -2.5117E 03 0.01130 -3.1283E 03 0.01180 -3.8360E 03 0.01230 -4.6430E 03 0.01280 -5.5466E 03 0.01330 -6.55236 03 0.01380 -7.66546 03 0.01430 -8.8785E 03 0.01480 -1.0190E 04 0,01530 -1. IS966 04 0.01580 -1,3093E 04 0.01630 -1.4678E 04 0.01680 -1.6341E 04 0.01 730 -I. 8080E 04 0.01780 -1.9884E 04 0.01830 -2.1746E 04 0.01880 -2.3671E 04 0.01930 -2.56516 04 0.01980 -2.7676E 04 0.02030 -2.9735E 04 0.02080 -3. 1813E 04 0.02130 -3.3903E 04 0.02180 -3.39906 04 ' 0.02230 -3.80578 04 0.02280 -4.0098E 04 0.02330 -4.2102E 04 0.02380 -4.40546 04 0.02430 -4.5346E 04 0.02480 -4.7772E 04 0.02800 -5.7371E 04 0.03300 -7.266SE 04 0.03800 -a.o634~ 04 ME VALlJE FUNCTION 0.00040 -1.2000E 00 0. oongo - 1 .2000~ 00 0.00140 -1.2000E 00 0.00190 -1.2000E 00 0.00240 -1.2000E 00 0.00290 -2.8000E 00 0.00340 -4.4000E 00 0.00390 -9.3000E 00 0.00440 -1.7500E 01 0.00490 -3.0590E 01 0.00540 -3.4600E 01 0.00590 -9.1200E 01 0.00640 -1.4600E 02 0.00690 -2.2260E 02 0.00740 -5.3490E 02 0.00790 -4.8130E 02 0.00840 -6.7660E 02 0.00890 -9.2620E 02 0.00940 -1.2388E 03 0.00990 -1.6212E 03 0.01040 -2.0829E 03 0.01090 -2.6271E 03 0.01140 -3.2611E 03 0.01190 -3.9910E 03 0.01240 -4.8151E 03 0.01290 -5.7391E 03 0.01340 -6.7662E 03 0.01390 -7.9000E 03 0.01440 -9.1319E 03 0.01490 -1.0463E 04 0.01540 -1.1888E 04 0.01590 -1.3403E 04 0.01640 -1.5003E 04 0.01690 -1.6683E 04 0.01740 -1.8436E 04 0.01 790 -2.0252E 04 0.01840 -2.2128E 04 0.01890 -2.4063E 04 0.01940 -2.60536 04 0.0 1 990 -2.8085E 04 0.02040 -3.015OE 04 0.02090 -3.2233E 04 0.02140 -3.4323E 04 0.02190 -3.64056 04 0.02240 -3.8469E 04 0.02290 -4.0504E 04 0.02340 -4.24966 04 0.02390 -4.4439E 04 0.02440 -4.63166 04 0.02490 -4.8127E 04 0.02900 -5.9642E 04 0 03400 -7.53836 04 0.03900 -8.0671E 04 TABLE 6 .A-12 (SHEET 12 OF 28 t-' N TABLE 6.A-12 (SHEET 13 OF 28) TIME FUNCTION NUMBER = ( 7) FUNCTION DESCRIPTION 8 I FORCING FUNCTION AT NODE 38 00>02S 0. Nt.IItCf C OF ABSCISSAE L 577) FUNCl IOt*! SCALE FACTOR 8 ( 1.000OE 001 TlME VALUE FUNCTION 0. -2.000OE-01 0.00050 -2.0000E-01 0.00100 -2.0000E-01 0.00150 -2.0000E-01 0.00206 -2.0000E-01 0.00250 -2.0000E-01 0.00300 -5.0000E-01 0.00350 -8.0000E-01 0.00400 - 1.7000E 00 0.00450 -3.5000E 00 0.00500 -6.4000E 00 0.00550 -1.0900E 01 0.00G00 -1.8300E 01 0.00F50 -2.9407E 01 0.00700 -4.4600E 01 0.00750 -6.6000E 01 0.00900 -9.4700E 01 0.00350 -1.3250E 02 0.00900 -1.8040E 02 0.00950 -2.4010E 02 ' 0.01000 -3.1330E 02 0.01050 -4.0110E 02 0.01100 -5.0430E 02 0.01150 -6.2460E 02 0.01200 -7 6140E 02 0.01250 -9.1G90E 02 0.01300 -1.0904E 03 0.01350 -1.28276 03 0.01400 -1.4945E 03 0.01450 -1.7245E 03 0.01500 -1.972hE 03 0.01550 -2.237GE 03 0.01600 -2.5190E 03 0.01650 -2.8157E 03 0.01700 -3.1270E 03 0.01750 -3.4517E 03 0.01800 -3.7871E 03 0.01850 -4.1336E 03 0.01900 -4.4913E 03 0.01950 -4.E506E 03 0.02000 -5.2333E 03 0.021150 -t .6136E 03 0.02100 -5.9962~ 03 ' 0.02150 -6.3798E 03 0.02200 -6.7617E 03 0.02250 -7.1400E 03 0.02300 -7.5123E 03 0.02350 -7.8766E 03 0.02400 -8.2309E 03 0.02450 -8.5735E 03 0.02sno -8.9032~ 03 0.03000 -1.1462E 04 0.03500 -1.4243E 04 TIHE VALUE FUNCTION 0.00010 -2.0000E-01 0.00060 -2.0000E-01 0.00110 -2.0000E-01 0.00160 -2.0000E-01 0.00210 -2.0000E-01 0.00260 -2.0000E-01 0.00310 -5.0000E-01 0.00360 -1.1000E 00 0.00410 -2.0000E 00 0.00460 -3.8000E 00 0.00510 -7.3000E 00 0.00560 -1.2300E 01 0.00610 -2.0300E 01 0.00660 -3.2000E 01 0.00710 -4.8400E 01 0.00760 -7.1300E 01 0.00810 -1.0170E 02 0.00860 -1.4100E 02 0.00910 -1.9130E 02 0.00960 -2.5370E 02 0.01010 -3.2980E 02 0.01060 -4.2060E 02 0.01110 -6.2710E 02 0.01160 -6.5030E 02 0.01210 -7.9130E 02 0.01260 -9.5020E 02 0.01310 -1.l272E 03 0.01360 -1.3236E 03 0.01410 -1.6391E 03 0.01460 -1.7728E 03 0.01510 -2.0242E 03 0.01560 -2.2927E 03 0.01610 -2.5775E 03 0.01660 -2.87696 03 0.01710 -3.19llE 03 0.01760 -3.5176E 03 0.01810 -3.8556E 03 0.01860 -4.2042E 03 0.01910 -4.5638E 03 0.01960 -4.9333E 03 0.02010 -6.3088E 03 0.02060 -5.6900E 03 0.02110 -6.0730E 03 0.02160 -6.4563E 03 0.02210 -6.8376E 03 0.02260 -7.2147E 03 0.02310 -7.5857E 03 0.02360 -7.Q484E 03 0.02410 -8.3004E 03 0.02460 -8.6406E 03 0.02600 -9.51906.03 0.03100 -1.2072E 04 0.03600 -1.4537E 04 TIHE VALUE FUNCTION 0.00020 -2.0000E-01 0.00070 -2.00OOE-01 0.00120 -2.0000E-01 0.00170 -2,0000E-01 0.00220 -2.0000E-01 0.00270,-2.0000E-01 0.00320 -6.0000E-01 0.00370 -1.1000E 00 0.00420 -2.3000E 00 0.00470 -4.4000E 00 0.00520 -7.900OE 00 0.00570 -1.3bOOE 01 0.00620 -2.230OE 01 0.00670 -3.4700E 01 0.00720 -5.22OOE 01 0.00770 -7.6800E 01 0.00820 -1. 086OE 02 0.00870 -1.5020E 02 0.00920 -2.0290E 02 0.00970 -2.675OE 02 0.01020 -3.465OE 02 0.01070 -4.4060E 02 0.01120 -6.5040E 02 0.01170 -6.7720E 02 0.01220 -6.2130E 02 0.01270 -9.8380E 02 0.01320 -1.1647E 03 0.01370 -1.3650E 03 0.01420 -1.S844E 03 0.01470 -1.8214E 03 0.01520 -2.0765E 03 0.01570 -2.3480E 03 0.01620 -2.6361E 03 0.01670 -2.93E9E 03 0.01720 -3.2554E 03 0.01770 -3.5845E 03 0.01820 -3.9245E 03 0.01870 -4.2758E 03 0.01920 -4.6372E 03 .0.01970 -5.0078E 03 0.02020 -5.3851E 03 0.02070 -5.7665E 03 0.02120 -6,1500E 03 0.02170 -6.8330E 03 0.02220 -6.9136E 03 0.02270 -7.2897E 03 0.02320 -7.6590E 03 0.02370 -6.OlSSE 03 0.02420 -8.3696E 03 0.02470 -8.7070E 03 0.02700 -1.0065E 04 0.03200 -1.2739E 04 0.03700 -1.4730E 04 TIHE VALUE FUNCTION 0.00030 -2.0000E-01 0.00080 -2.0000E-01 0.00130 -2.0000E-01 0.00180 -2.0000E-01 0.00230 -2.0000E-01 0.00280 -5.OOOOE-01 0.00330 -6.OOOOE-01 0.00380 -1.4000E 00 0.00430 -2.6000E 00 0.00480 -6.OOOOE 00 0.00530 -8.BOOOE 00 0.00580 -1.5000E 01 0.00630 -2.4400E 01 0.00680 -3.7700E 01 0.00730 -6.6900E 01 0.00780 -8.2300E 01 0.00830 -1.1610E 02 0.00880 -1.6010E 02 0.00930 -2.1520E 02 0.00980 -2.8250E 02 0.01030 -3.6440E 02 0.01080 -4.6130E 02 0.01130 -5.745OE 02 0.01180 -7.04SOE 02 0.01230 -8.8270E 02 0.01280 -1.0186E 03 0.01330 -1.2033E 03 0.01380 -1.4077E 03 0.01430 -1.630SE 03 0.01480 -1.0713E 03 0.01130 -2.1296E 03 0.01180 -2.4045E 03 0.01630 -2.6956E 03 0.01680 -3.0009E 03 0.01730 -3.3203E 03 0.01780 -3.6516E 03 0.01830 -3.9937E 03 0.01880 -4.3471E 03 0.01930 -4.7108E 03 0.01980 -5.0827E 03 0.02030 -8.4608E 03 0.02080 -5.6427E 03 0.02130 -6.2263E 03 0.02180 -6.6095E 03 0.02230 -6.9891E 03 0.02280 -7.364OE 03 0.02330 -7.7320E 03 0.02380 -1.0905E 03 0.02430 -8.4379E 03 0.02480 -8.7732E 03 0.02800 -1.0536E 04 0.03300 -1.3345E 04 0.03800 -1.4819E 04 TlHE VALUE FUNCTION 0.00040 -2.0000E-01 0.00090 -2.0000E-01 0 00140 -2.0000E-01 0.00190 -2.0000E-01 0.00240 -2.0000E-01 0.00290 -5.0000E-01 0.00340 -8.OOOOE-01 0.00390 -1.7000E 00 0.00440 -3.ZOOOE 00 0.00490 -5.6000E 00 0.00540 -1.0000E 01, 0.00590 -1.6800E 01 0.00640 -2.68OOE 01 0.00690 -4.0900E 01 0.00740 -6.85OOE 01 0.00790 -8.8400E 01 0.00840 -1.2430E 02 0.00890 -1.7010E 02 0.00940 -2.2750E 02 0.00990 -2.S770E 02 0.01040 -3.6250E 02 0.01090 -4.0230E 02 0.01140 -5.S890E 02 0.01190 -7.3290E 02 0.01240 -8.0430E 02 0,01290 -1.O540E 03 0.01340 -1.2426E 03 0.01390 -1.4508E 03 0.01440 -1.6771E 03 0.01490 -1.9215E 03 -0.01540 -2.1832E 03 0.01590 -2.4614E 03 0.01640 -2.7553E 03 0.01690 -3.0638E 03 0.01740 -3.3037E 03 0.01790 -3.7192E 03 0.01840 -4.0637E 03 0.01890 -4.4191E 03 0.01940 -4.7846E 03 0.01990 -5.1578E 03 0.02040 -5.5370E 03 0.02090 -5.8196E 03 0.02140 -6.3033E 03 0.02190 -6.6856E 03 0.02240 -7.064BE 03 0.02290 -7.4384E 03 0.02340 -7.6043E 03 0.02390 -8.1611E 03 0.02440 -8.5059E 03 0.02490 -8.8384E 03 0.02900 -1.0953E 04 0.03400 -1.3844E 04 0.03900 -1.4815E 04 0.04000 -1.4711E 04 0.04500 -1.3327E 04 0.05000 -1.1748E 04 0,05500 -1 .O684E 04 0.06000 -1.0514E 04 0.06500 -1.1202E 04 0.07000 -1.2237E 04 0.07500 -1.3317E 04 0.08000 - 1 .4057E 04 0.0e5oo -1.4172~ 04 0.09000 -1.3888E 04 0.09500 -1.3707.E 04 0.10000 -1.3889E 04 0.10500 -1.3923E 04 0.11000 -1.3596E 04 0.11500 -1.2932E 04 0.12000 -1.2110E 04 0.12500 -1.1308E 04 0.13000 -1.0615E 04 0.13500 -1.0285E 04 0.14000 -1.0327E 04 0.14500 -1.0$52E 04 0.15000 -1.OG44E 04 0.1 Cj!i00 - 1 .0773E 04 t3 0.16000 -1.0918E 04 E 0.16900 -1.lO66E 04 t' 0.17000 -1.0929E 04 M 0,17500 -1.OS78E 04 0.18000 -1.0137E 04 cn . 0.18500-1.0616E04 0 '1 9000 - 1 . OFGOE 04 5;, 0.19500 -1.0692E 04 I 0.20000 -1.0817E 04 I--' 0: 21000 -1. 1021E 04 h, 0.22000 -1.0917E 04 0.23000 -1.0754E 04 6.24000 -1.09G9E 04 0.25000 -1.1291E 04 0.26000 -1.1487E 04 0.27000 -1.1028E 04 0.28090 -1.0312E 04 0.29000 -1.0614E 04 0.30000 -1.102ZE 04 0.3lOpO -1.1237E 04 0.32000 - 1 . 1267E '04 0.33000 - 1 . 1060E 04 0.34000 - 1 .0387E 04 0.35000 - 1.0004E 04 0.3G000 -9.857EE 03 0.37000 -9.9302E 03 0.38000 -1.02316 04 0.39000 -9.9178E 03 0.40600 -9.8373E 03 0.41000 -9.9194E 03 0.42000 -1.0102E 04 0.43000 -1.05436 04 0.41000 -1.0488E 04 0.45000 -1.0395E 04 0.4600n -1.0372~ 04 0.47000 -1.0317E 04 0.48000 -1.OO8OE 04 0.49000 -9.450RE 03 TABLE 6.A-12 (SHEET 14 OF 28) 0.04200 -1.4301E 04 0.04300 -1.4006E 04 0.04700 -8.269IE 04 0.04800 -1.2340E 04 0.05200 -1.1260E 04 0.0530$l -1.1045E 04 b.05700 -1.0494E 04 0.05800 -1.0460E 04 0.06200 -1.0723E 04 0.06300 - 1.0860E 04 0.06700 -1.1586E 04 0.06800 -1.1806E 04 0.07200 -1.2654E 04 0.07300 -1.2873E 04 0.07700 -1.3689E 04 0.07800 -1.3827E 04 0.08200 -1.4170E 04 0.08300 -1.4194E 04 0.08700 -1.4073E 04 0.08800 -1.4009E 04 0.09200 -1.3787E 04 0.09300 -1.3761E 04 0.09?00 -1.3744E 04 0.09800 -1.3889E 04 0.10200 -1.3942E 04 0.10300 -1.3955E 04 0,10700 -1.3838E 04 0.10800 -1.3778E 04 0.11200 -1.3344E 04 0.11300 -1.3205E 04 0.11700 -1.2604E 04 0.11800 -1.2431E 04 0.12200 -1.1793E 04 0.12300 -1.1626E 04 0.12700 -1.0987E 04 0.12800 -1.0838E 04 0.13200 -1.0428E 04 0.13300 -1.O358E 04 0.13700 -1.O291E 04 0.13800 -1.0306E 04 0,14200 -1.0386E 04 0.14300 -1.0402E 04 0.14700 -1.0544E 04 0.14800 -1.0577E 04 0.15200 -1.0682E 04 0.18300 -1.0724E 04 0.15700 -1.0824E 04 0.15800 -1.0888E 04 0.16200 -1.0949E 04 0.16300 -1.098OE 04 0.16700 -1.0984E 04 0.16800 -1.0970E 04 0.17200 -1.0913E 04 0.17300 -1.O863E 04 0.17700 - 1 .0804E 04 0.17800 -1.07766 04 0.t8200 -1.0701E 04 0.18300 -1.0670E 04 0.18700 -1.0697E 04 0.18800 -1.0643E 04 0.19200 -1.0629E 04 0.19300 -1.0606E 04 0.19700 -1.06BOE 04 0.19800 -1.0845E 04 0.20400 -1.0886E 04 0.20600 -1.0834E 04 0.21400 -1.0946E 04 0.21600 -1.0938E 04 0.22400 -1.0913E 04 0.22600 -1.0916E 04 0.23400 -8.0735E 04 0.23600 -1.0794E 04 0.24100 -1.1203E 04 0.24600 -1.1294E 04 0,25400 -1.1266E 04 0.25600

-1.1381E 04 0.26400 -1.1456E 04 0.26600 -1.1326E 04 0.27400 -1.0984E 04

• 0.27600 -1.0904E 04 0.28400 -1.0264E 04 0.28600 -1.0323E 04 0.29400 -1.0704E 04 0.29600 -1.0705E 04 0,30400 -1.llBSE 04 0.30600 -t.flZlE 04 0.31400 -1.1231E 04 0.31600 -1.1360E 04 0.32400 -1.1391E 04 0.32600 -1.1362E 04 0.33400 -1.0712E 04 0.33600 -1.0598E 04 0.34400 -1.0285E 04 0.34600 -1.0136E 04 0.35400 -9.7808E 03 0.35600

-9.8599E 03 0.36400 -9.8472E 03 0.36600 -9.7282E 03 0.37400 -1.0061E 04 0.37600 -1.OlO5E 04 0.38400 -1.OOBSE 04 0.38600 -1.0102E 04 0.39400 -9.6435E 03 0.39600

-9.7887E 03 0.40400 -9.7230E 03 0.40600

-9.6969E 03 0.41400 -1.0017E 04 0.41600 -1.0029E 04 0.42400 -1.0275E 04 0.42600 -1.0359E 04 0.43400 -1.0532E 04 0.43600 -1.0486E 04 0,44400 -1,04726 04 0.44600 -1.0490E 04 0.45400 -1.0381E 04 0.45600 -1.0332E 04 0.46400 -1.0398E 04 0.46600 -1.0356E 04 0.47400 -1.0239E 04 0.47600 -1.0197E 04 0,48400 -9.9631E 03 0.48600 -0.7251E 03 0.49400-9.4228E03 0.49600-9.5013E03 TIME FUNCTION NUMBER I ( 8) TABLE 6.A-12 (SHEET 15 OF 28) I-' Lo M & FUNC' 1014 OESCR I PT l ON = f FORCINO FUNCTION AT NODE 39 00>02S 0. 96.4 ) NlJllCCR OF ABSCISSAE

= 577) FUI4CT IOIJ ZCALE FACTOR = I 1 .0000E 00) TIME VALUE FUNCTION 0. 9.6400E 01 0.00050 9.6400E 01 0.00100 9.5600E 01 0.00150 9.OPOOE 01 0.00200 7.6400E 01 0.00250 4.3EOOE 01 *0.00300 -1.5200E 01 0.00350 -1.lOlOE 02 0.00400 -2.4960E 02 0.004t3 -4.4630E 02 0.00500 - 7.2080E 02. 0.00550 -1.0933E 03 , 0.00600 -1.5829E 03 (3.00650 -2.2036E 03 0.00700 -2.9674E 03 0.00750 -3.6854E 03 t3 0.00300 - 4.96236 03 E 0.00850 -6.2004E 03 0.00900 -7.601QE 03 0.00950 -9.1712E 03 0.01000-1.0913E04 0.01050 -1.2826E 04 CT\ 0.01100 -1.4907E 04 0.01150 -1.7157E 04 ? 0.01200 -1.9571E 04 0.01250 -2.2145E 04 I-' 0.01300 -2.4874E 04 N 0.01350 -2.7748E 04 .0.01400 -3.0763E 04 0.01450 -3.3897E 04 0.OlB00 -3.7139E 04 0.0 1550 -4.0468E 04 0.01600 -4.38G5E 04 0.01650 -4.7307E 04 0.01700 -5.0781E 04 0.017.50 -5.4251E 04 0.01800 -5.76C3E 04 0.01850 -6.1071E 04 0.01900 -6.45236 04 0.01950 -6.7863E 04 0.02000 -7.1096E 04 0.02050 -7.4204E 04 0.02100 -7.7186E 04 0.02150 -8.0045E 04 0.02200 -8.2784E 04 0.02250 -8.5407E 04 0.02300 -8.7880E 04 0.02350 -9.0251E 04 0.02400 -9.2453E 04 0.02150 -9.4551E 04 0.02500 -9.6529E 04 0.03000 -1.0489E 05 0.03500 -1.0173E 05 TINE VALUE FUNCTION 0.00010 9.6400E 01 0.00060 9.6400E 01 0.00110 9.4800E 01 0.00160 8.8400E 01 0.00210 7.1600E 01 0.00260 3.4200E 01 0.00310 -3.1200E 01 0.00360 -1.3390E 02 0.00410 -2.8440E 02 0.00460 -4.9450E 02 0.00510 -7.8700E 02 0.00560 -1.162OE 03 0.00610 -1.6956E 03 0.00660 -2.3446E 03 0.00710 -3.1384E 03 0.00760 -4.O881E 03 0.00810 -5.1966E 03 0.00860 -6.4G69E 03 0.00910 -7.9016E 03 0.00960 -9.5051E 03 0.01010 -1.1282E 04 0.01060 -1.3229E 04 0.01110 -1.53446 04 0.01160 -1.7626E 04 0.01210 -2.0073E 04 0.01260 -2.2679E 04 0.01310 -2.5437E 04 0.01360 -2.P340E 04 0.01410 -3.1381E 04 0.01460 -3.4537E 04 0.01510 -3.7798E 04 0.01560 -4.1143E 04 0.01610 -4.4551E 04 0.01660 -4.8002E 04 0.01710 -5.1477E 04 0.01760 -5.4947E 04 0.01810 -5.8345E 04 0.01860 -6.1769E 04 0.01010 -6.5204E 04 0.019GO -6.8527E 04 0.02010 -7.1728E 04 0.02060 -7.4815E 04 0.02110 -7.7769E 04 0.02160 -8.0598E 04 0.02210 -8.3321E 04 0.02260 -8.5914E 04 0.02310 -8.83GBE 04 0.02360 -9.0688E 04 0.02410 -9.2883E 04 0.02460 -9.4956E 04 0.02600 -1.0012E 05 0.03100 -1.O5OSE 05 0.03600 -1.0038E 01 TIME VALUE FUNCTION 0,00020 9.6400E 01 0.00070 9.5600E 01 0.00.120 9.4000E 01 0.00170 8.6000E 01 0.00220 6.6000E 01 0.00270 2.3800E 01 0.003m -4.8600E 01 0.00370 -1.6OlOE 02 0.00420 -3.2020E 02 0.00470 -6.4570E 02 0.00520 -8.5720E 02 0.00570 -1.2746E 03 0.00620 -1.815lE 03 0.00670 -2.4912E 03 0.00720 -3.3159E 03 0.00770 -4.2939E 03 0.00820 -5.4379E 03 0.00$70 -6.7411E 03 0.00920 -8.2096E 03 0.00970 -9.8475E 03 0.01020 -1.1657E 04 0.01070 -1.3638E 04 0.01120 -1.5787E 04. 0.01170 -1.6103E 04 0.01220 -2.05e2~ 04 0.01270 -2.3218E 04 0.01320 -2.6006E 04 0.01370 -2.8938E 04 0.01420 -3.2003E 04 0.01470 -3.E182E 04 0.01520 -3.8461E 04 0.01670 -4.1820E 04 0.01620 -4.5239E 04 0.01670 -4.669GE 04 0.01720 -5.2173E 04 0.01770 -5.5637E 04 0.01820 -5.9029E 04 0.01870 -6.2460E 04 0.01920 -6.5876E 04 b.01970 -6.9166E 04 0.02020 -7.2353E 04 0.02070 -7.5398E 04 0.02120 -7.8347E 04 0.02170 -8.1163E 04 0.02220 -8.3890E 04 0.02270 -8.6416E 04 0.02320 -8.8843E 04 0.02370 -9.1143E 04 0.02420 -9.3307E 04 0.02470 -0.53E7E 04 0.02700 -1.0323E 05 0.03200 -1.0467E 05 0.03700 -9.8980E 04 TI HE VALUE FUNCT l ON 0.00030 9.6400E 01 0.00080 9.5600E 01 0.00130 9.3200E 01 0.00180 8.3600E 01 0.00230 5.9600E 01 0.00260 1.2600E 01 0.00330 -6.7000E 01 0.00380 -1.8710E 02 0.00430 -3.5960E 02 0.00480 -6.0030E 02 0.00550 -9.3140E 02 0.00580 -1.3721E 03 0.00630 -1.9388E 03 0.00680 -2.6438E 03 0.00730 -3.4991E 03 0.00780 -4.5121E 03 0.00830 -5.6854E 03 0.00880 -7.0212E 03 0.00930 -8.5229E 03 0.00980 -1.01S6E 04 0.01030 -1.2040E 04 0.01080 -1.4055E 04 0.01130 -1.6237E 04 0.01180 -1.8585E 04 0.01230 -2.1097E 04 0.01280 -2.3764E 04 0.01330 -2.6580E 04 0.01380 -2.9542E 04 0.01430 -3.2631E 04 0.01480 -3.5830E 04 0.01530 -3.9127E 04 0.01580 .-4.2500E 04 0.01630 -4.5928E 04 0.01680 -4.9391E 04 0.01730 -5.2867E 04 0.01780 -5.6327E 04 0.01830 -5.9706E 04 0.01880 -6.3165E 04 0.01930 -6.6545E 04 0.01980 -6.9816E 04 0.02030 -7.2972E 04 0.02080 -7.6000E 04 0.02130 -7.8918E 04 0.02180 -8.1694E 04 0.02230 -8.43766 04 0.02280 -6.6912E 04 0.02330 -8.0312E 04 0.02380 -9.lSBlE 04 0.02430 -9.3726E 04 0.02480 -0.8751E 04 0.02800 -1.0581E 05 0.03300 -1.0397E 05 0.03800 -9.7599E 04 TINE VALUE FUNCTION 0.00040 9 6400E 01 0.00090 9.6600E 01 0.00140 9.1600E 01 0.00190 8.0400E 01 0.00240 5.1800E 01 0.00290 -2.0000E-01 0.00340 -8.7600E 01 0.00390 -2.186OE 02 0.00440 -4.0160E 02 0.00490 -6.5870E 02 0.00540 -1.0107E 03 0.00590 -1.4754E 03' 0.00640 -2.0685E 03 0.00690 -2.8026E 03 0.00740 -3.6882E 03 0.00790 -4.7335E 03 0.00840 -5.9396E 03 0.00890 -7.3081E 03 0.00940 -8.8443E 03 0.00990 -1.0551E 04 0.01040 -1.2429E 04 0.01090 -1.4478E 04 0.01140 -1.6694E 04 0.01190 -1.9074E 04 0.01240 -2.1618E 04 0.01290 -2.43t6E 04 0.01340 -2.7160E 04 0.01390 -3.015OE 04 0.01440 -3.3262E 04 0.01490 -0.6483E 04 0.01540 -3.9796E 04 0.01690 -4.3182E 04 0.01640 -4.6619E 04 0.01690 -5.0086E 04 0.01740 -5.3561E 04 0.01790 -5.7014E 04 0.01840 -6.0368E 04 0.01890 -6.3838E 04 0.01940 -6.7206E 04 0.01990 -7.0460E 04 0.02040 -7.35676 04 0.02090 -7.6596E 04 0.02140 -7.9484E 04 0.02190 -8.2242E 04 0.02240 -8.4894E 04 0.02290 -8.7403E 04 0.02340 -8.9776E 04 0.02390 -9.2032E 04 0.02440 -9.4142E 04 0.02490 -9.6142E 04 0.02900 -1.0430E 05 0.03400 -1.0293E OS 0.03900 -9.6247E 04 TABLE 6.A-12 (SHEET 16 OF 28) 0.04200 -9.295OE 04 0.04700 -8.9282E 04 0.05200 -8.8609E 04 0.05700 -9.2981E 04 0.06200 -1.0014E 05 0.06700 -1.OBlBE 05 0.07200 -1.1435E 05 0.07700 -1.1305E 05 0.08200 - 1 . O823E 05 0.08700 -1.C304E 05 0.09200 -9.9706E 04 0.09700 -9.ESBBE 04 0.10200 -9.2873E 04 0.10700 -8.Q976E 04 0.11200 -8.7043E 04 0.11700'-8.4127E 04 0.12200 -8.10566 04 0.12700 -7.e245~ 04 0.13200 -7.5551E 04 0.13700 -7.2815E 04 0.14200 -7.0636E 04 0.14700 -6.9219E 04 0.15200 -6.8214E 04 0.15700 -6.782BE 04 0.16200 -6.7394E 04 0.16700 -6.6607E 04 0.17200 -6.5552E 04 0.17700 -6.4443E 04 0.18200 -6.3379E 04 0.18700 -6.2481~ 04 0.19200 -6.1833E 04 0.19700 -6.1126E 04 0.20400 -6.0020E 04 0.21400 -5.8949E 04 0.22400 -5.e579E 04 0.23400 -6.9137E 04 0.24400 -5.78S8E 04 0.25400 -5.7466E 04 0.26400 -5.5660E 04 0.27400 -5.5467E 04 0.28400 -6.6065E 04 0.29400 -5.5853E 04 0.30400 -8.4997E 04 0.31400 -5.4602E 04 0.32400 -5.3406E 04 0.33400 -5.3965E 04 0.34400 -5.4117E 04 0.35400 -6.4916E 04 ' 0.36400 -5.4438E 04 0.37400 -8.2622E 04 0.38400 -5.tll9E 04 0.39400 -4.9460E 04 0.40400 -4.9163E 04 0.41400 -4.8679E 04 0.42400 -4.8553E 04 0.43400 -4.8262E 04 0.44400 -4.0695E 04 0.45400 -4.8616E 04 0.46400 -4.0306E 04 0.47400 -4.8619E 04 0.48400 -4.9691E 04 0.49400 -4.0906E 04 TABLE 6.A-12 (SHEET 17 OF 28) FUNCTION OESCHiPTION

= t FORCING FUNCTION AT NODE 40 00,02S 0. FlUFlDER OF ABSCI SSAE 8 t 577) FUNCTION SCALE FACTOR = ( 1.0000E 00) TI ElE VALUE FUNCT l ON 0. 1.3630E 02 0.00050 1.3630E 02 0.00100 1.3510E 02 0.00150 t.2840E 02 0.00200 1.0800E 02 0.00250 6.190OE 01 0.00300 -2.1500E 01 0.00350 -1.5550E 02 0.00400 -3.5280E 02 0.00450 -F.3070E 02 0.00500 -1.0187E 03 0.00550 - 1.5450E 03 0.00600 -2.2370E 03 0.00650 -3.1140E 03 0.00700 -4.1936E 03 .O. 00750 -5.4908E 03 0.00800 -7.0126E 03 0.00850 -R. 7624E 03 0.00'00 - 1.0743E 04 0.00950 -1.29tlE 04 0.01000 -1.5422E 04 0.01050 -1.Ct26E 04 0.01100 -2.1067E 04 0.01150 -2.4247E 04 0.01200 -2.7658E 04 0.01250 -3.1296E 04 0.01300 -3.5151E 04 0.01350 -3.9213E 04 0.01400 -4.3474E 04 0.01450 -4.7904E 04 0.01500 -5.2484E 04 0.01550 -5.7189E 04 0.01600 -6.1989E 04 0.01650 -6.G855E 04 0.01700 -7.1764E 04 0.01750 -7.6671E 04 0.01800 -8.1490E 04 0.01850 -8.630GE 04 0.01900 -9.1184E 04 0.01950 -9.5903E 04 0.02000 -1.0047E 05 0.02050 -1.048GE 05 0.02100 -1.0908E 05 0.02150 -1.1312E 05 0.02200 -1.lG99E 05 0.02250 -1.2070E 05 0.02300 -1.P420E 05 0.02350 -1.27b4E 05 0.02400 -1.3065E 05 0.02450 -1.3362E 05 0.02300 -1.36418 05 0.03000 - 1 .4824E 05 0.03500 -1.4376E 05 TIME VALU~ FUNCTION 0.00010 1.3630E 02 0.00060 1.3630E 02 0.00110 1.3400E 02 0.001 60 1.2500E 02 0.00210 1.012OE 02 0.00260 4.8300E 01 0.00310 -4,4100E 01 0.00360 -1.8920E 02 0.00410 -4.0190E 02 0.00460 -6.9880E 02 0.00510 -1.1122E 03 0.00560 -1.6704E 03 0.00610 -2.3963E 03 0.00660 -3.3134E 03 0.00710 -4.4352E 03 0.00760 -5.7772E 03 0.00810 -7.3438E 03 0.00860 19.1390E 03 0.00910 -1.1167E 04 0.00960 -1.3433E 04 0.01010 -1.5944E 04 0.01060 -1.8695E 04 0.01110 -2.1684E 04 0.01160 -2.49096 04 0.01210 -2.8366E 04 0.01260 -3.2050E 04 0.01310 -3.5948E 04 0.01360 -4.005OE 04 0.01410 -4.4347E 04 0.01460 -4.8808E 04 0.01510 -5.3416E 04 0.01560 -5.8143E 04 0.01610 -6.2959E 04 0.01660 -6.7836E 04 0.01710 -7.2747E 04 0.01760 -7.7651E 04 0.01810 -8.2453E 04 0.01860 -8.7292E 04 0.0.1910 -9.2146E 04 0.01960 -9.6842E 04 0.02010 -1.0137E 05 0.02060 -1.0573E 05 0.02110 -1.0990E 05 0.02160 -1.13SOE 05 0.02210 -1.1775E 05 0.02260 -1.2141E 05 0.02310 -1.24BBE 05 0.02360 -1.2816E 05 0.02410 -1.31266 05 0.02460 -1.3419E 05 0.02600 -1.4149E 05 0.03100 -1.4846E 05 0.03600 -1.4185E 05 TIME VALUE FUNCTION 0.00020 1.3630E 02 0.00070 1.3JlOE 02 0.00120 1.32SOE 02 0.00170 1.2160E 02 0.00220 9.3300E 01 0.00270 ' 3.3600E 01 0.00320 -6.8700E 01 0.00370 -2.2610E 02 0.00420 -4.5250E 02 0.00470 -7.7120E 02 0.00520 -1.2114E 03 0.00570 -1.8012E 03 0.00620 -2.5651E 03 0.00670 -3.5206E 03 0.00720 -4.6861E 03 0.00770 -6.071OE 03 0.00820 - 7.6848E 03 0.00870 -9.5264E 03 0.00920 -1.1602E 04 0.00970 -1.3917E 04 0.01070 -1.6474E 04 0.01070 -1.6273E 04 0.01120 -2.2310E 04 0.01170 -2.55E3E 04 0.01220 -2.9087E 04 0.01270 -3.2812E 04 0.01320 -3.6751E 04 0.01370 -4.0895E 04 0.01420 -4.E226E 04 0.01470 -4.S720E 04 0.01520 -5.4352E 04 0.01570 -5.POe9E 04 0.0'1620 -6.3932E 04 0.01670 -6.8817E 04 0.01720 -7.3730E 04 0.01770 -7.8625E 04 0.01820 -8.3419E 04 0.01870 -8.8268E 04 . 0.01920 -9.3096E 04 0.01S70 -9.7745E 04 0.02020 -1.0225E 05 0.02070 -1.0655E 05 0.02120 -1.1072E 05 0.02170 -1.1470E 05 0.02220 -1.185OE 05 0.02270 -1.22t2E 05 0.02320 - 1.2555E 05 0.02370 -1.2880E 05 0.02420 -1.3186E 05 0.02470 -1.3476E 06 0.02700 -1.4589E 05 0.03200 -1.4792E 05 0.03700 -1.3988E 05 TIME VALUE FUNCTION 0.00030 1.3630E 02 0.00080 1.3510E 02 0.00130 1.3170E 02 0.00180 l.182OE 02 0.00230 8.4200E 01 0.00280 1.7800E 01 0.00330 -9.4700E 01 0.00380 -2.6440E 02 0.00430 -5.O82OE 02 0.00480 -8.4830E 02 0.00530 -1.3162E 03 0. 00580 - 1.9390E 03 0.00630 -2.7399E 03 0.00680 -3.7363E 03 0.00730 -4.9140E 03 0.00780 -6.3765E 03 0.00830 -8.0346E 03 O.OO8PO -9.9223E 03 0.00930 -1.204SE 04 0.00980 -1.4409E 04 0.01030 -1.7015E 04 0.01080 -1.9862E 04 0.01130 -2.2946E 04 0.01180 -2.5265E 04 0.01230 -2.9814E 04 0.01280 -3.3583E 04 0.01330 -3.7563E 04 0.01380 -4.1749E 04 0.01430 -4.6113E 04 0.01480 -5.0635E 04 0.01530 -5.5295E 04 0.01580 -6.0061E 04 0.01630 -6.4906E 04 0.01680 -6.9799E 04 0.01730 -7.4712E 04 0.01760 -7.9601E 04 0.01830 -8.4377E 04 0.01880 -8.9265E 04 0.01930 -9.4040E 04 0.01980 -9.8664E 04 0.02030 -1.0312E 05 0.02080 -1.0740E 05 0.02130 -1.1153E 05 0.02180 -1.1545E 05 0.02230 -1.1924E 05 0.02280 -1.2282E 05 0.02330 -1.2622E 05 0.02380 -1

.2942E 05 0.02430 -1.3245E 05 0.02480 -1.3532E 05 0.02800 -1.4953E 05 0.03300 -1.4693E 05 0.03800 -1.3793E 05 71ME VALUE FUNCTION 0.00040 1 .3630E 02 0.00090 1.3510E 02 0.00140 1.2950E 02 0.00190 1.1360E 02 0.00240 7.3200E 01 0.00290 -3.00OOE-01 0.00340 ' 1.2360E 02 0.00390 -3.0890E 02 0.00440 -5.6760E 02 0.00490 -9.3080E 02 . 0.00540 -1.4283E 03 0.00590 -2.0850E 03 0.00640 -2.9232E 03 0.00690 -3.9607E 03 0.00740 -5.2121E 03 0.00790 -6.6894E 03 0,00840 -8.3939E 03 0.00890 -1.0328E 04 0.00940 -1.2499E 04 0.00990 -1.4911E 04 0.01040 -1.7565E 04 0.01090 -2.0460E 04 0.01140 -2.3591E 04 0.01190 -2.6956E 04 0.01240 -3.0551E 04 0.01290 -3.4364E 04 0.01340 -3.8383E 04 0.01390 -4.2608E 04 0.01440 -4.7006E 04 0.01490 -5.1557E 04 0.01540 -5.6240E 04 0.01590 -6.1024E 04 0.01640 -6.5881E 04 0.01690 -7.0782E 04 0.01740 -7.5691E 04 0.01790 -8.0572E 04 0.01640 -8.5312E 04 0.01890 -9.0216E 04 0.01940 -9.4975E 04 0.01990 -9.9574E 04 0.02040 -1.0399E 05 0.02090 -1.0824E 05 0.02140 -1.1233E 05 0.02190 -1.1622E 05 0.02240 -1.1997E 05 0.02290 -1.2352E 05 0.02340 -1.2687E 05 0.02390 -1.3006E 05 0.02440 -1.33046 05 0.02490 -1.3587E 05 0.02900 -1.4739E 05 0.03400 -1.45d6E 05 0.03900 -1.3602E 05 H w m t-' I-' N I-' U) 03 & TABLE 6.A-12 (SHEET 18 OF 28) 0.04200 -1.31366 05 0.04700 -1.2617E 05 0.05200 - 1 .2522E 05 0.05700 -1.3140E OS 0.06200 -1.4152E 05 0.06700 - i .5288E OS 0.07200 -1.6160E 05 0.07700 -1.5976E 05 0.08200 -1. S295E 05 0.08700 -1.4562E 05 0.09200 -1.4090E 05 0.09700 -1.3650E 05 0.10200 -1.3125E 05 0.10700 -1.2716E 05 0.11200 -1.2301E 05 0.11700 -1.1889E 05 0.12200 -1.1455E OS 0.12700 -1.1058E 05 0.13200 -1.06776 05 0.13700 -1.0290E 05 0.14200 -9.9823E 04 0.14700 -9.7820E 04 0.16200 -9.6400E 04 0.15700 -9. bBE5E 04 0.16200 -9.J240E 04 0.16700 -9.4128E 04 0.17200 -9.2637E 04 0,17700 -9.1070E 04 0.18200 -8.9567E 04 0.18700 -8.8298E 04 0.19200 -8.73ex 04 0.19700 -8.6383E 04 0.20400 -8.4821E 04 0.21400 -8.3306E 04 0.22400 -8.2783E 04 0.23400 -8.3572E 04 0.24400 -8.1764E 04 0.25400 -8.1211E 04 0.26400 -7.86EBE 04 0.27400 -7.8385E 04 0.28400 -7.9231E 04 0.29400 -7.8931E 04 0.30400 -7.77216 04 0.31400 -7.7164E 04 0.32400 -7.5473E 04 0.33400 -7.6263E 04 0.34400 -7.6478E 04 . 0.35400 -7.7607E 04 0.36400 -7.6931E 04 0.37400 -7.4365E 04 0.38400 -7.2241E 04 0.36400 -6.9897E 04 0.40400 -6.9477E 04 0.41400 -6.8793E 04 0.42400 -6.8615E 04 0.43400 -6.8203E 04 0,44400 -6.8815E 04 0.45400 -6.6704E 04 0.46400 -6.8265E 04 0.47400 -6.8708E 04 0.48400 -7.0224E 04 0.49400 -6.Q114E 04

$ I t-' N TABLE 6.A-12 TI tiL FUNCT 1 ON NUMBER a I 10) (SHEET 19 OF 28) FUNCTION DESCRIPTION

= ( FORCINO FUNCTfON AT NODE 42 00,025 0. 314.2 ) NUIiL'LR OF ABSCISSAE

= t 577) FUN( TI ON SCALE FACTOR = t 1.0000E 00) T I PIE VALUE FUNCT I ON 0. 3.1420E 02 0.00050 -3.0042E 04 0.00100 -5.50786 04 0.00150 -8.6GRdE 04 0.00200 -1.16OBE 05 0.00P50 -1.4396E 05 0.00300 -1.7145E 05 0.00350 -1.9084E 05 0.00400 -2.2295E 05 0.00450 -2.5982E 05 0.00500 -2.97146 05 0.00St0 -3,3357E 05 0.00600 -3.6904E 05 0.00650 -4.0367E 05 0.00700 -4.3749E 05 0.00750 -4 .'7058E 05 0.00C00 -5.0272E 05 0.00850 -5.2282E 05 0.00900 -5.5260E 05 0.00Sf0 -5.8118E 05 0.01000 -6.0863E 05 0.01050 -6.3498E 05 0.01100 -6.G038E 05 0.01150 -6.P458E 05 0.01200 -7.C1602E 06 0.01.iSO -7.3011E 05 0.01300 -7.5108E 05 0.01350 -7.7102E 05 0.01400 -7.896EE 05 0.@1450 -8.070RE 05 0.01500 -8.2333E 05 (1.01550 -8.3913E 05 O.Olt00 -8.5356E 05 0.01650 -8.67726 05 0.01700 -6.819.1E 05 0.01750 -8.9548E 05 0.01800 -8.8976E 05 0.01850 -8.9992E 05 0.01900 -9.0949E 05 0.01950 -9.1848E 05 0.02000 -9.2700E 05 0.02050 -9.349tE 05 0.02100 -9.4188E 05 0.02'50 -9.4892E 05 0.02200 -E.S531E 05 0.02250 -9.6103E 05 0.02300 -9.6604E 05 0.02350 -9.7072E 05 0.02400 -$.7471E 05 0.02450 -9.7808E 05 0.02500 -9.8079E 05 0.03000 -9.7927E 05 0.03500 -9.7747E 05 ME VALUE FUNCTION 0.00020 -1.2815E 04 0.00070 -4.04716 04 0.00120 -6.7604E 04 0.00170 -9.872OE 04 0.00220 -3.2734E 05 0.00270 - 1 .6496E 05 0.00320 -1.8248E 05 0.00370 -2.03G7E 05 0.00420 -2.3716E 05 0.00470 -2.7489E 05 0.00520 -3.1182E 05 0.00570 -3.4782E 05 0.00620 -3.e305~ 05 0.00670 -4.1731E 05 0.00720 -4.E081E 05 0.00770 -4.8393E 05 0.00820 -5.0453E 05 0.00870 -5.3481E 05 0.00920 -5.6420E 05 0.00970 -5.9236E 05 0.01020 -6.1936E 05 0.01070 -6.4513E 05 ,0.01120 -6.7014E 05 0.01170 -6.P415E 05 0.01220 -7.1664E 05. 0.01270 -7.3873E 05 0.01320 -7.5915E 05 0.01370 -7.7864E 05 0.01420 -7.9677E 05 0.01470 -8.1340E 05 0.01520 -8.2976E 05 0.01570 -8.4508E 05 0.01620 -8.5945E 05 0.01670 -8.7315E 05 0.01720 -8.8742E 05 0.01770 -9.0040E 05 0.01820 -8.Q388E 05 0.01870 -9.0387E 05 0.01920 -9.12326 05 0.01970 -3.2194E 05 0.02020 -9.3025E 05 0,02070 -9.3780E 05 0.02120 -9.4473E 05 0.02170 -9.5154E 05 0.02220 -9.E768E 05 0.02270 -9.6290E 05 0.02320 -9.6799E 05 0.02370 -9.7239E 05 0.02420 -9.7614E 05 0.02470 -9.79275 05 0.02700 -9.9082E 05 0.03200 -9.7793E 05 0.03700 -9.7748E 05 TIHE VALUE FUNCTION 0.00030 -3.B818E 04 0.00080 -4.5462E 04 0.00130 -7.4140E 04 0.00180 -1.0458E 05 0.00230 -1.3291E 05 0.00280 -1.6046E 05 0.00390 - 1.7938E 05 0.00380 -2.0950E 05 0.00430 -2.446lE 05 0.00480 -2.8235E 05 0.00530 -3.19llE 05 0.00580 -3.6491E 05 0.00630 -3.0992E 05 0.00680 -4.2411E 05 0.00730 -4.5742E 05 0.00780 -4.9000E 05 0.00830 -5.1066E 05 0.00880 -5.4076E 05 0.00930 -6.6990E 05 0.00980 -1.9788E 05 0.01030 -6.2447E 05 0.01080 -6.5026E 05 0.01130 -6.7507E 05 0.01180 -6.9870E 05 0.01230 -7.2143E 05 0.01280 -7.4285E 05 0.01330 -7.6316E 05 0.01380 -7.8237E 05 0.01430 -8.0024E 05 0.01480 -8.1674E 05 0.01530 -8.3303E 05 0.01580 -8.4797E 05 0.01630 -8.6228E 05 0.01680 -8.7637E 05 0.01730 -8.QOOIE 05 0.01780 -9.0223E 05 0.01830 -8.9588E 05 0.01880 -9.O572E 05 0.01930 -S. 1481E 05 0.01980 -9.2366E 05 0.02030 -9.3183E 05 0.02080 -9.3923E 05 0.02130 -9.4612E 05 0.02180 -9.5282E 05 0.02230 -9.5883E 05 0.02280 -9.6398E 05 0.02330 -9.6893E 05 0.02380 -9.7319E 05 0.02430 -0.7681E 05 0.02480 -0.7961E 05 0.02800 -9.9553E 05 0.03300 -9.7755E 05 0.03800 -9.7768E 05 TIHE VALUE FUNCTION 0.00040 -2.4544E 04 0.00090 -5.0324E 04 . 0.00140 -8.0497E 04 0.00190 -1.1037E 05 0.00240 - 1.3044E 05 0.00290 -1.6595E 05 0.00340 -1.8502E 05 0.00390 -2.1612E 05 0.00440 -2.5225E 05 0.00490 -2.8976E 05 0.00540 -3.2633E 05, 0.00590 -3.6199E 05 0.00640 -3.9686E 05 0.00690 -4.3080E 05 ' 0.00740 -4.6400E 05 0,..00790

-4.9637E 05 0.00840 -5.1676E 05 0.00890 -5.4666E 05 0.00940 -5.7556E 05 0.00990 -6.0317E 05 0.01040 -6.2975E 05 0.01090 -6.5535E 05 0.01146 -6.7992E 05 0.01190 -7.0339E 05 0.01240 -7.2587E 05 0.01290 -7.4704E 05 0.01340 -7.670dE 05 0.01390 -7.8605E 05 0.01440 -8.0367E 05 0.01490 -8.2005E 05 0.01540 -8.3593E 05 0.01590 -8.5043E 05 0.01640 -8.6494E 05 0:01690 -8.7884E 05 0.01740 -8.9212E 05 0.01790 -8.8782E 05 0.01840 -8.9791E 05 0.01890 -9.0762E 05 0.01940 -9.1666E 05 0.01990 -9.2534E 05 0.02040 -9.3339E 05 0.02090 -9.4041E 05 0.02140 -9.4752E 05 0.02190 -9.MOBE 05 0.02240 -9.6994E 05 0.02290 -9.6503E 05 0.02340 -9.6984E 05 0.02390 -9.7396E 05 0.02440 -9.7746E 05 0.02490 -0.802lE 05 0.02900 -9.8077E 05 0.03400 -9.7743E 05 0.03900 -9.7797E 05 TABLE 6.A-1 (SHEET 20 OF 0.04200 -9.7963E 05 0.04700 -9.86C5E 05 o.os200 -1. ooe7~ 06 0.05700 - 1 .048GE 06 0.06200 -1.0835E 06 0.06700 -1.0971E 06 0.07200 -1.0936E 06 0.07700 -1.0876E 06 0.08200 -1.0848E 06 0.08700 -1.O831E 06 0.09200 -1.C799E 06 0.09700 - 1 .0752E 06 0.16200 -1.0698E 06 0.10700 -1.0645E 06 0.11200 -1.0589E 06 0.11700 -1.0537E 06 0.12200 -1.0487E 06 0.12700 -1.0435E 06 0.13200 -1.03e1~ 06 0.13700 -1.0326E 06 0.14200 -1.027OE 06 0.14700 -1.0216E 06 0.15200 -1.01G5E 06 0.15700 -1,012OE 06 0.16200 -1.0095E 06 0.16700 -1.0087E 06 0.17200 -1.0089E 06 0.17700 -1.009lE 06 0.18200 -1.0085E 06 0.18700 -1.0064E 06 0.19200 -1.0032E 06 0.19700 -9.e9llE 05 0.20400 -9.9268E 05 0.21400 -9.8371E 05 0.22400 -9.7424E 05 0.23400 -9.6549E 05 0.24400 -9.56e3E 05 0.25400 -9.4974E 05 0.26400 -9.4478E 05 0.27400 -9.4028E 05 0.28400 -9.3463E 05 0.29400 -9.2865E 05 0.30400 -9.2312E 05 0.31400 -9.1930E 05 0.32400 -9.1622E 05 0.33400 -9.13C7E 05 0.34400 -9.0886E 05 0.35400 -9.0427E 05 0.36400 -9.0004E 05 0.37400 -8.9700E 05 0.38400 -8.9541E 05 0.39400 -8.9367E 05 0.40400 -8.9189E 05 0.41400 "8.8964E 05 0.42400 -8.8728E 05 0.43400 -8.8505E 05 0.44400 -8.8327E 05 0.45400 -8.821OE 05 0.46400 -8.8068E 05 0.47400 -8.7820E 05 0.48400 -8.7501E 05 0.49400 -8.7181E 05 TABLE 6.A-12 (SHEET 21 OF 28) TI HE FUNCI 1014 NUMBER s ( 11) FUNCTION OES :RI PT1 ON = FORCINQ FUNCTlON AT NODE 43 00>02S 0. NUtlBCR OF ABSC I S5AE = ( 577) FUtJCTIOI4 SCALE FACTOR = f 1 .OOOOE 00) TIME VALUE FUNCTl ON 0. 1.3900E 01 0.00050 -1.3280E 03 0.00100 -2.4347E 03 0.00150 -3.8323E 03 0.00200 -5.1314E 03 0.00250 -6.3638E 03 0.00300 -7.5790E 03 0.00350 -8.4362E 03 0.00400 -9.8553E 03 0.00450 -1.1485E 04 0.00500 -1.3135E 04 0.00550 -1.4745E 04 0.00600 -1.6313E 04 0.00650 -1.7844E 04 0.00700 -1.9339E 04 0.00750 -2.0802E 04 0.00800 -2.2223E 04 0.00850 -2.3111E 04 0.00900 -2.4427E 04 0.00950 -2.5691E 04 0.01000 -2.6904E 04 0.01050 -2.8069E 04 0.01100 -2.9192E 04 0.01150 -3.0262E 04 0.01200 -3.12966 04 0.01250 -3.22746 04 0.01300 -3.3201E 04 0.01 350 -3:4083E 04 0.01400 -3.4907E 04 0.01450 -3.5675E 04 O.OlCOO -3.6395E 04 0.01550 -3.7093E 04 0.01600 -3.7733E 04 0.01650 -3.8357E 04 0.01700 -3.89BGE 04 0.01750 -3.9584E 04 0.01800 -3.P331E 04 0.01850 -3.9781E 04 0.01900 -4.0204E 04 0.01950 -4.0601E 04 0.-02000 -4.0978E 04 0.02050 -4.1328E 04 0.02100 -4.1636E 04 0.02150 -4.1947E 04 0.02200 -4.2229E 04 0.02250 -4.2482E 04 0.02300 -4.2704E 04 0.02350 -4.2910E 04 0.02400 -4.3087E 04 0.02450 -4.3236E 04 0.02500 -4.3356E 04 0.03000 -4.3288E 04 0.03500 -4.3209E 04 TIME VALUE FUNCTION 0.00010 -2.8G10E 02 0.00060 -1.5622E 03 0.00110 -2.6904E 03 0.00160 -4.1005E 03 0.00210 -5.3813E 03 0.00260 -6.6071E 03 0,00310 -7.8227E 03 0.00360 -8.7021E 03 0.00410 -1.0166E 04 0.00460 -1.l819E 04 0.00510 -1.3460E 04 0.00560 -1.6061E 04 0.00610 -1.6623E 04 0.00660 -1.8148E 04 0.00710 -1.9634E 04 0.00760 -2.1089E 04 0.00810 -2.2024E 04 0.00860 -2.3377E 94 0.00910 -2.4687E 04 0.00960 -2.5937E 04 . 0.01010 -2.7143E 04 0.01060 -2.8298E 04 0,01110 -2.9403E 04 0.01160 -3.0475E 04 0.01210 -3.1499E 04 0.01260 -3.2466.E 04 0.01310 -3.3383E 04 0.01360 -3.4253E 04 0.01410 -3.5065E 04 0.01460 -3.5823E 04 0.01510 -3.6538E 04 0.01560 -3.7220E 04 0.01610 -3.7860E 04 0.01660 -3.8488E 04 0.01710 -3.9100E 04 0.01760 -3.9666E 04 0.01810 -3.9424E 04 0.01860 -3.9068E 04 0,01910 -4.0285E 04 0.01960 -4.0677E 04 0.02010 -4.1050E 04 0.02060 -4.1390E 04 0.02110 -4.1699E 04 0.02160 -4.2007E 04 0.02210 -4.2283E 04 0.02260 -4.2529E 04 0.02310 -4.2747E 04 0.02360 -4.2948E 04 0.02410 -4.3119E 04 0.02460 -4.32E3E 04 0.02600 -4.3601E 04 0.03100 -4.3233E 04 0.03600 -4.3213E 04 TIME VALUE FUNCTION 0.00040 -1.0849E 03 0.00090 -2.2246E 03 0.00140 -3.5584E 03 0.00190 -4.a787E 03 0.00240 -6.1198E 03 0.00290 -7.3358E 03 0.00340 -6.1787E 03 0.00390 -P.S537E 03 0.00440 -1.llSlE 04 0.00490 -1.2809E 04 0.00540 -1.4425E 04 0.00590 - 1.6002E 04 0.00640 -1.7543E 04 0.00690 -1.9044E 04 0,00740 -2.O511E 04 0.00790 -2.1942E 04 0.00840 -2.2843E 04 0.00890 -2.4165E 04 0.00940 -2.5442E 04 0.00990 -2.6663E 04 0.01040 -2.7838E 04 0.01090 -2.6969E 04 0.01140 -3.0056E 04 0.01190 13.1093E 04 0.01240 -3.2087E 04 0.01290 -3.3023E 04 0.01340 -3.3907E 04 0.01390 -3.4747E 04 0.01440 -3.5526E 04 0.01490 -3.6250E 04 0.01540 -3.6952E 04 0.01590 -3.7593E 04 0.01640 -3.8234E 04 0.01690 -3.6849E 04 0.01740 -3.S436E 04 0.01790 -3.9246E 04 0.01840 -3.S692E 04 0.01690 -4.0121E 04 0.01940 -4.O521E 04 0,01990 -4.9905E 04 0.02040 -4.1260E 04 0.02090 -4.1571E 04 0.02140 -4.1885E 04 0.02190 -4.2175E 04 0.02240 -4.2434E 04 0.02290 -4.2659E 04 0.02340 -4.2871E 04 0.02390 -4.3034E 104 0.02440 -4.3208E 04 0.02490 -4.3330E 04 0.02900 -4.S365E 04 0.03400 -4.3207E 04 0.03900 -4.S231E 04 TABLE 6.A-12 (SHEET 22 OF 28)

TABLE 6.A-12 (SHEET 23 OF 28) TltlE FUNCTION NUMBER , = ( 121 cn 3: I-' h) FlttJCT 1014 OESCR l PT l ON = ( FORCING FUNCTION AT NODE 3 OOOOOOOOOVOOOOOOOOOO0000000)

NU~~F~R or PDSC I SSAE = ( 577) FUNLl16N SCALE FACTOR

= ( 1.0000E 00) T l HE VALUE 0. 0.00050 0.00100 0.0(?!50 0. (ltidoo 0.002 50 0.00~00 0.00:~50 0.00400 0.00450 0.00500 0.00550 0.00600 0.OOF.SO 0.00 /OCl 0.00750 0.00U00 0.00R50 0. OOPliO 0.00950 ? 0.01000 0.01050 0.01 100 0.01 150 0.01200 0.01250 0.01300 0.01350 0.01400 0.01450 0.01500 0.015:10 0.01G00 0.01C50 0.01700 0.01760 9. OlU00 0.01850 0.01900 0.01050 0.02000 0.020;0 0.02100 0.02160 0.02200 0.02250 0.02300 0.02330 0.02400 0.02459 0.02500 0.03000 0.03500 FUNCTION 1.1313E 04 1.1312E 04 1.1313E 04 1.1313E 04 1.1313E 01 1.1313E 04 1.1314E 04 1.1315E 04 1.131C;E 04 1.1319E 04 1.1324E 04 1.1333E 04 1.134PE 04 1 .1360E 04 1.1385E 04 1.1418E 04 I. 1465E 04 1.1527E 04 1 .1607E 04 1 . 1708E 04 1 .1833E 04 1.1087E 04 1.2167E 04 1.2384E 04 l.2642E 04 1 .293FE 04 1.327GE 04 1.. 36GGF. 04 1.4109E 04 1 .4G03E 04 1.5153E 04 1. '3760E 04 1.6430E 04 1.71GOE 04 1.7954E 04 1.ECtSE 04 1.973GE 04 2.0722E 04 2.1774E 04 2.2895E 04 2.4076E 04 2.5321E 04 2.6621E 04 2.7972E 04 2.9375E 04 3.0817E 04 ' 3.2294E 04 3.3804E 04 3.5333E 04 3.6880E 04 3.8433E 04 5.3102E 04 6.4040E 04 TIHE VALUE 0.00010 0.000EO 0.001 10 0.00160 0.00210 0.00260 0.00310 0.00360 0.00410 0.00460 0.00510 - 0.00560 0.00610 0.00660 0.00710 0.00760 0.00810 0.00860 0.00910 0.00960 0.01010 0.01060 0.01110 0.01 160 0.01210 0.01260 0.01310 0.01i60 0.01410 0.01460 0.01510 0.01560 0.01610 0.01 660 0.01710 0.01760 0.01810 0.01860 0.01910 0.01960 0.02010 0.02060 0.021 10 0.02160 0.02210 0.02260 0.02310 0.02360 0.02410 0.02460 0.02600 0.03100 0.03600 FUNCT I ON 1.1313E 04 1.1312E 04 1.1313E 04 1.1313E 04 1.1313E 04 $. 1313E 04 1.1314E 04 1.131SE 04 1.1318E 04 1.1319E 04 1.1325E 04 1.1335E 04 1. 1346E 04 1.1364E 04 1.1389E 04 1.1427E 04 1.1476E 04 l.lS42E 04 1.162GE 04 1 . 1730E 04 1.1860E 04 1 .2020E 04 1.2211E 04 1.2431E 04 1.2697E 04 1.3000E 04 1.3350E 04 1.3749E 04 1.4201E 04 1 ,470'JE 04 1.5269E 04 1.5888E 04 1.6572E 04 1.7313E 04 1.8121E 04 1.8990E 04 1.9927E 04 2.0926E 04 2.1995E 04 2.3126E 04 2.4323E 04 2. 5578E 04 2. C886E 04 2.8253E 04 2.9661E 04 3.1110E 04 3.2593E 04 3.4109E 04 3.5642E 04 3.7190E 04 4.1557E 04 5.5843E 04 6.5070E 04 TlHE VALUE 0.00020 0.00070 0.00120 0.00170 0.00220 0.00270 0.00320 ' 0.00370 0.00420 0.00470 0.00520 0.00570 0.00620 0.00670 0.00720 0.00770 0.00820 0.00870 0.00920 0.00970 0.01020 0.01070 0.01120 0.01170 0.01220 0.01270 0.01320 0.01370 0.01420 0.0147CI 6. or 520 0.01570 0.01620 0.01670 0.01720 0.01770 0.01820 0.01870 O.Ot920 0.01970 0.02020 0.02070 0.02120 0.02170 0.02220 0.02270 0.02320 0.02370 0.02420 0.02470 0.02700 0.03200 0.03700 FUNCT l ON 1.1312E 04 1.1312E 04 1.1313E 04 1.1313E 04 1.1313E 04 , 1.1314E 04 1.1315E 04 1.1315E 04 1.1317E 04 1.13PlE 04 1 .1327E 04 1.1336E 04 1.1348E 04 1.13E9E 04 1.1398E 04 1.1435E 04 1.14e9E 04 1.1556E 04 1.1644E 04 1.1754E 04 1.1888E 04 1.2053E 04 1.2253E 04 1.2483E 04 1 . P756E 04 1.3067E 04 1.3429E 04 1.3835E 04 1.4298E 04 1.4816E 04 1.5388E 04 t.6022E 04 1.6715E 04 1.7468E 04 1.B2SlE 04 l.9172E 04 2.0120E 04 2.1134E 04 2.2212E 04 2.3359E 04 2.4567E 04 2.5834E 04 2.7156E 04 2.8530E 04 2.9948E 04 3.1404E 04 3.2894E 04 3.4412E 04 3. E95OE 04 3.7501E 04 4.4611E 04 5.8451E 04 6.5688E 04 TlnE VALUE 0.00030 0.00080 0.00130 0.00180 0.00230 0.00280 0.00330 0.00380 0.00430 0.00480 0.00530 0.00580 0.00630 0.00680 0.00730 0.00780 0.00830 0.00880 0.00930 0.00980 0.01030 0.01030 0.01130 0.01180 0.01230 0.01280 0.01330 0.01380 0.01430 0.0t48O 0.01530 0.01500 0.01630 0.016CO 0.01730 0.01 780 0.01830 0.01880 0.01930 0.01980 0.02030 0.02080 0.02130 0.02180 0.02230 0.02280 0.02330 0.02380 0.02430 0.02480 0.02800 0.03300 0.03800 FUNCTION 1.1312E 04 1.1313E 04 1.1313E 04 1.1313E 04 1.1313E 04 1.1313E 04 1.1314E 04 1.1316E 04 1.1319E 04 1.13236 04 I. 1331E 04 1.1340E 04 ' 1.1356E 04 1.1379E 04 1.14108 04 1.1453E 04 1.1513E 04 1.15898 04 1.1685E 04 1.1804E 04 1.1952E 04 1.2128E 04 1.2337E 04 1.2589E 04 1.2674E 04 1.3206E 04 1.3583E 04 1.4014E 04 t.4499E 04 1.5039E 04 1.8636E 04 1.6293E 04 t.7007E 04 1.7789E 04 1.8635E 04 1.9546E 04 2.0318E 04 2.1558E 04 2.2664E 04 2.3837E 04 2.5065E 04 2.6356E 04 2.7700E 04 2.9092E 04 3.0525E 04 3.1995E 04 3.3499E 04 3.5028E 04 3.6572E 04 3.8123E 04 0.0320E 04 6.2589E 04 6.5695E 04 t-' IQ a, & TABLE 6.A-12 (SHEET 24 OF 28) 0.04200 6.2807E 04 0.04300 0.04700 6.1792E 04 0.04800 0.05200 3.6294E 04 0.05300 0.05700 2.1760E 04 0. 05800 0.06200 1 .4937E 04 0.06300 0.06700 1.

S061E 04 0.06800 0.07200 3.2886E 04 0.07300 0.07700 6.1204E 04 0.07800 0.08200 7.1605E 04 0.08300 0.08700 8.J874E 04 0.08800 0.09200 9.1546E 04 0.09300 0.09700 8.8652E 04 0.09800 0.10200 7.8317E 04 0.10300 0.10700 6.6638E 04 0.10800 0.11200 5.3426E 04 0.11300 0.11700 4.247.6E 04 0.11800 0.12200 3.5505E 04 0.12300 0.12700 3.3004E 04 0.12800 0.13200 3.4436E 04 0.13300 0.13700 3.8897E 04 0.13800 0.14200 4.4734E 04 0.14300 0.14700 5.0629E 04 0.14800 0.15200 5.5298E 04 0.15300 0.16700 5.8235E 04 0.15800 0.16200 6.9439E 04 0.16300 0.16700 5.9079E 04 0.16800 0.17200 5.7629E 04 0.17300 0.17700 5.5367E 04 0.17800 0.18200 5.3036E 04 0.18300 0.18700 6.1209E 04 0.18800 0.19200 4.9706E 04 0.19300 0.19700 4.Q030E 04 0.19800 0.20400 4.9525E 04 0.20600 0.21400 6.2210E 04 0.21600 0.22400 6.5713E 04 0.22600 0.23400 5.8214E 04 0.23600 0.24400 S.Bi34E 04 0.24600 0.25400 5.8845E 04 0.25600 0.26400 5.7946E 04 0.26600 0.27400 6.6820E 04 0.27600 0.28400 6.5509E 04 0.28600 0.29400 5.E561E 04 0.29600 0.30400 5.6720E 04 0.30600 0.31400 5.7827E 04 0.31600 0.32400 5 8926E 04 0.32600 0.33400 6.739OE 04 0.33600 0.34400 6.415GE 04 0.34600 0.35400 4.8374E 04 0.35600 0.36400 4.S612E 04 0.36600 0.37400 4.3903E 04 0.37600 0.38400 4.4549E 04 0.38600 0.39400 4.6552E 04 0.39600 0.40400 4.8852E 04 0.40600 0.41400 5.0899E 04 0.41600 0.42400 5.1978E 04 0.42600 0.43400 5.2485E 04 0.43600 0.44400 6.1931E 04 0.44600 0.45400 5.0834E 04 0.45600 0.46400 4.9384E 04 0.46600 0.47400 4.7738E 04 0.47600 0.48400 4.6286E 04 0.48600 0.49400 4.4619E 04 0.49600 TABLE 6.A-12 (SHEET 25'0~ 28) FUNCTION DESCRIPTION

= I FORCING FUNCTION AT NUDE 2 00>02S 0. 9776. 6 ) NUliL'Ci.

OF I.B.P,C I SSAE I: ( 677) FUNCTIDIJ SCALE FACTOR

= ( 1.0000E 00) T IHE VAL UE 0. 0.00050 0.001rJo 0.00150 0.00;r00 0.00250 0.00300 0.00380 0. 00.100 0.00450 0. 00f.00 0.005C.O 0.00G00 0.00650 0.00700 0.00750 0.00000 0.00850 0.00000 ! 0.00550 0.01000 0.01050 0.01 100 0.01 150 O.OlirO0 0.012tio 0.01300 0,01350 0.01400 0.014'50 0.01500 0.01850 0.01GOu 0.01C.:~o 0.01700 O.P17SG 0.0 1800 O.Ol8SO 0.01900 0.01050 0.02000 0.02050 0.02100 0.02150 0.02200 0.02250 0.02300 0.02350 0.02400 0.02450 0.02:.00 0.03000 0.03500 FUNCT 1 ON 9.7758E 03 9.0718E 03 1.03745: 04 1 . I5OPE 04 1.350YE 04 1.6558E 04 2.0738E 04 2.63675: 04 3.5183E 04 4.6793E 04 5.088BE 04 7.1085E 04 8.3443E 04 9.L9tilE 04 1.0858E 05 1.2165E 05 1.3d70E 05 1 .4SP6E 05 1.5716C 05 1.7052E 05 1. P402E 05 1.97G4E 05 2. l14SE 05 2.2536E 05 2. '2':'56E 05 2.5378E 05 2.6811E 05 2.8258E 05 2.9706E 05 3.1150E 05 3.2600E 05 3.4067E 05 3.6520E 05 3.69CEE 05 3.8400E 05 3.99GOE 05 4.0737E 05 4.2105E 05 4.3475E 05 4,4812E 05 4.6113E 05 4.73GFE 05 4.es~Cr 05 4.9746E 05 5.0873E 05 5.1949E 05 5.29Gf;E 05 5.3942E 05 5.4849E 05 5.5701E 05 5.6495E 05 6.0142E 05 6.0774E 05 TIME VALUE 0.00010 0.00060 0.001 10 0.00160 0.00210 0.00260 0.00310 0.00360 0.00410 0.00460 0.005\0 0.00560 .O. 006 10 0.006CO 0.00710 0.00760 0.00810 0.00860 0.00910 0.00960 0.01010 0.01060 . 0.01110 0.01160 0.01210 0.01260 0.01310 0.01360 0.01410 0.01460 0.015IO 0.01560 0.01610 0.01660 0.01710 0.01760 0.01810 0.01860 0.01910 0.019GO 0.02010 0.02060 0.021 10 0.02160 0.02210 0.02260 0.02310 0.02360 0.02410 0.02460 0.02600 0.03100 0.03600 FUNCTION 9.77i6E 03 9.9309E 03 1.0544E 04 1.1832E 04 1.4031E 04 1.7300E 04 2.1711E 04 2.7863E 04 3.7332E 04 4.9207E 04 6.1318E 04 7.35488 04 6.59318 04 9.8465E 04 1.1 125E 05 1.2426E 05 1.3333E 05 1 .4650E 05 1.5904E 05 1.7321E 05 1.8674E 05 2.0039E 05 2.1418E 05 2.2822E 05 2.4241E 05 2.6GG5E 05 2.7101E 05 2.8549E 05 P. 9905E 05 3.1439E 05 3.2894E 05 3.4354E 05 3.5814E 05 3.7289E 05 3.87718 05 4.0225E 05 4.1012E 05 4.2382E 05 4.37466 05 4.5075E 05 4.63G8E 05 4.76136 05 4.8808E 05 4.9976E 05 5.1093E 05 6.2157E 06 6.3167E 05 5.4125E 05 5.5024~ 05 5.6864E 05 5.7971E 05 6.0477E 05 6.0574E 05 TIME VALUE 0.00020 0.00070 0.001i0 0.00170 0.00220 0.00270.

0.00320 0.00370 0.00420 0.00470 0.00520 0.00570 0. OOG20 0.00670 0.00720 0.00770 0.00820 0.00870 0.00920 0.00970 0.01020 0.01070 0.01 120 0.01170 0.01220 0.01270 0.01320 0.01370 0.01420 0.01470 0.01620 0.01570 0.01620 0.01670 0.01720 0.01770 0.01820 0.01870 0.01920 0.01970 . 0.02020 0.02070 0.02120 0.02170 0.02220 0.02270 0.02320 0.02370 0.02420 0.02470 0.02700 0.03200 0.03700 FUNCTION 9.7836E 03 1 .0009E 04 1.0740E 04 1.2192E 04 1 .4597E 04 1 .8093E 04 2.27316 04 2.9492E 04 3. 9559E 04 5.1683E 04 6.3752E 04 7.6004E 04 8.6428E 04 1.01 03E 05 1.1384E 05 1.26eGE 05 1.3598E 05 1.4915E 05 1 .6250E 05 1.78928 05 1.8946E 05 2.0311E 05 2.1700E 05 2.3105E 05 2.4552E 05 2.5953E 05 2.7387E 05 2.8838E 05 3.02848 05 3.8721E 05 3.3187E 05 3.4G5OE 05 3.6112E 05 3.7573E 05 3.9073E 05 4.0538E 05 4.1285E 05 4. P658E 05 4.40071 05 4 .C335E 05 4.6621E 05 4.7851E 05 4.9044E 05 5.0204E 05 5.1311E 05 5.2356E 05 6.3363E 05 5.4311E 05 5.5197E 05 5.6025E 05 5.9251E 05 6.0744E 05 6.0287E 05 TINE VALUE 0.00030 0.00080 0.00130 0.00180 0.00230 0.00280 0.00330 0.00380 0,00430 0.00480 0.00530 0.00580 0.00630 0.00680 0.00730 0.00780 0,00830 0.00880 0.00930 0.00980 0.01030 0.01080 0.01130 0.01180 0.01230 0.01280 0.01330 0.01380 0.01430 0.01480 0.01530 0.01580 ' 0.01630 0.01680 0.01730 0.01780 0.01830 0.01880 0.01930 0.01980 0.02030 0.02080 0.02130 0.02180 0.02230 0.02260 0.02330 0.02380 0.02430 0.02480 0.02800 0.03300 0.03800 FUNCT l ON 9.8002E 03 1.OlOBE 04 1.09536 04 1.2591E 04 1.520GE 04 1.8929E 04 2.3803E 04 3.1260E 04 4.1979E 04 6.4041E 04 6.6187E 04 7.84766 04 9.0934E 04 1 .0358E 05 1.1643E 05 1.2951E 05 1.3860E 05 t.518lE 05 I .6517E 05 1.7863E 05 1.9214E 05 2.0589E 05 2.1981E 05 2.3385E 05 2.4808E 05 2.6237E 05 2.76778 05 2.9127E 05 3.0573E 05 3.2013E 05 3.34856 05 3.4940E 05 3.6406E 05 3.7885E 05 3.9365E 05 4.0804E 05 4.1557E 03 4.2933E 05 4.4277E 05 4.5596E 05 4.6872E 05 4.8093E 05 4.92796 05 5.0425E 05 5.1526E 05 6.2563E 05 5.355PE 05 6.4492E 05 5.5367E 05 5.6176E 05 6.035GE 05 6.0881E 05 5.9974E 05 TINE VALUE 0.00040 0.00090 0.00140 0.00190 0.00240 .O. 00290 0.00340 0.00390 0.00440 0.00490 0.00540 0.00590 0,00640 0-. 00690 0.00740 0.00790 0.00840 0.00890. 0.00940 0.00990 0.01040 0.01090 0.01 140 0.01190 0.01240 0.01290 0.01340 0.01390 0.01440 0.01490 0.01540 0.01590 0.01640 0.01690 0.01740 0.01790 0.01840 0.01890 0.01940 0.01090 0.02040 0.020?0 0.02140 0.02190 0.02240 0.02290 0.02340 0.02390 0.02440 0.02490 0 02900 0.03400 0.03900 FUNCTION 9.8292E 03 1.0228E 04 1.1219E 04 1.3029E 04 1.5859E 04 1.98lOE 04 2.6013E 04 3.3157E 04 4.4385E 04 6.6460E 04 6.0637E 8.0955E 04 04 9.3448E 04 . 1.061ZE 05 1.1904E 05 ' 1.3214E 05 1.4123E 05 1.6447E 05 1.6784E 05 1.8129E 05 1.9489E 05 2.0867E 05 2.2262E 05 2.3671E 05 2.5095E 05 2.652GE 05 2.7965E 05 2.9416E 05 3.0861E 05 3.23076 05 3.3770E 05 3.5218E 05 3.6694E 05 3.8171E 05 3.9639E 05 4.0476E 05 4.1826E 05 4.3203E 05 4.4545E 05 4.5856E 05 4.7120E 05 4.0327E 05 4.9513E 06 6.0651E 05 5.1739E 05 6.27676 05 5.3749E 05 6.4674E 05 5.6535E 05 1.6338E 05 6.9738E 05 6.0875E 05 6.9612E 05 t-' m n m TABL (SHEET 0.04200 0.04700 0.05200 0.05700 0.06200 0.06700 0.07200 0.07700 0.08200 0.08700 0.09200 0.09700 0.10200 0.10700 0.11200 0.11700 0.12200 0.12700 0.13200 0.13700 0.14200 0.14700 0.15200 0.15700 0.16200 0.16700 0.17200 0.17700 0.18200 0.18700 0.19200 0.19700 0,20400 0.21400 0.22400 0.23400 0.24400 0.25400 0.26400 0.27400 0.28400 0.29400 0.30400 0.3l400 0.32400 0.33400 0.34400 0.35400 0.36400 0.37400 0.38400 0,39400 0.40400 0.41400 0.42400 0.43400 0,44400 0.45400 0.46400 0.47400 0.48400 0.49400 TABLE 6.A-12 TI ME FUNCT I OIJ NUMBER = ( 14) (SHEET 27 OF 28) FUNCTION DESCRIPTION ( FORCINO FUNCTION AT NODE 1 00>02S 0. NU11BEE 01: PeSCI SSAE 5 ( 577) FUNCTION SCALE FACTOR

= ( 1.0000E 001 TlHE VALUE 0. 0.00050 0.00100 0.00150 0.00200 0.00250 0.00300 0.00350 0.00400 0.00450 0.00500 0.00550 0.00600 0.00650 0.00700 0.00750 0.00800 0.00050 0.00300 0. OOYZJO 0.01000 0.01050 0.01 100 0.01 150 0.01200 0.01250 0.01 300 0.01350 0.01400 0.01450 0.01tioo 0.015'iO 0.01GOO 0.01650 0.01700 0.01750 0.01~~00 '0.01U50 0.01R00 0.01950 0.02000 0.02050 0.02100 0.02150 0.02200 0.02250 0.02300 0.02350 .0.02400 0.02450 0.02500 0.03000 0.03500 FUNCT I PN - 1.4767E 04 - 1.4539E 04 -1.3366E 04 - 1.0726E 04 -6.1022E 03 8.8200E 02 1.0376E 04 2.3074E 04 4.3004E 04 6.9184E 04 9.6042E 04 1.225DE 05 1.4882E 05 1.7466E 05 2.0012E 05 2.2524E 05 2.4970E 05 2.6381E 05 2.8671E 05 3.0871E 05 3.2990E 05 3.5026E 05 3.6994E 05 3.8875E 05 4.0700E 05 4.2422E 05 4.4058E 05 4.5617E 05 4.7069E 05 4.8417E 05 4.9689E 05 5.0922E 05 5.2052E 05 5.3161E 05 5.4281E 05 5.5311E 05 5.4762E 05 5.5b47E 05 5.6284E 05 5.6981E 05 5.7634E 05 5.8234E 05 5.0757e 05 5.9286E 05 5.9761E 05 6.0178E 05 6.0543E 05 6.0876E 05 6. t15OE 05 6.1370E 05 6.1540E 05 6.0781E 05 5.9923E 05 FUNCT I ON - 1.4763E 04 - 1.4400E 04 -1.2970E 04 -9.9759E 03 -4.9013E 03 2.5779E 03 1.2575E 04 2.6451E 04 4.7861E 04 7.4583E 04 1.01 38E 05 1.2786E 05 1.5402E 05 t. 7980E 05 2.0518E 05 2.3019E 05 2.4498E 05 2.6843E 05 2.9122E 05 3.1301E 05 3.34066 05 3.64266 05 '3.7366E 05 3.9250E 05 4.1054E 05 4,2760E 05 4.4379E 05 4.5918E 05 4.73478 05 4.8675E 05 4,9942E 05 5,1144E 05 5.2277E 05 5.3394E 05 5.4483E 05 5.5402E 05 5.4924E 05 5.5700E 05 5. G426E 05 5.7112E 05 5.7759E 05 6.83408 05 5.88668 05 5.9388E 05 5.9849E 05 6.0254E 05 6.0614E 05 6.09356 05 6.1198E 05 6.1408E 05 6.1 885E 05 6.0537E 05 5.9792E 05 TlME VALUE 0.00020 0.00070 0.00120 0.00170 0.00220 0.00270 0.00320 0.00370 0.00420 0.00470 0.00520 0.00570 0.00620 0.00670 0.00720 0.00770 0.00820 0.00870 0.00920 0.00970 0.0t020 0.01070 0.01120 0.01170 0.01220 0.01270 0.01320 0.01370 0.01420 0.01470 Q.01520 0.01570 0.01620 0.01670 0.01720 0.01770 0.01820 0.01870 0.01920 0.01970 0.02020 0.02070 ' 0.02120 0.02170 0.02220 0.02270 0.02320 0.02370 0.02420 0.02470 0.02700 0.03200 0.03700 FUNCT l ON -1.4746E 04 -1.4216E 04 -1.2513E 04 -9.1409E 03 -3.6048E 03 4.3781E 03 1.4872E 04 3.0139E 04 5.2988E 04 7.99e7E 04 1.0670E 05 1.3311E 05 1.59POE 05 1.8490E 05 2.1022E 05 2.3513E 05 2.4980E 05 9.7303E 05 2.9563E 05 3.1732E 05 3.3816E 05 3.6813E 05 3.7753E 05 3.9619E 05 4.1395E 05 4.3094E 05 4.4688E 05 4.6210E 05 4.7620E 05 4.8914E 05 5.0192E 05 5.1385E 05 5.2512E 05 8.3586E 05 5.4711E 05 5. f733E 05 5.5081E 05 5.5852E 05 5.655OE 05 5.7246E 05 6.7881E 05 5.8448E 05 5.8972E 05 5. B482E 05 5.9935E 05 6.0313E 05 6.0683E 05 6.0992E 05 6.1244E 05 6.14446 05 6.2146E 05 6.03e9~ 05 5.9652E 05 TIME VALUE 0.00030 0.00080 0,00130 0.00160 0.00230 0.00280 0.00330 0.00380 0.00430 0.00480 0.00530 0.00580 0.00630 0.00680 0.00730 0.00780 0.00830 0.00880 0.00930 0.00980 0.01030 0.01080 0.01130 0.01180 0.01230 0.01280 0.01330 0.01380 0.01430 0.01480 0.01530 0.01580 0.01630 0.01680 0.01730 0.01780 0.01830 0.01880 0.01930 0.01980 0.02030 0.02080 0.02130 0.02100, 0.02230 0.02280 0.02330 0.02380 0.02430 0.02480 0.02800 0.03300 0.03800 FUNCTION - 1.4707E 04 -1.39C6E 04 -1.1991E 04 -8.2176E 03 -2.2093E 03 6.2797E 03 1.7288E 04 3.4136E 04 5.8358E 04 8.5338E 04 1.1200E 05 1.3835E 05 1.6437E 05 1.9000E 05 2.1524E 05 2.4005E 05 2.5449E 05 2.776pE 05 3.0002E 05 3.2157E 05 3.4213E 05 3.6211E 05 3.8136E 05 3.9974E 05 4.1 745E 05 4.3415E 05 4.5002E 05 4.6501E 05 4.789OE 05 4.9175E 05 5.0447E 05 5.1609E 05 5.2734E 05 5.3842E 05 5.4914E 05 5.5872E 05 5.5234E 05 5.5995E 05 5.6697E 05 5.7378E 05 5.8001E 05 5.8555E 05 5.9076E 05 5.9576E 05 6.0019E 05 6.0392E 05 6.0750E 05 6.1047E 05 6.12886 05 6.1463E 05 6.2424E 05 6.0217E 05 5.9542E 05 TIME VALUE FUNCTION 0.00040 -1.4639E 04 0.00090 -1.3703E 04 0.00140 -1.1396E 04 0.00190 -7.2071E 03 0.00240 -7.1460E 02 0.00290 8.2780E 03 0.00340 2.0016E 04 0.00390 3.8425E 04 0.00440 6.3773E 04 0.00490 S.0691E 04 0.00540 1.1730E 05 0.00590 1.4359E 05 ' 0.00640 1.6953E 05 0.00690 I. 9506E 05 0.00740 2.2025E 05 0.00790 2.4493E 05 0.00840 2.5916E 05 0.00890 2.8214E 05 0.00940 3.0438E 05 0.00990 3.2569E 05 0.01040 3.4622E 05 9.01090 3.6605E 05 0.01140 3.8512E 05 0.01190 4.0340E 05 0.01240 4.2090E 05 0.01290 4.3742E 05 0.01340 4.5305E 05 0.01390 4.6787E 05 0.01440 4.8155E 05 0.01490 4.9434E 05 0.01540 5.0671E 05 0.01590 5.1803E 05 0.01640 5.2941E 05 0.01690 5.4034E 05 ' 0.01740 5.5075E 05 0.01790 5.4614E 05 0.01840 5.5391E 05 0.01890 5.6140E 05 0.01940 5.6840E 05 0.01990 5.7507E 05 0.02040 5.8118E 05 0.02090 5.8645E 05 0.02140 5.9181E 05 0.02190 5.9670E 05 0.02240 6.0099E 03 0.02290 6.0469E 05 0.02340 6.0814E 05 0.02390 6.1100E 05 0.02440 6.1331E 05 0.02490 6. t503E 05 0.02900 6.1039E 05 0.03400 6.0061E 05 0.03900 5.9447E 05 TABLE 6.A-3.2 (SHEET 28 OF 28) 5.9324E 05 5.9345E 05 6.0512E 05 6.3609E 05 6.6733E 05 6.8339E 05 6.8294E 05 6.7711E 05 6.7227E 05 6.6875E 05 6.6499E Of 6.5982E 05 6.5410E 05 6.4862E 05 6.4288E 05 6.3740E 05 6.3215E 05 6.2727E 05 6.2168E 05 6.1624E 03 6.1099E 05 6.0593E 05 6.0137E 05 5.9714E 05 5.9154E 05 5.9330E 05 5.9288E 05 5.9263E Of, 5.9202E 05 5.902GE 05 5.8758E 05 5.84248 05 6.7965E 05 5.7187E 05 5.6392E 05 6.5654E Of, 5.4957E 05 6.4305E 05 5.3821E 05 5.3433E 05 5.3001E 05 5.2C99E 05 5.2002E 05 5.1669E 05 5.1S78E 05 5.1151E 05 5.0787E 05 6.0462E 05 5.0103E 05 4.9745E 05 4.9533E 05 4.9371E 05 4.9235E 05 4.9014E 05 4.8833E 05 4.8637E 05 4.8457E 05 4.8381E 05 4.6257E 05 4.8058E 05 4.7819E 05 4.7560E 05 LSCS-UFSAR REV. 13 ATTACHMENT 6.B RECIRCULATION SYSTEM SINGLE-LOOP OPERATION

LSCS-UFSAR 6.B-i REV. 15, APRIL 2004 ATTACHMENT 6.B TABLE OF CONTENTS PAGE 6.B RECIRCULATION SYSTEM SINGLE-LOOP OPERATION 6.B-1 6.B.1 INTRODUCTION AND

SUMMARY

6.B-1 6.B.1.1 GE Analysis 6.B.1.2 SPC Analysis 6.B.2 MCPR FUEL CLADDING INTEGRITY SAFETY LIMITS 6.B-1 6.B.2.1 Core Flow Uncertainty 6.B-1 6.B.2.2 Core Flow Measurement During Single-Loop Operation 6.B-1 6.B.2.3 Core Flow Uncertainty Analysis 6.B-2 6.B.2.4 TIP Reading Uncertainty 6.B-3

6.B.3 MCPR OPERATING LIMIT 6.B-4 6.B.3.1 Abnormal Operational Transients 6.B-4 6.B.3.2 Feedwater Controller Failure - Maximum Demand 6.B-5 6.B.3.2.1 Identification of Causes and Frequency Classification 6.B-5 6.B.3.2.2 Sequence of Events and Systems Operation 6.B-5 6.B.3.2.3 Effect of Single Failures and Operator Errors 6.B-6 6.B.3.2.4 Core and System Performance 6.B-6 6.B.3.2.5 Barrier Performance 6.B-7 6.B.3.2.6 Radiological Consequences 6.B-7 6.B.3.3 Generator Load Rejection Without Bypass with RPT 6.B-8 6.B.3.3.1 Identification of Causes and Frequency Classification 6.B-8 6.B.3.3.2 Sequence of Events and System Operation 6.B-8 6.B.3.3.3 Results 6.B-9 6.B.3.3.4 Barrier Performance 6.B-10 6.B.3.3.5 Radiological Consequences 6.B-10 6.B.3.4 Recirculation Pump Seizure Accident 6.B-10 6.B.3.4.1 Identification of Causes and Frequency Classification 6.B-10 6.B.3.4.2 Sequence of Events and Systems Operations 6.B-10 6.B.3.4.3 Systems Operation 6.B-11 6.B.3.4.4 Core and System Performance 6.B-11 6.B.3.4.5 Results 6.B-11 6.B.3.4.6 Barrier Performance 6.B-12 6.B.3.4.7 Radiological Consequences 6.B-12 LSCS-UFSAR 6.B-ii REV. 15, APRIL 2004 6.B.3.5 Summary and Conclusions 6.B-12 6.B.4 OPERATING MCPR LIMIT 6.B-12 6.B.5 STABILITY ANALYSIS 6.B-14 6.B.6 LOSS-OF-COOLANT ACCIDENT ANALYSIS 6.B-14 6.B.6.1 Break Spectrum Analysis 6.B-14 6.B.6.2 Single-Loop MAPLHGR Determination 6.B-14 6.B.6.3 Small Break Peak Cladding Temperature 6.B-15 6.B.7 REFERENCES 6.B-16 LSCS-UFSAR 6.B-iii REV. 13 ATTACHMENT 6.B

LIST OF TABLES

NUMBER TITLE 6.B-1 Input Parameters and Initial Conditions for Transients and Accidents (Analysis of Initial Core) 6.B-2 Sequence of Events for Figure 6.B-3 (Typical) 6.B-3 Sequence of Events for Figure 6.B-4 (Typical) 6.B-4 Sequence of Events for Figure 6.B-5 (Typical, GE) 6.B-5 Summary of Event Results (Typical)

LSCS-UFSAR 6.B-iv REV. 13 ATTACHMENT 6.B LIST OF FIGURES

NUMBER TITLE 6.B-1 Illustration of Single Recirculation Loop Operation Flows 6.B-2 Main Turbine Trip With Bypass Manual Flow Control (Typical) 6.B-3 Feedwater CF With One-Pump Operation (Typical) 6.B-4 Load Rejection With One Pump Operation 6.B-5 Seizure of One Recirculation Pump (Typical) 6.B-6 Decay Ratio vs. Power Curve for Two-Loop and Single-Loop Operation (Typical, GE) 6.B-7 Uncovered Time vs. Break Area - LSCS Units 1 and 2 Suction Break LPCS/DG Failure LSCS-UFSAR 6.B-1 REV. 15, APRIL 2004 6.B RECIRCULATION SYSTEM SINGLE-LOOP OPERATION 6.B.1 INTRODUCTION AND

SUMMARY

Sections 6.B.2, 6.B.3, 6.B.

4, and 6.B.5 describe the GE methodology for the MCPR safety limit calculation and single loop op eration transient analyses. The transient analyses presented in this chapter are for a specific cycle, and are not re-performed for each reload.

6.B.1.1 GE Analysis

Single-loop operation at reduced power is highly desirable in the event recirculation pump or other component maintenance renders one loop inoperative. To justify single-loop operation, accidents and abnor mal operational transients associated with power operations, as presented in Se ction 6.3 and Chapter 15.0, were reviewed for the single loop case with only one pump in operation.

Increased uncertainties in the core total flow and TIP readings resulted in a 0.01 incremental increase in the MCPR fuel cladding integrity safety limit during single-loop operation. This 0.01 increase is re flected in the MCPR operating limit. No other increase in this limit is required because all abnormal operational transients are bounded by the rated power/flow analyses performed. The least-stable power/flow condition, achieved by tripping both recirculation pumps, is not affected by one-pump operation.

6.B.1.2 FANP Analysis To support operation with a single recirculation loop, the MCPR safety limit, pressurization transients, and slow flow excursions were evaluated in Reference 3.

During single loop operation (SLO), the uncertainties of some of the core parameters increase. As a result, the potent ial exists that the MCPR safety limit to support SLO is greater than the base case MCPR safety limit.

FANP will perform analyses each cycle to support or establish both the two-loop and single-loop MCPR safety limits. Cycle specific analyses determine the adder to the base case MCPR safety limit to provide the necessary protection to ensure that 99.9% of the rods do not experience boiling transition during an anticipated operational occurrence.

FANP evaluates single loop operation (SLO) to provide limits to support SLO. This evaluation can be found in References 2 and 3. FANP has addressed the LOCA analysis (Section 6.B.6) in Reference 2 for ATRIUM-9B fuel and in Reference 7 for ATRIUM-10 fuel. FANP provides MCPR and MAPLHGR limits to support single loop operation.

LSCS-UFSAR 6.B-2 REV. 14, APRIL 2002 6.B.2 MCPR FUEL CLADDING INTEGRITY SAFETY LIMIT Except for core total flow and TIP reading, the uncertainties used in the statistical analysis to determine the MCPR fuel cladding integrity safety limit are not

dependent on whether coolant flow is provided by one or two recirculation pumps. A 6% core flow measurement uncertaint y has been established for single-loop operation (compared to 2.5% for two-loop operation). As shown below, this value conservatively reflects the one standard deviation (one sigma) accuracy of the core flow measurement system documented in Reference 1. The random noise component of the TIP reading uncertainty was revised for single recirculation loop operation to reflect the operating plant test results given in Subsection 6.B.2.4. This revision resulted in a single-loop operation process computer uncertainty of 6.8% for initial cores. Comparable two-loop process computer uncertainty values are 6.3% for initial cores. The net effect of these two revised uncertainties is a 0.01 incremental increase in the required MCPR fuel cladding integrity safety limit.

6.B.2.1 Core Flow Uncertainty

6.B.2.2 Core Flow Measurement During Single-Loop Operation The jet pump core flow measurement system is calibrated to measure core flow when both sets of jet pumps are in forward flow; total core flow is the sum of the indicated loop flows. For single-loop operation, however, some inactive jet pumps will be backflowing. Therefore, the measured flow in the backflowing jet pumps must be subtracted from the measured flow in the active loop. In addition, the jet pump coefficient is different for reverse flow than for forward flow, and the

measurement of reverse flow must be modi fied to account for this difference.

For single-loop operation the total core fl ow is derived by the following formula:

Where C (= 0.95) is defined as the ratio of "Inactive Loop True Flow" to "Inactive Loop Indicated Flow," and "Loop Indicated Flow" is the flow indicated by the jet pump "single-tap" loop flow summers an d indicators, which are set to indicate forward flow correctly.

The 0.95 factor was the result of a conservative analysis to appropriately modify the single-tap flow coefficient for reverse fl ow. (NOTE: The LSCS value of the "C" coefficient is 0.78 (+/-0.078) at reactor operat ing conditions.) If a more exact, less conservative core flow is required, special in-reactor calibration tests would have to be made. Such calibration tests would involve calibrating core support plate P versus core flow during two-pump operation along the 100% flow control line, operating on =FlowLoop Inactive CFlow IndicatedLoop Active FlowCore Total

LSCS-UFSAR 6.B-3 REV. 13 one pump along the 100% flow control line, and calculating the correct value of C based on the core derived from the core support plate P and the loop flow indicator readings.

6.B.2.3 Core Flow Uncertainty Analysis

The uncertainty analysis procedure used to establish the core flow uncertainty for one-pump operation is essentially the same as for two-pump operation, except for some extensions. The core flow uncertainty analysis is described in Reference 1. The analysis of one-pump core flow uncertainty is summarized below.

For single-loop operation, the total core flow can be expressed as follows (refer to Figure 6.B-1):

W C = W A - W I Where W C = total core flow, W A = active loop flow, and W I = inactive loop (true) flow.

By applying the "propagation of errors" method to the above equation, the variance of the total flow uncertainty can be approximated by:

1-a 1-a where W C = uncertainty in total core flow (%), W A = uncertainty in active loop flow (%), W I = uncertainty in inactive loop flow (%), and a = W I / W A The uncertainty of W A was analyzed to be 2.8%. A conservative, bounding value of 3.0% was used for W Ain the total flow uncertainty variance calculation. The 2 I W 2a-1 a 2 A W 2a-1 1 2 C W+=

LSCS-UFSAR 6.B-4 REV. 13 uncertainty, W I is comprised of the uncertainty in the "C" coefficient and random uncertainties such as jet pump P measurement uncertainty and instrumentation errors. The bounding value of 3.75% for W I was used in the determination of W I. Based on the above uncertainties an d a bounding value of 0.36 for a, the variance of the total flow uncertainty is approximately:

= (5.0%)2 When the effect of 4.1% co re bypass flow uncertainty at 12% (bounding case) bypass flow fraction is added to the above total core flow uncertainty, the active coolant

flow uncertainty is:

which is less than the 6% core flow uncerta inty assumed in the statistical analysis.

In summary, core flow during one-pump operation is established in a conservative way and its uncertainty has been conservatively evaluated.

6.B.2.4 TIP Reading Uncertainty To ascertain the TIP noise uncertainty for single recirculation loop operation, a test was performed at an operating BWR. The test was performed at a power level 59.3% of rated with a single recirculation pump in operation (core flow 46.3% of rated). A rotationally symmetric control rod pattern existed prior to the test.

Five consecutive traverses were made with each of five TIP machines, giving a total of 25 traverses. Analysis of their data resulted in a nodal TIP noise of 2.85%. Use of this TIP noise value as a component of the process computer total uncertainty results in a one-sigma process computer total uncertainty value for single-loop operation of 6.8% for initial cores.

()()2 2 2 2 23.75% 0.36-1 0.36 3.0% 0.36-1 1 W+C ()()()25.7% 24.1% 212.0112.0 25.0% 2 active=+=

LSCS-UFSAR 6.B-5 REV. 13 6.B.3 MCPR OPERATING LIMIT 6.B.3.1 Abnormal Operational Transients

The consequences of an Anticipated Operational Occurrence (AOO) initiated from Single Loop Operation (SLO) are no different than the consequences of the same event initiated from two-loop operation, given the same initial power/flow conditions. One transient analyzed only for single loop operation, the abnormal startup of an idle recirculation loop, results in more severe consequences at low power levels than similar cold water injection transients (i.e. feedwater controller failure) as analyzed for two loop operation.

An analysis of this event is given in Section 15.4.4. The fuel thermal-mechanical integrity and safety limit MCPR (as increased for SLO) are protected during a postulated AOO in SLO mode by adhering to thermal limits derived from the more limiting of either the two-loop operation AOO results or the re sults from the idle recircula tion loop startup event. Results of these analyses, and a discussion of the applicability of these analyses to SLO, may be found in the LaSalle Admini strative Technical Requirements and its associated references.

Figure 6.B-2 shows the consequences of a typical pressurization transient (turbine trip) as a function of power level. As can be seen, the consequences of operation at lower power (such as would occur during SLO) result in lower reactor pressurization and neutron flux levels. Therefore, in absolute terms of maximum pressure and flux, SLO results in a milder transient than two-loop operation.

The power and flow dependent thermal limits developed for two loop operation are applicable for SLO, except for portions of the thermal limits which must be adjusted for the more severe consequences of the idle recirculation loop startup event discussed above. The flow dependent thermal limits are based on the event where both recirculation loop controllers fail (in the case of SLO, this event bounds failure of one controller, as the flow and power increase would be less). However, for operation in SLO, the flow dependent ther mal limits are adjusted to also bound the results of the idle recirculation loop startup event. These thermal limits are found in the LaSalle Administrative Technical Requirements.

The power dependent thermal limits are based on pressurization transients, such as the load rejection without bypass event, and the feedwater controller failure event (which is also a cold water injection event). As described above, the two loop results bound the SLO results for these events. Therefore, these SLO thermal limits are only different from the dual loop thermal limits in that they have been adjusted to protect a MCPR safety limit that is 0.

01 higher than the dual loop value.

In the following sections, three of the most limiting transients of cold water increase, pressurization, and flow decrease events are analyzed for single-loop

operation. These analyses were performed for the initial cycle core. For reload LSCS-UFSAR 6.B-6 REV. 13 cores, the bounding two loop operation analysis results for even ts a and b below are found in the LaSalle Administrative Techni cal Requirements. The transients are, respectively:

a. feedwater flow controller failure (maximum demand),

b. generator load rejection with bypass failure, and

c. one pump seizure accident.

The plant initial conditions are given in Table 6.B-1.

6.B.3.2 Feedwater Controller Failure - Maximum Demand This section presents in itial cycle GE results.

6.B.3.2.1 Identification of Causes and Frequency Classification

This event is postulated on the basis of a single failure of a control device, specifically one which can directly cause an increase in coolant inventory by increasing the feedwater flow. The most severe applicable event is a feedwater controller failure during maximum flow demand. The feedwater controller is forced to its upper limit at the beginning of the event.

This event is considered to be an incident of moderate frequency.

6.B.3.2.2 Sequence of Ev ents and Systems Operation With excess feedwater flow the water level rises to the high-level reference point at which time the feedwater pumps and the main turbine are tripped and a scram is

initiated. Table 6.B-2 lists the sequence of events for Figure 6.B-3. The figure shows the changes in important variables during this transient.

Identification of Operator Actions

a. Observe that high feedwater pump trip has terminated the failure event.

b. Switch the feedwater controlle r from auto to manual control in order to try to regain a correct output signal.

c. Identify causes of the failure and report all key plant parameters during the event.

LSCS-UFSAR 6.B-7 REV. 14, APRIL 2002 Systems Operation In order to properly simulate the expected sequence of events, the analysis of this event assumes normal functioning of pl ant instrumentation and controls, plant protection and reactor protection systems.

Important system operational actions for this event are high level tripping of the main turbine, feedwater turbine, turbine stop valve scram trip initiation, recirculation pump trip (RPT), and low-water level initiation of the reactor core isolation cooling system and the high-pressure core spray system to maintain long-term water level control following tripping of feedwater pumps (not simulated).

6.B.3.2.3 Effect of Single Failures and Operator Errors In Table 6.B-2 the first sensed event to initiate corrective action to the transient is the vessel high-water level (L8) trip. Multiple level sensors are used to sense and detect when the water level reaches the L8 setpoint. At this point in the logic, a single failure will not initiate or prevent a turbine trip signal. Turbine trip signal transmission, however, is not built to single-failure criterion. The result of a failure at this point would have the effect of delaying the pressurization "signature."

However, high moisture levels entering the turbine will be detected by high levels in the moisture separators which are designed to trip the unit. In addition, excessive moisture entering the turbine w ill cause vibration to the point where it too will trip the unit.

Scram trip signals from the turbine are designed such that a single failure will neither initiate nor impede a reactor scram trip initiation.

6.B.3.2.4 Core and System Performance

Mathematical Model

The computer model described in Subsecti on 15.1.1.3 was used to simulate this event.

Input Parameters and Initial Conditions

The analysis has been performed with the plant condition tabulated in Table 6.B-1, except that the initial vessel water level is at level setpoint L4 for conservation. By lowering the initial water level, more feedwater will get in, hence higher neutron flux will be attained before Level 8 is reached.

The same void reactivity coefficient used for pressurization transient is applied since a more negative value conservatively increases the apparent severity of the power increase. End of cycle (all rods out) scram characteristics are assumed. The safety/relief valve action is conservatively assumed to occur with higher than LSCS-UFSAR 6.B-8 REV. 13 nominal setpoints. The transient is simulated by programming an upper limit failure in the feedwater system such that 135% feedwater flow occurs at design pressure of feedwater sparge rs (1075 psia). Since the re actor is initially operating at a lower power level, the feedwater sparger experiences a pressure which is much lower than the design pressure, hence th e feedwater runout capacity reaches 160%

of rated.

Results The simulated feedwater controller transient is shown in Figure 6.B-3 for the case of 78% power 63% core flow. The high-water level turbine trip and feedwater pump trip are initiated at approximately 5.46 se conds. Scram occurs simultaneously from stop valve closure, and limits the neutron flux peak and fuel thermal transient so that no fuel damage occurs. MCPR remains above safety limit and peak fuel center temperature increases less than 170

° F. The turbine bypass system opens to limit peak pressure in the steamline near the safety valves to 1103 psig and the pressure at the bottom of the vessel to about 1118 psig.

Consideration of Uncertainties

All systems utilized for protection in this event were assumed to have the poorest allowable response (e.g., relief setpoints, scram stroke time, and work characteristics). Expected plant behavior is , therefore, expected to lead to a less severe transient.

6.B.3.2.5 Barrier Performance

As noted above, the consequences of this event do not result in any temperature or pressure transient in excess of the criteri a for which the fuel, pressure vessel, or containment are designed; therefore, these barriers maintain integrity and function as designed.

6.B.3.2.6 Radiological Consequences

The consequences of this event do not result in any fuel failures; however, radioactive steam is discharged to the suppression pool as a result of SRV activation.

6.B.3.3 Generator Load Rejection Without Bypass With RPT

This section presents in itial cycle GE results.

LSCS-UFSAR 6.B-9 REV. 13 6.B.3.3.1 Identification of Causes and Frequency Classification Fast closure of the turbine control valves (TCV) is initiated whenever electrical grid disturbances occur which result in significant loss of electrical load on the generator. The turbine control valves are re quired to close as rapidly as possible to prevent overspeed of the turbine-generator rotor. Closure of the main turbine control valves will increase system pressure.

This event is categorized as an infrequent incident with the following characteristics:

Frequency: 0.0036/plant-year MTBE: 278 years Frequency basis: thorough searches of domestic plant operating records have revealed three instances of bypass failure during 628 bypass system operations. This gives a probability of bypass failure of 0.0048. Combining the actual frequency of a generator load rejection with the failure rate of the bypass yields a frequency of a generator load rejection with bypa ss failure of 0.0036 event/plant year.

6.B.3.3.2 Sequence of Events and System Operation

Sequence of Events

A loss of generator electrical load at 78% and 63% flow under single recirculation loop operation produces the sequence of ev ents listed in Table 6.B-3. Notice that the vessel level reaches L8 at 5.3 seconds. The trip of feedwater pumps on L8 is not simulated.

Identification of Operator Options

a. Verify proper bypass valve performance.

b. Observe that the pressure regulator is controlling reactor pressure at the desired value.

c. Record peak power and pressure.

d. Verify relief valve operation.

System Operation Turbine control valve (TCV) fast closure initiates a scram trip signal for power levels greater than or equal to 25% of rated core thermal power. In addition, LSCS-UFSAR 6.B-10 REV. 14, APRIL 2002 recirculation pump trip is initiated. Both of these trip signals satisfy single failure criterion and credit is taken for these protection features.

The pressure relief system which operates the relief valves independently when system pressure exceeds relief valve instrumentation setpoints is assumed to function normally during the time period analyzed.

All plant control systems maintain normal operation unless specifically designated to the contrary.

Mitigation of pressure increase, the basic nature of this transient, is accomplished by the reactor protection system functions. Turbine control valve trip scram and RPT are designed to satisfy the single failure criterion.

Mathematical Model

The computer model described in Subsecti on 15.1.1.3 was used to simulate this event.

Input Parameters and Initial Conditions These analyses have been performed, unless otherwise noted, with the plant conditions tabulated in Table 6.B-1.

The turbine electrohydraulic control syst em (EHC) power/load imbalance device detects load rejection before a measurable speed change takes place.

The closure characteristics of the turbine control valves are assumed such that the valves operate in the full arc (FA) mode and have a full stroke closure time, from fully open to fully closed, of 0.15 second.

Auxiliary power would normally be indepe ndent of any turbine-generator overspeed effect and continuously supplied at rated frequency as automatic fast transfer to auxiliary power supplies normally occurs. For the purposes of worst case analysis, the recirculation pumps are assumed to remain tied to the main generator and thus increase in speed with the T-G overspeed until tripped by the recirculation pump trip system (RPT).

The reactor is operating in the manual flow-control mode when load rejection occurs. Results do not significantly differ if the plant had been operating in the automatic flow-control mode.

LSCS-UFSAR 6.B-11 REV. 13 6.B.3.3.3 Results Figure 6.B-4 shows that, for the case of bypass failure, peak ne utron flux reaches about 135.6% of rated, average surface he at flux reaches 8% of rated. The calculated MCPR is 1.29, which is well above the safety limit.

Consideration of Uncertainties

The full-stroke closure rate of the tu rbine control valve of 0.15 second is conservative. Typically, the actual closure rate is more like 0.2 second. Clearly the less time it takes to close, the more severe the pressurization effect.

All systems utilized for protection in this event were assumed to have the poorest allowable response (e.g., relief setpoints, scram stroke time, and worth characteristics). Expected plant behavior is, therefore, expected to reduce the actual severity of the transient.

Peak pressure at the valves reaches 1128 ps ig. The peak nuclear system pressure reaches 1153 psig at the bottom of the vessel, well below th e nuclear barrier transient pressure limit of 1375 psig.

6.B.3.3.4 Barrier Performance

The consequences of these events do not result in any temperature or pressure transient in excess of the criteria for which the fuel, pressure ve ssel, or containment are designed and, therefore, these barrier s maintain their integrity as designed.

6.B.3.3.5 Radiological Consequences

The consequences of the events identified previously do not result in any fuel failures; however, radioactivity is nevertheless discharged to the suppression pool

as a result of SRV activation.

6.B.3.4 Recirculation Pump Seizure Accident This analysis presents initial cycle GE results.

6.B.3.4.1 Identification of Causes and Frequency Classification

The case of recirculation pump seizure re presents the extremely unlikely event of instantaneous stoppage of the pump motor shaft of one recirculation pump. This produces a very rapid decrease of core flow as a result of the large hydraulic resistance introduced by the stopped rotor.

This event is considered to be a limiting fault.

LSCS-UFSAR 6.B-12 REV. 14, APRIL 2002 Actual occurrence data is not available at this time.

6.B.3.4.2 Sequence of Ev ents and Systems Operations

Table 6.B-4 lists the sequence of events for this recirculation pump seizure accident.

Identification of Operator Actions

The operator should ascertain that the reactor scrams with the turbine trip resulting from reactor water level swell.

The operator should regain control of reactor water level through RCIC operation or by restart of a feedwater pump, and must monitor reactor water level and pressure control after shutdown.

6.B.3.4.3 Systems Operation

In order to properly simulate the expected sequence of events, the analysis of this event assumes normal functioning of pl ant instrumentation and controls, plant protection, and reactor protection systems.

Operation of HPCS and RCIC systems, though not included in this simulation, are expected to occur in order to maintain adequate water level.

6.B.3.4.4 Core and System Performance

Mathematical Model The nonlinear dynamic model described briefly in Subsection 15.1.1.3 is used to simulate this event.

Input Parameters and Initial Conditions

This analysis has been performed, unless otherwise noted, with plant conditions tabulated in Table 6.B-1. For the purpose of evaluating consequences to the fuel thermal limits this transient event is assumed to occur as a consequence of an unspecified, instantaneous stoppage of the active recirculation pump shaft while the reactor is operating at 78% NB rated power under SLO conditions. Also, the reactor is assumed to be operating at thermally limited conditions.

The void coefficient is adjusted to the most conservative value; that is, the least negative value in Table 6.B-1.

LSCS-UFSAR 6.B-13 REV. 13 6.B.3.4.5 Results Figure 6.B-5 presents the results of the accident. Core coolant flow drops rapidly, reaching a minimum value of 76.4 at about 1.09 second. The level swell produces a trip of both the main and feedwater turbines which, in turn, results in stop valve closure scram. The turbine trip, occurring after the time at which MCPR results, does not significantly retard the heat flux decrease and imposes no threat to fuel thermal limits. Considerations of uncertainties are included in the GETAB analysis.

6.B.3.4.6 Barrier Performance

The bypass valves and momentary opening of some of the safety/relief valves limit the pressure to well within the range allo wed by the ASME vessel code. Therefore, the reactor coolant pressure boundary is not threatened by overpressure.

6.B.3.4.7 Radiological Consequences

The consequences of this event do not result in any fuel failures; however, radioactivity is nevertheless discharged to the suppression pool as a result of SRV activation.

6.B.3.5 Summary and Conclusions

The transient results for these initial cycles analyses are summarized in Table 6.B-5. This table indicates that for the transient events analyzed here, the

MCPRs are well above the safety limit value of 1.06 (original analysis MCPR safety limit). It is concluded that the ther mal margin safety limits established for two-pump operation are also applicable to single-loop-operation conditions.

For pressurization, Table 6.B-5 indicates that the peak pressures are below the ASME code value of 1375 psig. Hence, it is concluded that the pressure barrier integrity is maintained under single-loop-operation conditions.

6.B.4 OPERATING MCPR LIMIT

For single-loop operation, the rated condition steady-state MCPR limit is increased by 0.01 to account for the increase in the fuel cladding integrity safety limit (Section 6.B.2). At lower flows, the steady-state operating MCPR limit is conservatively established by a flow depe ndent MCPR. The operating limit is the more conservative of the two. This en sures that the 99.9% statistical limit requirement is always satisfied for any postulated abnormal operational occurrence.

LSCS-UFSAR 6.B-14 REV. 15, APRIL 2004 6.B.5 STABILITY ANALYSIS The least stable power/flow condition attain able under normal conditions occurs at natural circulation with the control rods set for rated power and flow. This

condition may be reached following the trip of both recirculation pumps. As shown in Figure 6.B-5, operation along the minimum forced recirculation line with one pump running, at minimum speed, is more stable than operating with natural circulation flow only, but is less stable than operating with both pumps operating at minimum speed. Because of the increased flow fluctuation during one-recirculation-loop operation, the flow control should be left in manual operation to preclude unnecessary wear on the automatic controls.

6.B.6 Loss-of-Coolant Accident Analysis An analysis of single recirculation loop operation utilizing the models and assumptions documented in Reference 4 was performed for each LSCS unit. Using this method SAFER/GESTR-LOCA calculations were performed for the DBA. The SLO PCTs were calculated without a MAPLHGR reduction. GE determined the results were within the 10 CFR50.46 acceptance criteria. However, SLO without MAPLHGR reduction results in more limiting PCTs than the two loop LOCA.

Reference 5 concluded that if ARTS power and flow dependent MAPLHGR multipliers are applied, then SLO results of the LOCA analysis are less limiting than the two loop LOCA results.

A limited spectrum of LOCA/ECCS analyses with SLO unique assumptions was performed in Reference 2 to determine the LaSalle ATRIUM-9B SLO MAPLHGR limits. The two-loop break spectrum results were used to select potentially limiting SLO LOCAs for the analysis. The most important parameters for the LOCA analyses are the break size, break location and the ECCS single failure assumption.

Six different combinations were analyzed for SLO. These parameters are independent of the initial conditions. FANP determined that SLO analyses will show a similar limiting break size, break location and single failure to the two-loop break spectrum results.

The domain for Single Loop Operation (SLO) is not affected by Power Uprate to 3489 MWt. The current SLO analysis (Ref erence 4) is valid for uprated power conditions as evaluated in Reference 6.

A spectrum of LOCA/ECCS SLO analyses was performed for AT RIUM-10 fuel. The analyses determined the SLO multiplier to the two-loop MAPLHGR limits so that the limiting PCT for SLO is less than the limiting PCT for two-loop operation.

6.B.6.1.1 Break Spectrum Analysis

For GE Fuel, SAFER/GESTR-LOCA calculations were performed for LaSalle Units 1 and 2 for SLO using very conservative and bounding assumptions given in Section LSCS-UFSAR 6.B-14a REV. 15, APRIL 2004 5.3.1 of Reference 4. The most limiting SLO break was consistent with the limiting two-loop operation break, the DBA recirculation suction side break, single failure of the HPCS diesel generator.

For SPC Fuel, RELAX, FLEX, and HUXY calculations were performed for a limited spectrum using the assumptions given in Section 9.0 of Reference 2. Explicit analyses for DBA double-ended guillotine (DEG) break, and other breaks smaller

LSCS-UFSAR 6.B-15 REV. 15, APRIL 2004 and larger than the limiting case were used to confirm the limiting SLO case. The time of core reflood for the single loop operation of 184.4 seconds is consistent with the time of core reflood for the two-loop op eration of 189.5 seconds. Both the single-loop and two-loop operation cases were for the 1.1 ft 2 break of the pump discharge piping with a single failure of the HPCS diesel generator.

For ATRIUM-10 fuel, a spectrum of LOCA/ECCS SLO analyses was performed (Reference 7). The SLO analyses were performed with a 0.90 multiplier applied to the two-loop MAPLHGR limit. The lim iting SLO LOCA is the 1.0 DEG pump suction line break with a single failure of the LPCS diesel generator.

6.B.6.2.1 Single-Loop MAPLHGR Determination For GE Fuel, the limiting break determined for two-loop operation is analyzed for SLO to confirm that the SLO PCT (1490 o F Appendix K analysis basis from Reference 4) is less than the 10CFR50.46 limit. The SLO PCT for the limiting break is still well below the 10CFR50.46 limit assuming no MAPLHGR reduction for SLO. Therefore, no MAPLHGR reduction is required for LaSalle Units 1 and 2 under SLO for the GE fuel. Application of ARTS MAPLHGR multipliers assures that two loop LOCA results described in Se ction 6.3.3.9.1 remain the licensing basis for LaSalle Units 1 and 2.

For ATRIUM-9B fuel, the limiting break determined is analyzed for SLO to confirm that the SLO PCT is less than the two-loop operation PCT . Analyses were performed to confirm that the PCT tren ds for SLO and two-loop operation are generally the same. The MAPLHGR used fo r the Reference 2 SLO analyses is 13.5 kW/ft which is the same as the two-loop op eration value; however a 0.9 multiplier is used for the SLO MAPLHGR.

The SLO PCT is 1628 F with a 0.29% MWR. These results are less limiting than the maximum two-loop operation results; therefore, an SLO MAPLHGR multiplier of 0.9 is appropriate for ATRIUM-9B fuel.

For ATRIUM-10 fuel, the analyses show that the limiting two-loop LOCA results bound the limiting SLO LOCA results when a MAPLHGR multiplier of 0.90 is applied to the two-loop MAPLHGR limit.

6.B.6.3 Small Break Peak Cladding Temperature

Section 5.3.1 of Reference 4 discusses why the DBA break is more limiting than the smaller break sizes for SLO. Section 5.3.1 of Reference 4 also discusses the effect of the assumptions used in the one-pump op eration analysis and the duration of nucleate boiling. GE did not calculate sma ll break results for SLO because they are non-limiting.

FANP analyses for ATRIUM-9B and ATRIUM

-10 fuel used a spectrum of break sizes that include small breaks to identify the limiting break. The limiting break LSCS-UFSAR 6.B-15a REV. 15, APRIL 2004 for ATRIUM-9B fuel is a small break of 1.1 square feet in the recirculation pump discharge line. For ATRIUM-10, the limiting break is a double-ended guillotine break of the recirculation pump suction line. The analyses for both ATRIUM-9B and ATRIUM-10 show that the limiting two-loop LOCA results bound the SLO LOCA results when the SLO MAPLHGR multiplier of 0.90 is applied to the two-loop MAPLHGR limit.

LSCS-UFSAR 6.B-16 REV. 15, APRIL 2004 6.B.7 REFERENCES

1. General Electric BWR Therma l Analysis Basis (GETAB): Data, Correlation, and Design Application, General Electric Company (NEDO-10958-A), January 1977.

2. "LOCA Break Spectrum Analysis for LaSalle Units 1 and 2," EMF-2174(P). Siemens Power Corporation, March 1999.

3. "LaSalle Extended Operating Domain (EOD) and Equipment Out of Service (EOOS) Safety Analysis fo r ATRIUM-9B Fuel," SPC Report EMF-95-205(P).

4. "LaSalle County station Units 1 and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-32258P, General Electric Company, October 1993.

5. "LaSalle County Station Units 1 and 2 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31510P, General Electric Company, December 1987.

6. LaSalle County Station Power Uprate Project, Task 407, "ECCS Performance," GE-NE-A1300384 01, Revision 1, September 1999.

7. "LaSalle Units 1 and 2 LOCA Brea k Spectrum Analysis for ATRIUM-10 Fuel", EMF-2639(P), Revision 0, Framatome ANP, November 2001.

LSCS-UFSAR TABLE 6.B-1 (SHEET 1 OF 2)

TABLE 6.B-1 REV. 13 INPUT PARAMETERS AND INITIAL CONDITIONS FOR ANALYSIS OF INITIAL CORE TRANSIENTS AND ACCIDENTS FOR SINGLE-LOOP OPERATION (INITIAL CORE VALUES)**

1. Thermal Power Level, Analysis Value, % NBR 78 2. Steam Flow, lb/h 10.71 x 10 6 3. Core Flow, lb/h 68.26 x 10 6 4. Feedwater Flow Rate, lb/sec 2976

5. Feedwater Enthalpy, Btu/lb 367.3

6. Vessel Dome Pressure, psig 1001

7. Vessel Core Pressure, psig 1006

8. Turbine Bypass Capacity, % NBR 25

9. Core Coolant Inlet Enthalpy, Btu/lb 516.8

10. Turbine Inlet Pressure, psig 969.3

11. Fuel Lattice 8 x 8

12. Core Average Gap Conductance, Btu/sec-ft 2-°F 0.1662 13. Core Leakage Flow, %

12 14. Required MCPR Operating Limit 1.41

• 15. MCPR Safety Limit 1.06 16. Doppler Coefficient, -¢/°F Nominal EOC-1 Analysis Data 0.221 0.221 17. Void Coefficient, -¢/% Voids Nominal EOC-1 Analysis Data for Power Increase Events Analysis Data for Power Increase Events 7.429 12.63 7.01 18. Core Average Rated Void Fraction, % 0.414 19. Scram Reactivity, Analysis Data FSAR Figure 15.0-2 20. Control Rod Drive Speed, position versus time FSAR Figure 15.0-2

• Dual-pump operation operating limit for 63%

core flow, obtained by applying K f-curve to operating limit CPR at rated condition (1.24).

• • For cycle specific inputs, see the transient analysis input parameters.

LSCS-UFSAR TABLE 6.B-1 (SHEET 2 OF 2)

TABLE 6.B-1 REV. 13 (INITIAL CORE VALUES)

21. Jet Pump M Ratio 3.20 22. Safety/Relief Valve Capacity, % NBR at 1165 psig Manufacturer Quantity Installed 111.5 Crosby 18 23. Relief Function Delay, sec 0.1 24. Relief Function Response, sec 0.1 25. Setpoints for Safety/Relief Valves Safety Function, psig

Relief Function, psig 1150, 1175, 1185, 1195, 1205 1076, 1086, 1096, 1106, 1116

26. Number of Valve Groupings Simulated Safety Function, No. Relief Function, No.

5 5 27. Vessel Level Trips, Inches above Steam Dryer Skirt Bottom (Instrument Zero) Level 8 - (L8)

Level 3 - (L3) Level 2 - (L2) 55.5 12.5 -50 28. RPT Delay, sec 0.14 29. RPT Inertia Time Constant for Analysis, sec 6.0

LSCS-UFSAR TABLE 6.B-2 TABLE 6.B-2 REV. 13 SEQUENCE OF EVENTS FOR FIGURE 6.B-3 (INITIAL CORE RESULTS)

TIME (sec) EVENT 0 Initiate simulated failure of 160% upper limit on feedwater flow. 5.46 L8 vessel level setpoint trips main turbine and feedwater pumps.

5.47 Reactor scram trip actuated from main turbine stop valve position switches.

5.47 Recirculation pump (RPT) actuated by turbine stop valve position

switches.

5.57 Main turbine stop valves closed and main turbine bypass valves start to open. 8.01, 8.29 Relief valves actuated (groups 1, 2). 11.67, 12.23 Relief valves close (groups 2, 1). 29.32 Main turbine bypass valves closed. 48.35 Main turbine bypass valves start to open.

LSCS-UFSAR TABLE 6.B-3 TABLE 6.B-3 REV. 13 SEQUENCE OF EVENTS FOR FIGURE 6.B-4 (INITIAL CORE RESULTS)

TIME (sec) EVENT -0.015 (approx) Turbine-generator detection of loss of electrical load 0 Turbine-generator power load unb alance (PLU) devices trip to initiate turbine control valve fast closure 0 Turbine bypass valves fail to operate 0 Fast control valve closure (FCV) initiates scram trip 0 Fast control valve closure (FCV) initiates recirculation pump trip (RPT) 0.039 Turbine control valves closed 0.14 Recirculation pump motor circuit breakers open, causing decrease in core flow to natural circulation 1.98, 2.12, 2.27, 2.45, 2.74 Relief valves actuated (groups 1, 2, 3, 4, 5) 4.58, 4.91, 5.20 (est) Relief valves close (groups 5, 4, 3) 5.30 Vessel level reaches L8 setpoint, feed water pumps tripped (not simulated) 5.50, 5.84 (est) Relief valves close (groups 2, 1) 12.00 Relief valves actuated (group 1) 19.0 (est) Relief valves close (group 1) 33 2 Relief valves actuated (group 1) 38.0 (est) Relief valves close (group 1)

LSCS-UFSAR TABLE 6.B-4 TABLE 6.B-4 REV. 13 SEQUENCE OF EVENTS FOR FIGURE 6.B-5 (INITIAL CORE RESULTS)

TIME (sec) EVENT 0 Single pump seizure was initiated, core flow decreases to natural recirculation 1.23 Reverse flow ceases in the idle loop 4.93 High vessel water level (L8) trip initiates main turbine trip 4.93 High vessel water level (L8) trip initiates feedwater turbine trip 4.93 Main turbine trip initiates bypass operation

4.96 Main turbine valves reach 90% open position and initiate reactor scram trip 5.03 Turbine stop valves closed and turbine bypass valves start to open to regulate pressure 10.0 (est) Turbine bypass valves start to close 25.1 Turbine bypass valves closed 38.6 Turbine bypass valves reopen on pressure increase at turbine inlet

LSCS-UFSAR TABLE 6.B-5 TABLE 6.B-5 REV. 13

SUMMARY

OF EVENT RESULTS SINGLE RECIRCULATION LOOP OPERATION (Typical)

PARAGRAPH F IGURE DESCRIPTION M AXIMUM N EUTRON F LOW (% NBR) M AXIMUM D OME PRESSURE (psig) M AXIMUM VESSEL PRESSURE (psig) M AXIMUM S TEAMLINE PRESSURE (psig) M AXIMUM C ORE AVERAGE SURFACE H EAT F LUX (% of Initial)

MCPR FREQUENCY* C ATEGORY 6.B.3.2 6.B-3 Feedwater flow Controller Failure (Maximum Demand) 119.3 1112 1126 1103 108.8 1.26 a 6.B.3.3 6.B-4 Generator Load Rejection 135.6 1138 1153 1128 103.5 1.29 b 6.B.3.4 6.B-5 Seizure of Active Recirculation Pump 78.0 1021 1031 1018 100.0 1.17 c

__________________________

• a = incident of moderate frequency; b = infrequent incident; c = limiting faults