ML22144A131

6 to Updated Final Safety Analysis Report, Chapter 6, Engineered Safety Featres

ML22144A131
Person / Time
Site:	Beaver Valley
Issue date:	05/18/2022
From:	Energy Harbor Nuclear Corp
To:	Office of Nuclear Reactor Regulation
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ML22144A115	List: L-22-058, Submittal of Revision 26 to Updated Final Safety Analysis Report ML22144A114 ML22144A116 ML22144A117 ML22144A118 ML22144A119 ML22144A120 ML22144A121 ML22144A122 ML22144A123 ML22144A124 ML22144A126 ML22144A127 ML22144A128 ML22144A129 ML22144A130 ML22144A131 ML22144A132 ML22144A133 ML22144A134 ML22144A136
References
L-22-058
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Text

BVPS-2 UFSAR Rev. 26 CHAPTER 6 TABLE OF CONTENTS Section Title Page 6 ENGINEERED SAFETY FEATURES 6.1-1 6.1 ENGINEERED SAFETY FEATURE MATERIALS 6.1-1 6.1.1 Metallic Material 6.1-1 6.1.2 Organic Materials 6.1-4 6.1.3 References for Section 6.1 6.1-7 6.2 CONTAINMENT SYSTEMS 6.2-1 6.2.1 Containment Functional Design 6.2-1 6.2.2 Containment Heat Removal System 6.2-47 6.2.3 Secondary Containment Functional Design 6.2-57 6.2.4 Containment Isolation System 6.2-57 6.2.5 Combustible Gas Control in Containment 6.2-66 6.2.6 Containment Leakage Testing 6.2-73 6.2.7 Fracture Prevention of Containment Pressure Boundary Materials 6.2-76 6.2.8 References for Section 6.2 6.2-76 6.3 EMERGENCY CORE COOLING SYSTEM 6.3-1 6.3.1 Design Bases 6.3-1 6.3.2 System Design 6.3-2 6.3.3 Performance Evaluation 6.3-14 6.3.4 Inspection and Testing Requirements 6.3-20 6.3.5 Instrumentation Requirements 6.3-23 6.3.6 Reference for Section 6.3 6.3-26 6.4 HABITABILITY SYSTEMS 6.4-1 6.4.1 Design Bases 6.4-1 6.4.2 System Design 6.4-2 6.4.3 System Operational Procedures 6.4-5 6.4.4 Design Evaluation 6.4-6 6.4.5 Inspection and Testing Requirements 6.4-8 6.4.6 Instrumentation Requirements 6.4-8 6.4.7 References for Section 6.4 6.4-8 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5-1 6.5.1 Engineered Safety Feature Filter Systems 6.5-1 6.5.2 Containment Spray as a Fission Product Cleanup System 6.5-3 6.5.3 Fission Product Control Systems 6.5-11 6.5.4 References for Section 6.5 6.5-18 6-i

BVPS-2 UFSAR Rev. 26 TABLE OF CONTENTS (Cont)

Section Title Page 6.6 IN-SERVICE INSPECTION OF ASME CODE CLASS 2 AND CLASS 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 6.6-2 6.6.4 Inspection Intervals 6.6-2 6.6.5 Examination Categories and Requirements 6.6-3 6.6.6 Evaluation of Examination Results 6.6-3 6.6.7 System Pressure Tests 6.6-3 6.6.8 Augmented In-service Inspection to Protect Against Postulated Piping Failures 6.6-3 Appendix 6A - Generic Letter 2004-02 Containment Sump Evaluation 6-ii

BVPS-2 UFSAR Rev. 26 LIST OF TABLES Table Number Title 6.1-1 Typical Materials Employed for Components of Engineered Safety Features Systems Balance of Plant 6.1-2 Typical Materials Employed for Components of Engineered Safety Features Systems Nuclear Steam Supply System 6.1-3 Protective Coatings on Westinghouse-Supplied Equipment Inside the Containment 6.2-1 Thermophysical Properties of Passive Heat Sink Materials 6.2-2 Beaver Valley MAAP-DBA Parameter File Summary of Containment Nominal Volumes and Metal Heat Sinks 6.2-2A Beaver Valley MAAP-DBA Parameter File Summary of Containment Concrete Heat Sinks 6.2-3 Containment Design Evaluation Parameters 6.2-4 Key Input Data to MAAP-DBA (Peak Pressure Calculations) 6.2-5 Beaver Valley MAAP-DBA Parameter File Summary of Junction Flow Areas 6.2-6 Containment Peak Pressure Results for A Design Basis Large Break LOCA Beaver Valley 6.2-7 Double-Ended Hot-Leg Break Sequence of Events 6.2-8 BVPS-2 Double-Ended Pump Suction Break Minimum Safeguards Sequence of Events 6.2-9 Double-Ended Pump Suction Break Maximum Safeguards Sequence of Events (CIB Failure) 6.2-9B Deleted 6.2-10 MAAP-DBA Peak Pressure Results for a Design Basis Main Steam Line Break Beaver Valley 6.2-10a Deleted 6.2-11 MAAP-DBA Peak Temperature Results for a Design Basis Main Steam Line Break Beaver Valley 6.2-11a Deleted 6-iii

BVPS-2 UFSAR Rev. 26 LIST OF TABLES (Cont)

Table Number Title 6.2-12 Sequence of Events - Peak Containment Pressure Case 1.069 Ft 2 Double-Ended Rupture (DER) With a Main Steam Isolation Valve (MSIV) Failure at Thirty Percent Power (Case 11m) 6.2-13 Sequence of Events - Peak Containment Temperature Case 1.069 Ft2 Double-Ended Rupture (DER) With a Main Steam Isolation Valve (MSIV) Failure at Full Power (Case 1m) 6.2-14 System Parameters Initial Conditions for Thermal Uprate 6.2-14A Safety Injection Flow Minimum Safeguards 6.2-14B Safety Injection Flow Maximum Safeguards 6.2-14C Double-Ended Hot-Leg Break Blowdown Mass and Energy Releases 6.2-14D Double-Ended Pump Suction Break Blowdown Mass and Energy Releases (Same for all DEPS Runs) 6.2-14E Double-Ended Pump Suction Break Minimum Safeguards Reflood Mass and Energy Releases 6.2-14F Double-Ended Pump Suction Break - Minimum Safeguards Principle Parameters During Reflood 6.2-14G Double-Ended Pump Suction Break Minimum Safeguards Post-Reflood Mass and Energy Releases 6.2-14H Double-Ended Pump Suction Break Maximum Safeguards Reflood Mass and Energy Releases 6.2-14I Principle Parameters During Reflood Double-Ended Pump Suction Break - Maximum Safeguards 6.2-14J Double-Ended Pump Suction Break Maximum Safeguards Post-Reflood Mass and Energy Releases 6.2-14K Double-Ended Hot-Leg Break Mass Balance 6.2-14L Double-Ended Pump Suction Break Mass Balance Minimum Safeguards 6.2-14M Double-Ended Pump Suction Break Mass Balance Maximum Safeguards 6.2-14N Double-Ended Hot-Leg Break Energy Balance 6.2-14O Double-Ended Pump Suction Break Energy Balance Minimum Safeguards 6-iv

BVPS-2 UFSAR Rev. 26 LIST OF TABLES (Cont)

Table Number Title 6.2-14P Double-Ended Pump Suction Break Energy Balance - Maximum Safeguards) 6.2-14Q Double-Ended Pump Suction Break (SW Failure) - Maximum Safeguards Reflood Mass And Energy Releases 6.2-14R Double-Ended Pump Suction (SW Failure) - Maximum Safeguards Principle Parameters During Reflood 6.2-14S Double-Ended Pump Suction (SW Failure) Maximum Safeguards Post-Reflood Mass And Energy Releases 6.2-14T Double-Ended Pump Suction (SW Failure) Mass Balance Maximum Safeguards 6.2.14U Double-Ended Pump Suction (SW Failure) Energy Balance Maximum Safeguards 6.2.14V LOCA Mass And Energy Release Analysis ANS 1979 Core Decay Heat Power Fraction 6.2-15 Deleted 6.2-15a Deleted 6.2-16 Deleted 6.2-17 Deleted 6.2-18 Deleted 6.2-19 Deleted 6.2-20 Deleted 6.2-21 Pressurizer Subcompartment Vent Path Description 6.2-22 Pressurizer Subcompartment Nodal Description 6.2-23 Mass and Energy Release Rates Pressurizer Spray Line Double Ended Rupture 6.2-24 Mass and Energy Release Rates Pressurizer Surge Line Double-Ended Rupture 6.2-25 Summary of Subcompartment Peak Calculated Pressure Differential Used in Structural Analysis 6-v

BVPS-2 UFSAR Rev. 26 LIST OF TABLES (Cont)

Table Number Title 6.2-26 Deleted 6.2-27 Steam Generator Subcompartment Vent Path Description 6.2-28 Steam Generator Subcompartment Nodal Description 6.2-29 Mass and Energy Release Rates 360 in2 LDR at the Steam Generator Outlet Nozzle 6.2-30 Mass and Energy Flow Rates 6.2-31 Mass and Energy Release Rates 707 in2 Longitudinal Intrados Split at the Steam Generator Inlet ELBOW 6.2-32 Deleted 6.2-33 Deleted 6.2-34 Deleted 6.2-35 Deleted 6.2-36 Deleted 6.2-37 Deleted 6.2-38 Deleted 6.2-39 Deleted 6.2-40 Deleted 6.2-41 Deleted 6.2-42 Deleted 6.2-43 Deleted 6.2-44 Deleted 6.2-45 Deleted 6.2-46 Deleted 6.2-47 Deleted 6.2-48 Deleted 6-vi

BVPS-2 UFSAR Rev. 26 LIST OF TABLES (Cont)

Table Number Title 6.2-49 Deleted 6.2-50 Beaver Valley Power Station Unit 2 Initial Condition Assumptions MSLB Mass and Energy Releases Inside Containment 6.2-51 Beaver Valley Power Station Unit 2 Balance of Plant Assumptions MSLB Mass and Energy Releases Inside Containment 6.2-52 LBLOCA Mass and Energy Releases from BCL Vessel-Side 6.2-53 LBLOCA Mass and Energy Releases from BCL Vessel-Side 6.2-54 Active Heat Sinks for Minimum Containment Pressure Analysis 6.2-55 Large Break LOCA Containment Wall Data Used for Calculation of Containment Pressure 6.2-56 Quench Spray System Component Design Data 6.2-57 Component Design Data Recirculation Spray System 6.2-58 Supplementary Leak Collection and Release System Air Flow Rates 6.2-59 Net Positive Suction Head for Containment Heat Removal System 6.2-60 Containment Isolation Features 6.2-61 Deleted 6.2-62 Deleted 6.3-1 Emergency Core Cooling System Component Parameters 6.3-2 Emergency Core Cooling System Relief Valve Data 6.3-3 Motor-Operated Isolation Valves in Emergency Core Cooling Systems 6.3-4 Materials Employed for Emergency Core Cooling System Components 6.3-5 Failure Modes and Effects Analysis Emergency Core Cooling System - Active Components 6.3-6 Emergency Core Cooling System Recirculation Piping Passive Failure Analysis Long Term Phase 6-vii

BVPS-2 UFSAR Rev. 26 LIST OF TABLES (Cont)

Table Number Title 6.3-7 Sequence of Switchover Operation from Injection to Recirculation 6.3-8 Emergency Core Cooling System Shared Functions Evaluation 6.3-9 Normal Operating Status of Emergency Core Cooling System Components for Core Cooling 6.4-1 Control Room Envelope Ventilation Parameters 6.4-1a Control Room Envelope Ventilation Parameters Used for LOCA and CREA Analyses 6.4-2 Deleted 6.4-3 Locations of Control Room Air Intake, Toxic Gas Storage, and Radiation Release Points 6.5-1 Comparison of Engineered Safety Features Filter System Design Features with Regulatory Guide 1.52 6.5-2 Containment Thermodynamic Data - Loss of Coolant Accident 6.5-3 Deleted 6.5-4 Parameters for Calculating Minimum and Maximum Spray pH during Quench Spray Operation 6.5-5 Parameters for Minimum Ultimate Sump pH Calculation 6.5-6 Primary Containment Information 6.5-7 Supplementary Leak Collection and Release System Principal Components and Design Parameters 6.5-8 Supplementary Leak Collection and Release System Air Flow Rates 6-viii

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES Figure Number Title 6.2-1 MAAP-DBA Containment Nodalization 6.2-2 MAAP-DBA Containment Nodalization (Plan View) 6.2-3 MAAP5 Node and Junction Arrangement 6.2-4 Deleted 6.2-5 Containment Pressure Time-History for the DEHL Break Case 6.2-6 Containment Temperature Time-History for the DEHL Break Case 6.2-7 Containment Pressure Time-History for the DEPS Min SI Break Case 6.2-8 Containment Temperature Time-History for the DEPS Min SI Break Case 6.2-9 Containment Pressure Time-History for the DEPS Max SI (CIB Failure) Break Case 6.2-10 Containment Temperature Time-History for DEPS Max SI (CIB Failure) Break Case 6.2-10A Deleted 6.2-10B Deleted 6.2-11 Containment Sump Temperature Time-Histories for the DEPS Min SI and DEPS Max SI (CIB Failure) Break Case 6.2-11A Deleted 6.2-11B Deleted 6.2-12 Containment Pressure Time-History for the Limiting MSLB Pressure Case 6.2-13 Containment Temperature Time-History for the Limiting MSLB Pressure Case 6.2-14 Containment Pressure Time-History for the Limiting MSLB Temperature Case 6.2-15 Containment Temperature Time-History for the Limiting MSLB Temperature Case 6-ix

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES (Cont)

Figure Number Title 6.2-16 Deleted 6.2-17 Deleted 6.2-18 Staggered Mesh Control Volume Approximately for THREED 6.2-19 Computional Block Diagram for THREED 6.2-20 Pressurizer Subcompartment Nodalization Diagram 6.2-21 Pressurizer Cubicle Nodalization - Plan View El. 767'-10" - 23 Node Model 6.2-22 Pressurizer Cubicle Nodalization - Plan View El. 738'-10" - 23 Node Model 6.2-23 Pressurizer Relief Tank Cubicle Nodalization Plan View El.

718'-6" - 23 Node Model 6.2-24 Pressurizer Cubicle Elevation View Section 1-1 6.2-25 Pressurizer Cubicle Average Pressure vs Time-Spray Line DER Nodes 1 and 3 6.2-26 Pressurizer Cubicle Average Pressure vs Time-Spray Line DER Nodes 2 and 4 6.2-27 Pressurizer Cubicle Average Pressure vs Time-Spray Line DER Nodes 5 and 7 6.2-28 Pressurizer Cubicle Average Pressure vs Time-Spray Line DER Nodes 6 and 8 6.2-29 Pressurizer Cubicle Average Pressure vs Time-Spray Line DER Nodes 9 to 12 6.2-29A Pressurizer Cubicle Average Pressure vs Time-Spray Line DER Nodes 13 to 16 6.2-29B Pressurizer Cubicle Average Pressure vs Time-Spray Line DER Nodes 17 to 20 6.2-29C Pressurizer Cubicle Average Pressure vs Time-Spray Line DER Nodes 21 to 23 6.2-30 Pressurizer Cubicle Average Pressure vs Time-Surge Line DER at the Pressurizer Nozzle Nodes 1, 2, 3, and 4 6-x

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES (Cont)

Figure Number Title 6.2-31 Pressurizer Cubicle Average Pressure vs Time-Surge Line DER at the Pressurizer Nozzle Nodes 5, 6, 7, and 8 6.2-32 Pressurizer Cubicle Average Pressure vs Time-Surge Line DER at the Pressurizer Nozzle Nodes 9, 10, 11, and 12 6.2-33 Pressurizer Cubicle Average Pressure vs Time-Surge Line DER at the Pressurizer Nozzle Nodes 13 and 15 6.2-34 Pressurizer Cubicle Average Pressure vs Time-Surge Line DER at the Pressurizer Nozzle Nodes 14, 16, 17, and 19 6.2-35 Pressurizer Cubicle Average Pressure vs Time-Surge Line DER at the Pressurizer Nozzle Node 18 6.2-36 Pressurizer Cubicle Average Pressure vs Time-Surge Line DER at the Pressurizer Nozzle Nodes 20, 21, 22, and 23 6.2-37 Pressurizer Cubicle Average Pressure vs Time Surge Line DER in the Pressurizer Relief Tank Cubicle Nodes 1, 2, 3, and 4 6.2-38 Pressurizer Cubicle Average Pressure vs Time Surge Line DER in the Pressurizer Relief Tank Cubicle Nodes 5, 6, 7, and 8 6.2-39 Pressurizer Cubicle Average Pressure vs Time Surge Line DER in the Pressurizer Relief Tank Cubicle Nodes 9, 10, 11, and 12 6.2-40 Pressurizer Cubicle Average Pressure vs Time Surge Line DER in the Pressurizer Relief Tank Cubicle Node 13 6.2-41 Pressurizer Cubicle Average Pressure vs Time Surge Line DER in the Pressurizer Relief Tank Cubicle Nodes 14, 15, and 16 6.2-42 Pressurizer Cubicle Average Pressure vs Time Surge Line DER in the Pressurizer Relief Tank Cubicle Nodes 17, 18, and 19 6.2-43 Pressurizer Cubicle Average Pressure vs Time Surge Line DER in the Pressurizer Relief Tank Cubicle Nodes 20, 21, 22, and 23 6.2-44 Steam Generator Subcompartment Nodalization Diagram 6.2-45 Steam Generator Subcompartment Nodalization Plan View Elevation 718'-6" to 727'-2" 6.2-46 Steam Generator Subcompartment Nodalization Plan View El.

727'-0" to 740'-3" 6-xi

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES (Cont)

Figure Number Title 6.2-47 Steam Generator Subcompartment Nodalization Plan View El.

740'-3" to 767'-10" 6.2-48 Steam Generator Subcompartment Nodalization Plan View El.

767'-10" 6.2-49 Steam Generator Subcompartment Nodalization Elevation Views Sections 1-1 and 2-2 6.2-50 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq Break Nodes 1 and 5 6.2-51 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 2 and 6 6.2-52 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 3 and 4 6.2-53 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 7 and 11 6.2-54 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 8 and 12 6.2-55 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 9 and 10 6.2-56 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 13 and 17 6.2-57 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 14 and 18 6.2-58 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 15 and 16 6.2-59 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 19 and 23 6.2-60 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 20 and 24 6.2-61 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 21 and 22 6.2-62 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 25 and 27 6-xii

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES (Cont)

Figure Number Title 6.2-63 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 26 and 28 6.2-64 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 29 and 30 6.2-65 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 320 Sq In Break Nodes 31, 32, and 33 6.2-66 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 1 and 5 6.2-67 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 2 and 6 6.2-68 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 3 and 4 6.2-69 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 7 and 11 6.2-70 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 8 and 12 6.2-71 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 9 and 10 6.2-72 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 13 and 17 6.2-73 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 14 and 18 6.2-74 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 15 and 16 6.2-75 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 19 and 23 6.2-76 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 20 and 24 6.2-77 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 21 and 22 6.2-78 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 25 and 27 6-xiii

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES (Cont)

Figure Number Title 6.2-79 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 26 and 28 6.2-80 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 29 and 31 6.2-81 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 180 Sq In Break Nodes 30, 32, and 33 6.2-82 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 1 and 5 6.2-83 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 2 and 6 6.2-84 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 3 and 4 6.2-85 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 7 and 11 6.2-86 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 8 and 12 6.2-87 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 9 and 10 6.2-88 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 13 and 17 6.2-89 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 14 and 18 6.2-90 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 15 and 16 6.2-91 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 19 and 23 6.2-92 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 20 and 24 6.2-93 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 21 and 22 6.2-94 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 25 and 27 6-xiv

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES (Cont)

Figure Number Title 6.2-95 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 26 and 28 6.2-96 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 29 and 31 6.2-97 BV-2 33 Node Steam Generator Cubicle Absolute Pressure vs Time

- 707 Sq In Steam Generator Inlet Elbow Split Nodes 30, 32 and 33 6.2-98 Deleted 6.2-99 Deleted 6.2-100 Deleted 6.2-101 Deleted 6.2-102 Deleted 6.2-103 Deleted 6.2-104 Deleted 6.2-105 Deleted 6.2-106 Deleted 6.2-107 Deleted 6.2-108 Deleted 6.2-109 Deleted 6.2-110 Deleted 6.2-111 Deleted 6.2-112 Deleted 6.2-113 Deleted 6.2-114 Deleted 6.2-115 Deleted 6-xv

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES (Cont)

Figure Number Title 6.2-116 Deleted 6.2-117 Deleted 6.2-118 Schematic Diagram of Secondary System 6.2-119 Lower Bound Containment Pressure 6.2-120 Deleted 6.2-121 Quench Spray System 6.2-122 Recirculation Spray System 6.2-123 Containment Quench and Recirculating Spray Piping 6.2-124 Tandem Mechanical Seal Arrangement for Recirculation Pumps 6.2-125 Spatial Droplet Size Distribution 6.2-126 Quench Spray Coverage on Operating Floor, El. 767'-10" 6.2-127 Quench Spray Coverage in Containment Dome 6.2-128 Deleted 6.2-129 Energy Absorbed by Passive Heat Sinks - Containment Peak Pressure Analysis for LOCA (HLDER & PSDER Min. ESF-3300 Sec.)

6.2-130 Containment Depressurization Energy Removed by Recirculation Coolers for LOCA (PSDER - Min. ESF) 6.2-130A Energy Removed by Recirculation Spray Coolers, Subatmospheric Peak Pressure (PSDER - Min. ESF) 6.2-131 Combustible Gas Control System 6.2-132 Deleted 6.2-133 Deleted 6.2-134 Deleted 6.2-135 Deleted 6.2-136 Deleted 6-xvi

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES (Cont)

Figure Number Title 6.2-137 Deleted 6.2-138 Containment Internal Structure 6.2-139 Expected Long-term Circulation Patterns in Containment 6.2-140 Deleted 6.2-141 Deleted 6.3-1 Safety Injection System 6.3-2 Safety Injection Accumulators 6.3-3 DELETED 6.3-4 Pump Head Characteristic Curve - Low Head - Safety Injection Pumps 6.3-5 Pump Head Characteristic Curve - High Head - Centrifugal Charging Pumps 6.3-6 Process Flow Diagram Safety Injection System 6.4-1 DELETED 6.4-2 Section 1-1 6.4-3 Section 2-2 6.4-4 Section 3-3 6.4-5 Control Room Air Intakes, Toxic Gas Storage, and Radiation Release Points 6.5-1 DELETED 6.5-2 Supplementary Leak Collection and Release System 6.5-3 Aerosol Removal Rates Within Sprayed Region 6.5-4 Aerosol Removal Rates Within Unsprayed Region 6-xvii

BVPS-2 UFSAR Rev. 26 CHAPTER 6 ENGINEERED SAFETY FEATURES 6.1 ENGINEERED SAFETY FEATURE MATERIALS 6.1.1 Metallic Material 6.1.1.1 Materials Selection and Fabrication 6.1.1.1.1 Balance of Plant Materials are specified in accordance with ASME III, NC-2160 and NC-3120, on the basis of their compatibility with the core and containment spray solutions described in Section 6.1.1.2. General corrosion, intergranular corrosion, caustic, and chloride stress corrosion have been considered.

Mechanical properties of the materials used in the engineered safety features (ESF) are in accordance with ASME Boiler and Pressure Vessel Code Section II, Parts A, B, and C.

The specifications for the principal pressure-retaining materials for components of the ESF are compiled in Table 6.1-1.

During a loss-of-coolant accident (LOCA), negligible general corrosion of stainless steel is anticipated (Griess and Baccarella 1969). The probability of stress corrosion is reduced due to chloride control (Scharfstein and Brindley 1958), the short time at elevated temperatures (Griess and Creek 1969; Uhlig 1948; National Association of Corrosion Engineers (NACE) 1967) and the low concentration of free caustic (NACE 1967; Berry 1971). Westinghouse Electric Corporation (Westinghouse) has evaluated the integrity of the construction materials for ESF equipment when exposed to post-design basis accident (DBA) conditions (Whyte and Picone 1971). Based on the results of this investigation, as well as testing by Oak Ridge National Laboratory (ORNL) and others, the behavior of austenitic stainless steels in the post-DBA environment is acceptable.

Intergranular corrosion in the heat-affected zone of austenitic stainless steel welds will not occur due to controlled welding processes to limit sensitization (Section 1.8), limited time at elevated temperature, and the fact that the chemicals used at low concentrations are not significant intergranular corrosives (Uhlig 1948; Perry and Chilton 1973). Because all austenitic stainless steel base materials are provided in the solution annealed condition, they are immune to intergranular corrosion.

Negligible attack of carbon and low alloy steels is anticipated during a LOCA because these materials are resistant to basic solutions.

The use of stainless steel is in accordance with Regulatory Guide 1.44, as discussed in Section 1.8. Contamination and cleanliness control are provided consistent with Regulatory Guides 1.37 and 1.44, and as discussed in Section 1.8.

6.1-1

BVPS-2 UFSAR Rev. 26 Insulation for austenitic stainless steel is in accordance with Regulatory Guide 1.36 as discussed in Section 1.8. The amounts of leachable chloride, fluoride, sodium, and silicates are comparable to the values listed in Regulatory Guide 1.36. To avoid hot cracking, all production welding on austenitic stainless steel is in accordance with Regulatory Guide 1.31, as discussed in Section 1.8.

Cold worked austenitic stainless steels exhibiting a yield strength in excess of 90,000 psi are not used.

6.1.1.1.2 Westinghouse Scope of Supply Typical materials specifications used for components in the ESF are listed in Table 6.1-2. Materials utilized conform with the requirements of the ASME Boiler and Pressure Vessel Code,Section III, plus applicable and appropriate Addenda and Code Cases.

The welding materials used for joining the ferritic base materials of the ESF conform, or are equivalent, to ASME Material Specifications SFA-5.1, SFA-5.2, SFA-5.5, SFA-5.17, SFA-5.18, and SFA-5.20. The welding materials used for joining nickel-chromium-iron alloys in similar base material combination and in dissimilar ferritic or austenitic base material combination conform to ASME Material Specifications SFA-5.11 and SFA-5.14.

The welding materials used for joining the austenitic stainless steel base materials conform to ASME Material Specifications SFA-5.4 and SFA-5.9.

These materials are qualified to the requirements of the ASME Code Section III and Section IX and are used in procedures which have been qualified to these same rules. The methods utilized to control delta ferrite content in austenitic stainless steel weldments are discussed in Sections 5.2.3 and 1.8.

All parts of components in contact with borated water are fabricated of, or clad with, austenitic stainless steel or equivalent corrosion-resistant material. The integrity of the safety-related components of the ESF is maintained during all stages of component manufacture. Austenitic stainless steel is utilized in the final heat treated condition as required by the respective ASME Code Section II material specification.

Furthermore, austenitic stainless steel materials used in the ESF components are required to be handled, protected, stored, and cleaned according to recognized and accepted methods designed to minimize contamination which could lead to stress corrosion cracking. Section 5.2.3 discusses the rules covering these controls stipulated in Westinghouse specifications and provides additional information concerning austenitic stainless steel, including the avoidance of sensitization and the prevention of intergranular attack. No cold worked austenitic stainless steels having yield strengths greater than 90,000 psi are used for components of the ESF within the Westinghouse standard scope.

Westinghouse supplied ESF components within the containment that would be exposed to core cooling water and containment sprays in the event of a LOCA utilize materials listed in Table 6.1-2. These components are manufactured primarily of stainless steel or other corrosion-resistant material. Westinghouse has conducted a test program to evaluate the integrity of construction materials for ESF equipment when exposed to 6.1-2

BVPS-2 UFSAR Rev. 26 post-DBA conditions wherein the test conditions conservatively represented post-DBA conditions (Whyte and Picone 1971). The test program considered spray and core cooling solutions of the design chemical compositions, as well as the design chemical compositions contaminated with corrosion and deterioration products which may be transferred to the solution during recirculation. The effects of sodium (free caustic), chlorine (chloride),

and fluorine (fluoride) on austenitic stainless steels were considered.

Based on the results of this investigation, as well as testing by ORNL and others, the behavior of austenitic stainless steels in the post-DBA environment is acceptable. No cracking is anticipated on any equipment even in the presence of postulated levels of contaminants, provided the core cooling and spray solution pH is maintained at an adequate level.

The inhibitive properties of alkalinity (hydroxyl ion) against chloride cracking and the inhibitive characteristic of boric acid on fluoride cracking have been demonstrated.

Section 1.8 provides information concerning compliance with Regulatory Guides 1.31, 1.37, and 1.44.

The integrity of safety-related balance of plant and primary plant components of the ESF has been maintained throughout component manufacture and installation.

6.1.1.2 Composition, Compatibility, and Stability of Containment Spray and Safety Injection Coolants The method used for controlling the recirculated sump solution pH is discussed in Section 6.2.2.

Following an accident that initiates sprays, the aqueous phase inside containment is maintained in the long term by the containment sump pH control system using sodium tetraborate consistent with the description in Section 6.5.2.

Stress corrosion cracking of stainless steel piping in simulated pressure-suppression and fission product absorption sprays was investigated by Griess and Creek (1971) for the USAEC. It was found that the higher pH borate solutions (pH of 6.5 and 7.5) caused little or no stress corrosion cracking of this material.

Hydrogen generation due to the corrosion of metals and the control of the hydrogen within the containment following a LOCA are discussed in Section 6.2.5.

The vessels used for storing ESF coolants include the safety injection accumulators, and the refueling water storage tank (RWST).

The accumulators are filled with borated water and pressurized with nitrogen gas. The accumulators are carbon steel clad with austenitic stainless steel. Principal design parameters of the accumulators are listed in Table 6.3-1.

6.1-3

BVPS-2 UFSAR Rev. 26 The RWST is the source of borated cooling water for quench spray and safety injection. The RWST is austenitic stainless steel. Principal design parameters of the RWST are given in Section 6.2.2.2.

The containment sump pH control system contains sodium tetraborate in stainless steel baskets. Principal design parameters are given in Section 6.2.2.2.

Significant corrosive attack on the vessels used for storing the ESF coolants is not expected because of the corrosion resistance of the materials used and the absence of chlorides.

The quantity and identity of all soluble acids and bases within containment, and an estimate of the time-history of the pH of the aqueous phase in the containment sump, are identified in Section 6.5.

6.1.2 Organic Materials 6.1.2.1 Balance of Plant (Inside Containment)

The organic materials existing inside the containment consist principally of paint, coatings, and insulation. There are no significant quantities of wood, plastics, lubricants, or other organic materials inside the containment.

The containment is composed of concrete and steel, neither of which is subject to radiolytic or pyrolytic decomposition. All concrete surfaces of the containment structure are coated with paint with a normal dry film thickness of 0.042 in. All steel surfaces of the containment structure, except for approximately 4600 ft.2 of steel, are coated with paint with a normal dry film thickness of 0.008 in. Approximately 4600 ft.2 of steel are coated with only primer paint. The primed surfaces comprise a small percentage and will not adversely affect plant operability or significantly impact decontaminability, since steel sufaces can be easily washed down. The following criteria utilized for the selection of protective coatings and paints for use within the containment:

1. Major surfaces

a. This category includes the containment liner, structural steel, large equipment supports, hangers, embedments, the polar crane, restraints, etc.

b. The coatings comply with Regulatory Guide 1.54, as discussed in Section 1.8.

c. The coatings are tested in an environment at least as severe (in terms of maximum temperature and pressure and their transient gradients) as those anticipated within the containment in the event of a DBA to demonstrate their ability to maintain their integrity.

d. The DBA simulation tests conducted for the purpose of validating the acceptability of the coatings to be used are, in general, in accordance with ANSI N101.2-1972.

6.1-4

BVPS-2 UFSAR Rev. 26

2. Minor surfaces

a. This category includes routine touchup of damaged coatings, spot priming of bare areas, damaged galvanizing, bolt heads, nuts, miscellaneous fasteners, tack and stud welds, concrete patches, color codings, surfaces inaccessible for optimal processing, etc.

b. Protective coating systems, which have been independently laboratory type-tested to LOCA or other applicable DBAs, are applied whenever feasible. The significant process control parameters verified by testing (for example, surface preparation and film thickness) are imposed.

However, imposing the extensive, explicit quality assurance (QA) criteria of ANSI N101.4-1972 is deemed impractical.

Appropriate quality control (QC) criteria are imposed and verified via QA surveillance. Although tested paint provides a high degree of assurance of post-DBA film integrity when applied under proper processing conditions and monitored to assure a quality application, surfaces painted in this manner are recorded as unidentified coatings in accordance with the following item 3.

3. Miscellaneous surfaces

a. This category includes components such as valve bodies, handwheels, electrical cabinetry, control panels, loudspeakers, emergency light cases, and miscellaneous off-the-shelf components.

b. Because of the impracticability of imposing the Regulatory Guide requirements on the standard shop processes used in painting these items, Regulatory Guide 1.54 is not invoked when shop priming and, subsequently, when finish painting, since the total surface of such items is relatively small when compared to the total surface area for which QA requirements are imposed. The total estimated surface area covered by unqualified paint (that is, coating work which is not in accordance with any particular material, process, or QC criteria) is recorded. This category includes certain portions and subassemblies of components and equipment generally painted in accordance with the full QA requirements outlined in the previously mentioned item 1 (for example, it is not practical to procure pipe hanger spring cans with anything other than unqualified paint).

c. In general, stainless steel and corrosion-resistant alloys are not painted.

d. No special QA requirement is imposed when painting surfaces which will be insulated.

6.1-5

BVPS-2 UFSAR Rev. 26 The amounts of such coatings are evaluated to ensure that the coatings do not prevent the ESF from performing their intended function.

Final choice of paint type depends on the results of testing. Tests to date indicate neither radiolytic nor pyrolytic decomposition.

Overcoating is not intended during the service life. However, when specific areas require it, the total coating thickness expected to be accumulated will not be in excess of the thicknesses qualified by simulated DBA tests or these areas will be evaluated as discussed previously for miscellaneous surfaces.

Plastics and elastomers used in ESF components are selected based on their ability to maintain satisfactory properties after exposure to design radiation levels. No adverse interactions with ESF are likely as a result of materials released by radiation decomposition or chemical reaction of the organic materials in the post-accident environment.

6.1.2.2 Westinghouse Scope of Supply Compared with the total painted surfaces inside the containment, the painted surfaces of Westinghouse-supplied equipment comprise a small percentage. Table 6.1-3 quantifies the significant amounts of protective coatings on Westinghouse-supplied components located inside the containment building.

For large equipment requiring protective coatings (Table 6.1-3),

Westinghouse specifies or approves the type of coating systems utilized; requirements with which the coating system must comply are stipulated in Westinghouse process specifications which supplement the equipment specifications. For these components, the generic types of coatings used are zinc-rich silicate or epoxy-based primer with or without chemically-cured epoxy or epoxy-modified phenolic top coat.

The remaining equipment, which requires protective coatings on much smaller surface areas, is procured from numerous vendors. For this equipment, Westinghouse specification require that high quality coatings be applied using good commercial practices. Table 6.1-3 identifies the typical types of equipment and the approximate quantities of protective coatings on such equipment.

Westinghouse has conducted tests to evaluate the suitability, during post-DBA conditions, of protective coatings to be used in the reactor containment (Westinghouse 1971). Tests have shown that certain epoxy and modified phenolic systems are satisfactory for use in the containment.

This evaluation considered resistance to high temperature and chemical conditions anticipated during a LOCA, as well as high radiation resistance.

Information regarding compliance with QA requirements for protective coatings has been submitted to the U.S. Nuclear Regulatory Commission (USNRC) for review (Westinghouse 1977) and has been accepted as satisfactory (USNRC 1977).

6.1-6

BVPS-2 UFSAR Rev. 26 6.1.3 References for Section 6.1 Berry, W.E. 1971. Corrosion in Nuclear Applications. John Wiley and Sons, New York, N.Y.

Griess, J.C. and Baccarella, A.L. 1969. Design Considerations of Reactor Containment Spray Systems, The Corrosion of Materials in Spray Solutions.

USAEC Report ORNL-TM-2412, Part III, Oak Ridge National Laboratory.

Griess, J.C. and Creek, G.E. 1969. Design Considerations of Reactor Containment Spray Systems, Corrosion Tests with Low pH Spray Solution.

USAEC Report ORNL-TM-2412, Oak Ridge National Laboratory.

Griess, J.C., and Creek, G.E. 1971. Design Considerations of Reactor Containment Spray Systems, The Stress Corrosion Cracking of Types 304 and 316 Stainless Steel in Boric Acid Solutions. USAEC Report ORNL-TM-2412, Part X, Oak Ridge National Laboratory.

National Association of Corrosion Engineers (NACE) 1967. Proceedings of Conference: Fundamental Aspects of Stress Corrosion Cracking. Ohio State University.

Perry, R.H. Chilton, C.H. 1973. Chemical Engineers Handbook. Fifth Edition, McGraw-Hill Book Co., New York, N.Y.

Scharfstein, L.R. and Brindley, W.F. 1958. Corrosion, Vol. 14, National Association of Corrosion Engineers, Houston, Texas, 558t.

Uhlig, H.H. 1948. Corrosion Handbook. John Wiley and Sons, New York, N.Y.

U.S. Nuclear Regulatory Commission (USNRC) 1977. Personal communication between C.J. Heltemes, Jr., USNRC, Quality Assurance Branch and C.

Eicheldinger, Westinghouse, PWRSD, letter-dated April 27, 1977.

Westinghouse Electric Corporation (Westinghouse) 1971. Evaluation of Protective Coatings for Use in Reactor Containment. WCAP-7825.

Westinghouse 1977. Personal communication between C. Eicheldinger Westinghouse, PWRSD, Nuclear Safety Department and C. J. Heltemes, Jr.

USNRC, Quality Assurance Branch), letter NS-CE-1352 dated February 1, 1977.

Whyte, D.D. and Picone, L.F. 1971. Behavior of Austenitic Stainless Steel in Post Hypothetical Loss of Coolant Environment. WCAP-7803, Westinghouse.

6.1-7

BVPS-2 UFSAR Tables for Section 6.1

BVPS-2 UFSAR Rev. 26 TABLE 6.1-1 TYPICAL MATERIALS EMPLOYED FOR COMPONENTS OF ENGINEERED SAFETY FEATURES SYSTEMS BALANCE OF PLANT Component and Material(1) Material(1)Specification Piping Stainless steel SA-312 Tp. 304, SA-376 Tp. 316, SA-312 Tp. 316, SA-358 Tp. 304 Cl. 1, SA-376 Tp.

304, SA-358 Tp. 316 Cl. 1 Carbon Steel SA-106, Gr. B, SA-155 Gr. KC 70 C1. 1, SA-155 Gr. C 55 Cl. 1 Fittings, connections and flanges Stainless steel SA-403 WP 304, SA-182 F 316, SA-403 WP 316 W, SA-403 WP 316, SA-182 F 304, SA-403 WP 304 W Carbon Steel SA-105, SA-234 WPB, SA-181 Gr. 1, SA-234 WPC, SA-216 WCB Bolting Studs SA-193 Gr. B6, SA-193 Gr. B7 Nuts SA-194, Gr. 6, SA-194 Gr. 2H, SA-194 Gr. 7 Containment spray nozzles SA-351 CF8, 304 S/S Recirculation spray pumps Suction casing SA-358 Tp. 304 Cl. 1 and SA-240 304L Discharge column SA-312 Tp. 304L Discharge head` SA-312 Tp. 304L, SA-182 F 304L, SA-240 Tp. 304L Discharge flange SA-182 F 304L Bolting SA-193 G4. B8 Cl. 1 Bowls A-351 CF8 Mechanical seal Tungsten carbide, carbon, SA-182-304L, SA-312-304L, SA-340-304L 1 of 3

BVPS-2 UFSAR Rev. 26 TABLE 6.1-1 (Cont)

Component and Material(1) Material(1)Specification ESF sumps Plate SA-240 Tp. 304 Trash racks 304 stainless steel Perforated plate 304 stainless steel Valves Stainless steel Bonnet studs SA-193 Gr. B6, SA-193 Gr. B8, SA-193 Gr.

B8M Valve stems 17-4 PH, 316 S/S Body castings SA-351 CF8, SA-351 CF8M Body forgings SA-182 F304, SA-182 F316 Packing John Crane 187-I-CR or equal Bonnet nuts SA-194 Gr. 8, SA-194 Gr. 6, SA-194 Gr. 7 Carbon steel Bonnet studs SA-193 Gr. B7 Bonnet nuts SA-194 Gr. 2H/7 Valve stems SA-182 F6 tempered, forged 13% Cr Body castings SA-216 WCB Body forgings SA-105 Gr. II Packing John Crane 187-I-CR or equal Seat seals and seal rings Metallic Austenitic SS, SA-182 F6, 12% Cr Plastic Polyethylene, nylon Elastomers Rubber, viton Recirculation spray coolers Tubes SA-249 Tp. 304 Shell 304 S/S Tube sheet 304 S/S Cross and long baffles 304 S/S Channel and channel Carbon steel cover Quench Pumps Casing SA-351 CF8 Impeller SA-351 CF8 Impeller rings 304 S/S overlayed with Colmonoy 6 Casing rings 304 S/S overlayed with Colmonoy 4 Shaft 17-4pH - condition H1100 2 of 3

BVPS-2 UFSAR Rev. 26 TABLE 6.1-1 (Cont)

Component and Material(1) Material(1)Specification Welding material Ferritic steels ASME SFA-5.1, 5.5, 5.17, 5.18, and 5.20 Austenitic stainless ASME SFA-5.4 and 5.9 steel Ferritic to austenitic Field installation ASME SFA-5.11 and 5.14 or ASME SFA-5.4 and 5.9 Type 309 S/S Shop fabrication ASME SFA-5.11 and 5.14 or ASME SFA-5.4 and 5.9 Type 309 S/S NOTE:

(1) Materials listed in this table may have been replaced with materials of equivalent design characteristics. The term equivalent is described in UFSAR Section 1.12, Equivalent Materials.

3 of 3

BVPS-2 UFSAR Rev. 7 TABLE 6.1-2 TYPICAL MATERIALS EMPLOYED FOR COMPONENTS OF ENGINEERED SAFETY FEATURES SYSTEMS NUCLEAR STEAM SUPPLY SYSTEM Component and Material(1) Material(1)Specification Valves Bodies SA-182, Tp. F316; SA-351, Gr. CF8 or CF8M, SA-105, Gr. II, SA-476, Tp. 316; A216 Gr. WCB Bonnets SA-182, Tp. F316; SA-351, Gr. CF8 or CF8M, SA-479; Tp. 316; A216 Gr. WCB; Haynes alloy No. 6B; SA-240, Tp. 316 Discs SA-182, Tp. F316; SA-564, Gr. 630; SA-351, Gr. CF8 or CF8M, SA-479, Tp. 316; Stellite No. 6 Pressure-retaining bolting SA-453, Gr. 660, SA-193 Gr. B7 Pressure-retaining nuts SA-453, Gr. 660; SA-194, Gr.

6; SA-193, Gr. B6, SA-194, C12H, SA-194 Gr. 7 Auxiliary heat exchangers Heads SA-240, Tp. 304; SA-515-70 Nozzle necks SA-182, Gr. F304; SA-312, Tp. 304; SA-240, Tp. 304 Tubes SA-213, Tp. 304; SA-249, Tp. 304 Tubesheets SA-182, Gr. F304; SA-240, Tp. 304; SA-516, Gr. 70 with Stainless Steel Cladding A-7 Analysis Shells SA-240 and SA-312, Tp. 304 LS SA-285C 1 of 2

BVPS-2 UFSAR Rev. 7 TABLE 6.1-2 (Cont)

Component and Material(1) Material(1)Specification Auxiliary pressure vessels, tanks, filters, etc.

Shells and heads SA-351, Gr. CF8A; SA-240, Tp. 304; SA-264 Clad Plate of SA-537, Cl 1 with SA-240, Tp. 304 Clad and Stainless Steel Weld Overlay A-8 Analysis Flanges and nozzles SA-182, Gr. F304; SA-350, Gr.

LF2 with SA-240, Tp. 304 and Stainless Steel Weld Overlay A-8 Analysis Piping SA-312 and SA-240, Tp. 304 or Tp. 316 Seamless Pipe fittings SA-403, Tp. 304 Seamless Closure bolting and nuts SA-193, Gr. B7 and SA-194, Gr. 2H Auxiliary pumps Pump casing and heads SA-182, Gr. F304 or F316; SA-351, Gr. CF8 Flanges and nozzles SA-182, Gr. F304 or F316; SA-403, Gr. WP 316L Seamless Stuffing or packing box SA-182, Gr. F304; SA-351, cover Gr. CF8 or CF8M; SA-240, Tp. 304 or 316 Closure bolting and nuts SA-193, Gr. B6 and B7; SA-453, Gr. 600; and Nuts, SA-194, Gr. 6 and 7 SA-193, Gr. B6X; SA-194, Gr. 2H Piping SA-312, Tp. 304 or 316 Seamless Pipe fittings SA-403, Gr. WP316L Seamless SA-182, F316 NOTES (1) Materials listed in this table may have been replaced with materials of equivalent design characteristics. The term equivalent is described in UFSAR Section 1.12, Equivalent Materials.

2 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.1-3 PROTECTIVE COATINGS ON WESTINGHOUSE-SUPPLIED EQUIPMENT INSIDE THE CONTAINMENT Painted Surface Area Component (ft(2)

Reactor coolant pump motors 2,500 Accumulator tanks 4,200 Manipulator crane 2,600 Other refueling equipment 2,125 Remaining equipment 1,300 (such as valves, auxiliary tanks, heat exchanger supports, transmitters, alarm horns, and small instruments) 1 of 1

BVPS-2 UFSAR Rev. 26 6.2 CONTAINMENT SYSTEMS 6.2.1 Containment Functional Design 6.2.1.1 Containment Structure 6.2.1.1.1 Design Bases The containment structure is designed in accordance with General Design Criteria (GDC) 13, 16, 38, 50, and 64 (Section 3.1). The criteria are amplified as follows:

1. The peak calculated containment pressure following the design basis accident (DBA) remains below the containment design pressure of 45 psig. The loss-of-coolant accident (LOCA),

which results in the highest calculated containment pressure, is the DBA for the containment structure (containment integrity DBA) design.

2. A spectra of accidents and single failures, combined with simultaneous occurrences such as seismic events and loss of offsite power (LOOP), are considered to establish the containment peak pressure and depressurization time.

3. The containment minimum design pressure is the minimum calculated pressure that results from inadvertent actuation of the containment depressurization system. Section 6.2.1.1.3.5 provides a description of the analysis.

4. The design bases for the containment internal structures (subcompartments) are given in Section 6.2.1.2.

5. The sources and rates of mass and energy released into the containment for the LOCA and the main steam line break (MSLB) accidents are described in Sections 6.2.1.3 and 6.2.1.4, respectively.

6. The design bases of the containment depressurization system as an energy removal system are described in Section 6.2.2.

7. The calculated containment pressure is reduced to less than half of the peak pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

8. After depressurization, the containment pressure is maintained below atmospheric pressure for at least 30 days. The accident conditions which result in the maximum subatmospheric pressure is the DBA for maintaining subatmosphere pressure.

9. The capability for post-accident pressure reduction and energy removal from the containment under various single failure conditions in the engineered safety features (ESF) is discussed in Section 6.2.2.

6.2-1

BVPS-2 UFSAR Rev. 26

10. The containment system is designed to limit fission product leakage following a LOCA. Chapter 15 discusses the analysis.

11. The bases for the containment back pressure analysis used for the emergency core cooling system (ECCS) analysis are discussed in Section 6.2.1.5.

12. Instrumentation capable of operating in the post-accident environment is provided to monitor the containment atmosphere pressure and temperature and the sump water temperature and level following a LOCA. Section 7.5 describes the equipment and the parameters recorded. Section 3.11 gives a discussion of the qualification of the instrumentation for post-accident environment operation.

13. The containment structure is designated Seismic Category I.

(Section 3.2),

6.2.1.1.2 Design Features The containment structure is a cylindrical, carbon steel-lined, reinforced concrete structure which encloses the components and major piping within the reactor coolant pressure boundary. The structure is designed to contain the radioactive fluids and fission products that may result from postulated LOCAs inside the containment.

The containment is an atmospheric type containment. During normal operation, the containment structure is maintained near atmospheric pressure (typically 12.8 to 14.2 psia).

Arrangements and crossections of the containment structure are shown on Figures 3.8-1, 3.8-2, 3.8-3, 3.8-4, 3.8-5, 3.8-6 and 3.8-7. The structural design is described in Section 3.8. The design provisions to protect the containment structure and ESF systems against loss of function from dynamic effects (for example, missiles and pipe whip) that could occur following postulated LOCAs are described in Sections 3.5 and 3.6.

Applicable codes and standards are identified in Section 3.8.1.2.

The containment structure is designed to withstand internal pressurization from high energy pipe breaks within the structure (Section 6.2.1.1.3.2) and external pressurization due to inadvertent actuation of the containment depressurization system (Section 6.2.1.1.3.5). The containment maximum internal design pressure is 45.0 psig and the minimum internal design pressure is 8 psia.

The internal design of the containment structure allows air to circulate freely. All cubicles and compartments within the containment are open at their tops to allow air circulation. Convective mixing in conjunction with containment spray assures a uniform mixture of hydrogen in the containment. Section 6.2.5 discusses combustible gas control in the containment.

The containment structure is equipped with a containment sump located at the outer wall of the containment (Figure 6A-1). Sufficient openings 6.2-2

BVPS-2 UFSAR Rev. 26 exist in the upper floors and structure of the containment to allow water generated from accident conditions to drain to the containment sump.

Spray water that falls into the refuelinq cavity will drain into either the fuel transfer canal or the reactor cavity. The entrapment of this water in these areas has been accounted for in the DBA analysis as discussed in Section 6A.2.7.

Section 6.2.2.3.2 discusses the net positive suction head (NPSH) requirements for the recirculation spray pumps.

6.2.1.1.3 Design Evaluation This section describes the method used to evaluate the functional capability of the containment design. It also describes the computer code MAAP-DBA developed by Fauske and Associates (FAI 2005) that is utilized to evaluate the spectrum of pipe ruptures.

Sections 6.2.1.1.3.6 through 6.2.1.1.3.8 provide the results of analyzing a spectrum of pipe ruptures for the primary and secondary systems.

6.2.1.1.3.1 Internal Pressures A pressure peak occurs near the end of the initial blowdown of the RCS after a double-ended rupture (DER) of either a hot leg or in the crossover leg at the pump suction. Its magnitude is a function of the following parameters:

1. The containment free volume,

2. The mass of air inside the containment structure (a function of initial pressure and temperature),

3. The amount of mass and energy that flows out of the break during the initial blowdown of the RCS, and

4. The rate of heat removal from the containment atmosphere by the containment heat sinks.

The highest peak pressure occurs after a DER of a hot leg. This event releases the most energy to the containment atmosphere during the initial blowdown, since the hot leg pipe size is larger than that of an RCS pump suction, and there is no resistance to flow due to an RCS pump as is the case with a pump suction DER. The magnitude of this peak pressure is independent of the active ESF (minimum or normal) because they do not become effective until after the first peak is reached. However, the accumulators do have a small effect on the first peak.

Following the core reflooding period, the containment depressurization systems and the containment passive heat sinks remove energy from the containment atmosphere at a rate sufficient to rapidly reduce the pressure. The depressurization time is a function of the following parameters:

6.2-3

BVPS-2 UFSAR Rev. 26

1. The containment free volume,

2. The mass of air inside the containment structure,

3. The rate of heat removal (or addition) from (or to) the containment atmosphere by the passive heat sinks within the containment structure,

4. The rate of heat removal from the containment atmosphere by the CHRS,

5. The rate of mass and energy release to the containment from the break following the end of core reflooding, and

6. The mass of nitrogen added to the containment from the accumulators gas space.

After the containment is depressurized, the systems continue to remove energy from the containment at a rate sufficient to maintain the containment depressurization. The passive heat sinks may add energy back to the containment atmosphere following depressurization. The containment experiences two other pressure peaks when the capacity of the depressurization systems are reduced. The first pressure peak occurs after one-half of the operating recirculation pumps are switched from containment spray to core injection. The second pressure peak occurs when the refueling water storage tank (RWST) empties (termination of quench spray).

6.2.1.1.3.2 Containment Analysis Analytical Model The MAAP-DBA code was developed to allow the calculation of containment response attributes for a spectrum of postulated LOCA and main steam line break sequences as part of design basis calculations for BVPS-2 containment. The containment assessment for a design basis application is implemented in a manner consistent with the NRC guidance provided in the Standard Review Plan. This includes the use of Tagami and Uchida heat transfer correlations for the quantification of the passive heat sink responses. The spectrum of containment response attributes to be quantified include the peak containment pressure, the short and long-term containment temperature, the containment liner temperature, the long-term sump water temperature, the available NPSH for ECCS and containment spray pumps, and the maximum service water outlet temperature for the containment heat removal heat exchanger. To address this set of containment response attributes for the spectrum of loss of coolant accident break sizes, both single node and multiple node containment models are used. The single node models apply for those design basis sequences and attributes that employ the Tagami and Uchida heat transfer correlations. For the multiple node applications, a heat and mass transfer analogy based on natural convection is used.

A single node model is used to calculate peak containment pressure and containment liner temperature as well as post accident containment global gas temperature profiles for equipment qualification. A multi-node model is used for NPSH and sump water temperature. This provides improved 6.2-4

BVPS-2 UFSAR Rev. 26 accountability of water hold up for NPSH and debris transport calculations.

The design bases events analyzed and evaluated include the rupture of a pipe in the Reactor Coolant System (LOCA) and the Main Steam Line Break (MSLB) between the top of the steam generator and the penetration through the containment wall. See Section 6.2.1.1.3.6 for the LOCA results and Section 6.2.1.1.3.7 for the MSLB results. LOCA breaks were evaluated at hot leg, cold leg, and pump suction locations. All these design bases events assumed a 2917.4 MWt core power. Evaluations for the limiting containment design basis events were evaluated to assess the peak containment pressure, peak containment gas temperature, long term temperatures within the containment, and the peak liner temperatures attributes. To ensure that the most conservative value of each of the attributes was identified and evaluated the most conservative value (max or min) of each input parameter for each attribute was selected.

The mass and energy released to the containment can also vary depending upon a combination of variables such as break size, break location, single active failure, power level, and containment air pressure at the time of the break. The consequences of the breaks can further vary dependent on a variety of possible single active failures that may occur concurrent with the breaks and affect the availability of engineered safety features (ESFs). Single active failures that were considered to identify the worst single failure that maximizes the challenge to the containment integrity include:

1. the failure of a single train of engineered safety features such as might occur with the failure of an Emergency Diesel Generator (EDG) coincident with a loss of off-site power,

2. the single failure of the containment isolation phase B signal (CIB), which would result in the failure of one complete train of quench and recirculation sprays to start, which means that the remaining train of sprays would be available to cool the containment atmosphere,

3. the failure of a service water pump to supply cooling water to one train of the recirculation spray heat exchangers (two heat exchangers) which are part of the containment heat removal system,

4. the failure of a timer start relay which would result in the failure of one train of recirculation spray,

5. the failure of a main steam isolation valve (MSLB only),

6. the failure of a main feedwater isolation valve (MSLB only),

7. the failure of one train of quench spray Operational conditions in the reactor coolant system including the reactor and steam generators were also examined for the worst possible conditions that could influence the mass and energy releases from the break.

6.2-5

BVPS-2 UFSAR Rev. 26 Section 6.2.1.3 discusses the spectrum of LOCA mass and energy releases and Section 6.2.1.4 discusses the spectrum of MSLB mass and energy releases used as input to the containment analysis.

Thus, the containment analyses were performed in a manner that ensured that the evaluations identified and examined the most severe challenges to successful operation of the containment and its supporting mechanical and electrical safety systems.

6.2.1.1.3.2.1 Application of MAAP-DBA to Containment Analysis The MAAP-DBA Generalized Containment Model (GCM), which is discussed in detail in BVPS-2 Licensing Amendment 153, was used for DBA evaluations.

6.2.1.1.3.2.2 Parameter File/Nodalization Nodalization The application of the MAAP-DBA containment model to a commercial nuclear power plant begins with the characterization of the containment building geometry, emergency safeguard systems, etc., in the plant specific parameter file. This parameter file includes specifications for the entire plant, with primary emphasis on the containment information.

Secondary information, such as generic data, including the reactor core and the reactor coolant system, are also included in the parameter file.

The mass and energy releases are specified as external input data.

When formulating a containment parameter file, the most important decision lies in the specification of the number of nodes used to represent the building. To be consistent with the previous BVPS DBA analyses, the evaluations for peak pressure and temperature are performed using single node models. However, those evaluations which are sensitive to potential water accumulation (holdup) in various locations within the building are performed with multi-node models.

There are a few guidelines to be followed for multi-node models.

1. Each building region which is a separate room or compartment with limited connections (flow paths) to the remainder of the building should be treated as a separate node. For example, the reactor cavity in a typical PWR large dry containment is generally separated from the remainder of the building by a thick concrete biological shield. Furthermore, the walls around the in-core instrument tubes that penetrate through the bottom of the reactor vessel in most PWR designs segregate the region from other compartments. Hence, this region should be one of the nodes. Furthermore, specific rooms such as the in-core instrumentation seal table room that may also be compartmentalized for shielding purposes should also be a separate node.

2. Typically the design basis accident conditions include analyses for a large break RCS LOCA as well as evaluations for a main steam line break. For those accident analyses requiring a multi-node model such as maximum recirculation sump temperature 6.2-6

BVPS-2 UFSAR Rev. 26 following a large break LOCA, the containment nodalization should include the region surrounding the reactor coolant system, the loop compartment(s), and the region above the operating deck as individual nodes. In this regard, the LOCA conditions considered include any sensitivities related to whether the LOCA is postulated to occur in any of the reactor coolant loops. Consequently, if the reactor coolant loops are in one large compartment, a single node is sufficient.

Conversely, if the loop compartments and other RCS components, such as the pressurizer, are in individual rooms, then the nodalization scheme should be expanded to include each of these compartments as a separate node.

3. An important parameter of the DBA evaluations is the sump temperature under accident conditions. Thus, that region in the bottom of the containment which includes the recirculation sump and the floor of the containment outside the reactor cavity should be considered as a separate node.

4. The nodalization scheme needs to be sufficient to represent the potential for light gas stratification in the top of the containment building. Consequently, there should be at least two nodes (one above the other) in the region above the operating deck where light gases, such as steam could accumulate. This is the only region of the BVPS containment model that uses multiple nodes to represent an open region.

Comparisons with large scale experiments have shown that the number of nodes required to represent a large dry containment building is between 10 and 30 nodes depending upon the extent of compartmentalization inside the containment. In general, with the relatively open configuration above the operating deck and in the annular region, the number of nodes to be used to represent a typical large dry containment is less than 20. The nodalization used for Beaver Valley Unit 2 included 17 nodes.

Flow Paths/Junctions The junctions or flow paths connecting the various nodes are also defined in the TOPOLOGY section of the parameter file. These junctions include those paths which are doorways, hatchways, open areas, grating, etc., as well as those flow paths which could be identified for the ventilation systems such as the Containment Air Recirculation Fans. These junctions enable the major flow transport paths to be clearly specified and quantified with respect to their available area, their potential to be flooded by water accumulation, the potential for water accumulation within containment nodes, etc. Hence, this topology entry in the parameter file is very important in providing a realistic characterization, including the potential for global and countercurrent natural circulation, of the containment response to DBA conditions.

Structural Heat Sinks Structural heat sink information including the surface areas, thicknesses, materials, whether they are steel lined, whether the outer surface is 6.2-7

BVPS-2 UFSAR Rev. 26 painted, etc., is also described in the parameter file. During DBA conditions the heat sink response is typically sufficiently slow that only a few heat sinks have the thermal conduction developed through the entire width of the heat sink. Nonetheless the MAAP-DBA parameter file has the capability for all of these heat sinks to be identified as two-sided structures, thereby enabling the parameter file to be used for DBA evaluations as well as for accident analyses evaluations over an extended time period, i.e., hours or days. To accomplish this, the node facing each heat sink surface is identified in the parameter file, i.e., a heat sink face is pointed to the specific node with which it interacts, and its opposite face is pointed to another node.

Engineered Safeguards Engineered safeguards that are specific to the containment are also defined in the parameter file, including the containment spray pumps, and the heat exchangers that are used to remove decay heat from the containment during recirculation. The configuration of the ECCS and containment spray pumps must be specified in terms of:

1. those pumps which take suction only from the Refueling Water Storage Tank (RWST),

2. those pumps which take suction from the RWST and are switched over to take suction from the containment sump at containment recirculation,

3. those pumps which only take suction from the containment sump under recirculation conditions.

Heat removal capabilities are identified with the type of pumping system.

Issues related to the single failure criterion are addressed in the input decks assembled for each sequence. The parameter file is meant to represent the nominal operating condition for specific systems. As part of this, the configuration also defines whether any pumps are piggybacked to the discharge of a lower pressure pump to increase their discharge pressure.

6.2.1.1.3.2.3 Treatment of the Mass and Energy Releases As discussed in Section 6.2.1.3 and Section 6.2.1.4, there are a number of LOCA and MSLB accident conditions that are analyzed for the containment response. The discharge from the break location is the mass and energy source that is input into the containment analysis.

There are two means of treating the mass and energy releases input to MAAP-DBA for LOCA. For instance, the evaluations for the maximum temperatures within containment following an accident focus on those set of conditions which result in the hottest steam being released to the containment atmosphere, i.e., a double-ended break where the mass and energy streams from the two sides of the break (the hot water flow rate from the cold leg side and the steam flow from the steam generator side) are discharged into the containment atmosphere as separate streams.

Conversely, the evaluations for the minimum available NPSH focus on those 6.2-8

BVPS-2 UFSAR Rev. 26 conditions which could result in the maximum sump temperature and the largest recirculation flow rate to maximize the frictional losses. In this case, the mass and energy releases from the two sides of the guillotine break are mixed together before entering the containment such that there is minimal steam released to the containment environment and the temperature of the water added to the containment sump is maximized.

Therefore, from this description, the mass and energy releases for a similar type of break are manipulated to cover the potential uncertainties related to the break configuration and how this influences the specific attributes that must be evaluated to ensure that the containment is capable of remaining within its design basis envelope for all of the accident conditions considered.

For those accident sequences, which result in the long term response of the containment, after containment recirculation, the mass and energy releases from the RCS are dependent upon the temperature of the containment sump due to the recirculated and injected water. (The sump water may pass through a heat exchanger prior to this injection). Since the sump temperature changes with time, long term evaluations require feedback from the containment evaluation. Specifically, the mass and energy releases need the sump water temperature history such that long term analyses properly incorporate the decreasing temperature of the containment sump. Because of this, the long-term (greater than 3600 seconds) mass and energy release calculations are performed with the MAAP-DBA code. These input functions are used to incorporate the sump water temperature history, and are consistent with the methodology discussed in Section 6.2.1.3. These user defined functions are characterized for the long term discharge from the break for both a mixed discharge and for an unmixed discharge of steam and water. In both cases, the flow rates that are used are those calculated with the methodology discussed in Section 6.2.1.3 and only the specific enthalpies of the discharge flows are calculated to represent the influence of the time dependent RCS injection temperature as the containment cools.

6.2.1.1.3.2.4 Influence of Varying Containment Operating Conditions Another aspect of the evaluation is the spectrum of operating conditions that could be experienced by the containment at the time that the accident is initiated. For example, the containment pressure may vary between 12.8 and 14.2 psia. Furthermore, the containment atmosphere temperature could be at its maximum value or its minimum value. These types of operating parameters have an influence on the specific attribute being evaluated, and the different boundaries of these operating conditions were investigated to determine the set of conditions which maximizes the challenge to the attributes being evaluated.

6.2.1.1.3.2.5 Input Parameters, Assumptions and Model Table 6.2-1 through Table 6.2-2a list the heat sink input data and net free volumes utilized in the containment analysis. Table 6.2-3 lists the design evaluation parameters. Table 6.2-4 lists the key input data to MAAP-DBA for pressure calculations.

6.2-9

BVPS-2 UFSAR Rev. 26 Based on detailed drawing reviews and site visits and considering the plant-specific features, it was determined that the containment would most appropriately be represented with a 17 node model of the containment (see Figures 6.2-1, 6.2-2 and 6.2-3). This scheme enables the model to represent the individual compartments for each of the three Reactor Coolant System (RCS) loops, the recirculation sump region, the reactor cavity region, the annular region outside of the cooling loops, and three nodes above the operating deck. Using multiple nodes above the operating deck enables the stratification of light gases to be calculated when this is part of the evaluation. The physical flow paths between the containment compartments are also included, as junctions, in the containment model. The junction areas and loss coefficients are based on the plant dimensions and are summarized in Table 6.2-5. MAAP-DBA calculates the quasi steady state nodal pressure distribution at each time step such that inertial effects due to flow acceleration are not required nor calculated due to the relatively slow containment pressurization (containment pressure benchmarks have demonstrated this behavior). Thus, inertial coefficients for each junction are not included in the parameter file.

Major parts of the parameter file include the individual nodal volumes that make up the total containment volume, the volume vs. height function of these nodes such that water accumulation can be properly evaluated, the structural and containment heat sinks within these individual nodes, the surface characterization of the heat sink in terms of whether the surface is painted, how it is painted (number of layers, their thicknesses and the thermal conductivity of each layer), whether the heat sink is concrete, steel or steel lined concrete, etc. Furthermore, the setpoints for system actuation, pump curves, heat exchanger capacities, etc., are also contained in the unit-specific parameter file. The containment node volumes, metal heat sink areas and masses, and concrete heat sink areas and thicknesses included in the containment model, are tabulated in Tables 6.2-2 and 6.2-2a. The heat sinks include structural steel, concrete liners, ventilation ducts and supports, pipes, pipe supports and restraints, and heavy equipment.

The characteristics of the containment spray systems (header elevations and flow rates) are also included in the parameter file. A quench spray (QS) system is actuated on a containment high-high pressure signal and after a start delay, directs cold water from the RWST to the quench spray ring header in containment. A variable quench spray flow rate in gpm is used per train as determined by the pump curve.

After a RWST Low Level coincident with a Containment Pressure High-High Signal, a recirculation spray (RS) system is actuated, which directs water from the containment recirculation sump, through a heat exchanger, and then to the recirculation spray ring header in containment. Containment heat removal is accomplished by the RS heat exchanger. The recirculation spray flow rate as determined by the pump curve is input per train.

6.2.1.1.3.2.6 Acceptance Criteria An acceptance criterion was developed for each of the types of analyses being performed. These are as follows:

6.2-10

BVPS-2 UFSAR Rev. 26

1. peak containment pressure less than 45 psig and the pressure is less than half the peak pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

2. peak containment temperature less than the equipment qualification curve,

3. peak liner temperature less than 280F.

6.2.1.1.3.3 Mass and Energy Releases to Containment Loss-of-Coolant Accident The rates of mass and energy release to the containment during the blowdown, reflood, and post-reflood periods are discussed in Section 6.2.1.3 for pipe failures at the following locations:

1. Hot leg (between vessel and steam generator)

2. Pump suction (between steam generator and pump)

Main Steam Line Break Accident Section 6.2.1.4 discusses the mass and energy release analysis for secondary system pipe rupture inside the containment.

6.2.1.1.3.4 Description of Passive Heat Sinks The passive heat sinks include the containment structure, internal concrete, and miscellaneous metal equipment within the containment. The metal heat sinks are distributed within the 17 containment nodes and the concrete heat sinks are modeled with >75 elements for the purpose of this analysis. The thermal properties of the heat sink materials are given in Table 6.2-1.

A description of the sinks used in the containment analysis with a listing of the metal mass and surface area for the metal heat sinks and the slab thicknesses and surface areas for the concrete heat sinks is given in Table 6.2-2.

Each concrete heat sink except the reactor cavity and lower compartment floors are treated as two sided. All of the metal heat sinks are exposed to the containment atmosphere and are treated as single sided slabs.

Resistance to heat transfer at the liner-concrete interface is considered in the containment analysis by use of a conservatively low value of thermal contact conductance of 100 Btu/hr-ft2-F. Since the steel liner is used as a form for pouring of the concrete, and since the concrete mix is very wet, the liner is, in effect, in good thermal contact with the concrete.

The model considers transient heat conduction to the containment structure through the composite thermal resistance made up of the paint film on the steel liner, the liner itself, the liner-concrete interface, and the 6.2-11

BVPS-2 UFSAR Rev. 26 concrete. See Section 6.2.1.1.3.2.2.1 for further discussion of the treatment of the structural heat sinks.

6.2.1.1.3.5 External Pressure Inadvertent operation of the containment depressurization system will cause a decrease in the pressure inside the containment, thereby increasing the normal external pressure differential on the containment structure.

The analysis of maximum external differential pressure assumes inadvertent actuation of the quench spray system caused by a single spurious containment isolation Phase B (CIB) signal.

The maximum external pressure differential is calculated by determining the minimum attainable pressure inside the containment and subtracting this value from the average barometric pressure (14.36 psia).

The minimum pressure possible is calculated to be 11.4 psia, based on the following assumptions:

1. Minimum initial air partial pressure,

2. Maximum initial containment temperature, and

3. Final containment temperature which equals the minimum RWST, temperature.

Table 6.2-3 depicts the containment design evaluation parameters.

6.2.1.1.3.6 Loss-of-Coolant Accident Results The LOCA containment transient analysis was performed with the MAAP-DBA computer code (Section 6.2.1.1.3.2) for a spectrum of pipe break locations. Containment analysis were conducted for large break LOCAs including a double-ended hot leg break (DEHL) and a double-ended pump suction break (DEPS).

Sensitivity analyses were conducted to determine the most challenging set of plant conditions related to the specific attribute being evaluated (i.e., the peak calculated containment pressure). The initial containment conditions which yield the highest peak calculated containment pressure are as follows:

1. Initial containment pressure of 14.2 psia,

2. Initial containment temperature of 108°F, and

3. Initial containment relative humidity of 15%.

The mass and energy releases used in this analysis are discussed in Section 6.2.1.3.

6.2-12

BVPS-2 UFSAR Rev. 26 Table 6.2-6 summarizes the peak containment pressures for the large break LOCA cases. The DEHL break case results in the peak calculated pressure of 44.8 psig. The pressures are reported as psig and referenced to an atmospheric pressure of 14.3 psi. As illustrated by this table, all of these sequences result in a pressurization which is less than the design basis value of 45 psig.

Figure 6.2-5 and Figure 6.2-6 illustrate the containment pressure and temperature time history for the DEHL break case. Figure 6.2-7 and 6.2-8 illustrate the containment pressure and temperature time histories for the DEPS break with minimum containment safeguards (Min SI) and Figure 6.2-9, and 6.2-10 illustrate the containment pressure and temperature time histories for the DEPS break with maximum containment safeguards (Max SI) cases. The containment sump temperature transients for each of the DEPS cases are given on Figure 6.2-11.

The sequence of events are summarized in Tables 6.2-7, 6.2-8 and 6.2-9.

See Section 6.2.1.3 for an illustration of the energy distribution in the nuclear steam supply system (NSSS) prior to the break, at the end of the primary system blowdown, and during other periods of the postulated accident.

As mentioned previously, the acceptability of the results for the containment pressure is that the design basis break conditions analyzed using design basis methodology for the mass and energy releases to the containment must be less than the design basis structural capability of 45 psig. As demonstrated by analyses for various types of break conditions and mass and energy releases, these results meet the acceptance criteria of less then 45 psig. Also, the calculated pressure transients demonstrate that the containment pressure is reduced to below one-half of the peak pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6.2.1.1.3.7 Main Steam Line Pipe Break Results The MSLB containment transient analysis was performed with the MAAP-DBA computer code (Section 6.2.1.1.3.2). The program is used to calculate the thermodynamic state of the containment due to the mass and energy addition to the containment atmosphere.

Main steam line breaks can be postulated to occur with the plant in any operating condition, ranging from hot shutdown to full power. Because of the opposing effects of changing power level on MSLB mass and energy releases, no single power level can be singled out as a worst initial condition for the MSLB. Therefore a spectra of power levels, spanning the operating range (100.6 percent, 70 percent, and 30 percent), as well as the hot shutdown condition, have been analyzed. A spectra of MSLB accidents covering different break areas, single-active failures and reactor operating power levels are analyzed. The mass and energy releases are discussed in Section 6.2.1.4.

The single-active failures addressed include:

1. The failure of the containment isolation phase B (CIB), which results in the failure of one complete train of quench spray to 6.2-13

BVPS-2 UFSAR Rev. 26 start (this failure also results in the failure of one train of recirculation spray, however no credit is taken in the MSLB containment analysis for the recirculation spray system),

2. the failure of a main steam isolation valve (MSIV),

3. the failure of a main feedwater isolation valve (MFIV), and

4. the failure of an Emergency Diesel Generator (EDG) which results in the failure of one train each of Safety Injection, Quench Spray, and Service Water.

In addition, the containment initial conditions (pressure, temperature and relative humidity) are also factors in selecting the governing MSLB cases.

Sensitivity analyses were conducted to determine the most challenging set of plant conditions related to the specific attribute being evaluated (i.e., the peak calculated containment temperature). Table 6.2-3 presents the input data used in the peak pressure analysis.

The containment pressure history was analyzed for each sequence using the mass and energy releases, assuming that the main steam line break occurred in the region immediately above the operating deck where the steam lines exit the steam generator and then run horizontally toward the containment wall.

More than 20 Break cases were analyzed to cover the different possibilities of power level, break type and size, single active failure, and initial conditions. Results show that the peak containment pressure occurs following a 1.069 ft2 Double-ended Rupture (DER) with a Main Steam Isolation Valve (MSIV) Failure at 30 percent Power. The peak calculated containment pressure for this case is 39.3 psig. Results show that the peak containment temperature occurs following a 1.069 ft2 Double-ended Rupture (DER) with a Main Steam Isolation Valve (MSIV) Failure at Full Power. The peak calculated containment temperature for this case is 343.9°F.

Table 6.2-10 and Table 6.2-11 summarize the peak containment pressure and peak containment temperature results respectively. Figures 6.2-12, 6.2-13, 6.2-14 and 6.2-15 illustrate the containment pressure and containment temperature transients for the two limiting cases.

The qualification of safety-related equipment inside the containment to the pressure and temperature cases resulting from a MSLB is discussed in Section 3.11.

The chronology of events for the limiting containment pressure and temperature cases is given in Tables 6.2-12 and 6.2-13.

6.2.1.1.3.8 Feedwater Pipe Break Results The feedwater pipe break is not as severe as the main steam line break since the break effluent is at a lower specific enthalpy. Therefore, the feedwater pipe break is not analyzed.

6.2-14

BVPS-2 UFSAR Rev. 26 6.2.1.2 Containment Subcompartments 6.2.1.2.1 Design Bases The containment subcompartments are designed in accordance with GDC 4 and 50.

Break locations, types, and areas used for the design of containment subcompartments are as follows and described in Section 6.2.1.2.3.1:

1. Pressurizer subcompartments

a. Upper pressurizer cubicle - spray line DER,

b. Lower pressurizer cubicle - surge line DER, and

c. Pressurizer relief tank cubicle - surge line DER.

2. Steam generator subcompartments - RCS 707 in2 longitudinal intrados split break at the steam generator inlet elbow.

3. Reactor cavity subcompartment - RCS 150 in2 cold leg limited displacement rupture (LDR) at the reactor vessel nozzle.

Justification for each break, size, and location, and the use of pipe restraints to limit the break area are described in Section 3.6.

6.2.1.2.2 Design Features Figures 3.8-1, 3.8-2, 3.8-3, 3.8-4, 3.8-5, 3.8-6 and 3.8-7 provide detailed plan and elevation drawings of the containment subcompartments showing the component and equipment locations. The volume and vent areas for each subcompartment are discussed in Section 6.2.1.2.3.3.

6.2.1.2.3 Design Evaluation Containment subcompartment analyses are performed to calculate the pressure gradient transient across major equipment, supports, and walls that will result from postulated pipe ruptures. The resulting pressure gradients are used to calculate the loads and moments on major equipment and supports. The maximum differential pressures across the subcompartment walls are used as the design basis for the structures.

A model is developed for each subcompartment to predict a conservative transient pressure response. Each subcompartment is subdivided into a network of control volumes or nodes. Boundaries between control volumes, which represent junctions or vent paths, are located at physical discontinuities where geometric influences are expected to create a pressure differential. A detailed description of each subcompartment model is given in Section 6.2.1.2.3.3.

The assumed initial conditions for the subcompartment volumes are conservatively chosen to maximize the resultant differential pressure responses. The values selected are as follows:

6.2-15

BVPS-2 UFSAR Rev. 26 Maximum temperature (F) 105 Minimum air partial pressure (psia) 8.9 Minimum relative humidity (%) 50 The containment subcompartment design evaluations used the THREED computer program (Boyle 1975, Meyer 1981), which considers two-phase, two-component (steam and water-air) flow through the vents and accounts for the fluid inertial effects. Section 6.2.1.2.3.2 provides a detailed description of THREED.

The critical flow correlation typically selected for each vent path is the homogeneous equilibrium model (HEM).

The description of, and justification for, the subsonic and sonic flow model and the degree of entrainment used in the vent flow calculations are given in Section 6.2.1.2.3.2.

In those situations, however, where the component is most vulnerable to a loading induced by the rupture of a pipe not immediately adjacent to the component or where the worst loading results from an overturning moment created by loads away from the break, the Moody choked flow correlation (Moody 1965), with a discharge coefficient of 1.0, is used to yield corresponding high values of flow.

The vent loss coefficients used in the subcompartment analyses depend on the geometry of the particular vent. The values of the total loss coefficients for both forward and reverse flow directions are simply the sum of the head losses for the separate parts of the system. These head losses consist of the following:

1. Contraction and expansion losses are determined as a function of the ratio of the upstream and/or downstream cross-sectional area to the cross-sectional area of the vent.

2. Bend losses resistance is determined by the angle and length of the bend and the hydraulic diameter of the vent.

3. Friction losses, although generally very small, are calculated as an fl/d term.

4. Form losses are due to objects in the flow path such as grating, and are calculated based on the methods by Idel'chik (1966).

The previous list of losses are defined specifically by Idel'chik. The values of the loss coefficients used in the subcompartment analyses are given in Section 6.2.1.2.3.3.

The RCS mass and energy release rates are provided by the NSSS vendor for each break. The release rates are computed by the SATAN Program (Shepard 6.2-16

BVPS-2 UFSAR Rev. 26 et al 1975). The initial BVPS-2 operating conditions are selected to yield the maximum calculated blowdown.

6.2.1.2.3.1 Break Type Definitions and Areas Two types of breaks are used to analyze containment subcompartments. The first type is a guillotine break, which results in complete pipe separation. A guillotine break which results in a break flow area of two pipe cross sections is called a double ended rupture (DER). In some subcompartments, pipe restraints limit the displacement of two broken ends of the pipe so that the break flow area is less than two pipe cross-sectional areas. This type of break is called a limited displacement rupture (LDR).

The second type of break is a longitudinal split which is equivalent to a hole in the pipe.

The break type(s), location(s), and area(s) to which the subcompartment walls and equipment supports are designed are listed in Table 6.2-25. All breaks analyzed within a particular subcompartment are described in Section 6.2.1.2.3.3. Pipe restraints will limit the break area to an equal or smaller area than that analyzed.

The RCS break type(s) and location(s) which are considered in the subcompartment analyses are given by Clout (1973).

Double-ended ruptures are considered in the analysis for the pressurizer cubicle. Breaks with less than two cross-sectional flow areas are used in the analyses for the reactor cavity and steam generator subcompartments.

6.2.1.2.3.2 Subcompartment Analytical Model Functional description of THREED code The THREED computer program is used to calculate the transient conditions of pressure, temperature, and humidity in various subcompartments following a postulated rupture in a moderate or high energy pipeline. The results obtained from THREED analyses are used to calculate loads on structures and to define environmental conditions for equipment qualification.

The THREED computer program is similar to RELAP4 (Moore and Rettig 1974; Aerojet Nuclear Company (ANC) 1976) and will give the same results as RELAP4 if similar options are chosen. THREED was formulated to perform subcompartment analyses with capabilities and operations extended beyond those available in RELAP4. A significant improvement in THREED is that the HEM has been extended to include two-phase, two-component flow which is encountered in subcompartment analysis.

Description of the model The THREED computer code can be viewed as a numerical integrator for the macroscopic form of the basic field equations describing the conservation of mass, energy, and momentum. The conservation equations, along with the 6.2-17

BVPS-2 UFSAR Rev. 26 equation of state for the fluid, give a complete solution to the fluid flow phenomena. THREED solves a stream tube form of the field equations based on the assumptions of one-dimensional, homogeneous, thermal-equilibrium flow. Although THREED does not prohibit the use of multidimensional flow paths, the flow paths are modeled to approximate a one-dimensional equation.

Subcompartments are modeled in THREED as a hydraulic network that consists of a series of interconnecting, user-defined nodes (mass and energy control volumes). Nodes are connected by internal junctions (momentum control volumes) with the internodal flow rates determined by the solution of the momentum equation. An internal junction control volume is defined as the composite volume between the centers of adjacent nodes. This inconsistency in control volumes (a different control volume for momentum than for mass and energy) is illustrated on Figure 6.2-18. This staggered mesh approximation is necessary for purposes of solving the equations.

Fill junctions are dissimilar to internal junctions in that they have no initial node, and their flow rate is dependent only on the junction area and time. These junctions are used to simulate flow originating external to the network (blowdown). Mathematically, they are treated as boundary conditions.

THREED numerically solves finite difference equations which account for mass and energy flows into and out of a node. Figure 6.2-19 summarizes the computational approach used in THREED.

The fluid conservation equations used by THREED can be obtained by integrating the stream tube equations over a fixed volume, V. The mass and energy equations are developed for the generalized i-th node, while the momentum equations are developed for the generalized j-th internal junction connecting nodes K and L. Neglecting kinetic energy effects, the resulting equations are as follows:

Conservation of Mass: The mass equation is (ANC 1976).

dM i Wij (6.2-14) dt j where:

Mi = Total mass in node i: (Mi = Mwi + Mai)

Mwi = Total mass of water in node i Mai = Total mass of air in node i Wij = Mass flow rate into node i from junction j 6.2-18

BVPS-2 UFSAR Rev. 26 Conservation of Energy: The energy equation for homogeneous flow is (ANC 1976):

dU i Wij ( hij Z ij Z i ) (6.2-15) dt j where:

Ui = Total fluid internal energy of water in node i hij = Local enthalpy at junction j of the fluid entering or leaving node i Zi j Zi = Elevation change from the center of mass in node i at Z to junction j i

Conservation of Momentum: The incompressible equation for homogeneous flow is (ANC 1976):

dW j Ij PK PKgj PL PLgj fj (6.2-16) dt where:

Ij = Geometric inertia for junction j Wj = Mass flow rate in junction j PK = Total static pressure in node K (at center)

PKgi = Gravity pressure differential from the center of node K to junction j PL = Total static pressure in node L (at center)

PLgi = Gravity pressure differential from junction j to the center of node L Fj = Static pressure change term Equation of State: The functional form of the equation of state is:

Pi f (U i , M wi , M ai ) (6.2-17) 6.2-19

BVPS-2 UFSAR Rev. 26 where:

Pi = Total static pressure in node i The following assumptions are made in deriving the equation of state:

1. The components of water and air form a homogeneous mixture with a uniform temperature.

2. Water, if present, occupies the entire volume. Air, if present, occupies the same volume as the water vapor according to the Gibbs-Dalton Law. Air is assumed to be insoluble in water, and there can be no air present if the volume is filled with liquid water.

3. Air is treated as a perfect gas,

4. If air and liquid water are present, the atmosphere is saturated with water vapor (relative humidity of 100 percent).

5. If air is present, the liquid water conditions are the saturated conditions for Pwi. A more accurate model would have liquid water at the subcooled conditions corresponding to Pi and Ti. This assumption is made to limit calls to the water property routines to one per iteration. If no water is present in the volume (Mw = 0), the detailed form of the equation of state is:

Ui = Mai Cva Ti (6.2-18)

M ai Ra Ti Pi (6.2-19)

Vi where:

Cva = Constant volume heat capacity of air Ti = Temperature of node i Ra = Gas constant of air Vi = Volume of node i If water is present in the volume (Mw 0), the detailed form of the equation state is:

6.2-20

BVPS-2 UFSAR Rev. 26 Vwi M wi / Vi (6.2-20)

U i M wi U wi (Ti ,Vwi ) M ai Cva Ti (6.2-21)

M ai Ra Ti Pai (6.2-22)

X i M wi Vgi ( Ti , Vwi )

Pi Pwi ( Ti , Vwi ) Pai (6.2-23) where:

Vwi = Specific volume of water in node i Uwi = Specific internal energy of water in node i Pai = Partial pressure of air in node i Xi = Quality of node i Vgi = Specific volume of water vapor in node i Pwi = Partial pressure of water in node i It should be noted that the internal code calculations are done in Systeme Internationale units. The reference temperature used for the calculation of the internal energy of air is zero degrees Kelvin. The properties of steam are based on the 1967 ASME formulation of the properties of steam.

Fill Junctions: These are normally used to input blowdown (mass and energy release) into a node(s). Their functional form is:

Wj = f(t) (6.2-24) hij = f(t) (6.2-25)

Fan Junctions: These junctions may be used to model ventilation fan operation in situations where such modeling is appropriate. Their functional form is:

Wj = f(Hj) (6.2-26) where:

Hj = Head difference across the fan junction 6.2-21

BVPS-2 UFSAR Rev. 26

1. Choked Flow Options for Internal Junctions Since an incompressible flow model has no mechanism to restrict flow through a junction to the maximum allowable (choked) flow rate, it is necessary to use a separate calculation to restrict the flow rate. To determine if the flow is choked, the momentum Equation 6.2-16, is solved using a forward finite difference approximation and compared with a calculated choked flow (HEM or Moody). The lesser flow is selected as the junction flow rate for the time step.

Both the HEM and the Moody (1965) flow model are based on stagnation properties. Since it is not usually possible to calculate the velocity in a node, it is assumed that the static and stagnation properties in a node are the same (neglecting kinetic energy effects). This may result in an underprediction of the choked flow rate, which is conservative in most cases.

2. Homogeneous Equilibrium Model The HEM is approximated in THREED using an ideal gas approximation.

That is, the choked isentropic ideal gas flow equation is utilized and the isentropic exponent is modified to accommodate two-phase, two-component flow. The isentropic exponent is defined as:

i Vwi Pi

( VP )

i s (6.2-27) wi where:

i = Isentropic exponent in node i The equation utilized by THREED to calculate the HEM is:

W j 12 A j ( 2 1) b g c i Pai Vai (6.2-28) i where:

b = (i+1)/2(i-1) (6.2-28A)

Aj = Flow area of junctions j (ft2) i = Isentropic exponent of source node i gc = Proportionality constant - 32.17 (ft-lbm /lbf-sec2)

Pai = Stagnation pressure in source node i (psia) 6.2-22

BVPS-2 UFSAR Rev. 26 Vai = Stagnation specific volume of air source node i (ft3/lbm)

Wj = Mass flow in junction j (lbm/sec)

3. Moody Choked Flow Model The Moody flow model used in THREED is based on the interpolation of tables from RELAP4/MOD 5 (ANC 1976). The model is for one-component flow and, when air is present, the tables are accessed with the total pressure and average enthalpy of the node.

4. Junction Check Valves A valve may be modeled in any non-fan internal junction as follows:

Normally closed - trips open instantaneously Normally open - trips closed instantaneously

5. Time Step Control If the automatic time step control option is selected, the maximum time step will be limited by the following calculation based on the nodal conditions (ANC 1976):

(6.2-29) where:

i = Node number, from 1 to n DT = Time step size Pi = dPi /dt Assumptions employed in THREED The following assumptions are employed in THREED:

1. Lumped parameter (control volume) approach is utilized,

2. Adiabatic process,

3. Independent inflow (blowdown),

4. Thermodynamic equilibrium in each node,

5. One-dimensional formulation,

6. Staggered mesh for the conservation equations, 6.2-23

BVPS-2 UFSAR Rev. 26

7. Incompressible form of the momentum equation,

8. Kinetic energy effects are neglected,

9. For choked flow models, static properties in the nodes considered to be stagnation properties, and

10. Valves open or close instantaneously.

6.2.1.2.3.3 Containment Subcompartment Analysis Results Pressurizer Cubicle and Pressurizer Relief Tank Cubicle The pressurizer cubicle and the pressurizer relief tank cubicle are analyzed according to the nodalization diagrams shown on Figure 6.2-20.

Plan and elevation views depicting the nodal arrangement are shown on Figures 6.2-21, 6.2-22, 6.2-23, and 6.2-24. This nodalization models all significant physical obstructions to flow and is used for predicting loads on the subcompartment walls, pressurizer, and supports.

Vent data and nodal net volumes for the model are listed in Tables 6.2-21 and 6.2-22, respectively.

The following three RCS breaks, listed with the corresponding blowdown distribution, are analyzed:

1. Spray line DER in the upper pressurizer cubicle (100 percent of the blowdown deposited into node 7).

2. Surge line DER at the pressurizer nozzle (100 percent of the blowdown deposited into node 18).

3. Surge line DER in the pressurizer relief tank cubicle (100 percent of the blowdown deposited into node 15).

The mass and energy release rates for the spray line DER and the surge line DER are given in Tables 6.2-23 and 6.2-24, respectively.

The nodal pressure responses for each break described previously are presented on Figures 6.2-25 through 6.2-43.

The peak calculated differential pressures across the pressurizer cubicle walls for the three breaks analyzed are tabulated in Table 6.2-25. These differential pressures are used in the structural analysis of the pressurizer cubicle.

A simultaneous rupture of three 6-inch safety lines in the upper pressurizer cubicle is enveloped by the spray line DER.

Steam Generator Cubicle Steam generator cubicle 2 is nodalized using a 33 node model. The three steam generator cubicles are similar in design, thus the results obtained 6.2-24

BVPS-2 UFSAR Rev. 26 from analyzing pipe ruptures in cubicle 2 will be representative of the other steam generator cubicles.

The nodalization schematic used in the steam generator subcompartment analysis is shown on Figure 6.2-44. Plan and elevation views of the steam generator cubicle depicting the nodal arrangement are shown on Figures 6.2-45, 6.2-46, 6.2-47, 6.2-48, and 6.2-49. This nodalization models all significant physical obstructions to flow and is used for predicting loads on both the subcompartment walls and major components within the subcompartment.

Vent data and net volumes for the model are listed in Tables 6.2-27 and 6.2-28, respectively.

Three RCS breaks are analyzed to calculate transient pressure responses for the evaluation of loads on components and structures.

The following three RCS breaks, listed with the corresponding blowdown distribution, are analyzed:

1. A 320 in2 LDR at the steam generator outlet nozzle. One hundred percent of the blowdown is deposited into node 32. The mass and energy release rates are given in Table 6.2-29.

2. A 180 in2 LDR at the reactor coolant pump (RCP) outlet nozzle.

Fifty percent of the blowdown is deposited into both nodes 8 and 9. The mass and energy release rates are given in Table 6.2-30.

3. A 707 in2 longitudinal intrados split break at the steam generator inlet elbow. Fifty percent of the blowdown is deposited into node 32 and 25 percent into both nodes 7 and 8.

The mass and energy release rates are given in Table 6.2-31.

Main steam lines are not routed through any portion of the steam generator cubicle and are not considered in the analysis.

The preceding three breaks are chosen to evaluate loads on the subcompartment walls and component supports. These breaks were chosen from the nine breaks listed in a report by Clout (1973) as limiting cases which envelop conditions resulting from all nine breaks.

The nodal pressure responses for each break analyzed are given on Figures 6.2-50, 6.2-51, 6.2-52, 6.2-53, 6.2-54, 6.2-55, 6.2-56, 6.2-57, 6.2-58, 6.2-59, 6.2-60, 6.2-61, 6.2-62, 6.2-63, 6.2-64, 6.2-65, 6.2-66, 6.2-67, 6.2-68, 6.2-69, 6.2-70, 6.2-71, 6.2-72, 6.2-73, 6.2-74, 6.2-75, 6.2-76, 6.2-77, 6.2-78, 6.2-79, 6.2-80, 6.2-81, 6.2-82, 6.2-83, 6.2-84, 6.2-85, 6.2-86, 6.2-87, 6.2-88, 6.2-89, 6.2-90, 6.2-91, 6.2-92, 6.2-93, 6.2-94, 6.2-95, 6.2-96 and 6.2-97.

The peak calculated pressure differentials (with respect to the bulk containment pressure) are tabulated in Table 6.2-25. These pressure differentials are used in the structural analysis of the steam generator cubicle.

6.2-25

BVPS-2 UFSAR Rev. 26 The design of the steam generator cubicle walls above the operating floor (el 767 feet-10 inches) are based on the pressure transients that result from analyzing a 707 in2 longitudinal intrados split break at the steam generator inlet elbow. The Moody flow correlation with a 1.0 discharge coefficient is used to conservatively calculate high flow rates from the area of the break into the nodes above the operating floor.

The peak calculated pressure differentials (with respect to the bulk containment pressure) are tabulated in Table 6.2-25 for the nodes above the operating floor. These pressure differentials are used in the structural analysis of the steam generator cubicle.

Reactor Cavity The reactor cavity is analyzed utilizing the nodalization schematic shown on Figure 6.2-98. Plan and elevation views of the reactor cavity depicting the nodal arrangement are shown on Figures 6.2-99, 6.2-100, and 6.2-101. This nodalization models all significant physical obstructions to flow and is consistent with the recommended reactor cavity nodalization presented in the LASL (1979) Report.

The insulation on the ruptured inlet pipe is assumed to block the shield wall pipe penetration vent area instantaneously, thus causing higher pressure differential across the reactor vessel. The neutron shielding material draped over the ruptured pipe is assumed to be ejected from the reactor cavity.

Insulation on the remaining pipes is unaffected by the rupture. However, the neutron shielding material on the outlet pipes on either side of the break are assumed to be displaced, lodging under the other two inlet pipes, thus blocking flow below the cold legs coincidentally with the break occurrence and further increasing the differential pressure across the reactor vessel.

High pressure is assumed to collapse the neutron shielding material that surrounds the reactor vessel in nodes 3, 4, 5, 9, 10, and 11 (Figure 6.2-100), thus preventing the flow from the reactor cavity (nodes 3, 4, and 5) into the reactor annular region (nodes 15, 16, and 17). This assumption is conservative in that it creates a greater overturning moment on the reactor vessel.

The reactor cavity water seal limits air flow out of the top of the annulus and directs the normal ventilation flow through the pipe penetrations to provide cooling for the concrete.

Blowout panels are also located in the incore instrumentation tunnel to prevent overpressurization. The blowout panels are designed to open when the differential pressure across the panels reaches 1.5 to 2.0 psi.

Panels are membranes of approximately 1 lb each, and the probability of impacting critical items of larger mass is extremely small. If impact occurs, no damage to the critical item will result.

6.2-26

BVPS-2 UFSAR Rev. 26 The vent area from the upper reactor cavity into the incore instrumentation tunnel (via the reactor annulus and lower reactor cavity) is limited by the neutron shield design. The net vent area out of the incore instrumentation tunnel (52 ft2) after the panels open is more than adequate to prevent exceeding the compartment design pressure. This additional vent area, available after the incore instrumentation tunnel blowout panels open, is not included in the reactor cavity model to conservatively calculate the pressure transients.

A 150 in2 LDR at the reactor vessel inlet nozzle is analyzed to evaluate loads on the subcompartment walls and component supports. This rupture envelops a 150 in2 LDR at the reactor outlet nozzle. The blowdown from the break is deposited equally into nodes 4, 5, 10, and 11.

The peak calculated and differentials (with respect to the bulk containment) are tabulated in Table 6.2-25. These pressure differentials are used in the structural analysis of the reactor cavity.

6.2.1.3 Mass and Energy Release Analyses for Postulated Loss-of-Coolant Accidents The LOCA mass and energy releases for the containment are generated using the Westinghouse March 1979 LOCA mass and energy release model. These releases are used in the containment response calculation. The Westinghouse March 1979 LOCA mass and energy release model has been reviewed and approved by the NRC for use on Westinghouse-designed PWRs, including the conversion to atmospheric containment for the Beaver Valley Units.

The Westinghouse generated LOCA mass and energy releases for the first hour are used in the MAAP-DBA containment response analysis. After this time the Westinghouse rates are still used, however the break enthalpy is calculated, along with the containment response, by MAAP-DBA. This is done in order to account for the influence of the time-dependent safety injection temperature during the recirculation mode of operation.

This section describes the LOCA mass and energy release calculation methodology for the hypothetical double-ended pump suction (DEPS) and double-ended hot-leg (DEHL) break cases. It also explains that the analysis of the DEPS and DEHL LOCAs bounds all current licensing basis LOCAs, including the double-ended cold leg (DECL) break.

6.2.1.3.1 Input Parameters and Assumptions The mass and energy release analysis is sensitive to the characteristics of various plant systems, in addition to other key modeling assumptions.

Where appropriate, bounding inputs are utilized and instrumentation uncertainties are included. Nominal parameters are used in certain instances.

All input parameters are chosen consistent with accepted analysis methodology. Some of the most critical items are the RCS initial conditions, core decay heat, safety injection flow, and primary and secondary metal mass and steam generator heat release modeling. Specific 6.2-27

BVPS-2 UFSAR Rev. 26 assumptions concerning each of these items are discussed next. Tables 6.2-14, 6.2-14a, and 6.2-14b present key data assumed in the analysis.

Higher RCS operating temperatures to bound the highest average coolant temperature range were used as bounding analysis conditions. The use of higher temperatures is conservative because the initial fluid energy is based on coolant temperatures that are at the maximum levels attained in steady state operation. Additionally, an allowance to account for instrument error and deadband is reflected in the initial RCS temperatures.

The rate at which the RCS blows down is initially more severe at the higher RCS pressure. Additionally the RCS has a higher fluid density at the higher pressure (assuming a constant temperature) and subsequently has a higher RCS mass available for releases.

The selection of the fuel design features for the long-term mass and energy release calculation is based on the need to conservatively maximize the energy stored in the fuel at the beginning of the postulated accident (i.e., to maximize the core stored energy).

The nominal RCS volume is increased because this assumption helps maximize the initial RCS mass and energy.

A minimum uniform steam generator tube plugging (SGTP) level is modeled.

This assumption maximizes the reactor coolant volume and fluid release by considering the RCS fluid in all SG tubes. The SGTP assumption maximizes heat transfer area and therefore, the transfer of secondary heat across the SG tubes. Additionally, this assumption reduces the reactor coolant loop resistance, which reduces the pressure drop upstream of the break for the pump suction breaks and increases break flow.

The initial steam generator fluid mass is calculated at full power, and then increased to cover uncertainties. The steam generator water mass and metal mass used conservatively bounds the Model 51 steam generator.

Portions of the SG secondary metal, such as the upper elliptical head, upper shell, and miscellaneous upper internals, have poor heat transfer due to their location in the steam region. The mass of this metal is approximately 216,300 lbm per SG. The stored energy in this metal will be transferred to the RCS and released to the containment at a much slower rate and is not considered during the first hour of the LOCA mass and energy release calculation for the double-ended pump suction breaks. The stored energy in the rest of the SG secondary metal and fluid is released to the containment within the first hour.

After one hour, the Westinghouse LOCA mass and energy calculation has extracted all of the stored energy from the RCS, except for the stored metal energy in the steam generator upper internals and upper elliptical heads. This energy is assumed to be removed at a constant rate over the next six hours and is added to the core decay heat as an energy source for the long-term steaming rate calculation.

6.2-28

BVPS-2 UFSAR Rev. 26 Regarding safety injection flow, the mass and energy release calculation considered configurations/failures to conservatively bound respective alignments. These cases include (1) a Minimum Safeguards case (one Charging/High Head Safety Injection pump (CH/HHSI) and one Low Head Safety Injection (LHSI) pump) and (2) a Maximum Safeguards case (two CH/HHSI and two LHSI pumps).

In summary, the following assumptions were employed so that the LOCA mass and energy releases are conservatively calculated, thereby maximizing the energy release to containment:

1. The nominal RCS volume is increased by 3 percent (1.6-percent allowance for thermal expansion and 1.4 percent for uncertainty)

2. The reactor is assumed to be operating at full core rated thermal power (2900 MWT) and an allowance for calorimetric error (+0.6 percent of power) is added.

3. Core-stored energy is based on the time in life for maximum fuel densification. The assumptions used to calculate the fuel temperatures for the core-stored energy calculation account for appropriate uncertainties associated with the models in the PAD code (e.g., calibration of the thermal model, peller densification model, and cladding creep model). In addition, the fuel temperatures for the core-stored energy calculation account for appropriate uncertainties associated with manufacturing tolerances (e.g., pellet as-built density). The total uncertainty for the fuel temperature calculation is a statistical combination of these effects and is dependent upon fuel type, power level, and burn up.

4. The RCS is assumed to be at the maximum expected full power operating temperature and an allowance for temperature measurement uncertainty of +4.0°F is added. These uncertainties conservatively include both deadband and bias.

5. The RCS is assumed to be at the nominal RCS pressure and an allowance for pressure measurement uncertainty of +42 psi was added.

6. Conservatively high heat transfer coefficients (i.e., steam generator primary/secondary heat transfer and reactor coolant system metal heat transfer) are modeled. The SG secondary stored energy is released in one hour. All of the additional stored energy in the upper elliptical head, upper shell, and miscellaneous upper internals, is released at a constant rate over the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

7. The LOCA back-pressure is assumed to remain at the containment design pressure (45 psig). This assumption determines the end of the blowdown phase and minimizes the safety injection flow rate during the reflood phase.

6.2-29

BVPS-2 UFSAR Rev. 26

8. A uniform SGTP level of 0% is assumed. This assumption:

• Maximizes reactor coolant volume and fluid release,

• Maximizes heat transfer area across the SG tubes,

• Reduces coolant loop resistance, which reduces the P upstream of the break for the pump suction breaks and increases break flow.

9. The full power SG level is used to calculate the initial secondary mass and 10% is added to cover uncertainty.

Thus, based on the previously discussed conditions and assumptions, a bounding analysis was made for the release of mass and energy releases from the RCS in the event of a LOCA at the future uprated conditions.

6.2.1.3.2 Description of Analyses The evaluation model used for the long-term LOCA mass and energy release calculations is the March 1979 model. This evaluation model has been reviewed and approved generically by the Nuclear Regulatory Commission (NRC). The approval letter is included with WCAP-10325-P-A, Westinghouse, 1983.

6.2.1.3.3 LOCA Mass and Energy Release Phases The containment system receives mass and energy releases following a postulated rupture in the RCS. These releases continue over a time period, which, for the LOCA mass and energy release analysis, is typically divided into four phases.

Blowdown - the period of time from accident initiation (when the reactor is at steady state operation) to the time that the RCS and containment reach an equilibrium state.

Refill - the period of time when the lower plenum is being filled by accumulator and Emergency Core Cooling System (ECCS) water. At the end of blowdown, a large amount of water remains in the cold legs, downcomer, and lower plenum. To conservatively consider the refill period for the purpose of containment mass and energy releases, it is assumed that this water is instantaneously transferred to the lower plenum along with sufficient accumulator water to completely fill the lower plenum. This allows an uninterrupted release of mass and energy releases to containment. Thus, the refill period is conservatively neglected in the mass and energy release calculation.

Reflood - begins when the water from the lower plenum enters the core and ends when the core is completely quenched.

Post-reflood (Froth) - describes the period following the reflood phase.

For the pump suction break, a two-phase mixture exits the core, passes through the hot legs, and is superheated in the steam generators prior to exiting the break as steam. After the broken loop steam generator cools, 6.2-30

BVPS-2 UFSAR Rev. 26 the break flow becomes two-phase.

6.2.1.3.4 Computer Codes The March 1979 model mass and energy release evaluation model is comprised of mass and energy release versions of the following codes: SATAN VI, WREFLOOD, FROTH, and EPITOME.

SATAN VI calculates blowdown, the first portion of the thermal-hydraulic transient following break initiation, including pressure, enthalpy, density, mass and energy flow rates, and energy transfer between primary and secondary systems as a function of time.

The WREFLOOD code addresses the portion of the LOCA transient where the core reflooding phase occurs after the primary coolant system has depressurized (blowdown) due to the loss of water through the break and when water supplied by the ECCS refills the reactor vessel and provides cooling to the core. The most important feature of WREFLOOD is the steam/water mixing model, discussed in subsection 6.2.1.3.8.2.

FROTH models the post-reflood portion of the transient. The FROTH code calculates the heat release from the energy stored in the secondary fluid and metal masses, excluding the upper internals and upper elliptical head.

This part of the steam generator metal mass is not actively cooled by the two-phase fluid circulating through steam generator tubes and takes longer to cooldown.

EPITOME continues the FROTH post-reflood portion of the transient from the time at which the secondary equilibrates to containment design pressure to the end of the transient (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />). It also compiles a summary of data on the entire transient, including formal instantaneous mass and energy release tables and mass and energy releases balance tables with data at critical times.

After one hour, the Westinghouse LOCA mass and energy releases calculation has extracted all of the stored energy from the RCS, except for the stored metal energy in the steam generator upper internals and upper elliptical heads. This energy is assumed to be removed at a constant rate over the next six hours and is added to the core decay heat as an energy source for the long-term steaming rate calculation. See Section 6.2.1.1.3.2.2.2.

6.2.1.3.5 Break Size and Location Generic studies (March 1979 model, WCAP-10325-P-A, Westinghouse, 1983, Section 3) have been performed to determine the effect of postulated break size on the LOCA mass and energy releases. The double-ended guillotine break has been found to be limiting due to larger mass flow rates during the blowdown phase of the transient. During the reflood and post-reflood phases, the break size has little effect on the releases.

Three distinct locations in the reactor coolant system loop can be postulated for pipe rupture for any release purposes:

6.2-31

BVPS-2 UFSAR Rev. 26

1. Hot leg (between vessel and steam generator)

2. Cold leg (between pump and vessel)

3. Pump suction (between steam generator and pump)

The DEHL break location yields the highest blowdown mass and energy release rates (March 1979 model, WCAP-10325-P-A, Westinghouse, 1983, Section 3.3). Although the core flooding rate would be the highest for this break location, the amount of energy released from the steam generator secondary side is minimal because the majority of fluid that exits the core vents directly to containment, bypassing the steam generators. As a result, the reflood mass and energy releases are reduced significantly as compared to either the pump suction, or cold-leg break locations where the core exit mixture must pass through the steam generators before venting through the break. Studies have confirmed that there is no reflood peak (i.e., from the end of the blowdown period the containment pressure would continually decrease) for the hot leg break.

Therefore, the mass and energy releases for the blowdown phase of the hot-leg break are calculated and used in the containment peak pressure and temperature response calculation.

Studies have determined that the blowdown transient for the DECL break is, in general, less limiting than that for the pump suction break (March 1979 model, WCAP-10325-P-A, Westinghouse, 1983, Section 3.3). The cold leg blowdown is faster than that of the pump suction break, and more mass is released into the containment. However, the core heat transfer is greatly reduced, and this results in a considerably lower energy release into containment. The flooding rate during the reflood phase is greatly reduced, and the energy release rate into the containment is reduced.

Therefore, the cold-leg break is bounded by other breaks and no further evaluation is necessary.

The pump suction break combines the effects of the relatively high core-flooding rate, as in the hot-leg break, and the additional stored energy in the steam generators. As a result, the pump suction break yields the highest energy flow rates during the post-blowdown period by including all of the available energy of the RCS in calculating the releases to containment.

Therefore, the break locations that were analyzed for this program were the DEPS rupture (10.5 ft2) and the DEHL rupture (9.2 ft2). LOCA mass and energy releases have been calculated for the blowdown, reflood, and post-reflood phases for the DEPS cases. For the DEHL case, the releases were calculated only for the blowdown phase with this methodology.

6.2.1.3.6 Application of Single-Failure Criterion The mass and energy release calculation assumes a complete loss of all offsite power coincident with the LOCA. The emergency diesel generators are actuated to provide power for the safety injection system. The combination of signal delay plus diesel delay and additional delays in starting the ECCS pumps results in the delivery of SI after the end of blowdown.

6.2-32

BVPS-2 UFSAR Rev. 26 Two cases are analyzed to assess the effects of a single failure in the mass and energy release calculation. The first case assumes a single failure of one of the emergency diesel generators, resulting in the loss of one train of safeguards equipment. This, in combination with other conservative assumptions (maximum resistances, minimum pump head-flow curves), minimizes the safety injection flow rate. The second case assumes a failure in the containment spray system. The safety injection flow rate for this case is maximized by assuming both trains of safeguards equipment are operating and by including other conservative assumptions (minimum resistances, maximum pump head-flow curves). In addition to these two cases, a third case that assumes a failure of one service water (SW) train was analyzed.

6.2.1.3.7 Acceptance Criteria A large break LOCA is classified as an ANS Condition IV event, an infrequent fault. To satisfy the NRC acceptance criteria presented in the Standard Review Plan Section 6.2.1.3, the relevant requirements are as follows:

1. 10 CFR 50, Appendix A

2. 10 CFR 50, Appendix K, paragraph I.A To meet these requirements, the following must be addressed:

1. Sources of energy

2. Break size and location

3. Calculation of each phase of the accident 6.2.1.3.8 Results 6.2.1.3.8.1 Blowdown Mass and Energy Release Data The SATAN-VI code is used for computing the blowdown transient. The code utilizes the control volume (element) approach with the capability for modeling a large variety of thermal fluid system configurations. The fluid properties are considered uniform, and thermodynamic equilibrium is assumed in each element. A point kinetics model is used with weighted feedback effects. The major feedback effects include moderator density, moderator temperature, and Doppler broadening. A critical flow calculation for subcooled (modified Zaloudek), two-phase (Moody), or superheated break flow is incorporated into the analysis. The methodology for the use of this model is described in WCAP-10325-P-A, Westinghouse, 1983.

Table 6.2-14C presents the calculated mass and energy release for the blowdown phase of the DEHL break. For the hot-leg break mass and energy release tables, break path 1 refers to the mass and energy releases exiting from the reactor vessel side of the break; and break path 2 refers to the mass and energy releases exiting from the steam generator side of the break.

Table 6.2-14D presents the calculated mass and energy releases for the blowdown phase of the DEPS break with either minimum or maximum ECCS 6.2-33

BVPS-2 UFSAR Rev. 26 flows. For the pump suction breaks, break path 1 in the mass and energy release tables refers to the mass and energy releases exiting from the steam generator side of the break; break path 2 refers to the mass and energy releases exiting from the pump side of the break.

6.2.1.3.8.2 Reflood Mass and Energy Release Data The WREFLOOD code is used for computing the reflood transient. The WREFLOOD code consists of two basic hydraulic modelsone for the contents of the reactor vessel and one for the coolant loops. The two models are coupled through the interchange of the boundary conditions applied at the vessel outlet nozzles and at the top of the downcomer. Additional transient phenomena, such as pumped safety injection and accumulators, reactor coolant pump performance, and steam generator releases are included as auxiliary equations that interact with the basic models as required. The WREFLOOD code permits the capability to calculate variations during the core reflooding transient of basic parameters such as core flooding rate, core and downcomer water levels, fluid thermodynamic conditions (pressure, enthalpy, density) throughout the primary system, and mass flow rates through the primary system. The code permits hydraulic modeling of the two flow paths available for discharging steam and entrained water from the core to the break, the path through the broken loop and the path through the unbroken loops.

A complete thermal equilibrium mixing condition for the steam and ECCS injection water during the reflood phase has been assumed for each loop receiving ECCS water. This is consistent with the usage and application of the March 1979 mass and energy release evaluation model in recent analyses, for example, D.C. Cook (Docket 50-315). Even though the March 1979 model credits steam/water mixing only in the intact loop and not in the broken loop, the justification, applicability, and NRC approval for using the mixing model in the broken loop has been documented (D.C. Cook Docket 50-315). Moreover, this assumption is supported by test data and is further discussed below.

The model assumes a complete mixing condition (i.e., thermal equilibrium) for the steam/water interaction. The complete mixing process, however, is made up of two distinct physical processes. The first is a two-phase interaction with condensation of steam by cold ECCS water. The second is a single-phase mixing of condensate and ECCS water. Since the steam release is the most important influence to the containment pressure transient, the steam condensation part of the mixing process is the only part that needs to be considered. (Any spillage directly heats only the sump.)

The most applicable steam/water mixing test data has been reviewed for validation of the WREFLOOD steam/water mixing model. This data, generated in 1/3-scale tests (EPRI report 294-2), are the largest scale data available and thus, most clearly simulate the flow regimes and gravitational effects that would occur in a Pressurized Water Reactor (PWR). These tests were designed specifically to study the steam/water interaction for PWR reflood conditions.

6.2-34

BVPS-2 UFSAR Rev. 26 A group of 1/3-scale tests corresponds directly to the reflood conditions.

The injection flow rates for this group cover all phases and mixing conditions calculated during the reflood transient. The data from these tests were reviewed and discussed in detail in WCAP-10325-P-A, Westinghouse, 1983. For all of these tests, the data clearly indicate the occurrence of very effective mixing with rapid steam condensation. The mixing model used in the WREFLOOD calculation is therefore wholly supported by the 1/3-scale steam/water mixing data. Descriptions of the test and test results are contained in Docket 50-315, Amendment No. 126, June 1989 and EPRI report 294-2, June 1975.

The calculated DEPS reflood phase LOCA mass and energy releases are given in Table 6.2-14E for the minimum safeguards case and in Table 6.2-14H for the maximum safeguards case and Table 6.2-14Q for the service water failure case. The transient responses of the principal parameters during reflood are given in Table 6.2-14F for the DEPS minimum safeguards case and in Table 6.2-14I for the DEPS maximum safeguards case and Table 6.2-14R for the service water failure case.

6.2.1.3.8.3 Post-Reflood Mass and Energy Release Data The FROTH code, as described by Shepard, et. al., is used for computing the post-reflood transient. The FROTH code calculates the heat release rates resulting from a two-phase mixture present in the steam generator tubes. The mass and energy releases that occur during this phase are typically superheated due to the depressurization and equilibration of the broken-loop and intact-loop steam generators. During this phase of the transient, the RCS has equilibrated with the containment pressure, but the steam generators contain a secondary inventory at an enthalpy that is much higher than the primary side. Therefore, there is a significant amount of reverse heat transfer that occurs. Steam is produced in the core due to core decay heat. For a pump suction break, a two-phase fluid exits the core, flows through the hot legs, and becomes superheated as it passes through the steam generator. Once the broken loop cools, the break flow becomes two-phase. During the FROTH calculation, ECCS injection is addressed for both the injection phase and the recirculation phase. The FROTH code calculation stops when the secondary side equilibrates to the saturation temperature (Tsat) at the containment design pressure. After this point, the EPITOME code completes the SG depressurization. The methodology for the use of this model is described in the March 1979 topical. (See subsection 6.2.1.3.8.5 and subsection 6.2.1.3.8.6 for additional information.)

Table 6.2-14G presents the two-phase post-reflood mass and energy release data for the double-ended pump suction case minimum safeguards case.

Table 6.2-14J presents the two-phase post-reflood mass and energy release data for the double-ended pump suction maximum safeguards case. Table 6.2-14S presents the release data for the double-ended pump suction service water failure case.

6.2.1.3.8.4 Decay Heat Power Model The American Nuclear Society Standard ANSI/ANS-5.1 1979 has been used for the determination of decay heat in the mass and energy release analysis.

6.2-35

BVPS-2 UFSAR Rev. 26 Table 6.2-14V, lists the generic decay heat curve used in the Beaver Valley mass and energy release calculations applying the March 1979 LOCA mass and energy release methodology.

Significant assumptions in the generation of the decay heat curve for use in the LOCA mass and energy release analysis include the following:

1. Decay heat sources considered are fission product decay and heavy element decay of U-239 and Np-239.

2. Decay heat power from the following fissioning isotopes is included: U-238, U-235, and Pu-239.

3. Fission rate is constant over the operating history of maximum power level.

4. The factor accounting for neutron capture in fission products has been taken from Equation 11 (ANSI/ANS-5.1 1979, August 1979) for up to 10,000 seconds and from Table 10 (ANSI/ANS-5.1 1979, August 1979) for beyond 10,000 seconds.

5. The fuel has been assumed to be at full power for 108 seconds.

6. The number of atoms of U-239 produced per second has been assumed to be equal to 70 percent of the fission rate.

7. The total recoverable energy associated with one fission has been assumed to be 200 MeV/fission.

8. Two-sigma uncertainty (two times the standard deviation) has been applied to the fission product decay.

Based upon NRC staff review, Safety Evaluation Report (SER) of the March 1979 evaluation model, use of the ANS Standard ANSI/ANS-5.1-1979 decay heat model was approved for the calculation of mass and energy releases to the containment following a LOCA.

6.2.1.3.8.5 Steam Generator Equilibration and Depressurization Steam generator equilibration and depressurization is the process by which secondary side energy is removed from the steam generators in stages. The FROTH computer code calculates the heat removal from the secondary mass until the secondary temperature is the saturation temperature (T sat) at the containment design pressure. After the FROTH calculations, the EPITOME code continues the FROTH calculation for SG cooldown removing steam generator secondary energy at different rates (i.e., first and second stage rates). The first stage rate is applied until the steam generator reaches Tsat at the user specified intermediate equilibration pressure, when the secondary pressure is assumed to reach the containment pressure.

Then the second stage rate is used until the final depressurization, when the secondary reaches the reference temperature of Tsat at 14.7 psia, or 212°F. The heat removal of the broken-loop and intact-loop steam generators is calculated separately.

6.2-36

BVPS-2 UFSAR Rev. 26 During the FROTH calculations, steam generator heat removal rates are calculated using the secondary side temperature, primary side temperature, and a secondary side heat transfer coefficient determined using a modified McAdam's correlation. Steam generator energy is removed during the FROTH transient until the secondary side temperature reaches saturation temperature at the containment design pressure.

The constant heat removal rate used during the first heat removal stage is based on the final heat removal rate calculated by FROTH. The SG energy available to be released during the first stage interval is determined by calculating the difference in secondary energy available at the containment design pressure and that at the (lower) user specified intermediate equilibration pressure, assuming saturated conditions. The intermediate equilibrium pressures are chosen as discussed in Shepard, et.

al., Sections 2.3 and 3.3. This energy is then divided by the first stage energy removal rate, resulting in an intermediate equilibration time.

At this time, the rate of energy release drops substantially to the second stage rate. The second stage rate is determined as the fraction of the difference in secondary energy available between the intermediate equilibration and final depressurization at 212°F, and the time difference from the time of the intermediate equilibration to the user-specified time of the final depressurization at 212°F. With current methodology (WCAP-10325-P-A, Westinghouse, 1983), all of the secondary energy remaining after the intermediate equilibration is conservatively assumed to be released by imposing a mandatory cooldown and subsequent depressurization down to atmospheric pressure at 3600 seconds, i.e., 14.7 psia and 212°F.

6.2.1.3.8.6 Long Term Mass & Energy Releases The long-term (greater than 3600 seconds) mass and energy release calculations are performed through user defined input functions which is an option in the MAAP-DBA code. The MAAP-DBA code was used for convenience. This method of determining the long-term mass and energy releases is consistent with past applications of Westinghouse methodology.

These user defined functions are characterized for the long term discharge from the break for (a) a mixed discharge and (b) for an unmixed discharge of steam and water. In both cases, the flow rates that are used are those calculated by the EPITOME code and only the specific enthalpies of the discharge flows are calculated to represent the influence of the time dependent RCS injection temperature as the containment cools.

6.2.1.3.8.7 Sources of Mass and Energy The sources of mass considered in the LOCA mass and energy release analysis are given in Tables 6.2-14K, and 6.2-14L and 6.2-14M and 6.2-14T.

These sources are the reactor coolant system, accumulators, and pumped safety injection.

The energy inventories considered in the LOCA mass and energy release analysis are given in Tables 6.2-14N and 6.2-14O and 6.2-14P and 6.2-14U.

The energy sources are listed below.

1. RCS water

2. Accumulator water (all three inject) 6.2-37

BVPS-2 UFSAR Rev. 26

3. Pumped SI water

4. Decay heat

5. Core stored energy

6. RCS metal (includes the reactor vessel and internals, hot and cold leg piping, SG inlet and outlet plenums, and SG tubes)

7. SG metal (includes transition cone, shell, wrapper, and other internals)

Note: The DEHL cases also conservatively include the upper internals and upper elliptical head.

8. SG secondary energy (includes fluid mass and steam mass)

9. Secondary transfer of energy (feedwater into, and steam out of, the SG secondary)

The energy reference points are as follows.

1. Available energy: 212°F; 14.7 psia

2. Total energy content: 32°F; 14.7 psia The mass and energy inventories are presented at the following times, as appropriate:

1. Time zero (initial conditions)

2. End of blowdown time

3. End of refill time

4. End of reflood time

5. Time of broken loop steam generator equilibration to pressure setpoint

6. Time of intact loop steam generator equilibration to pressure setpoint

7. Time of full depressurization (3600 seconds)

The Zirc-water reaction energy was not considered in the mass and energy release data presented because the clad temperature was not assumed to increase high enough for the rate of the Zirc-water reaction to be of any significance.

6.2.1.3.9 Conclusions Plant specific LOCA mass and energy release analyses were developed using approved design basis methodology. The analysis bounds core operation at 6.2-38

BVPS-2 UFSAR Rev. 26 uprated conditions with the current SGs. The results of this analysis were provided for use in the containment analysis.

The consideration of the various energy sources in the long-term mass and energy release analysis provides assurance that all available sources of energy have been included in this analysis. Thus, the review guidelines presented in Standard Review Plan Section 6.2.1.3 have been satisfied.

6.2.1.4 Mass and Energy Release Analysis for Postulated Main Steam Line Break Inside Containment Main Steam Line Breaks occurring inside a reactor containment structure may result in significant releases of high-energy fluid to the containment environment and elevated containment temperatures and pressures. The magnitude of the releases following a steam line rupture is dependent upon the plant initial operating conditions and the size of the rupture as well as the configuration of the plant steam system and the containment design.

These variations make it difficult to determine the absolute worst cases for either containment pressure or temperature evaluation following a steam line break. The main steam line break (MSLB) analysis considers a variety of postulated pipe breaks encompassing wide variations in plant operation, safety system performance, and break size in determining the mass and energy (M&E) releases for use in containment analysis. A spectrum of MSLB accidents, covering different break areas and reactor operating power levels, is analyzed and discussed in the following sections. As stated in Section 6.2.1.1.3.8, a feedwater line break is not analyzed since an MSLB is the most limiting, conservative case with regard to containment design, integrity of the containment pressure boundary, and the resulting containment environmental conditions.

6.2.1.4.1 Mass and Energy Release Data To determine the effects of plant power level and break area on the mass and energy releases from a ruptured steamline, spectra of both variables have been evaluated. At nominal full NSSS power levels of 100.6 percent, 70 percent, 30 percent, and 0 percent of nominal full-load power, two break types have been defined. These breaks are defined as the following.

1. A full double-ended rupture (DER) downstream of the steamline flow restrictor, which is integral with the steam generator nozzle. For this case, the actual break area equals the cross-sectional area of the steamline, but the blowdown from the steam generator with the broken line is controlled by the flow restrictor throat area (1.069 ft2). The reverse flow from the intact steam generators is controlled by the total flow restrictor throat area in the intact loops. Since these flow restrictors are part of the steam generator nozzle, no pipe breaks can occur with a flow area greater than the throat area of 1.069 ft2.

2. A split rupture that represents the largest break that will neither generate a steamline isolation signal from the Westinghouse Solid-State Protection System (SSPS) nor result in water entrainment in the break effluent. Reactor protection 6.2-39

BVPS-2 UFSAR Rev. 26 and safety injection actuation functions are obtained from containment pressure signals.

6.2.1.4.2 Single-Failure Assumptions To avoid unnecessary conservatism, bounding multiple failure assumptions have not been made for most of the MSLB cases in the analysis. Most cases analyzed considered only one single failure. One of these failures is considered as part of the containment response analysis as discussed in Section 6.2.1.1.3.7. The postulated single failures (discussed also in Land 1976) that increase the MSLB M&E releases to containment are discussed below.

a. Failure of the Main Steam Isolation Valve (MSIV) in the Faulted Loop The main steamline isolation function is accomplished via the MSIV in each of the three steamlines. Each valve closes on an isolation signal to terminate steam flow from the associated steam generator. The Main Steam Line Break upstream of this valve, as postulated for the inside-containment analysis, creates a situation in which the steam generator on the faulted loop cannot be isolated, even when the MSIV successfully closes.

The break location allows a continued blowdown from the faulted-loop steam generator until it is empty and all sources of feedwater and auxiliary feedwater addition are terminated. If the faulted-loop MSIV fails to close, blowdown from more than one steam generator is prevented by the closure of the corresponding MSIV for each intact-loop steam generator.

Therefore, there is no failure of a single MSIV that could cause continued blowdown from multiple steam generators.

In addition to the continued blowdown from the faulted-loop steam generator after MSIV closure, the steam in the unisolable section of the steamline needs to be considered. An MSIV failure can impact the mass and energy releases, since a failed MSIV will result in a larger unisolable steamline volume.

b. Failure of the Main Feedwater Isolation Valve (MFIV) in the Faulted Loop If the MFIV in the feedwater line to the faulted steam generator is assumed to fail in the open position, backup isolation is provided via the main feedwater flow control valve (FCV) closure. The additional inventory between the MFIV and the FCV in the faulted loop would be available to be released to containment.

6.2.1.4.3 Initial Conditions Main Steam Line Breaks can be postulated to occur with the plant in any operating condition ranging from hot shutdown to full power. Since steam generator water mass decreases with increasing power level, breaks occurring at lower power levels will generally result in a greater total 6.2-40

BVPS-2 UFSAR Rev. 26 mass release to the containment. However, because of increased stored energy in the primary side of the plant, increased heat transfer in the steam generators, and additional energy generation in the fuel, the energy release to the containment from breaks postulated to occur during full-power, or near full-power, operation may be greater than for breaks occurring with the plant in a low-power, or hot-shutdown, condition.

Additionally, pressure in the steam generators changes with increasing power and has a significant influence on the rate of blowdown.

Because of the opposing effects on mass versus energy release for the MSLB due to a change in initial power level, a single power level cannot be specified as the worst case for either the containment pressure cases or the containment temperature cases. Therefore, representative power levels including 100.6 percent, 70 percent, 30 percent, and 0 percent of nominal full NSSS power conditions have been investigated for BVPS-2 based on the information in Land (1976).

In general, the plant initial conditions are assumed to be at the nominal value corresponding to the initial power for that case, with appropriate uncertainties included. Table 6.2-50 identifies the values assumed for RCS vessel average temperature, RCS flow, RCS pressure, pressurizer water volume, feedwater enthalpy, steam generator pressure, and steam generator water level corresponding to each power level analyzed. Main steam line break mass and energy releases assuming an RCS average temperature at the high end of the Tavg window are conservative with respect to similar releases at the low end of the Tavg window. At the high end, there is more mass and energy available for release into containment. The thermal design flowrate has been used for the RCS flow input consistent with the assumptions documented in Land (1976).

Uncertainties on the initial conditions assumed in the analysis for the BVPS-2 power uprating analysis program have been applied only to the RCS average temperature, the steam generator mass and the power fraction at full power. Nominal values are adequate for the initial conditions associated with pressurizer pressure and pressurizer water level.

Uncertainty conditions are only applied to those parameters that could increase the amount of mass or energy discharged into containment.

6.2.1.4.4 Description of the Blowdown Model The LOFTRAN code (Burnett, et al., 1984) calculates mass and energy releases to the containment following a Main Steam Line Break, as specifically described in this report and which is summarized as follows.

1. Primary system fluid temperatures and pressures calculations The LOFTRAN code is used for studies of transient response of a PWR system to specified perturbations in process parameters.

LOFTRAN is a versatile program suited to both accident evaluations and control system studies. LOFTRAN simulates a multiloop system by a model containing a reactor vessel, hot and cold leg piping, steam generators (tube and shell sides), and the pressurizer. The pressurizer heaters, spray, relief and safety valves are considered in the program. Point-model 6.2-41

BVPS-2 UFSAR Rev. 26 neutron kinetics and reactivity effects of the moderator, fuel, boron, and control rods are included. The secondary side of the steam generator uses a homogeneous, saturated mixture for the thermal transients and a water level correlation for indication and control. Core decay heat generation assumed in calculating the MSLB mass and energy releases is based on the ANS (1979) decay heat + 2 model.

Blowdown mass and energy releases determined using LOFTRAN include the effects of core power generation, main and auxiliary feedwater additions, engineered safeguards systems, reactor coolant system thick-metal heat storage including steam generator thick-metal mass and tubing, and reverse steam generator heat transfer.

The use of the LOFTRAN code for the analysis of the MSLB M&E releases is documented in Osborne and Love (1985) and has been reviewed and approved by the NRC (1986) for this application.

2. Steam generator fluid mass A maximum initial steam generator mass in the faulted-loop steam generator has been used in the analysis of the MSLB inside containment. The use of a high faulted-loop initial steam generator mass maximizes the steam generator inventory available for release to containment. The initial mass has been calculated as the value corresponding to the programmed level

+7 percent narrow-range span and assuming 0 percent tube plugging. The initial mass uncertainty is a conservative value with respect to the BVPS-2 plant-specific value. This assumption is conservative with respect to the RCS cooldown through the faulted-loop steam generator resulting from the Main Steam Line Break.

3. Steam generator reverse heat transfer Once the steamline isolation is complete, the steam generators in the intact loops become sources of energy that can be transferred to the steam generator with the broken steamline.

This energy transfer occurs via the primary coolant. As the primary plant cools, the temperature of the coolant flowing in the steam generator tubes could drop below the temperature of the secondary fluid in the intact steam generators, resulting in energy being returned to the primary coolant. This energy is then available to be transferred to the steam generator with the broken main steam line.

4. Reactor coolant system metal heat capacity As the primary side of the plant cools, the temperature of the reactor coolant could drop below the temperature of the reactor coolant piping, the reactor vessel, the reactor coolant pumps, and the steam generator thick-metal mass and tubing. As this occurs, the heat stored in the metal is available to be 6.2-42

BVPS-2 UFSAR Rev. 26 transferred to the steam generator with the broken line. The effects of this RCS metal heat are included in the results using conservative thick-metal masses and heat transfer coefficients.

5. Break flow model Blowdown properties are determined using the Moody (1965) correlation with a discharge coefficient of 1.0. The quality of the blowdown is input as a function of time for mass and energy release calculations. An input quality of 1.0 is used for all main steam line breaks inside containment. The assumption of saturated steam being released for all breaks is a conservative assumption that maximizes the energy release into containment.

6. Loss of offsite power Loss of offsite power is not assumed in the MSLB analysis. The assumption of a trip of all the reactor coolant pumps (RCPs) coincident with reactor trip is less limiting than with offsite power available since the mass and energy releases are reduced due to the loss of forced reactor coolant flow, resulting in less primary-to-secondary heat transfer (Land 1976). Therefore, all MSLB M&E release cases are analyzed with the RCPs continuing to operate.

7. Core reactivity coefficients Since the main steam line break is a cooldown event, it is conservative to use large negative moderator coefficients and low Doppler coefficients as characteristic of end-of-cycle (EOC) life. Most limiting core reactivity coefficients at EOC are used to maximize the reactivity feedback effects resulting from the MSLB. Use of maximum reactivity feedback results in higher power generation if the reactor returns to criticality, thus maximizing heat transfer to the secondary side of the steam generators. Also, for all MSLBs, the most reactive control rod is assumed to be stuck out of the core.

6.2.1.4.5 Energy Inventories The rapid depressurization that occurs following a main steam line break typically results in large amounts of water being added to the steam generators through the main feedwater system. Rapid-closing FIVs or FCVs in the main feedwater lines limit this effect. The feedwater addition that occurs prior to closing of the FIVs or FCVs influences the steam generator blowdown in several ways. First, because the water entering the steam generator is subcooled, it lowers the steam pressure thereby reducing the flowrate out of the break. As the steam generator pressure decreases, some of the fluid in the feedwater lines downstream of the isolation valves will flash into the steam generators providing additional secondary fluid which may exit out of the break. Secondly, the increased flow causes an increase in the total heat transfer from the primary to secondary systems resulting in greater integrated energy being released out of the break.

6.2-43

BVPS-2 UFSAR Rev. 26 Following the initiation of the MSLB, main feedwater flow is conservatively modeled by assuming an increase in feedwater flow prior to reactor trip. The initial increase in feedwater flow (until fully isolated) is in response to the feedwater control valve opening up in response to the steam flow/feedwater flow mismatch, or the decreasing steam generator water level as well as due to a lower backpressure on the feedwater pump as a result of the depressurizing steam generator. This maximizes the total mass addition prior to feedwater isolation. The feedwater isolation response time, following the safety injection signal, is assumed to account for delays associated with signal processing plus MFIV stroke time. For the circumstance in which the MFIV in the faulted loop fails to close, there is no effect on the feedwater isolation response time since the total delay for the FCV closure is also 7 seconds.

Following feedwater isolation, as the steam generator pressure decreases, some of the fluid in the feedwater lines downstream of the isolation or control valve may flash to steam if the feedwater temperature exceeds the saturation temperature. This unisolable feedwater line volume is an additional source of fluid that can increase the mass discharged out of the break. The unisolable volume in the feedwater lines is maximized for the faulted loop. Feedwater line piping volumes available for steam flashing in this analysis are shown in Table 6.2-51.

Generally, within the first minute following a main steam line break, the auxiliary feedwater (AFW) system is initiated on any one of several protection system signals. Addition of auxiliary feedwater to the steam generators will increase the secondary mass available for release to containment as well as increase the heat transferred to the secondary fluid. The auxiliary feedwater flow to the faulted and intact steam generators has been assumed to be a function of the backpressure on the AFW pumps as a result of the depressurizing steam generator in the MSLB analysis inside containment. A range of cavitating venturi sizes in each of the AFW supply lines to the steam generators has been assumed that maximizes flow to the faulted-loop steam generator and minimizes flow to the intact-loop steam generators. Auxiliary feedwater flow to the faulted-loop steam generator has been assumed up until the time of operator action to isolate the flow to the steam generator near the break location. Auxiliary feedwater system assumptions that have been used in the analysis are presented in Table 6.2-51.

6.2.1.4.6 Additional Information Required for Confirmatory Analyses For the DER cases, the forward-flow cross-sectional area from the faulted-loop steam generator is limited by the integral flow restrictor area of 1.069 ft2, which is less than the actual area of 4.9 ft2 for the main steam piping inside containment. The cross-sectional area of the steam piping at this location is larger than the sum of the flow restrictors in the intact-loop steam generators. Therefore, the larger cross-sectional area of the ruptured steamline expels steam faster than the smaller cross-sectional area of the intact-loop steam generator flow restrictors can fill it. Thus, the blowdown of the initial steam in the steamline header piping is modeled in the first few seconds of the event, followed by the reverse-flow blowdown from the intact-loop steam generators until MSIV 6.2-44

BVPS-2 UFSAR Rev. 26 closure. At the time of MSIV closure, the steam flow from the intact-loop steam generators is terminated, but it is assumed that all steam that has exited the steam generator prior to steamline isolation is released through the break.

The contribution to the mass and energy releases from the steam in the secondary plant main steam loop piping and header has been included in the mass and energy release calculations. The initial flowrate is determined using the Moody correlation, the pipe cross-sectional area, and the initial steam pressure. A conservative steam piping volume is used in this blowdown calculation representing the main steam piping from the steam generator to the turbine throttle valve.

The full DER represents the break producing the highest mass flowrate from the faulted-loop steam generator. Smaller DER break sizes are represented by a reduction in the initial steam blowdown rate at the time of the break. Therefore, no other DER break sizes have been considered other than the full DER.

For the split-break MSLB cases, the break area is smaller than the area of a single integral flow restrictor. The flowrate from all steam generators prior to MSIV closure and the flowrate from a single steam generator after MSIV closure supply the steam flow to the break. The steam in the unisolable portion of the steamline does not affect the blowdown until the time of steam generator dry out, when the flowrate from the steam generator would decrease below the critical flowrate out of the break. At this point, the additional steam in the piping begins to have an effect on break flowrate until the steamline piping is empty. To model this effect, the mass of the unisolable steam in the steamline is added to the initial mass of the faulted steam generator. This accurately reflects both the total mass and energy that will be released from the break, and the timing of the effect of the unisolable steamline volume on the blowdown.

Steamline isolation is assumed in all three loops to terminate the blowdown from the two intact steam generators. A time accounting for delays associated with signal processing plus MSIV stroke time, with unrestricted steam flow through the valve during the valve stroke, has been assumed.

6.2.1.5 Minimum Containment Pressure Analysis for Performance Capability Studies on Emergency Core Cooling System The containment backpressure used for the limiting case double-ended cold leg guillotine break for the ECCS analysis presented in Section 15.6.5 is shown on Figure 6.2-119. The containment backpressure is calculated, using the methods and assumptions described in Appendix A by Bordelon (et al 1974a). Input parameters, including the containment initial conditions, net free containment volume, passive heat sink materials, thicknesses, and surface areas, and starting time and number of containment cooling systems used in the analysis, are described in the following paragraphs.

6.2-45

BVPS-2 UFSAR Rev. 26 6.2.1.5.1 Mass and Energy Release Data The mass and energy releases to the containment during the transient includes the break flow and ECCS spill through the broken cold leg from the vessel side and loop side, and are presented in Tables 6.2-52 and 6.2-53, respectively.

The mathematical models which calculate the mass and energy releases to the containment are described in Section 15.6.5 and conform to 10 CFR 50, Appendix K, ECCS Evaluation Models. A break spectrum analysis is performed (Section 15.6.5) that considers various Moody discharge coefficients for the double-ended cold leg guillotines. During refill, the break mass and energy released to the containment is assumed to be zero, which minimizes the containment pressure. During reflood, the effect of steam water mixing between the safety injection water and the steam flowing through the RCS intact loops reduces the available energy released to the containment vapor spaces and therefore tends to minimize containment pressure.

6.2.1.5.2 Initial Containment Internal Conditions The following initial values were used in the analysis:

1. A minimum containment pressure of 14.3 psia,

2. A nominal containment temperature of 70F,

3. A minimum RWST temperature of 45F,

4. A minimum outside air temperature of -20F,

5. A minimum outside ground temperature of 32,

6. A maximum relative humidity of 100 percent.

6.2.1.5.3 Containment Volume The volume used in the analysis was 1.804 x 106 ft3.

6.2.1.5.4 Active Heat Sinks The containment spray system operates to remove heat from the containment.

Pertinent data for this system which were used in the analysis are presented in Table 6.2-54. No containment air coolers were modeled in this analysis. In addition, heat transfer between the sump water and the containment vapor space was not considered in the analysis.

6.2-46

BVPS-2 UFSAR Rev. 26 6.2.1.5.5 Steam-Water Mixing Water spillage rates from the broken loop accumulator are determined as part of the core reflooding calculation and are included in the containment code calculational model.

6.2.1.5.6 Passive Heat Sinks The passive heat sinks used in the analysis, with their thermophysical properties, are given in Table 6.2-55. The passive heat sinks and their thermophysical properties were determined to be representative of plant conditions which would produce conservatively high heat removal.

6.2.1.5.7 Other Parameters No other parameters have a substantial effect on the minimum containment pressure analysis.

6.2.1.6 Inspection and Testing Requirements Containment testing is classified in two categories as follows:

1. Structural acceptance integrity testing verifies the structural integrity of the reactor containment exterior structure. This is described in Section 3.8.1.7.

2. Containment leak rate testing verifies the leakage rates are within allowable limits. This is described in Section 6.2.6.

6.2.1.7 Instrumentation Requirements Indicators are provided in the main control room to monitor containment atmosphere temperature and humidity. These are also monitored by the Beaver Valley Power Station - Unit 2 (BVPS-2) computer.

A detailed discussion of the instrumentation, actuations, and logic functions may be found in Section 7.1, 7.2, 7.3, and 7.5.

6.2.2 Containment Heat Removal System The CHRS, which is also called the containment depressurization system, is composed of two systems: the quench spray system (QSS) and the recirculation spray system (RSS).

The containment depressurization system reduces the containment temperature and returns the containment pressure to subatmospheric following a break in either the primary or secondary system piping inside the containment.

Heat that is removed from the containment atmosphere by the QSS and RSS is transferred to the containment sump. Heat is then removed from the containment by the service water system via the recirculation spray coolers.

6.2-47

BVPS-2 UFSAR Rev. 26 6.2.2.1 Design Bases The containment depressurization system is designed in accordance with the following criteria:

1. General Design Criterion 38, with respect to containment heat removal.

2. General Design Criterion 39, with respect to permitting periodic inspection of the containment depressurization system.

3. General Design Criterion 40, with respect to permitting periodic testing of the containment depressurization system.

4. General Design Criterion 41, as it relates to the control of fission products. Section 6.5.2 discusses iodine removal.

5. Regulatory Guide 1.1, as it relates to the NPSH available (NPSHA) to the ECCS and containment depressurization system pumps.

6. Regulatory Guide 1.26, as it relates to quality group standards. The system is designed in accordance with ASME Section III Class 2, and is designated Safety Class 2 except as indicated on CHRS figures referenced in Table 3.2-3.

7. Regulatory Guide 1.29, as it relates to seismic classification.

The system is designed to Seismic Category I.

8. Regulatory Guide 1.82, as it relates to the design of sumps for ECCS and CSSs. Design exception and justifications are discussed in Section 6.2.2.3.1.

9. The containment spray headers are capable of delivering spray water to the containment atmosphere in sufficient quantity and with an average droplet diameter to ensure adequate heat removal.

10. The containment depressurization system will perform its design function properly, assuming the worst single active failure in the short term or an active or passive failure in the long term (after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />).

11. The potential for surface fouling of the secondary sides of the recirculation spray heat exchangers is considered in calculating heat transfer rates.

12. Instrumentation is provided to monitor the containment depressurization system and system component performance under normal and accident conditions to satisfy the design requirement of GDC 38.

13. Provisions are made to allow drainage of spray and ECCS water to the containment sump. The sources and quantities of energy 6.2-48

BVPS-2 UFSAR Rev. 26 that must be removed from the containment to meet the design bases are given in the energy distribution tables in Section 6.2.1.

14. The containment atmosphere pressure is reduced to less than half of the peak pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Section 6.2.1).

15. The RSS of the containment depressurization system is capable of operating in the post-accident environment for 30 days following the DBA.

6.2.2.2 System Design The containment depressurization system consists of two systems: the QSS and the RSS. The design of each system is described in the following sections.

The entire containment depressurization system is constructed of corrosion-resistant materials, primarily stainless steel.

The components of the containment depressurization system have been selected so that the conditions of service (pressure, temperature, and fluid composition) do not prevent the system from performing its intended function. Section 3.11 discusses the environmental design conditions of the containment depressurization system components.

6.2.2.2.1 Quench Spray System The QSS, shown on Figure 6.2-121, is composed of two redundant parallel trains: Train A and Train B. Each train contains one quench spray pump which draws water independently from the RWST and is capable of providing 100 percent of the required spray capacity. Separate emergency diesel generators provide power to the electrically-operated components of Train A and Train B. The two trains connect to two common 360 degree spray headers connected in parallel, with risers 180 degrees apart. Component design data for the QSS are given in Table 6.2-56. The quench spray pumps are activated after receipt of a CIB signal and are effective 87.6 seconds after a design basis LOCA (85.5 seconds after a CIB signal setpoint).

This delay (assuming design basis LOCA coincident with LOOP and emergency diesel generator start signal) is made up of the following delays:

Emergency diesel generator start-up 10 seconds Sequencer delay 16.5 seconds Pump acceleration to rated speed 3 seconds Piping fill time (maximum) 45 seconds Total delay 74.5 seconds Each quench spray pump operating alone is capable of supplying approximately 3,000 gpm of spray flow to the quench spray headers. Both quench spray pumps operating together can supply approximately 4,500 gpm 6.2-49

BVPS-2 UFSAR Rev. 26 to the spray headers. The actual flow rate is determined by the difference between the containment total pressure and the RWST water level. This system flow curve is based upon the pump head versus capacity curve supplied by the pump manufacturer, the pressure losses in the lines, headers, and nozzles, and elevation differences in the system.

The quench spray pumps are located in separate cubicles adjacent to both the containment structure and the RWST and are flood- and missile-protected. Each quench spray pump discharge line contains a weight-loaded check valve inside containment and a motor-operated valve (MOV) outside containment.

The two quench spray ring headers are located at el 78 ft-6 in and 103 ft-8 in, respectively, above the operating floor, with centerline diameters of 100 ft-6 in and 31 ft-6 in, also respectively.

There are a total of 159 nozzles on the two quench spray ring headers:

120 nozzles on the lower header and 39 nozzles on the upper header. The orientation of the quench spray nozzles on each header is shown on Figure 6.2-123. The nozzle spacing and direction of the spray are designed to maximize spray coverage. The quench spray coverage is discussed in Section 6.2.2.3.1.

The mean diameter of the spray droplets is less than 1,000 microns.

The quench spray nozzles are manufactured by Spray Engineering Company (SPRACO) and are Model 1713A. These nozzles have no moving parts and are of a design which minimizes clogging. The orifice size of these nozzles is 3/8-inch diameter. The analysis of the spray effectiveness is based on information provided on this model nozzle (SPRACO undated).

The RWST is a Seismic Category I vertical, cylindrical tank with a flat bottom and hemispherical top, mounted on and secured to a reinforced concrete foundation. The tank is fabricated of Type 304L stainless steel plates. The tank has a vent that is sized to provide adequate protection against overpressurization and excessive internal vacuum. The RWST is vented via a 12-inch stainless steel vent pipe located at the highest point on the tank roof, well above the overflow line. The layout of the vent line has a 180 return shape with no low points which could become clogged. The stainless steel material is corrosion resistant and the entire length of the vent line is heat traced to prevent freezing.

Component design data for the RWST is given in Table 6.2-56.

The water temperature in the RWST is controlled by circulating it through a heat exchanger which uses chilled water. The tank is insulated to limit the temperature rise of 50F water to 0.5F per 24-hour period with the outside mean temperature at its maximum value for the site of 104F. The tank is provided with a manhole for access into the tank for inspection during the refueling periods and an overflow to the nuclear equipment vent and drain system.

The tank is designed as a Seismic Category I component to withstand design seismic loadings. An evaluation is made to establish that there is no loss of function following the safe shutdown earthquake conditions. The RWST is not itself required for post-earthquake safe shutdown. Piping of 6.2-50

BVPS-2 UFSAR Rev. 26 the connecting systems, including the connection to the fuel pool, is also designed to withstand seismic loading to ensure the functioning of the systems.

The connecting piping from the RWST is isolated from the non-safety Fuel Pool Purification System by two series isolation valves that receive a safety injection signal to close during a DBA requiring use of the RWST as a borated water source.

6.2.2.2.2 Recirculation Spray System The RSS, shown on Figure 6.2-122, consists of two 360-degree spray ring headers and four pumps and heat exchangers. Each 360-degree spray ring header is fed by two risers, where each riser originates from one of the recirculation coolers. The two redundant recirculation spray pumps that are connected to the same spray ring header are supplied with emergency power from separate emergency diesel generators. The design recirculation flow from the spray headers is maintained following failure of an emergency bus.

The design of the RSS is sufficiently independent and redundant so that an active failure in the recirculation spray mode, or an active or passive failure in the cold leg or hot leg recirculation mode of ECCS, has no effect on its ability to perform its safety function.

The recirculation spray pumps are started automatically after receipt of a CIB signal coincident with a RWST level low signal. The spray becomes effective approximately 77 seconds after pump start. When the water in the RWST has reached a predetermined extreme low level, two of the recirculation spray pumps are automatically switched to the cold leg recirculation mode of ECCS, as discussed in Section 6.3.

The two recirculation spray ring headers are located at el 81 ft-2 in and 84 ft-5 in, respectively, above the operating floor, with centerline diameters of 97 ft-6 in and 93 ft, also respectively.

Each recirculation spray ring header contains 292 nozzles. The nozzles are Model 1713A, manufactured by SPRACO. The orientation of the recirculation spray nozzles is shown on Figure 6.2-123.

The four recirculation spray pumps take suction from the containment sump, which is enclosed by a protective screen assembly. A description of the containment sump strainer assembly is provided in Section 6A.1.2.

The containment sump pH control system improves the iodine removal capability of the recirculation spray system. A minimum of 13,980 pounds of sodium tetraborate is installed in 6 stainless steel baskets located around the perimeter of the containment bottom floor elevation of 690-11. These basket structures are seismically designed. Upon flooding of the containment, the NaTB in solution with the borated water provides a minimum containment sump water pH of 7.0.

The four recirculation spray pumps and associated motors are located in separate cubicles outside the containment. The four pumps are of the vertical deep-well type and have shaft extensions to permit locating the pump suctions at a level below the containment sump, with the motors at an elevation slightly above grade. Each pump has a design capacity of approximately 3,500 gpm.

6.2-51

BVPS-2 UFSAR Rev. 26 The pumps are fitted with a tandem mechanical seal arrangement, which provides a seal fluid between the double mechanical seals on the recirculation pumps to preclude outleakage of radioactive fluid during pump operation. The recirculation pump tandem mechanical seal is shown on Figure 6.2-124. The seal arrangement consists of two mechanical face seals (Items No. 2 and No. 7) arranged to enclose a nonradioactive seal fluid. After the pump is started, the pressure between the seals is maintained higher than the pressure outside either seal by an accumulator (Item No. 5) with a weight-loaded diaphragm. The accumulator is fitted with two sealed level indicator alarms (Item No. 9) for annunciation in the main control room. The seal fluid is cooled by being pumped through an air cooler (Item No. 4) by a pumping ring (Item No. 6). The accumulator is conservatively sized to allow for sufficient volume of demineralized water outleakage during operation of the pump.

The accumulator and piping are constructed of stainless steel, are designed for service at peak operating conditions, and are designed in accordance with ASME Section III, Class 2.

The tandem mechanical seal arrangement provides a double barrier against leakage of radioactive fluid from the seals of the recirculation pumps. The seal arrangement also provides adequate lubrication and cooling to prevent scoring of the seals. Level alarms are provided to ensure that adequate accumulator water volume is available and to indicate the failure of either seal.

All portions of the recirculation spray pumps which come in contact with pumped fluid are constructed of austenitic stainless steel or other materials of superior corrosion resistance. The recirculation spray pumps are designed in accordance with the criteria in Tables 6.2-57 and 6.5-1.

The recirculation spray pumps are designed to accommodate the anticipated debris present in the containment sump. Parts of the pump running with close tolerances are provided with surface hardness and finish to withstand particulate matter. Specifically, the bearings of the pump are carbon/graphite type with flushing grooves. The pump's shaft sleeves are Type 304L stainless steel with hardfacing. The combination of the two provides a highly reliable bearing surface in the presence of particulates. Similarly, the pump's wearing rings, both impeller and bowl, are hardened to provide wear resistance if particulate matter is in pumpage. The impeller wearing ring is hardfaced with a carbide material and the bowl ring is a heat treated stainless steel brought to a hardness to provide a difference in Brinell hardness between the stationary and rotating ring.

An endurance test has been performed to verify that the required performance will not be degraded during extended operation. The endurance test consisted of a total run time of 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> in a 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> on/1 hour off cyclic operation. The following parameters were monitored throughout the test period: flow rate, suction and discharge pressure, and vibration. Pump performance throughout the test did not vary outside the acceptable performance criteria of 2 percent of the pump performance curve established at the start of the test. After completion of the test, inspection of the internal surfaces of the pump indicated that only minimal wear was experienced during the duration of the test.

The recirculation water flows through recirculation coolers, where it is cooled by service water (Ohio River) (Section 9.2) flowing at 6.2-52

BVPS-2 UFSAR Rev. 26 approximately 5,500 gpm per cooler. The overall heat transfer coefficients for each recirculation cooler include fouling factors for the tube side and shell side of 0.0003 hour-F-ft2/Btu. Since the recirculation water pressure in the coolers is greater than the service water pressure, only outleakage can occur and dilution of the borated water by service water in the containment is not possible. This ensures that the necessary cold shutdown margin by boron is maintained. Component design data for the RSS are given in Table 6.2-57.

6.2.2.3 Design Evaluation 6.2.2.3.1 Heat Removal System Performance The effectiveness of the containment sprays to cool the containment atmosphere following a LOCA is greater than 99 percent for the quench spray and greater than 95 percent for the recirculation spray. The small droplets of the quench and recirculation sprays approach 100 percent of thermal equilibrium with the containment atmosphere while falling through the containment atmosphere. These results are based on the (Parsly) 1970 method. It is conservatively assumed in all containment design analyses that the spray effectiveness for heat removal is 95 percent.

The droplet size spectrum for the Model 1713A nozzle was determined by SPRACO, utilizing the procedure described as follows.

Droplet size spectrum tests were accomplished by using high speed photography. A special chamber was used to house the photographic equipment. A slot on the roof of this chamber made it possible for a portion of the spray to project into the photographic chamber. This photographic equipment was mounted on a traversing rack which traversed outward from the spray axis. Photographs were taken in each size zone of the quadrant. The images of stopped motion droplets recorded on the negatives were measured and counted. Histograms, which are the incremented frequency plots, were constructed for each test condition (SPRACO undated). A typical histogram is shown on Figure 6.2-125.

Figures 6.2-126 and 6.2-127 are plan and elevation views, respectively, of the containment which show expected spray patterns and the spray overlap.

These spray patterns are for the quench spray only with one quench spray pump operating. With only one quench spray pump operating, approximately 100 percent of the containment operating floor area is covered and approximately 79 percent of the containment free volume is sprayed. Table 6.2-58 lists the containment subvolumes covered by the quench spray. The quench spray coverage is determined from the typical coverage charts from the nozzle manufacturer for the SPRACO Model 1713A nozzle at 40 psig across each nozzle, corrected for elevation, pressure and temperature.

For the recirculation spray subsystems, a strainer assembly is installed over the containment sump to provide protection against debris for downstream components. The performance of the containment sump has been evaluated in accordance with Generic Letter 2004-02. Descriptions of the containment sump strainer and Generic Letter 2004-02 evaluations are provided in Appendix 6A.

6.2-53

BVPS-2 UFSAR Rev. 26 Heat is transferred to the structural steel and concrete inside the containment by conduction and condensation. The energy absorption rate by the passive heat sinks is shown on Figure 6.2-129.

Heat is also removed from the containment atmosphere by the quench and recirculation sprays and is transferred to the containment sump water.

The energy in the containment sump water is transferred to the service water via the recirculation spray coolers. The energy removal rate by the recirculation spray coolers from sump water is shown on Figures 6.2-130 and 6.2-130A. Minimum ESF is assumed for this calculation.

6.2.2.3.2 Net Positive Suction Head Available to Recirculation Pumps Sufficient NPSH is available to the recirculation pumps during both the recirculation spray mode and the recirculation mode of low head safety injection. The following equation is used to calculate the NPSHA.

NPSHA = Pc + Z - Hf - Pv where:

Pc = Containment atmosphere total pressure Z = Elevation head of water above first stage impeller Hf = Head loss from friction in pump suction pipe Pv = Vapor pressure of sump liquid (saturation pressure at liquid temperatures)

All preceding parameters are expressed in feet of head.

It should be noted that NPSH is referenced to the inlet of the first stage impeller.

Table 6.2-59 compares DBA LOCA NPSHA to required NPSH for recirculation pump at start of recirculation spray, assuming minimum ESF (one train).

Required NPSH includes a 15% increase to account for air ingestion in accordance with Regulatory Guide 1.82.

The DBA LOCA NPSHA is conservatively low; the actual NPSHA will always be higher for the following reasons:

1. As time progresses, containment water level used to calculate the elevation head will continually increase due to quench spray and break effluent release inside containment.

2. Spray and condensed water hold up at different floors and cavities inside containment is conservatively calculated and subtracted from the sump water inventory. Condensed water film on heat sink surfaces is also subtracted from the sump water inventory.

6.2-54

BVPS-2 UFSAR Rev. 26

3. The time delays for spray, drainage, and condensed water to reach the sump are considered.

For these reasons, i.e., conservatively low NPSHA and conservatively high NPSH required, a higher margin between available and required NPSH is assured.

Sensitivity analyses are used to determine the limiting input parameters, break locations and single failures which predict the minimum NPSH available.

6.2.2.3.3 Iodine Removal by Containment Spray System The spray nozzles are selected to provide adequate iodine removal capability and containment coverage. The resulting iodine removal coefficients are evaluated in Section 6.5.2.

6.2.2.3.4 Failure Analysis A failure modes and effects analysis (FMEA) to determine if the instrumentation and controls (I&C) and electrical portions meet the single failure criterion, and to demonstrate and verify how the GDC and IEEE Standard 279-1971 requirements are satisfied, has been performed on the QSS and RSS. The FMEA methodology is discussed in Section 7.3.2. The results of this analysis can be found in the separate FMEA document (Section 1.7).

6.2.2.4 Inspection and Testing Requirements Preoperational tests are performed on the containment depressurization system as described in Section 14.2.12. Section 6.6 describes the in-service periodic inspection and system pressure tests.

On January 11, 2008, the NRC issued Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems. Generic Letter 2008-01 requested licensees to evaluate the licensing basis, design, testing, and corrective action programs for the emergency core cooling, decay heat removal, and containment spray systems.

As a result, the company performed evaluations that included the review of gas susceptible piping locations, the development of activities to monitor various piping locations as appropriate based on industry experience and plant specific experience, and acceptance of some generic locations that normally accumulate voids that do not adversely affect the design function(s) of the system.

The company established a gas accumulation prevention and management program to ensure that gas accumulation is reasonably prevented or maintained less than the amount that challenges the functionality of the applicable systems and that appropriate action is taken when conditions adverse to quality are identified.

6.2.2.5 Instrumentation Requirements 6.2-55

BVPS-2 UFSAR Rev. 26 Control switches with indicating lights are provided in the main control room for the quench spray pumps. These pumps can be started manually or automatically. While in the automatic mode of operation, a diesel loading sequence signal combined with a CIB signal, or a CIB signal without a loss of power, will initiate the QSS. These pumps can be stopped manually after CIB is reset.

These pumps deliver cold water from the RWST to the spray headers inside the containment.

Control switches with indicating lights are provided in the main control room for the quench spray pumps suction and discharge MOVs. A CIB signal being present provides a signal to open these valves even though the quench spray pumps suction and discharge valves are normally open during plant operation. When a CIB signal is not present and a respective quench spray pump is not running, these valves can be closed manually.

Ammeters are provided on the main board in the main control room for each of the quench spray pumps.

Pressure indicators are provided in the main control room for each quench spray pump discharge and suction pressures.

Annunciation is provided in the main control room for the quench spray pumps auto start/stop and QSS trouble, which consists of the quench spray pumps low flow. These are also monitored by the plant computer system.

Control switches with indicating lights are provided in the main control room for the recirculation spray pumps. These pumps may be started manually or automatically. During the automatic mode of operation, the pumps are started when a CIB signal is present coincident with a RWST low level with normal or emergency ac power. Two of the operating recirculation spray pumps (C and D) are automatically switched to the cold leg recirculation mode of the ECCS when an extreme low level is reached in the RWST. The recirculation pumps may be stopped when a CIB signal is not present. The recirculation spray pumps take suction from the containment sump.

Control switches with indicating lights are provided in the main control room for the recirculation pump suction valves. These suction valves open automatically from a CIB and are normally open during plant operation.

Control switches with indicating lights are provided in the main control room for the recirculation pump discharge valves. Two of these discharge valves (A and B) operate the same as the suction valves.

The other two discharge valves (C and D) open when a CIB signal is present, and close when a recirculation mode initiation signal is present, and a respective safety injection discharge header valve is open. All of the discharge valves are normally open during plant operation.

Minimum flow recirculation valves are provided for two of the recirculation spray pumps (C and D). These valves will open automatically 6.2-56

BVPS-2 UFSAR Rev. 26 when a low flow signals from the recirculation spray pumps is present.

These valves will close when a normal discharge flow is present from the recirculation spray pumps. These valves are used during recirculation mode of safety injection to maintain minimum flow of the recirculation spray pumps.

Pressure indicators are provided in the main control room for each recirculation spray pump discharge pressure. Flow indicators are provided in the main control room for each recirculation spray pump discharge flow.

Temperature indicators are provided in the main control room for each recirculation spray pump discharge temperature. Ammeters are provided in the main control room for each recirculation spray pump. Level indicators are provided in the main control room for the containment sump level and the recirculation spray sump level.

Temperature indicators are provided in the main control room for the containment sump temperature. A temperature recorder is provided in the main control room for the containment sump temperature. A level recorder is provided in the main control room for the containment sump water level. Flow recorders are provided in the main control room for the recirculation spray pump discharge flow.

Annunciation is provided in the main control room for recirculation spray system trouble, which consists of: recirculation spray pumps seal water level off normal/low, recirculation spray pumps clean-out pit water level high, and recirculation spray pumps vibration high. Annunciation is also provided in the main control room for incore instrument room/containment sump level high/valve not reset, recirculation spray pumps auto start/stop, and recirculation spray sump level high. These conditions are monitored by the BVPS-2 computer system.

6.2.3 Secondary Containment Functional Design Beaver Valley Power Station - Unit 2 does not have a dual containment structure, therefore this section does not apply.

6.2.4 Containment Isolation System The purpose of the containment isolation system (CIS) is to isolate piping lines which penetrate the containment and to prevent the release of radioactive materials from the primary containment in the event of a LOCA.

6.2.4.1 Design Bases The design of the CIS is in accordance with the following criteria:

1. All fluid system components, which are part of the CIS, are designed, fabricated, installed, and tested in accordance with Quality Assurance Category I, Seismic Category I, and Safety Class 2 requirements described in Section 3.2 and in the particular fluid system sections referenced in Table 6.2-60.

General Design Criterion 1 is met by implementation of QA Category I requirements.

6.2-57

BVPS-2 UFSAR Rev. 26

2. All components of the containment isolation fluid, electrical, and controls systems are located in Seismic Category I structures, protected from the effects of natural phenomena, thus meeting GDC 2.

3. All containment isolation fluid system components as well as electrical and control components required for initiation are evaluated for the effects of postulated missiles as described in Section 3.5. This analysis plus the pipe break effects analysis described in Section 3.6 ensures that containment isolation can be achieved, thus meeting the requirements of GDC 4.

4. General Design Criterion 16 as it relates to providing a leaktight barrier against the uncontrolled release of radioactivity to the environment.

General Design Criteria 54 through 57 and Regulatory Guide 1.11 requirements are met, as applicable, in the design bases that follows.

5. Containment isolation consists of at least two barriers between the atmosphere outside containment and: 1) the atmosphere inside the containment; 2) the RCS; and 3) systems which could become connected to either the containment atmosphere or the RCS as a result of, or subsequent to, a LOCA.

The two barriers consist of one of the following arrangements :

a. One normally closed, administratively controlled isolation valve inside containment, and one normally closed, administratively controlled isolation valve outside containment,

b. One automatic isolation valve inside containment and one normally closed, administratively controlled isolation valve outside containment,

c. One normally closed, administratively controlled isolation valve inside containment and one automatic isolation valve outside containment; however, a simple check valve may not be used as the automatic isolation valve outside containment,

d. One automatic isolation valve inside containment and one automatic isolation valve outside containment; however, a simple check valve may not be used as the automatic isolation valve outside containment,

e. A sealed system inside containment and one isolation valve outside containment which is either automatic, normally closed, administratively controlled, or remote manually operated (a sealed system is one which is neither part of 6.2-58

BVPS-2 UFSAR Rev. 26 the reactor coolant pressure boundary nor connected directly to the containment atmosphere), or

f. In the case of the containment sump suction pipe and valve arrangements, a conservatively designed and fabricated single valve and suction pipe arrangement that prevents gross system leakage.

Check valves are weight-loaded or spring-loaded to have positive closure in the direction of flow. This ensures that these check valves will remain closed when the inside containment atmosphere returns to subatmospheric conditions following a DBA.

6. Containment isolation valves outside containment are located so as to require a minimum length of piping between the isolation valves and their respective penetrations.

7. The containment penetrations are designed such that operational test procedures, when used in conjunction with test connections (where required), can be used to check the leaktightness of each automatic containment isolation valve, including weight-loaded and spring-loaded check valves in accordance with 10 CFR 50, Appendix J, and as described in Section 6.2.6.

8. Determination of the extent of fuel failure (source term, Section 15.6.5) used in radiological calculations is in accordance to 10 CFR 50, Appendix K.

9. Instrumentation and adjunct control circuits associated with automatic valve closure shall cause the valves to fail in the position that provides greater safety upon loss of voltage or control air. Circuits which control redundant automatic valves are redundant to the extent that no single failure will preclude isolation.

10. Means are provided to periodically test the functions of the automatic isolation equipment, such as the set point of sensors, speed of response, and operability of fail-safe features.

11. The CIS is designed to be in compliance with Regulatory Guides 1.26, 1.29, and 1.141.

12. All remotely actuated valves of the CIS have their positions indicated in the main control room by separate limit switches installed directly on the valve actuator.

13. The CIS meets the intent of NUREG-0737, Action Item II.E.4.1, requiring dedicated hydrogen penetrations for recombiners to be located external to the containment (Section 6.2.5).

14. The CIS meets the intent of NUREG-0737, Action Item II.E.4.2, with regard to containment isolation dependability. The CIS is 6.2-59

BVPS-2 UFSAR Rev. 26 activated automatically when monitored system variables exceed pre-established set points. It is neither necessary nor desirable that every containment isolation valve close simultaneously upon a containment isolation signal. The plant design allows selected valves in the ESF systems, which are essential to mitigate the effects of an accident, to remain in or move to their open position. These pre-selected valves are remote, manually-operated, and are operated from the main control room. Table 6.2-60 lists the position of each valve in the CIS, under various plant conditions, and the valve position in response to containment isolation signal. The initiating conditions for containment isolation are presented in Section 7.3 in discussions of circuits, isolation functions and setting, logics, resetting, and redundancy. Section 9.4.7.3 discusses the containment purge system and its valve operations in relation to the requirements of the action item.

15. Section 8.3.1 discusses Class lE electrical power supplies utilized for initiating and completing valve closures. All controls and electrical systems for containment isolation are Class lE, and environmentally qualified in accordance with Section 3.11.

16. Spare penetrations are welded closed inside the containment.

17. A detailed pipe break and dynamic effects analysis, including pipe break exclusion and crack exclusion as applicable, is done for all piping systems inside and outside containment and is described in Section 3.6 to ensure that all essential safety-related components can perform their safety function following a postulated pipe rupture.

18. The CIS meets the single failure criterion described in Section 3.1.1.

6.2.4.2 System Description The CIS is initiated by diverse parameters, such as containment pressure and various reactor parameters. The initiation signals are train-operated with redundant and electrical separation being supplied from the initiating device to the individual isolation valve. Control circuits for initiation of containment isolation and the resetting of these initiation signals are discussed in Section 7.3.

Containment isolation is accomplished in two phases. Initiation of a containment isolation Phase A (CIA) signal shuts most of the automatic isolation valves as shown in Table 6.2-60. Selected other critical lines such as component cooling water to the reactor coolant pumps (RCPs),

service water to the containment air recirculation coolers or control rod drive mechanism shroud coolers are considered important enough to delay until the next phase - CIB. Initiating signals for this phase closes all required automatic valves.

6.2-60

BVPS-2 UFSAR Rev. 26 Following a DBA, the ambient temperatures in the containment may rise as high as 333.3F. Although such high temperatures are short-lived, it is possible that water trapped between isolation barriers may expand due to the thermal effects, creating pressures greater than the piping design limit. Overpressure protection is provided by: 1) relief valves between the isolation valves with set points in excess of the containment design pressure by 50 percent (minimum) but at or below the piping design limit.

These valves are designed to reseat when the overpressure conditions subside; 2) a check valve located on a bypass line, around the isolation valve inside the containment, to relieve pressure between the isolation valves; 3) it is inherent in lines which utilize a weight-loaded or spring-loaded check valve as an isolation valve inside containment since a high pressure condition between the check valve and the outside containment isolation valve will open the check valve to relieve pressure; or 4) isolation barriers and associated piping arrangements that are conservatively designed.

Containment isolation valves and valve operators are chosen based on the following requirements: fluid type, closure time, actuator failure position, secondary mode of operation, and overall space envelope. As an example, lines which are part of an ESF are isolated by MOV rather then air-operated valves (AOVs). They also fail as-is. An AOV, however, provides a very rapid (10 seconds or less) closure time for lines which will not be used post-accident. The maximum closure time for any containment isolation valve is 60 seconds or less. Solenoid-operated valves are utilized in place of AOVs when greater reliability post-accident use and safe failure positions are required. No secondary mode of operation is provided. The containment purge system isolation valves are only open during plant shutdown. The valves, however, will close automatically within 10 seconds upon receipt of a high radiation signal (Section 9.4.7.3).

Mechanical redundancy is provided by designing two isolation barriers between the RCS or atmosphere inside containment and the atmosphere outside containment. Electrical redundancy, including control circuitry, is provided by two Class lE power sources. For example, in a containment isolation arrangement with an MOV inside containment and an MOV outside containment, each valve is powered from a different Class lE source. The electrical or control signal power received by the containment isolation valves (including limit switches, etc) is designated as a Class lE component, as discussed in Section 8.3.1, thereby ensuring containment integrity under the most severe anticipated environment. Environmental qualification of Class lE components is discussed in Section 3.11.

Table 6.2-60 lists each fluid system penetrating the containment structure and indicates the isolation criteria to which it conforms. The details of containment isolation arrangements which differ in some manner from the specific arrangements allowed by GDC 54, 55, 56, and 57 are indicated as follows:

1. Emergency core cooling system safety injection lines, The safety injection portion of the ECCS must be operable after a DBA to keep the reactor core covered with water. The RCS 6.2-61

BVPS-2 UFSAR Rev. 26 cold leg injection lines are either opened immediately or during the transfer to ECCS recirculation mode (short-term),

following a DBA. The RCS hot leg injection lines are opened for long-term recirculation. Section 6.3 describes ECCS operation in detail. Table 6.2-60 lists the ECCS containment penetrations which must be opened. All ECCS lines penetrating the containment, which must be open at any time following an accident, include a remotely operable MOV outside containment and a weight-loaded check valve inside containment. None of the MOVs receive a containment isolation signal to close since post-accident opening is required.

2. Recirculation spray pump suction lines.

The suction line and valve arrangement for the recirculation spray pumps are conservatively designed to prevent significant system leakage. The major portion of this piping is buried in the reinforced concrete base mat of the containment.

Approximately 10 inches of open-ended pipe is exposed above the containment sump bottom. The piping between the outside of the containment wall and the isolation valve (including the valve) is contained within a specially designed encapsulation. These isolation valves, one in each line, are normally open, remotely controlled, and motor-operated. These isolation valves do not receive an automatic safety signal for closure.

The isolation valves do receive a CIB signal to ensure an open position. This arrangement ensures that water can be taken from the containment sump for recirculation following a DBA in addition to providing the required containment isolation function. The valves can be remote manually closed by the operator, if necessary.

3. Containment vacuum pump and hydrogen recombiner suction lines.

These lines are normally provided with two remotely-controlled, solenoid-operated isolation valves in series outside containment (Section 6.2.5). The containment vacuum system valves are normally open, the valves of the hydrogen control system (HCS) are normally closed. The valves are located as close as possible to the containment wall. The piping is designed in accordance with the break/crack exclusion criteria set forth in Branch Technical Position MEB 3-1, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containments, and as described in Section 3.6B. Leakage detection is not required as the valves are hermetically sealed. The containment vacuum system containment isolation valves close automatically upon receipt of a CIA signal. The post-DBA HCS is an ESF and the system's normally closed containment isolation valves must be opened manually by the operator for long-term hydrogen control.

4. Containment depressurization.

6.2-62

BVPS-2 UFSAR Rev. 26 The containment depressurization systems must be capable of operating following a DBA to reduce containment pressure to near initial atmospheric conditions. The isolation valves in these systems open upon receipt of a CIB signal, if not already open.

Each of the quench spray pump discharge and recirculation spray pump discharge lines is provided with a normally open, remotely-controlled, motor-operated isolation valve outside containment and a weight-loaded check valve inside containment.

These containment isolation physical arrangements comply with GDC 56, and also allow the depressurization systems to perform their post-DBA design function.

5. Containment leakage monitoring system This system consists of four open-tap instrument lines. These lines are used for either normal operation or accident condition containment pressure monitoring. Each line is provided with a flow restriction orifice to limit the amount of release in case of line rupture, and a normally open, remotely-controlled, solenoid-operated isolation valve. The valves are operated by control switches in the main control room and have indicating lights and alarms. In addition, the transmitters used for sensing containment pressure are seismically qualified. This arrangement complies with Regulatory Guide 1.11, Instrument Lines Penetrating Primary Reactor Containment.

6. Fuel transfer tube.

The fuel transfer tube penetrates the containment and is used to transfer spent fuel between the reactor and the fuel pool (Section 9.1.2). The tube is provided with a blank flange with dual O-rings inside containment and a normally closed, manually-operated valve outside containment. This arrangement will be tested to Type B leak test requirements of 10 CFR 50, Appendix J (Section 6.2.6). In addition, this tube is enclosed in another penetration encapsulation which is provided with a leak detection drain just inside the containment wall.

7. Auxiliary feedwater.

Auxiliary feedwater is used for long term cooldown post-accident and is provided with electro-hydraulic hand control valves (HCV) outside containment and check valves inside containment. The outside containment valves do not receive an automatic closure signal but may be closed by the operator if necessary.

8. Personnel air lock and equipment hatch.

The personnel air lock (PAL) is supplied with doors both inside and outside containment. The doors are secured by a rotating locking ring and sealed with a double O-ring gasket system.

6.2-63

BVPS-2 UFSAR Rev. 26 Each door is equipped with two normally closed, manually operated pressure equalizing valves (one from each side of the door). Each door is also equipped with an 18" breech design manway which incorporates double O-ring seals that seal using a wedged rotating locking ring. All the valves are administratively closed. Leak test connections are supplied and the air lock is tested to Type B leak test requirements in accordance with 10 CFR 50, Appendix J (Section 6.2.6).

The equipment hatch is supplied with a bolted hatch cover which may only be opened from inside the containment. The equipment hatch seals are similar to those used for the personnel hatch.

Bolted to the equipment hatch is the emergency air lock (EAL),

which is equivalent in sealing, security, and testing to the personnel hatch. The bolted connection between the EAL and equipment hatch has seals and leak test connections. The EAL and equipment hatch are Type B leak-tested in accordance with 10 CFR 50, Appendix J. The EAL has a manually-operated, administratively closed, isolation valve inside and outside containment.

9. Reactor coolant pump seal injection.

Each seal injection line to the RCPs includes a remote manual MOV outside containment and a weight-loaded check valve inside containment. The valve outside containment does not receive an automatic closure signal since seal injection should be maintained to the RCPs as long as the pumps are operating.

When the pumps have stopped operating, the MOVs are closed, remote-manually, by the operator for the long term.

10. Hydrogen recombiner return lines.

The hydrogen recombiner return lines from each recombiner contain a motor-operated and a manual isolation valve in series outside containment. The manual valves are administratively controlled and remain in the shut position until the recombiners are required for operation. The piping is designed in accordance with the crack exclusion criteria set forth in Branch Technical Position MEB 3-1, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containments, and as described in Section 3.6B.2.1.2.2.

11. Reactor Vessel Level Instrumentation System.

The Reactor Vessel Level Instrumentation System (RVLIS) provides indication of the reactor vessel fluid level or relative void content. Connections at the reactor vessel head, hot legs A and B, and the seal table provide the RVLIS sensing points. Tubing from these connections runs to high volume sensors which isolate the reactor coolant system from the remainder of the RVLIS tubing. The remainder of RVLIS tubing is filled with deaerated demineralized water.

6.2-64

BVPS-2 UFSAR Rev. 26 Capillary tubing runs from the high volume sensors, penetrates the containment, and runs to the hydraulic isolators. Tubing then connects the hydraulic isolators to differential pressure transmitters.

Containment isolation is provided by the hydraulic isolator and connecting capillary tubing. The hydraulic isolator located outside containment serves as an air isolation valve on the containment building.

Within the high volume sensor is a check valve which will close under reactor coolant system pressure if the connecting capillary tubing fails.

6.2.4.3 Safety Evaluation The design pressure of piping and components within the isolation boundaries afforded by the CIS are equal to, or greater than, the design pressure of the containment structure. Piping, valves, and penetrations of the CIS are designed, constructed, and installed in accordance with QA Category I, Seismic Category I, and ASME Section III, Class 2.

Systems penetrating the reactor containment are designated as "essential" or "nonessential," as shown in Table 6.2-60. The designation of a system as "essential" is based on the requirement that it remains operable during and after a design basis accident to mitigate the effects of the accident.

All "nonessential" systems are automatically isolated, as delineated in Table 6.2-60.

Essential lines penetrating containment that utilize single or double isolation valves outside containment employ conservative piping and valve design in accordance with SRP 3.6.2.

System design also includes check valves to preclude flow from containment to atmosphere from leakage outside containment. Leakage detection is also facilitated by monitoring of system parameters indicative of moderate system leakage, i.e., flow, pressure, and level indication. A discussion of the specific system parameters monitored for each listed essential system can be found in these sections:

Quench and Recirculation Spray Systems Section 6.2.2.5 Auxiliary Feedwater System Section 10.4.9.5 Main Steam System Section 10.3.7 Charging and Volume Control System Section 6.3 Combustible Gas Control System Section 6.2.5.5 Safety Injection System Section 6.3 Leakage Monitoring System Section 6.2.4.2 The ESF lines containing isolation valves that are open following a DBA are monitored so that containment integrity is ensured following DBA conditions. This monitoring is provided in the form of safety-related sump level indication (outside containment) as described in Section 7.3.

A rise in sump level would indicate a possible problem in a functioning 6.2-65

BVPS-2 UFSAR Rev. 26 ESF line(s), which would require isolation, at least until the source of water was identified.

Containment isolation valve operability and containment isolation barrier leakage rate testing are described in Section 6.2.6.

A FMEA to determine if the I&C and electrical portions meet the single failure criterion has been performed on the CIS. The results of this analysis can be found in the separate FMEA document (Section 1.7).

6.2.4.4 Inspection and Testing Requirements Prior to initial unit operation and during each major refueling period, as specified in 10 CFR 50, Appendix J, tests are conducted to ascertain the leaktightness and operability of the CIS.

Section 6.2.6 provides a description of the Type B and Type C test programs. In addition, fail-safe positions, sensor accuracy and set points, and response time are all tested and verified to be in accordance with technical requirements. The inspection and testing requirements of the ESF actuation system is discussed in Section 7.1. The mechanical components and pressure boundaries of all fluid system containment penetrations will be tested and inspected in accordance with the ASME OM Code and ASME Section XI.

6.2.5 Combustible Gas Control in Containment The combustible gas control system (CGCS) mixes and monitors the hydrogen concentration within the containment and maintains this concentration at a safe level following DBA.

The CGCS, which includes subsystems capable of monitoring the combustible gas concentrations within containment regions and reducing this combustible gas concentration, is shown on Figure 6.2-131.

Mixing of hydrogen in the containment following a postulated LOCA results from three mechanisms: 1) momentum transfer from the fluid jet exiting the break; 2) forced and natural convection flows within the containment atmosphere; and 3) molecular diffusion. All of these mechanisms will work together to enhance mixing within the containment to provide a homogeneous gas mixture and prevent local accumulation of hydrogen.

Good containment compartment mixing will occur during the blowdown period of the postulated LOCA due to the break effluent. The momentum of the jet from the break will cause turbulent mixing within the containment. This was demonstrated in a test performed for a high velocity jet source (Bloom 1982). Results from this test showed that "when the jet was initiated, local gas velocities, even far from the source, increased by a factor of three to five times over background velocities caused by natural convection and fan-induced recirculation." Although this test was performed for an ice condenser lower compartment geometry, the test results would be applicable to subcompartments (e.g., steam generator cubicle, pressurizer cubicle) which are open to the containment.

6.2-66

BVPS-2 UFSAR Rev. 26 Forced convection in the containment atmosphere will be generated by the containment spray systems which are designed to cool the containment atmosphere (Section 6.2.2). Approximately 3,500 gpm long-term recirculation spray flow rate (assuming minimum safeguards) is provided.

The spray will induce mixing by imparting momentum to the containment atmosphere. Air entrainment by the spray causes bulk mass motion which creates both large- and small-scale turbulence. Therefore, complete mixing should occur within a few minutes following a LOCA with containment spray operation (Sandia 1983).

In addition, steam condensation and cooling of the containment atmosphere by the sprays will result in flow to low pressure regions. This does not result in significant mixing within individual compartments, although significant intercompartment fluid transfer can occur (IDCOR 1983).

Natural convection due to density differences (buoyant effects) is another source which will cause mixing to occur in the containment atmosphere.

Gas flow occurs whenever there is a temperature difference between the wall and the bulk atmosphere. Gases heated or cooled by the walls will rise or fall, respectively, due to the density differences between the gas and the surrounding atmosphere. This buoyant force imparts momentum to the gas and significant turbulent mixing will result.

The presence of large heat sinks in the containment, such as internal walls, together with localized heat sources, such as hot equipment surfaces, will be expected to set up large-scale natural circulation cells. These circulation cells will help decrease any stratification that may occur in areas with the absence of jet induced or forced convection flows. Tests conducted during the containment systems experiment (CSE) program in a steam/air atmosphere indicated that natural convection caused good mixing in a large vessel (Hillard and Coleman 1970, Knudsen and Hillard 1969).

After completion of the blowdown period of the postulated LOCA, natural convection flows within the containment atmosphere will also be developed due to the break effluent. Cooling water is injected into the reactor core by the ECCS (Section 6.3). The injected water will exit the break as steam/water mixture. Buoyancy forces will cause the released steam to rise. This upward steam flow will generate containment mixing due to the entrainment of the atmosphere gases in the steam plume. The extent of mixing in areas away from the break due to the buoyant thermal plume discharging into the containment is a function of geometry, plume to atmosphere density ratio, and ratio of momentum to buoyancy forces (IDCOR 1983).

Molecular diffusion is another mechanism which would provide mixing within the containment following a postulated LOCA. Diffusion occurs due to concentration gradients. The rate of diffusion is too slow to expect mixing of large containment volumes in short times by itself, although molecular diffusion would add to the other mixing processes previously discussed.

6.2-67

BVPS-2 UFSAR Rev. 26 The containment internal structures (Section 3.8.1) are designed to be as open as practical to allow the circulation and mixing mechanisms to function. The volume above the operating floor, which comprises the majority of the containment volume, does not have significant barriers to obstruct mixing from the various mechanisms.

The steam generator and pressurizer subcompartments, the annulus between the crane wall and containment wall, and the hoisting spaces are open at the top (near the top for the upper pressurizer cubicle) and bottom and connect with each other at various elevations (Figure 6.2-138). Extensive use is made of gratings at intermediate levels within the compartments.

The quench and recirculation spray nozzles are located and oriented to cover as much area as possible. This design arrangement enhances mixing by establishing air movements and flow paths.

Containment compartments have direct circulation between each other as follows:

1. Dome - Fully open to operating floor and annulus

2. Operating Floor - Open to annulus, dome, steam generators (1, 2, and 3), pressurizer, RHR, hoist space, reactor vessel annulus, and upper pressurizer

3. Annulus - Open to dome, operating floor, steam generator 1, pressurizer, RHR hoist, RHR, hoist space, and basement

4. Steam Generator 1 - Open to operating floor, annulus, RHR hoist, RHR, and basement

5. Steam Generator 2 - Open to operating floor, pressurizer, hoist space, and basement

6. Steam Generator 3 - Open to operating floor, pressurizer, RHR hoist, RHR, and basement

7. Pressurizer - Open to operating floor, annulus, steam generator 2, steam generator 3, and upper pressurizer

8. Upper pressurizer - Open to pressurizer and operating floor

9. RHR Hoist - Open to annulus, steam generator 1, and steam generator 3

10. RHR - Open to annulus, steam generator 1, steam generator 3, and instrumentation tunnel

11. Hoist Space - Open to operating floor, annulus, steam generator 2, and basement

12. Basement - Open to annulus, steam generator 1, steam generator 2, steam generator 3, RHR, and hoist space 6.2-68

BVPS-2 UFSAR Rev. 26

13. Incore Instrumentation Tunnel - Open to reactor vessel annulus and open to the basement through ventilation ducts

14. Reactor Vessel Annulus - Open to incore instrumentation tunnel and open to operating floor through reactor cavity water seal access ports As discussed above, each compartment has openings to allow free circulation and there are no dead end compartments. Since there is a direct or indirect circulation between each compartment, an effective mixing will be achieved. Figure 6.2-139 shows the expected long-term circulation patterns within the containment that are caused by recirculation spray.

In summary, the design of the internal containment structure allows free circulation and mixing of gases through numerous large openings in the ceilings, floors, and walls of each compartment while the spray system enhances the circulation process throughout the containment.

Hydrogen generation from oxidation of zircaloy fuel cladding, radiolysis of the water in the core, and hydrogen present in the reactor coolant system would be released through the break opening to the containment.

Local accumulation of hydrogen within the compartment where the break occurred due to these sources would not occur. This is due to the combined action of the mixing mechanisms and the internal design of the containment structures which allows air to circulate freely. Hydrogen generation from the radiolysis of water in the sump and corrosion of metals (i.e., aluminum and zinc) by the containment spray would be generated over long periods of time. Because of the slow rates of release, diffusion and convection mechanisms will be more than enough to prevent hydrogen accumulation (IDCOR 1983).

It can be concluded that following a postulated LOCA with spray operation, good mixing will occur within the containment to prevent the localized accumulation of hydrogen.

6.2.5.1 Design Bases The design of the CGCS is in accordance with the following criteria:

1. General Design Criterion 41, as it relates to the control of hydrogen concentration in the containment atmosphere following an accident.

2. General Design Criterion 42, as it relates to periodic inspection of the containment cleanup system to assure its integrity and capability.

3. General Design Criterion 43, as it relates to periodic testing of the atmosphere cleanup system.

4. Deleted 6.2-69

BVPS-2 UFSAR Rev. 26

5. Regulatory Guide 1.26, as it relates to the system being designed to Safety Class 2 and ASME Section III, Class 2, standards, except the purge system which is non-nuclear safety (NNS) class and nonseismic and process sample tubing connected to the hydrogen analyzers which is Quality Assurance Category I, Seismic Category I, Safety Class 2.

6. Regulatory Guide 1.29, with respect to the CGCS being Seismic Category I except for the purge system. The CGCS is protected from the effects of tornadoes, external missiles, pipe ruptures, pipe whip, and jet impingement forces.

7. Regulatory Guide 1.97, as it relates to the capability of the instrumentation to monitor and sample the combustible gas concentrations within the containment during normal and post-accident conditions.

8. Deleted

9. The containment atmosphere is maintained uniformly mixed by the containment spray to prevent excessive stratification of combustible gases (Kundson and Hilliard 1969; Hilliard et al 1970).

10. Deleted

11. The CGCS is designed for all environmental conditions (Section 3.11).

12. Deleted

13. The CGCS meets the intent of NUREG-0737, Action Item II.E.4.1, that requires dedicated hydrogen penetrations for recombiners to be located external to the containment.

14. The CGCS meets the intent of NUREG-0737, action item II.F.1.6, concerning the monitoring of hydrogen concentration of the containment atmosphere.

15. The QA Category I portions of the CGCS suction and discharge piping are exempt from consideration of passive failures in accordance with NUREG 0800, Section 6.2.5 "Combustible Gas Control in Containment."

6.2.5.2 System Description The CGCS is shown in a simplified manner on Figure 6.2-131.

The containment penetration design is discussed in Section 6.2.4.

The two hydrogen analyzers shown on Figure 6.2-131 are located in the cable vault area. A remote control panel located in the service building is provided with each analyzer. The analyzers may be controlled locally at the remote control panel, but this control will be overridden manually 6.2-70

BVPS-2 UFSAR Rev. 26 from a control switch located on the main control board or automatically by a safety injection signal. Opening of the associated containment isolation valves is controlled in the same manner. Indicators and a recorder (one channel only) mounted on the main control board provide continuous monitoring of containment hydrogen concentration when the analyzers are operating. Annunciation is provided in the main control room when either or both analyzers are unavailable; in addition, the unavailability is monitored by the main plant computer system.

Containment hydrogen concentration indication is also provided on the remote control panel.

6.2.5.3 Design Evaluation The five major sources of hydrogen generation assumed following a DBA are:

1. Metal-water reaction involving the zirconium fuel cladding and the reactor coolant.

The reaction of the zirconium clad with water is assumed to occur instantaneously after a DBA and the hydrogen evolved is added to the containment atmosphere immediately. Section 15.6.5 discusses the calculated amount of hydrogen generated from the chemical reaction.

2. Pressurizer gas space and RCS water.

The hydrogen present in the pressurizer gas space and reactor coolant is assumed to be released to the containment atmosphere instantaneously after a DBA.

3. Radiolytic generation in the water collected on the containment floor.

Radiolytic decomposition of water is given by the following chemical equation:

2H2O + Energy 2H2 + O2 The energy for this reaction is supplied by the decay of fission products which have escaped from the reactor core and are dissolved in the sump water. The fission product distribution and energy absorption used in the analysis are in conformance with Regulatory Guide 1.7.

4. Radiolytic generation in the reactor core.

The chemical reaction presented for sump radiolysis is also applicable to radiolytic decomposition of water in the reactor core. In this case, the energy is supplied by the decay of fission products in the fuel.

Sump and core radiolysis is based on American Nuclear Society (ANS) draft standard ANSI 56.1, Combustible Gas Control dated June, 1976.

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5. Corrosion of metals by solutions used for containment spray.

The use of aluminum and zinc inside the containment is minimized. These materials are not utilized in the manufacture of safety-related components which are in contact with the core cooling fluid.

Aluminum and zinc bearing materials, such as galvanized steel, within the containment may react with the containment spray and produce hydrogen. The inventories of aluminum and zinc inside containment are documented and administratively controlled by plant procedures.

The volume control tank (VCT) is automatically isolated from the charging pumps by redundant valves on receipt of a safety injection signal. The valves close within 10 seconds after receipt of the signal. The liquid inventory in the tank is sufficient to prevent hydrogen from reaching the charging pump suction before isolation of the tank by virtue of a water seal. Thus, the VCT hydrogen is unable to reach the containment after a LOCA and is not considered a source of hydrogen in the analysis.

The total amount of hydrogen in the containment is calculated by summing the hydrogen produced by each of the five hydrogen generation sources.

The partial pressure of the hydrogen is calculated assuming a perfect gas.

The total pressure in the containment is then calculated by summing the partial pressures of the vapor and the noncondensables. The volume percent of the hydrogen is calculated from the ratio of the partial pressure of the hydrogen to the total containment pressure.

The internal design of the containment structures allows air to circulate freely. All cubicles and compartments within the containment are provided with openings near the top as well as openings in the floor to allow air circulation. Convective mixing in conjunction with containment spray assures a uniform mixture of hydrogen in the containment.

Containment system experiment tests (Knudsen and Hilliard 1969; Hilliard et al 1970) have verified that adequate mixing of the containment atmosphere is achieved by the CSS. The design of the BVPS-2 containment is similar to those of the Surry Power Station, Unit 1 and Unit 2, for which it has been concluded by the U.S. Atomic Energy Commission (USAEC 1972b) that there is adequate mixing of hydrogen in the post-LOCA environment.

The inlet isolation valve (2HCS*MOV110A) for the containment atmosphere purge subsystem has its electrical power supply deenergized during normal and post-accident operation to prevent a spurious signal from affecting the hydrogen control system safety function. Section 8.3.1.2.1 provides details on the method.

A FMEA to determine if the I&C and electrical portions meet the single failure criterion, and to demonstrate and verify how the GDC and IEEE Standard 279-1971 requirements are satisfied, has been performed on the 6.2-72

BVPS-2 UFSAR Rev. 26 CGCS. The FMEA methodology is discussed in Section 7.3,2. The results of this analysis can be found in the separate FMEA document (Section 1.7).

6.2.5.4 Inspection and Testing Requirements Chapter 14 provides a detailed discussion on preoperational testing performed on the CGCS. In-service periodic inspection and system pressure tests are discussed in Section 6.6.

6.2.5.5 Instrumentation Requirements Control switches are provided locally on the hydrogen recombiner panels in the safeguards area for the positive displacement blower.

Controls are provided at the control panel in the service building (near the hydrogen analyzers) for all automatic valves associated with the purge blower.

Annunciation is provided in the main control room for hydrogen control system local panel trouble. The input to this annunciator from the local panels is also monitored by the BVPS-2 computer system. Indicators are provided in the main control room to monitor hydrogen gas concentrations.

A recorder for hydrogen gas concentrations. A recorder for hydrogen gas concentration (channel A only) is provided.

The following controls and instruments are located on the hydrogen analyzer panel: a stream selector switch for stream to be analyzed, indicating lights for reference/zero gas pressure, or calibration/sample gas pressure low alarm, and high gas concentration.

6.2.6 Containment Leakage Testing The containment leakage rate tests are performed in accordance with 10 CFR 50, Appendix J, Option B, and GDC 52, 53, and 54.

The purpose of the containment leakage test program is to assure that leakage through the reactor containment, systems, and components penetrating the containment boundary does not exceed the allowable leakage rate values as specified in the Technical Specifications (Chapter 16) or other design base documents.

The containment leak testing program includes the performance of Type A tests to measure the containment overall integrated leak rate; Type B tests to detect local resilient seal leakage at electrical penetrations, equipment hatch, personnel hatch, emergency escape trunk, and fuel transfer tube flange; and Type C tests to measure containment isolation valve leakage rates.

6.2.6.1 Containment Integrated Leak Rate Test - Type A The periodic Type A leakage rate test will be conducted in accordance with 10 CFR 50, Appendix J, Option B. Pretest requirements will be identified and included as part of the Type A test procedure to ensure that the 6.2-73

BVPS-2 UFSAR Rev. 26 necessary preparations, precautions, and temporary modifications have been completed prior to Type A test commencement. Such pretest requirements will include unit status, instrumentation requirements, support systems status, temporary test or measurement equipment requirements, supplementary testing requirements, general containment inspection requirements prior to containment closeout, personnel assignment, shift briefings, etc.

In accordance with the Containment Leakage Rate Testing Program (CLRTP), a general inspection of the accessible interior and exterior surfaces of the containment structure will be performed for the purpose of identifying evidence of deterioration which may affect the containment structural integrity or leaktightness. Visual inspection will be performed to detect and observe: gross deformations of the interior surfaces of steel containment liner; paint failure due to massive rusting, electrolysis, or abrasion; evidence of exterior concrete spalling or cracking; high stress areas of the containment concrete such as equipment hatch, personnel hatch, electrical and valve penetration areas; accessible areas at the bend line; shake space integrity, etc. Should evidence of containment degradation be found, the Type A or structural acceptance test will not be performed until an evaluation has been performed and repairs made, if required. Such structural deterioration and subsequent corrective actions taken will be reported in accordance with the CLRTP.

System Venting and Draining To place the primary reactor containment system as close to post-accident conditions as possible, those portions of the fluid systems that are part of the reactor containment boundary that may be opened directly to the containment or outside atmosphere under post-accident conditions will be opened or vented to the appropriate atmosphere during the test.

Those lines which are normally fluid-filled and which may be drained or have the fluid driven off by the accident, including portions of systems inside or outside containment that penetrate the containment and may rupture as a result of a LOCA, will be drained to the extent necessary to expose the containment isolation valve seats to the containment atmosphere, except as noted by the following. Systems that are required for proper conduct of the test or to maintain BVPS-2 in a safe condition during the test shall be operable in their normal mode and need not be vented or drained. Additionally, systems that are normally filled with water and operable under post-accident conditions, such as the CHRS, need not be vented or drained. The CLRTP may provide additional exceptions for not venting or draining penetrations. Systems that are not vented or drained during the Type A test and which could become exposed to the containment atmosphere during a leakage DBA will be Type C tested, and the Type C test leakage rate for the penetration path will be added to the upper confidence limit.

The test pressure to which the containment is subjected during the Type A test is equivalent to the calculated peak containment pressure following the design basis accident. Temporary air compressors will be utilized to raise containment pressure. When the containment has reached test pressure, containment temperature will be monitored for a period of not 6.2-74

BVPS-2 UFSAR Rev. 26 less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> until stabilization criteria have been met. Once stabilized, the containment parameters of temperature, pressure, and vapor pressure will be observed and recorded for the duration of testing. The duration of the test period will be sufficient to enable adequate data to be accumulated and analyzed so that a leakage rate and upper confidence unit can be accurately determined. During this period, the containment leak rate will be calculated by the mass point or total time analysis technique to verify that it is within the limits of the BVPS-2 Technical Specifications requirements. Upon determination of an acceptable leakage rate, a verification test will be performed to confirm the capability of the method and the test instrumentation used to determine the containment leakage rate. Having met all test criteria, the containment will be vented and reduced to atmospheric conditions.

The as left acceptance criteria for an acceptable leakage rate test requires that containment leakage be less than 0.75 La, as defined by the CLRTP. A superimposed leak test will be conducted immediately following the Type A test. The results from this test will be considered acceptable provided the superimposed leak test data are within the acceptance criteria specified in ANSI/ANS 56.8 2002.

6.2.6.2 Containment Penetration Leakage Rate Test - Type B Type B containment penetration leakage tests are conducted in accordance with the CLRTP. Type B leakage tests are intended to detect local leakage and to measure leakage across containment electrical penetrations, equipment and personnel hatches, emergency escape trunk, and fuel transfer tube flange. A list identifying all containment penetrations is provided in Table 6.2-60.

The makeup air method of testing, which will primarily be used to measure Type B leakage, consists of the pressurization of a component with air or nitrogen and measuring leakage using a flowmeter installed in the pressurization line.

The test pressure to which Type B tests will be conducted is identical to that specified in Section 6.2.6.1 for Type A testing.

The periodic retest schedule for Type B testing will be in accordance with the CLRTP.

6.2.6.3 Containment Isolation Valve Leak Rate Tests - Type C Type C testing is performed on containment isolation valves to verify their sealing capability and leaktightness. All testing will be performed in accordance with the requirements of the CLRTP.

Type C tests will be performed by local pressurization applied in the same direction as that when the valve would be required to perform its safety function, unless it can be demonstrated that testing in a reverse direction is as conservative. Each valve to be tested will be closed by its normal means, that is, motor, solenoid, diaphragm, handwheel, etc, and will receive no additional adjustments (hand-tightening after closure by motor) or preliminary exercising.

6.2-75

BVPS-2 UFSAR Rev. 26 The containment isolation valves will be tested by local pressurization to the pressure specified in Section 6.2.6.1 for the Type A test. The test method will be to vent and drain a system, or portions thereof, and to pressurize across one, or a series of valves with air or nitrogen using primarily the makeup air method described in Section 6.2.6.2. Test connections located on both the inlet and outlet sides of a valve, or pair of valves, are provided to facilitate system draining and/or pressurization. Leakage will be measured using an installed flow meter in the pressure supply line. On multiple valve penetrations, only the highest leaking valve shall be recorded as the as left penetration leak rate. Valves, and their respective system status which must be Type C tested, are listed in Table 6.2-60. Test vents, drains, and connections located between isolation valves will have two barriers (valve with cap, and valve with flange) and will be administratively controlled. These connections will not be leak tested.

The test pressure will be as specified in Section 6.2.6.1 for Type A testing.

The acceptance criteria for allowable leakage associated with Type B and Type C combined leakages is to be in accordance with the CLRTP.

Scheduling for each periodic Type C test will be in accordance with the CLRTP.

6.2.6.4 Scheduling and Reporting of Periodic Tests The schedules for periodic tests are in accordance with the CLRTP. Report preparation for periodic Type A, B, and C testing will be in accordance with the CLRTP.

6.2.7 Fracture Prevention of Containment Pressure Boundary Materials A summary of the fracture toughness characteristics of the containment pressure boundary materials and the confirmation of compliance to GDC 51 can be found in the DLC transmittal to the NRC (Woolever 1983).

6.2.8 References for Section 6.2 Aerojet Nuclear Company (ANC) 1976. RELAP4/MOD 5: A Computer Program for Transient Thermal Hydraulic Analysis of Nuclear Reactors and Related Systems. User's Manual Vol I-III, Report ANCR-NUREG-1335.

American National Standards for Containment System Leakage Testing Requirements. ANSI/ANS-56.8-2002. (This document used only as a guideline.)

American Nuclear Society (ANS) 1978. Decay Heat Power in Light Water Reactors. ANS Proposed Standard 5.1, Revised September 1978.

Anderson, T.M. (Westinghouse) 1979. Personal Communication (Letter NS-TMA-2075 dated April 25, 1979) to J.F. Stolz, USNRC. Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version.

6.2-76

BVPS-2 UFSAR Rev. 26 ANSI/ANS-5.1 1979, "American National Standard for Decay Heat Power in Light Water Reactors," August 1979.

Bloom, G. R., et al. 1982. Hydrogen Distribution in a Containment with a High Velocity Hydrogen-Steam Source, presented at the Second International Workshop on the Impact of Hydrogen on Water Reactor Safety, Albuquerque, New Mexico.

Bordelon, F.M., Massie, H.W., Sr., Zordan, T.A. 1974a. Westinghouse Emergency Core Cooling Evaluation Model Summary. WCAP-8339.

Bordelon, F.M., et al 1947b. SATAN-VI Program: Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant, WCAP-8302 (Proprietary) and WCAP-8306.

Boyle, J.C. 1975; Meyer, S.P. 1981. Subcompartment Transient Response Code THREED. NU-092 (Proprietary).

Burnett, T. W. T, McIntyre, C. J., and Buker, J. C. 1984, LOFTRAN Code Description, WCAP-7907-P-A (Proprietary) and WCAP-7907-A (Nonproprietary).

Clout, R.L. 1973. Pipe Breaks for the LOCA Analysis of Westinghouse Primary Coolant Loop. WCAP-8082 (Proprietary) and WCAP-8172.

Collier G., et al 1974. Calculational Model for Core Reflooding After a Loss-of-Coolant Accident (WREFLOOD Code). WCAP-8170.

Docket No. 50-315, Amendment No. 126 Facility Operating License No. DPR-58 (TAC No. 7106) for D. C. Cook Nuclear Plant Unit 1, June 9, 1989.

Docket No. 50-472, Amendment No. 153 Facility Operating License No. NPF-73 for BVPS-2, February 2006.

EPRI 294-2, Mixing of Emergency Core Cooling Water with Steam: 1/3 Scale Test and Summary, (WCAP-8423), Final Report, June 1975.

Hilliard, R. K. and Coleman, L. F. 1970. Natural Transport Effects on Fission Product Behavior in the Containment Systems Experiment, BNWL-1474, Battelle Pacific Northwest Laboratories, Richland, Washington.

Hilliard, R. K., Coleman, L.F; Linderoth, C.E; McCormack, J.D; and Pastma, A.K. 1970. Removal of Iodine and Particles from Containment Atmosphere by Sprays. Battelle - Northwest, Richland, Wash., BNWL-1244.

Idaho National Engineering Laboratory 1975. CONTEMPT LT/026. A Computer Code for Predicting the Containment Pressure Temperature Response to a Loss-of-Coolant Accident.

IDCOR 1983. IDCOR Program Report, Technical Report 12.2, Hydrogen Distribution in Reactor Containment Building.

6.2-77

BVPS-2 UFSAR Rev. 26 Idel'chik, I.E. 1966. Handbook of Hydraulic Resistances. U.S. Department of Commerce, National Bureau of Standards, Institute for Applied Technology, U.S. Atomic Energy Commission Report AEC-TR-6630.

Knudsen, J. G. and Hilliard, R. K. 1969. Fission Product Transport by Natural Processes in Containment Vessels. Battelle - Northwest, Richland, Wash., BNWL-943.

Land, R. E. 1976, Mass and Energy Releases Following a Steam Line Rupture, WCAP-8822 (Proprietary) and WCAP-8860 (Nonproprietary).

Los Alamos Scientific Laboratory (LASL) 1979. Subcompartment Analysis Procedures Report. NUREG/CR-1199, LA-8169-MS.

McAdams, W.H. 1954. Heat Transmission. Third Edition, p. 44.

Moody, L.J. 1965. Maximum Flow Rate of a Single Component, Two-Phase Mixture. Journal of Heat Transfer Transactions, ASME Vol. 87, p 134-142.

Moore, K.V. and Rettig, W.H. 1974. RELAP4 - A Computer Program for Thermal Hydraulic Analysis. Aerojet Nuclear Company Report ANCR-1127.

Norberg, J.A.; Bingham, G.E.; Schmidt, R.C.; Waddops, D.A. 1969.

Simulated Design Basis Accident Tests of the Carolinas Virginia Tube Reactor Containment - Preliminary Results. IN-1324, Idaho Nuclear Corporation.

Osborne, M. P., and Love, D. S. 1985, Mass and Energy Releases Following a Steam Line Rupture, Supplement 1 - Calculations of Steam Superheat in Mass/Energy Releases Following a Steam Line Rupture, WCAP-8822-S1-P-A (Proprietary) and WCAP-8860-S1-A (Nonproprietary).

Parsly, L.F. 1970. Design Considerations of Reactor Containment Spray Systems - Part VI, The Heating of Spray Drops in Air/Steam Atmosphere.

ORNL-TM-2412.

Sandia National Laboratories and General Physics 1983. NUREG/CR-2726, SAND82-1137, R3, Light Water Reactor Hydrogen Manual.

Schmidt, R.C.; Bingham, G.E.; Norberg, J.A.; 1970. Simulated Design Basis Accident Tests of the Carolina Virginia Tube Reactor Containment - Final Report. UC-80, Idaho Nuclear Corporation.

Shepard R.M.; Massie, H.W.; Mark, R.H.; and Docherty, P.J. 1975.

Westinghouse Mass and Energy Release Data for Containment Design. WCAP-8312-A, Revision 2, WCAP-8264-P-A (Proprietary).

Slaughterbeck, D.C. 1970. A Review of Heat Transfer Coefficients for Condensing Steam in a Containment Building Following a Loss-of-Coolant Accident. Interim Task Report, Subtask 4.2.2.1, Idaho Nuclear Corporation.

Spray Engineering Company (undated). Spray Analysis on SPRACO Model 1713A Nozzles. Nashua, N.H.

6.2-78

BVPS-2 UFSAR Rev. 26 Stone & Webster Engineering Corporation (SWEC) 1971. LOCTIC - A Computer Code to Determine the Pressure and Temperature Response of Dry Containment to a Loss-of-Coolant Accident. SWND-1, Letter from W.J.L. Kennedy to P.A.

Monis, et al, Boston, Mass, 1967.

Uchida, H.; Oyama A.; and Togo, Y. 1964. Evaluation of Post-Incident Cooling Systems of Light-Water Power Reactors. Proceedings of the Third International Conference on the Peaceful Uses of Atomic Energy, Geneva, Switzerland, August 31-September 9, 1964, Volume 13, New York United Nations93-104.

U.S. Atomic Energy Commission (USAEC) 1970. Safety Evaluation by the Division of Reactor Licensing in the Matter of Virginia Electric and Power Company. North Anna Power Station Units 1 and 2, Docket Nos. 50-338 and 50-339.

U.S. Atomic Energy Commission 1972a. Safety Evaluation by the Division of Reactor Licensing in the Matter of Virginia Electric and Power Company.

Surry Power Station Units 1 and 2, Docket Nos. 50-280 and 50-281.

U.S. Atomic Energy Commission 1972b. Safety Evaluation by the Division of Reactor Licensing in the Matter of Virginia Electric and Power Company.

Maine Yankee Atomic Power Station, Docket No. 50-309.

U.S. Atomic Energy Commission 1974a. Evaluation Report by the Directorate of Licensing in the Matter of Duquesne Light Company, Toledo Edison Company, Pennsylvania Power Company. Beaver Valley Power Station Unit 1, Docket No. 50-334.

U.S. Atomic Energy Commission 1974b. Supplement No. 1 to Safety Evaluation Report by the Directorate of Licensing in the Matter of the Millstone Point Company, et al. Millstone Nuclear Power Station Unit 3, Docket No. 50-423.

U.S. Nuclear Regulatory Commission 1976b. Safety Evaluation Report Related to the Preliminary Design of the SWESSAR-PI Standard PWR Reference Nuclear Power Plant and Its Relationship to the RESAR-41 Standard Reference System. Docket No. STN 50-495.

U.S. Nuclear Regulatory Commission 1977. Safety Evaluation Report Related to the Preliminary Design of the SWESSAR-PI Standard PWR Reference Nuclear Power Plant and Its Relationship to the RESAR-3S Standard Reference System. Docket No. STN 50-495.

WCAP-10325-P-A (Proprietary), WCAP 10326-A (Non-Proprietary),

Westinghouse LOCA Mass & Energy Release Model for Containment Design, March 1979 Version, May 1983.

Westinghouse 1978. ECCS Evaluation Model, February 1978 Version WCAP-9220 (Proprietary).

Woolever, E. J. 1983. Letter from E. J. Woolever, DLC, to H. R. Denton, NRC, November 14, 1983.

6.2-79

BVPS-2 UFSAR Tables for Section 6.2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-1 THERMOPHYSICAL PROPERTIES OF PASSIVE HEAT SINK MATERIALS Specific Thermal Heat Conductivity Capacity Density Material (Btu/hr/ft/F) (Btu/lbm/F) (lbm/ft3)

Concrete 0.8 0.21 144 Stainless 11 0.11 491 steel Carbon steel 31 0.11 489 Paint is assumed to absorb no heat and to merely present a resistance to heat transfer with a heat conductance in the range of 42 to 252 Btu/hr/ft2/F in concrete and 266 to 2000 Btu/hr/ ft 2/F for steel.

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BVPS-2 UFSAR Rev. 16 TABLE 6.2-2 BEAVER VALLEY MAAP-DBA PARAMETER FILE

SUMMARY

OF CONTAINMENT NOMINAL VOLUMES AND METAL HEAT SINKS Metal Heat Sinks Surface Net Mass Area Free Volume (ft3) (lbm) (ft2) 1 Reactor cavity 11,826 425,023 2,975 2 Lower compartment 200,063 655,459 60,526 3 Instrument room 30,872 1,094 216 4 RHR platform 31,264 17,446 431 5 Loop C compartment 52,311 264,057 8,744 6 PZR compartment 48,637 50,933 2,431 7 Loop B compartment 49,141 267,633 11,410 8 RV head laydown area 45,542 17,075 2,898 9 Loop A compartment 51,429 284,087 11,462 10 Lower annulus north half 85,457 299,948 41,642 11 Lower annulus south half 85,663 280,947 37,869 12 Upper annulus north half 80,082 148,581 24,295 13 Upper annulus south half 80,294 224,330 23,716 14 Refueling cavity 36,620 131,960 5,522 15 Upper compartment 347,071 481,486 15,340 cylindrical section 16 Upper compartment lower 413,523 583,731 34,062 dome region 17 Upper compartment upper 108,635 0 0 dome region TOTAL 1,758,480 4,133,790 283,539 NOTES:

1 Metal heat sinks do not include major equipment, such as steam generators or RCS loop piping. Realistic heat sink values without any uncertainty included.

2 The containment steel liner mass is included with concrete heat sinks, therefore the liner mass is not reflected in the metal heat sink summary.

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BVPS-2 UFSAR Rev. 16 Table 6.2-2A Beaver Valley MAAP-DBA Parameter File Summary of Containment Concrete Heat Sinks Total No. Sides Heat Thickness One-Sided Inside Total Sink# Description ft Area ft2 Ctmt Area ft2 1 Shield wall to lower 4.50 1,524 2 3,049 2 Refuel cavity wall to loop 4.00 958 2 1,915 B

3 Shield wall to loop C 4.50 195 2 389 4 Refuel cavity wall to PZR 4.00 924 2 1,848 5 Shield wall to loop B 4.50 211 2 421 6 Shield wall to RV laydown 4.50 62 2 124 7 Shield wall to loop A 4.50 155 2 309 8 Instrument tunnel to lower 3.00 954 2 1,908 9 Refuel cavity wall to loop 4.00 826 2 1,653 C

10 Reactor cavity floor (1) 10.00 621 1 621 11 14 crane wall support 2.00 2,352 2 4,704 columns 12 Lower compartment floor 10.00 10,094 1 10,094 13 Lower compartment outer 4.50 8,445 1 8,445 wall(1) 14 Instrument room floor 4.00 1,017 2 2,034 15 Instrument room wall to 3.25 929 2 1,858 loop C 16 Instrument room wall to 3.25 704 2 1,408 loop A 17 Instrument room crane wall 2.00 1,545 2 3,091 18 Instrument room ceiling 2.00 923 2 1,846 19 Loop C floor 4.50 912 2 1,824 20 Loop C wall to PZR 3.00 1,043 2 2,085 21 Loop C crane wall 2.75 2,290 2 4,579 22 SG cubicle support columns 3.50 633 2 1,265 23 Loop C ceiling 2.00 923 2 1,846 24 PZR floor 2.00 880 2 1,761 1 of 4

BVPS-2 UFSAR Rev. 16 Table 6.2-2A (Cont)

Beaver Valley MAAP-DBA Parameter File Summary of Containment Concrete Heat Sinks Total No. Sides Heat Thickness One-Sided Inside Total Sink# Description ft Area ft2 Ctmt Area ft2 25 PZR wall to loop B 3.00 1,253 2 2,506 26 PZR crane wall 2.00 2,568 2 5,136 27 PZR intermediate deck 4.00 984 2 1,969 28 PZR ceiling 2.00 923 2 1,846 29 Loop B floor 4.50 914 2 1,828 30 Loop B wall to RV head 3.00 717 2 1,435 laydown 31 Loop B crane wall 2.75 1,768 2 3,536 32 Loop B intermediate roof 6.00 131 2 261 33 Loop B ceiling 2.00 923 2 1,846 34 RV head laydown wall to 4.00 885 2 1,771 fuel transfer canal 35 RV head laydown crane wall 2.75 915 2 1,829 36 RV head laydown ceiling 2.00 923 2 1,846 37 Loop A floor 4.50 1,082 2 2,165 38 Loop A crane wall 2.75 2,114 2 4,228 39 Loop A wall to fuel 4.00 1,623 2 3,247 transfer canal 40 Loop A interior walls 2.00 271 2 543 41 Loop A ceiling 2.00 923 2 1,846 42 Lower annulus south half 4.50 8,432 1 8,432 outer wall(1) 43 Lower annulus north half 4.50 8,432 1 8,432 outer wall(1) 44 Upper annulus south half 2.75 8,226 2 16,452 crane wall 45 Upper annulus south half 4.50 7,569 1 7,569 outer wall(1) 46 Upper annulus north half 2.75 9,038 2 18,076 crane wall 47 Upper annulus north half 4.50 7,569 1 7,569 outer wall(1) 2 of 4

BVPS-2 UFSAR Rev. 16 Table 6.2-2A (Cont)

Beaver Valley MAAP-DBA Parameter File Summary of Containment Concrete Heat Sinks Total No. Sides Heat Thickness One-Sided Inside Total Sink# Description ft Area ft2 Ctmt Area ft2 48 Fuel transfer canal floor 4.00 471 2 942 49 Lower dome outer wall 2.50 9,929 1 9,929 50 Upper dome outer wall 2.50 8,774 1 8,774 51 Pressurizer interior walls 2.00 527 2 1,054 52 Instrument room interior 1.25 147 2 295 wall 53 RV laydown to Loop A misc 3.00 164 2 328 wall 54 Support beam at 718'-6" 4.50 274 2 548 55 Cubicle walls above op. 1.50 3,262 2 6,523 deck 56 RHR room wall to loop C 3.25 539 2 1,078 57 RHR room wall to loop A 3.25 568 2 1,135 58 RHR room crane wall 2.75 1,342 2 2,684 59 Refuel cavity wall to 4.00 1,033 2 2,066 upper annulus south half 60 Refuel cavity wall to 4.00 1,181 2 2,361 upper annulus north half 61 Containment shell sections 4.50 338 1 338 with embedment plates in lower compartment(2) 62 Containment shell sections 4.50 492 1 492 with embedment plates in lower south annulus(1) 63 Containment shell sections 4.50 492 1 492 with embedment plates in lower north annulus(1) 64 Containment shell sections 4.50 525 1 525 with embedment plates in upper south annulus(1) 3 of 4

BVPS-2 UFSAR Rev. 16 Table 6.2-2A (Cont)

Beaver Valley MAAP-DBA Parameter File Summary of Containment Concrete Heat Sinks Total No. Sides Heat Thickness One-Sided Inside Total Sink# Description ft Area ft2 Ctmt Area ft2 65 Containment shell sections 4.50 525 1 525 with embedment plates in upper north annulus(1) 66 Lower dome sections with 2.50 3,313 1 3,313 embedment plates 67 Upper dome sections with 2.50 2,922 1 2,922 embedment plates 68 Wall Adjacent to Reactor 3.00 705 2 1,410 Enclosure 69 Cubicle 1.50 1,590 2 3,179 70 Elevator Pit 1.00 94 2 188 71 Unlined portion of lower 4.50 640 1 640 compartment outer wall 72 Unlined portion of lower 4.50 836 1 836 annulus south half outer wall 73 Unlined portion of lower 4.50 836 1 836 annulus north half outer wall 74 Unlined portion of upper 4.50 856 1 856 annulus south half outer wall 75 Unlined portion of upper 4.50 856 1 856 annulus north half outer wall TOTAL 218,566 Note:

(1) Includes painted carbon steel liner and gap resistance between liner and concrete.

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BVPS-2 UFSAR Rev. 17 TABLE 6.2-3 CONTAINMENT DESIGN EVALUATION PARAMETERS General Information - Containment Interior minimum design pressure (psia) 8.0 Internal design pressure (psig) 45 Design temperature (F) 280 Minimum free volume (ft3) 1.75 x 106 Design leak rate (vol %/day 0.1 Initial Containment Conditions Containment Pressure (psia) 12.8 to 14.2 Inside temperature (F)70-108 Outside temperature (F) (-20) to 103 Relative humidity (%) 15 to 100 Service water temperature (F) 32 to 89 Refueling Water Storage Tank Useable volume (gal) 866,592 RWST temperature (F) 45 to 65 1 of 1

BVPS-2 UFSAR Rev. 17 TABLE 6.2-4 KEY INPUT DATA TO MAAP-DBA (PEAK PRESSURE CALCULATIONS)

Containment volume 1,750,867 ft3 Initial containment pressure 14.2 psia Initial containment temperature 108°F Initial containment relative humidity 15%

Steel liner to concrete gap effective 100 BTU/hr/ft2/°F heat transfer coefficient Paint thickness on carbon steel heat 0.009 inches sinks Effective heat transfer coefficient 266 BTU/hr/ft2/F for the paint on the carbon steel Paint thickness on concrete heat sinks 0.0105 inches (walls/ceiling),

0.036 inches (floors)

Effective heat transfer coefficient 50 BTU/hr/ft2/F for the paint on the concrete heat 42 BTU/hr/ft2/F (floor),

sinks 144 BTU/hr/ft2/F (ceiling/walls)

Zinc thickness on carbon steel 0.00234 in RWST temperature 65F Containment high-high quench spray 26.8 psia (max); 24.0 psia setpoint (min)

Containment high (SI actuation, FW 22 psia (max); 18 psia (min) isolation and CIA) safety analysis limit setpoint range Containment intermediate high-high 24 psia (max); 20 psia (min)

(steam line isolation) safety analysis limit setpoint Start delay for quench spray 74.5 seconds Quench spray flow rate Determined by pump curve variable gpm Start delay for recirculation spray Variable, determined by RWST drawdown 1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-5 BEAVER VALLEY MAAP-DBA PARAMETER FILE

SUMMARY

OF JUNCTION FLOW AREAS Junction Junction Upstream Node Downstream Junction Flow Loss Number Number Node Number Area (ft2) Coefficient 1 1 4 3.14 .594 2 1 5 6.0 .583 3 1 7 6.0 .583 4 1 9 6.0 .583 5 2 4 417.0 .510 6 2 5 160.0 .535 7 2 6 33.5 .526 8 2 7 155.0 .535 9 2 8 976.0 1.0 10 2 9 140.6 .535 11 2 10 1166.0 .894 12 2 11 1166.0 .894 13 3 5 28.0 .547 14 12 13 903.2 1.0 15 4 5 28.0 .547 16 4 10 517.8 .511 17 5 6 56.0 .547 18 5 15 384.8 .590 19 6 7 56.0 .547 20 6 11 56.0 .547 21 6 15 64.2 .617 22 7 15 389.4 .584 23 8 11 800.0 .516 24 8 15 673.4 .894 25 9 10 56.0 .547 26 9 15 389.4 .584 27 10 11 493.3 1.0 28 10 12 1257.3 .894 1 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-5 (Cont.)

BEAVER VALLEY MAAP-DBA PARAMETER FILE

SUMMARY

OF JUNCTION FLOW AREAS Junction Junction Upstream Node Downstream Junction Flow Loss Number Number Node Number Area (ft2) Coefficient 29 11 13 1257.3 .894 30 12 15 792.3 .519 31 13 15 130.0 .522 32 17 16 7980.0 1.0 33 16 15 8120.0 1.0 34 16 13 1822.0 1.0 35 16 12 1822.0 1.0 36 15 14 1105.0 1.0 37 14 2 0.0 .756 (Refueling canal drain path) 38 14 1 20.04 .538 39 * - - -

40 * - - -

41 1 2 .785 .474 Note: Junction 39 represents design basis leakage and 40 is the containment failure junction set to add when containment pressure exceeds a pre-set value.

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BVPS-2 UFSAR Rev. 17 TABLE 6.2-6 CONTAINMENT PEAK PRESSURE RESULTS FOR A DESIGN BASIS LARGE BREAK LOCA BEAVER VALLEY Description Single Failure Peak Pressure (psig) (2)

DEPS MIN SI DG 42.3 DEPS MAX SI CIB 42.3 DEHL None 44.8 Single Failures - Failed Equipment DG One train each, SI, QSS, RSS CIB One train each, QSS, RSS (1) Only blowdown was quantified. Peak pressure occurs before any active failure could occur.

(2) Gauge pressure is referenced to 14.3 psi atmospheric pressure.

Reported peak pressure is for break node.

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BVPS-2 UFSAR Rev. 16 TABLE 6.2-7 DOUBLE-ENDED HOT-LEG BREAK SEQUENCE OF EVENTS Time (sec) Event Description 0.0 Break Occurs, Reactor Trip and Loss of Offsite Power are assumed 1.8 Containment High-High Setpoint is reached 3.0 Low Pressurizer Pressure SI Setpoint is reached (1760 psia) 11.5 Broken Loop Accumulator Begins Injecting Water 11.6 Intact Loop Accumulator Begins Injecting Water 18.2 Peak Containment Pressure During Blowdown 19.2 End of Blowdown Phase 1 of 1

BVPS-2 UFSAR Rev. 17 Table 6.2-8 BVPS-2 Double-Ended Pump Suction Break Minimum Safeguards Sequence of Events Time (sec) Event Description 0.0 Break Occurs, Reactor Trip and Loss of Offsite Power are assumed 1.7 Containment High-High Setpoint is reached 3.0 Low Pressurizer Pressure SI Setpoint is reached (1760 psia) 12.9 Broken Loop Accumulator Begins Injecting Water 13.0 Intact Loop Accumulator Begins Injecting Water 17.4 Peak Containment Pressure During Blowdown 21.4 End of Blowdown Phase 27.0 Safety Injection Begins 63.2 Accumulator Water Injection Ends 76.5 Quench Spray is initiated 214.2 End of Reflood Phase 3134.2 Recirculation Spray is initiated 3462.2 ECCS Recirculation Begins 3600(1) Transient Modeling Terminated (1) Except for long term attributes such as EQ profiles, sump and water temperature.

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BVPS-2 UFSAR Rev. 23 TABLE 6.2-9 DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS SEQUENCE OF EVENTS (CIB FAILURE)

Time (sec) Event Description 0.0 Break Occurs, Reactor Trip and Loss of Offsite Power are assumed 1.7 Containment High-High Setpoint is reached 3.0 Low Pressurizer Pressure SI Setpoint is reached (1760 psia) 12.9 Broken Loop Accumulator Begins Injecting Water 13.0 Intact Loop Accumulator Begins Injecting Water 17.4 Peak Containment Pressure During Blowdown 21.4 End of Blowdown Phase 27.0 Safety Injection Begins 63.9 Accumulator Water Injection Ends 76.5 Quench Spray is initiated 213.4 End of Reflood Phase 2472.0 Recirculation Spray is initiated 2732.0 ECCS Recirculation Begins 3600(1) Transient Modeling Terminated (1) Except for long term attributes such as EQ profiles, sump and water temperature.

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BVPS-2 UFSAR Rev. 26 TABLE 6.2-10 MAAP-DBA PEAK PRESSURE RESULTS FOR A DESIGN BASIS MAIN STEAM LINE BREAK BEAVER VALLEY Power Level, Single* Peak Pressure Description  % Failure (psig) 1M-1.069 ft2 DER 100.6 MSIV 36.8 2M-1.069 ft2 DER 100.6 MFIV/CIB 36.6 3M-0.753 ft2 Split 100.6 CIB 31.5 4M-0.753 ft2 Split 100.6 MSIV 32.3 5M-0.753 ft2 Split 100.6 MFIV 31.5 6M-1.069 ft2 DER 70 MSIV 37.0 7M-1.069 ft2 DER 70 MFIV/CIB 36.8 8M-0.757 ft2 Split 70 CIB 32.8 9M-0.757 ft2 Split 70 MSIV 33.6 10M-0.757 ft2 Split 70 MFIV 32.4 11M-1.069 ft2 DER 30 MSIV 39.3 12M-1.069 ft2 DER 30 MFIV/CIB 38.6 13M-0.756 ft2 Split 30 CIB 35.4 14M-0.756 ft2 Split 30 MSIV 36.3 15M-0.756 ft2 Split 30 MFIV 34.6 16M-1.069 ft2 DER 0 MSIV 38.9 17M-1.069 ft2 DER 0 MFIV/CIB 36.7 18M-0.608 ft2 Split 0 CIB 33.3 1 of 2

BVPS-2 UFSAR Rev. 26 TABLE 6.2-10 (Cont.)

MAAP-DBA PEAK PRESSURE RESULTS FOR A DESIGN BASIS MAIN STEAM LINE BREAK BEAVER VALLEY Power Level, Single* Peak Pressure Description  % Failure (psig) 19M-0.608 ft2 Split 0 MSIV 32.7 20M-0.608 ft2 Split 0 MFIV 31.0 Single Failures - Failed Equipment CIB One train QSS DG One train each, SI, QSS, SW MSIV One main steam isolation valve MFIV One main feedwater isolation valve

• Some of these cases assumed two active failures, one for M&E release and the other for containment response. This is a conservatism that helps control the number of cases in the run matrix. The M&Es for the double failure cases were not significantly different than if only a single failure had been assumed.

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BVPS-2 UFSAR Rev. 19 TABLE 6.2-11 MAAP-DBA PEAK TEMPERATURE RESULTS FOR A DESIGN BASIS MAIN STEAM LINE BREAK BEAVER VALLEY Power Level, Single* Peak Description  % Failure Temperature(°F) 1M-1.069 ft2 DER 100.6 MSIV 343.9 2M-1.069 ft2 DER 100.6 MFIV/CIB 343.9 3M-0.753 ft2 Split 100.6 CIB 315.0 4M-0.753 ft2 Split 100.6 MSIV 315.4 5M-0.753 ft2 Split 100.6 MFIV 314.9 6M-1.069 ft2 DER 70 MSIV 335.8 7M-1.069 ft2 DER 70 MFIV/CIB 335.8 8M-0.757 ft2 Split 70 CIB 312.7 9M-0.757 ft2 Split 70 MSIV 313.1 10M-0.757 ft2 Split 70 MFIV 312.7 11M-1.069 ft2 DER 30 MSIV 333.5 12M-1.069 ft2 DER 30 MFIV/CIB 333.5 13M-0.756 ft2 Split 30 CIB 309.6 14M-0.756 ft2 Split 30 MSIV 310.1 15M-0.756 ft2 Split 30 MFIV 309.6 16M-1.069 ft2 DER 0 MSIV 336.8 17M-1.069 ft2 DER 0 MFIV/CIB 335.1 18M-0.608 ft2 Split 0 CIB 300.1 19M-0.608 ft2 Split 0 MSIV 301.1 20M-0.608 ft2 Split 0 MFIV 300.1 Single Failures - Failed Equipment CIB One train QSS DG One train each, SI, QSS, SW MSIV One main steam isolation valve MFIV One main feedwater isolation valve

• Some of these cases assumed two active failures, one for M&E release and the other for containment response. This is a conservatism that helps control the number of cases in the run matrix. The M&Es for the double failure cases were not significantly different than if only a single failure had been assumed.

• • These two sequences were analyzed using a maximum spray setpoint of 36.7 psia.

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BVPS-2 UFSAR Rev. 16 TABLE 6.2-12 SEQUENCE OF EVENTS - PEAK CONTAINMENT PRESSURE CASE 1.069 FT2 DOUBLE-ENDED RUPTURE (DER) WITH A MAIN STEAM ISOLATION VALVE (MSIV) FAILURE AT THIRTY PERCENT POWER (Case 11m)

Time (sec) Event Description 0.0 Accident occurs; ruptured loop steam generator and turbine plant piping blowdown into containment begins.

1.26 Low steamline pressure setpoint for closing the MSIV and FWIV is reached.

1.59 Turbine plant piping blowdown is complete; intact loop steam generators begin blowdown into containment.

3.26 Containment pressure setpoint for spray initiation is reached 8.26 MSIV and MFIV are fully closed; intact loop steam generators end blowdown into containment.

32 Peak containment temperature is reached.

77.75 Containment quench spray enters containment atmosphere.

216 Peak containment pressure is reached.

1800/1803 Auxiliary feedwater to ruptured steam generator manually isolated; steam release to containment ends.

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BVPS-2 UFSAR Rev. 16 TABLE 6.2-13 SEQUENCE OF EVENTS - PEAK CONTAINMENT TEMPERATURE CASE 1.069 FT2 DOUBLE-ENDED RUPTURE (DER) WITH A MAIN STEAM ISOLATION VALVE (MSIV) FAILURE AT FULL POWER (Case 1m)

Time (sec) Event Description 0.0 Accident occurs; ruptured loop steam generator and turbine plant piping blowdown into containment begins.

1.58 Turbine plant piping blowdown is complete; intact loop steam generators begin blowdown into containment.

3.40 Low steamline pressure setpoint for closing the MSIV and FWIV is reached.

4.73 Containment pressure setpoint for spray initation is reached 10.40 MSIV and MFIV are fully closed; intact loop steam generators end blowdown into containment.

22 Peak containment temperature is reached.

79.25 Containment quench spray enters containment atmosphere.

166 Peak containment pressure is reached.

1800/1804 Auxiliary feedwater to ruptured steam generator manually isolated; steam release to containment ends.

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BVPS-2 UFSAR Rev. 18 TABLE 6.2-14 SYSTEM PARAMETERS INITIAL CONDITIONS FOR THERMAL UPRATE Parameters Value Core Thermal Power (MWt)* 2917.4 Reactor Coolant System Total Flowrate (lbm/sec) 27583.3 Vessel Outlet Temperature (°F)* 621 Core Inlet Temperature (°F)* 547.1 Vessel Average Temperature (°F)* 584 Initial Steam Generator Steam Pressure (psia) 826 Steam Generator Design 51 Steam Generator Tube Plugging (percent) 0 Initial Steam Generator Secondary Side Mass (lbm)* 127881 Assumed Maximum Containment Backpressure (psia) 59.7 Accumulator Water Volume (ft3) per accumulator 1127.8 N2 Cover Gas Pressure (psia) 575 Temperature (°F) 105**

Notes:

• The Core Power, RCS Temperatures, and Secondary Side Mass values listed above include uncertainty allowance.

• • This value is lower than the containment maximum average temperature limit of 108F, however, it is a conservative value for the accumulators which are located in the lowest part of the containment structure.

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BVPS-2 UFSAR Rev. 16 TABLE 6.2-14A SAFETY INJECTION FLOW MINIMUM SAFEGUARDS RCS Pressure Total Flow (psig) (GPM)

Injection Mode (Reflood Phase) 0 4254.9 20 3895.7 50 3264.7 95 1679.6 100 1412.5 150 388.8 200 383.6 400 362.8 600 341.8 Cold Leg Recirculation Mode 0 3767 Note:

A maximum RWST temperature of 65°F was used during the Injection Phase. A maximum recirculation temperature of 120°F was used during the Recirculation Phase.

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BVPS-2 UFSAR Rev. 16 TABLE 6.2-14B SAFETY INJECTION FLOW MAXIMUM SAFEGUARDS RCS Pressure Total Flow (psig) (GPM)

Injection Mode (Reflood Phase) 0 6148.5 20 5696.8 50 5019.3 100 3265.5 130 1481.0 150 847.9 200 840.0 400 803.7 600 771.0 Cold Leg Recirculation Mode 0 6228.6 Note:

A maximum RWST temperature of 65°F was used during the Injection Phase. A maximum recirculation temperature of 120°F was used during the Recirculation Phase for the CIB failure case 150°F was used for the SW failure case.

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BVPS-2 UFSAR Rev. 16 TABLE 6.2-14C DOUBLE-ENDED HOT-LEG BREAK BLOWDOWN MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec)

.00000 .0 .0 .0 .0

.00110 46506.1 29844.9 46504.5 29843.1

.00214 46366.6 29754.2 46133.9 29599.7

.101 41311.9 26879.3 26481.2 16957.0

.201 34980.7 22842.8 23912.4 15242.1

.301 34192.1 22255.7 21041.5 13262.6

.401 33317.1 21650.8 19607.0 12161.5

.502 32470.5 21091.0 18714.2 11411.2

.602 32391.2 21031.5 18148.0 10894.0

.701 32410.5 21055.5 17608.8 10426.7

.801 32014.1 20841.8 17255.8 10094.8

.902 31474.9 20557.3 16960.8 9818.1 1.00 31197.4 20470.2 16684.8 9571.9 1.10 30928.9 20413.4 16497.6 9391.0 1.20 30573.5 20308.0 16344.4 9239.6 1.30 30169.3 20163.3 16252.2 9131.9 1.40 29740.4 20003.3 16213.2 9060.6 1.50 29287.3 19827.8 16220.9 9021.2 1.60 28761.6 19606.8 16261.8 9003.7 1.70 28147.8 19322.1 16325.4 9002.8 1.80 27476.6 18995.0 16405.4 9013.9 1.90 26804.3 18663.2 16493.0 9032.5 2.00 26155.8 18342.7 16579.9 9054.3 2.10 25493.2 18003.5 16664.1 9077.9 1 of 5

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14C (Cont.)

DOUBLE-ENDED HOT-LEG BREAK BLOWDOWN MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 2.20 24793.6 17624.1 16742.3 9101.3 2.30 24103.8 17236.2 16809.8 9122.1 2.40 23435.8 16850.3 16867.0 9140.0 2.50 22788.1 16467.0 16913.2 9154.2 2.60 22157.1 16083.7 16946.5 9163.7 2.70 21562.7 15712.5 16966.1 9167.6 2.80 20999.5 15344.6 16971.5 9165.5 2.90 20499.3 15010.2 16963.7 9157.6 3.00 20034.1 14686.4 16942.7 9143.7 3.10 19628.5 14391.5 16909.2 9124.1 3.20 19285.5 14129.6 16865.5 9099.9 3.30 18980.9 13882.7 16811.7 9070.9 3.40 18721.7 13659.4 16748.3 9037.4 3.50 18506.3 13459.3 16677.2 9000.3 3.60 18320.5 13273.7 16597.5 8959.0 3.70 18171.4 13109.6 16511.1 8914.5 3.80 18047.1 12959.4 16417.7 8866.6 3.90 17946.8 12824.1 16317.5 8815.3 4.00 17867.0 12702.5 16211.5 8761.2 4.20 17776.0 12507.4 15983.3 8645.0 4.40 17808.4 12393.5 15726.1 8514.1 4.60 17927.7 12343.5 15429.9 8363.2 4.80 18115.9 12329.6 15074.2 8181.1 5.00 18396.0 12352.9 14740.5 8012.5 5.20 18767.1 12418.9 14296.3 7783.7 5.40 19352.6 12607.5 13848.4 7553.6 2 of 5

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14C (Cont.)

DOUBLE-ENDED HOT-LEG BREAK BLOWDOWN MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 5.60 14305.2 10429.7 13391.3 7319.0 5.80 14241.6 10255.5 12931.5 7083.4 6.00 14364.3 10266.5 12515.2 6872.0 6.20 14439.6 10292.3 12097.5 6659.0 6.40 14689.4 10301.9 11637.4 6421.2 6.60 15049.1 10399.0 11192.0 6191.2 6.80 14999.6 10249.0 10769.6 5973.6 7.00 15365.5 10330.1 10353.6 5758.9 7.20 15683.6 10402.0 9948.6 5549.8 7.40 15971.8 10469.1 9571.7 5355.7 7.60 16258.9 10543.6 9214.4 5171.8 7.80 16594.6 10651.0 8875.1 4997.0 8.00 17158.0 10907.0 8556.8 4833.1 8.20 17053.0 10792.2 8246.6 4673.3 8.40 16725.6 10538.7 7949.2 4520.3 8.60 14694.9 9424.6 7660.0 4371.6 8.80 13878.9 8964.1 7380.7 4228.8 9.00 13826.1 8916.4 7115.7 4094.4 9.20 13789.1 8883.5 6866.0 3969.2 9.40 13740.1 8844.0 6639.3 3856.8 9.60 13672.9 8787.6 6416.5 3745.9 9.80 13461.2 8649.8 6204.1 3640.7 10.0 12878.5 8314.2 5996.5 3538.1 10.2 12189.5 7927.5 5796.5 3440.0 10.4 11803.9 7705.9 5599.5 3344.4 10.4 11799.7 7703.4 5596.6 3342.9 10.4 11796.0 7701.2 5594.1 3341.7 3 of 5

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14C (Cont.)

DOUBLE-ENDED HOT-LEG BREAK BLOWDOWN MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 10.6 11564.5 7566.0 5411.4 3254.3 10.8 11330.9 7432.4 5230.6 3168.7 11.0 11053.8 7277.7 5055.7 3086.6 11.2 10712.5 7091.0 4886.8 3007.9 11.4 10331.8 6886.4 4722.8 2931.8 11.6 9964.0 6691.8 4566.3 2859.7 11.8 9625.7 6515.9 4413.7 2789.9 12.0 9308.6 6355.2 4265.5 2722.9 12.2 8992.6 6198.7 4122.2 2658.5 12.4 8650.4 6032.5 3981.1 2595.1 12.6 8260.6 5847.8 3839.5 2531.7 12.8 7824.8 5648.0 3689.7 2464.3 13.0 7355.1 5441.4 3528.5 2392.9 13.2 6851.5 5229.8 3350.6 2316.8 13.4 6334.7 5023.8 3157.0 2236.8 13.6 5794.9 4815.8 2947.5 2153.0 13.8 5241.0 4606.0 2732.2 2068.6 14.0 4683.9 4393.7 2518.7 1983.8 14.2 4106.9 4114.4 2321.9 1903.4 14.4 3655.6 3782.3 2145.6 1826.3 14.6 3390.8 3538.3 1997.8 1757.7 14.8 3204.0 3361.8 1873.2 1695.8 15.0 3043.7 3210.9 1771.0 1645.7 15.2 2853.9 3053.5 1677.5 1598.2 15.4 2616.4 2882.7 1596.2 1554.9 15.6 2347.8 2697.4 1518.1 1516.8 15.8 2093.8 2499.3 1435.1 1485.2 4 of 5

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14C (Cont.)

DOUBLE-ENDED HOT-LEG BREAK BLOWDOWN MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 16.0 1893.8 2304.3 1350.3 1450.4 16.2 1707.1 2093.3 1263.9 1416.2 16.4 1547.9 1912.1 1182.6 1368.2 16.6 1412.2 1757.7 1132.2 1346.8 16.8 1340.1 1676.1 1022.5 1239.5 17.0 1261.1 1583.5 953.6 1166.7 17.2 1162.6 1461.4 899.6 1105.5 17.4 1067.4 1343.3 858.0 1056.8 17.6 998.4 1256.2 811.0 1000.7 17.8 915.2 1153.4 726.6 898.5 18.0 836.1 1056.2 630.5 782.0 18.2 758.4 959.6 562.2 698.7 18.4 664.0 841.2 523.1 651.9 18.6 593.8 753.6 405.0 503.7 18.7 559.5 710.5 369.6 461.6 18.8 317.2 402.1 344.0 430.2 19.0 .0 .0 160.9 201.8 19.2 .0 .0 .0 .0

• Mass and Energy exiting from the RV side of the break

• • Mass and Energy exiting from the SG side of the break 5 of 5

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14D DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES (SAME FOR ALL DEPS RUNS)

Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec)

.00000 .0 .0 .0 .0

.00109 88813.2 47922.9 40351.7 21722.1

.101 40209.7 21715.8 20435.8 10994.1

.202 41011.2 22298.7 22609.3 12171.8

.301 44814.9 24586.6 23088.2 12441.9

.401 45435.3 25200.1 22751.7 12274.7

.501 44833.6 25178.3 22031.3 11896.1

.602 44232.4 25156.0 21364.1 11542.5

.702 44540.0 25626.6 20891.4 11291.2

.801 44177.2 25680.9 20541.8 11106.2

.901 43156.3 25326.6 20278.1 10967.8 1.00 41991.4 24872.3 20086.3 10867.0 1.10 40887.0 24443.5 19955.7 10798.7 1.20 39821.3 24030.8 19897.4 10768.6 1.30 38772.9 23621.4 19905.5 10774.0 1.40 37783.2 23230.6 19936.5 10791.0 1.50 36908.7 22887.7 19940.9 10792.8 1.60 36142.5 22591.2 19917.5 10778.9 1.70 35446.0 22327.1 19885.1 10759.9 1.80 34731.2 22053.6 19850.2 10739.8 1.90 33965.9 21761.5 19793.1 10707.8 2.00 33145.0 21449.3 19693.9 10653.2 2.10 32124.1 21021.3 19534.0 10565.6 2.20 30989.8 20520.6 19217.8 10393.1 1 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14D (Cont).

DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES (SAME FOR ALL DEPS RUNS)

Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 2.30 29718.8 19926.4 18898.7 10220.1 2.40 28370.0 19265.1 18585.8 10050.1 2.50 26271.4 18046.5 18237.6 9861.0 2.60 22526.4 15619.5 17936.3 9698.1 2.70 20341.8 14248.6 17630.1 9533.0 2.80 19301.7 13612.7 17291.8 9350.7 2.90 18053.8 12758.2 16997.9 9193.2 3.00 17127.4 12122.9 16751.9 9062.3 3.10 16396.6 11622.5 16516.2 8937.0 3.20 15708.0 11149.0 16291.8 8818.0 3.30 15096.6 10734.6 16088.7 8710.8 3.40 14548.9 10370.0 15901.7 8612.5 3.50 14063.1 10050.5 15727.2 8520.8 3.60 13636.3 9772.8 15539.9 8422.0 3.70 13269.5 9536.7 15426.3 8364.0 3.80 12948.4 9329.5 15291.4 8293.5 3.90 12609.6 9105.9 15123.9 8205.2 4.00 12288.5 8895.2 14986.8 8133.8 4.20 11754.8 8549.6 14723.3 7996.4 4.40 11324.6 8258.2 14466.5 7862.6 4.60 10987.7 8022.4 14240.5 7745.6 4.80 10702.1 7816.8 14000.8 7621.4 5.00 10461.2 7642.5 13753.4 7492.8 5.20 10236.1 7478.4 13395.5 7303.5 5.40 10103.0 7378.3 13027.1 7109.3 2 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14D (Cont).

DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES (SAME FOR ALL DEPS RUNS)

Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 5.60 10296.5 7512.6 14648.7 8004.6 5.80 10519.3 7740.8 14501.0 7929.5 6.00 9862.2 7865.5 14373.1 7867.2 6.20 8477.2 7510.2 14106.8 7727.0 6.40 7876.2 7245.0 13799.5 7564.8 6.60 7591.4 7038.1 13470.1 7390.4 6.80 7492.7 6810.1 13184.9 7239.5 7.00 7587.5 6652.4 12948.4 7113.8 7.20 7790.2 6576.2 12740.3 6999.6 7.40 7953.2 6499.0 12637.0 6938.7 7.60 8022.6 6416.0 12505.7 6858.8 7.80 8024.0 6352.5 12337.7 6758.3 8.00 7926.6 6244.8 12148.7 6647.7 8.20 7812.6 6139.8 12007.4 6565.3 8.40 7690.2 6041.2 11864.5 6482.9 8.60 7561.1 5950.0 11693.4 6385.2 8.80 7428.3 5864.5 11518.0 6285.8 9.00 7289.7 5779.9 11351.7 6192.3 9.20 7145.6 5694.1 11181.4 6097.4 9.40 6996.8 5606.2 11002.6 5998.2 9.60 6847.7 5517.9 10828.1 5901.7 9.80 6698.6 5429.1 10659.4 5808.7 10.0 6555.3 5346.1 10488.0 5714.3 10.2 6411.9 5258.8 10311.2 5617.0 10.4 6275.6 5171.5 10141.4 5524.2 10.6 6144.5 5084.6 9972.5 5432.1 3 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14D (Cont).

DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES (SAME FOR ALL DEPS RUNS)

Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 10.8 6017.4 4999.1 9805.3 5340.9 11.0 5892.8 4915.0 9643.4 5252.7 11.2 5771.6 4833.4 9483.7 5165.6 11.4 5647.4 4750.4 9325.8 5079.5 11.6 5513.4 4661.3 9141.1 4978.6 11.8 5368.9 4563.2 8967.0 4884.5 12.0 5226.4 4458.3 8806.8 4798.4 12.2 5093.0 4350.7 8629.6 4702.5 12.4 4969.0 4243.7 8462.8 4612.5 12.6 4846.9 4136.6 8287.8 4518.0 12.8 4723.7 4028.9 8117.7 4426.6 13.0 4602.9 3923.7 7878.7 4296.7 13.2 4490.6 3826.3 7607.2 4150.3 13.4 4390.7 3739.2 7478.0 4077.9 13.6 4297.3 3660.8 7252.9 3927.2 13.8 4205.1 3592.2 7209.6 3851.5 14.0 4112.3 3533.8 7192.1 3772.8 14.2 4017.5 3484.4 7146.4 3671.8 14.4 3913.0 3439.8 6964.3 3502.6 14.6 3800.2 3402.0 6677.3 3287.1 14.8 3675.5 3368.9 6392.0 3081.2 15.0 3543.1 3342.6 6195.1 2929.4 15.2 3403.2 3321.2 5974.9 2779.0 15.4 3253.9 3306.0 5754.2 2635.7 15.6 3057.1 3259.2 5532.0 2497.5 15.8 2763.2 3138.5 5062.7 2251.5 4 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14D (Cont).

DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES (SAME FOR ALL DEPS RUNS)

Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 16.0 2468.0 2958.4 4680.7 2042.7 16.2 2250.5 2759.1 4428.1 1892.9 16.4 2048.5 2529.8 4163.1 1746.1 16.6 1874.5 2323.3 3778.8 1555.6 16.8 1723.6 2142.6 3381.0 1362.8 17.0 1594.6 1986.4 3022.3 1189.2 17.2 1469.3 1833.8 2783.7 1067.8 17.4 1352.3 1690.7 2650.6 992.2 17.6 1228.0 1538.0 2602.1 952.7 17.8 1122.4 1407.9 2570.5 921.5 18.0 1022.8 1284.4 2615.4 915.7 18.2 911.4 1146.5 2760.6 941.6 18.4 809.0 1019.0 2941.7 976.9 18.6 704.7 888.5 3034.8 983.8 18.8 603.4 761.4 3003.2 952.9 19.0 509.2 643.1 2786.6 867.7 19.2 424.2 536.2 2538.9 778.1 19.4 349.8 442.5 2287.9 691.3 19.6 287.7 364.2 2025.4 604.1 19.8 250.1 316.8 1762.7 519.4 20.0 199.4 252.7 1510.9 440.4 20.2 125.5 159.4 1244.0 359.1 20.4 48.7 62.0 966.5 276.8 20.6 70.3 89.8 709.7 202.1 20.8 .0 .0 462.7 131.4 5 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14D (Cont).

DOUBLE-ENDED PUMP SUCTION BREAK BLOWDOWN MASS AND ENERGY RELEASES (SAME FOR ALL DEPS RUNS)

Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 21.0 .0 .0 214.2 60.9 21.2 .0 .0 32.0 9.1 21.4 .0 .0 .0 .0

• Mass and Energy exiting the SG side of the break

• • Mass and Energy exiting the pump side of the break 6 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14E DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 21.4 .0 .0 .0 .0 21.9 .0 .0 .0 .0 22.1 .0 .0 .0 .0 22.2 .0 .0 .0 .0 22.3 .0 .0 .0 .0 22.4 .0 .0 .0 .0 22.4 .0 .0 .0 .0 22.5 57.4 67.5 .0 .0 22.6 20.7 24.4 .0 .0 22.7 18.2 21.4 .0 .0 22.9 21.6 25.4 .0 .0 23.0 31.8 37.4 .0 .0 23.1 37.0 43.6 .0 .0 23.2 43.0 50.6 .0 .0 23.3 48.3 56.8 .0 .0 23.4 53.7 63.3 .0 .0 23.5 58.3 68.6 .0 .0 23.6 61.7 72.7 .0 .0 23.7 65.1 76.7 .0 .0 23.8 66.8 78.7 .0 .0 23.8 68.4 80.6 .0 .0 23.9 71.6 84.3 .0 .0 24.0 74.7 87.9 .0 .0 24.1 77.6 91.4 .0 .0 1 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14E (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 24.2 80.5 94.8 .0 .0 24.3 83.3 98.1 .0 .0 24.4 86.0 101.3 .0 .0 25.4 109.9 129.5 .0 .0 26.4 129.7 152.9 .0 .0 27.4 155.3 183.0 551.6 47.2 28.4 372.2 440.4 3820.8 431.3 28.5 374.6 443.4 3840.0 436.1 29.5 374.3 443.0 3826.9 440.5 30.5 367.7 435.1 3756.9 434.8 31.5 360.8 426.8 3683.8 428.3 32.5 353.9 418.7 3610.7 421.8 33.2 349.2 413.1 3560.3 417.2 33.5 347.3 410.7 3538.9 415.3 34.5 340.8 403.1 3468.9 408.9 35.5 334.6 395.7 3400.9 402.6 36.5 328.7 388.6 3335.0 396.5 37.5 322.9 381.8 3271.1 390.6 38.5 317.4 375.3 3209.3 384.8 39.5 312.2 369.0 3149.4 379.3 40.5 307.1 363.0 3091.4 373.9 41.5 302.2 357.1 3035.2 368.6 42.5 297.5 351.6 2980.7 363.5 43.5 293.0 346.2 2927.8 358.5 44.5 288.6 341.0 2876.4 353.7 45.5 284.4 336.0 2826.4 349.0 2 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14E (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 46.5 280.3 331.1 2777.8 344.4 46.9 278.7 329.2 2758.7 342.6 47.5 276.4 326.4 2730.5 339.9 48.5 272.5 321.9 2684.4 335.6 49.5 268.8 317.5 2639.5 331.3 50.5 265.2 313.2 2595.7 327.2 51.5 261.8 309.1 2553.0 323.1 52.5 258.4 305.1 2511.2 319.1 53.5 255.1 301.2 2470.4 315.2 54.5 251.9 297.4 2430.5 311.4 55.1 250.0 295.2 2407.0 309.2 55.5 248.8 293.7 2391.5 307.7 56.5 245.8 290.2 2353.3 304.0 57.5 242.8 286.7 2315.9 300.4 58.5 240.0 283.3 2279.2 296.9 59.5 237.2 279.9 2243.3 293.4 60.5 234.4 276.7 2208.0 290.0 61.5 231.8 273.5 2173.4 286.6 62.5 229.1 270.4 2139.5 283.3 63.5 218.0 257.2 232.5 100.6 64.5 239.8 283.0 238.2 111.6 65.5 236.1 278.7 237.1 109.8 66.5 232.4 274.3 235.9 107.9 67.5 228.7 269.9 234.7 106.0 68.5 225.1 265.6 233.5 104.2 69.5 221.5 261.4 232.4 102.5 3 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14E (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 70.5 218.1 257.3 231.3 100.8 71.5 214.6 253.2 230.3 99.1 72.5 211.2 249.2 229.2 97.5 73.3 208.5 246.0 228.4 96.2 73.5 207.8 245.2 228.1 95.8 74.5 204.6 241.4 227.1 94.3 75.5 201.8 238.1 226.1 93.0 76.5 199.0 234.8 225.2 91.6 77.5 196.3 231.5 224.2 90.4 78.5 193.6 228.4 223.3 89.1 79.5 191.0 225.3 222.4 87.9 80.5 188.5 222.3 221.6 86.7 81.5 185.9 219.3 220.7 85.6 82.5 183.5 216.4 219.9 84.5 84.5 178.8 210.8 218.3 82.3 86.5 174.2 205.5 216.8 80.3 88.5 169.9 200.4 215.4 78.4 90.5 165.9 195.6 214.1 76.6 92.5 162.0 191.0 212.9 74.9 94.5 158.3 186.6 211.7 73.3 95.0 157.4 185.6 211.4 73.0 96.5 154.9 182.6 210.6 71.9 98.5 151.6 178.7 209.6 70.5 100.5 148.5 175.1 208.6 69.2 102.5 145.6 171.6 207.7 68.0 104.5 142.9 168.4 206.9 66.9 4 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14E (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 106.5 140.4 165.4 206.1 65.8 108.5 138.0 162.6 205.4 64.8 110.5 135.8 160.0 204.7 64.0 112.5 133.7 157.6 204.1 63.1 114.5 131.8 155.4 203.5 62.4 116.5 130.1 153.3 203.0 61.7 118.5 128.5 151.4 202.5 61.0 120.5 127.0 149.6 202.1 60.4 120.8 126.8 149.4 202.0 60.3 122.5 125.6 148.0 201.7 59.9 124.5 124.4 146.6 201.3 59.4 126.5 123.2 145.2 201.0 58.9 128.5 122.2 144.0 200.7 58.5 130.5 121.3 142.9 200.4 58.1 132.5 120.4 141.9 200.1 57.8 134.5 119.7 141.0 199.9 57.5 136.5 119.0 140.2 199.7 57.2 138.5 118.4 139.5 199.5 57.0 140.5 117.9 138.9 199.3 56.7 142.5 117.4 138.3 199.2 56.5 144.5 117.0 137.8 199.0 56.4 146.5 116.6 137.4 198.9 56.2 148.5 116.3 137.1 198.8 56.1 149.8 116.2 136.9 198.8 56.0 150.5 116.1 136.8 198.7 55.9 152.5 115.9 136.5 198.6 55.8 5 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14E (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 154.5 116.0 136.7 198.7 55.9 156.5 116.2 136.9 199.3 56.0 158.5 116.3 137.0 200.2 56.3 160.5 116.3 137.1 201.7 56.6 162.5 116.4 137.1 203.5 57.0 164.5 116.4 137.2 205.7 57.5 166.5 116.4 137.1 208.2 58.1 168.5 116.3 137.0 211.0 58.7 170.5 116.1 136.8 214.1 59.3 172.5 115.8 136.5 217.4 60.0 174.5 115.5 136.0 220.9 60.7 176.5 115.0 135.4 224.6 61.4 178.5 114.3 134.7 228.4 62.2 180.5 113.6 133.8 232.4 62.9 180.6 113.5 133.7 232.6 62.9 182.5 112.6 132.7 236.6 63.7 184.5 111.6 131.5 241.0 64.4 186.5 110.4 130.0 245.4 65.2 188.5 109.0 128.4 250.1 66.0 190.5 107.4 126.5 254.9 66.8 192.5 107.0 126.0 258.2 67.2 194.5 106.7 125.7 261.2 67.5 196.5 106.4 125.4 263.8 67.8 198.5 106.1 125.0 266.3 68.0 200.5 105.7 124.6 268.7 68.1 202.5 105.4 124.1 270.8 68.1 6 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14E (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 204.5 104.9 123.6 272.8 68.1 206.5 104.5 123.1 274.7 68.1 208.5 104.1 122.6 276.4 68.0 210.5 103.6 122.0 278.1 67.9 212.5 103.1 121.5 279.7 67.7 214.2 102.7 121.0 280.9 67.6

• Mass and Energy exiting the SG side of the break

• • Mass and Energy exiting the pump side of the break 7 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14F DOUBLE-ENDED PUMP SUCTION BREAK - MINIMUM SAFEGUARDS PRINCIPAL PARAMETERS DURING REFLOOD Flooding Injection Core Downcomer Total Accum Spill Time Temp Rate Carryover Height Height Flow Enthalpy (sec) (°F) (in/sec) Fraction (ft) (ft) Frac (lbm/sec) (Btu/lbm) 21.4 169.0 .000 .000 .00 .00 .333 .0 .0 .0 .00 22.2 166.4 22.694 .000 .67 1.40 .000 6117.9 6117.9 .0 74.50 22.4 165.1 24.146 .000 1.07 1.32 .000 6067.6 6067.6 .0 74.50 22.7 164.5 2.530 .113 1.32 1.96 .298 5952.2 5952.2 .0 74.50 22.9 164.4 2.633 .130 1.35 2.40 .310 5934.5 5934.5 .0 74.50 23.0 164.4 2.595 .163 1.37 2.69 .361 5887.9 5887.9 .0 74.50 23.2 164.4 2.604 .205 1.41 3.34 .390 5842.1 5842.1 .0 74.50 23.8 164.4 2.494 .302 1.50 4.99 .427 5710.0 5710.0 .0 74.50 24.4 164.5 2.425 .386 1.59 6.83 .442 5574.6 5574.6 .0 74.50 28.4 164.9 4.268 .635 2.01 15.61 .663 4786.7 4404.9 .0 71.19 30.5 165.1 3.918 .684 2.25 15.62 .658 4491.7 4109.2 .0 70.97 33.2 165.6 3.627 .710 2.51 15.62 .651 4221.1 3828.9 .0 70.64 39.5 167.5 3.232 .731 3.00 15.62 .635 3718.3 3307.3 .0 69.91 46.9 170.4 2.942 .739 3.50 15.62 .617 3263.8 2837.3 .0 69.08 55.1 174.1 2.711 .741 4.00 15.62 .599 2862.7 2423.8 .0 68.14 62.5 177.6 2.547 .741 4.42 15.62 .584 2560.4 2113.3 .0 67.25 1 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14F (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK - MINIMUM SAFEGUARDS PRINCIPAL PARAMETERS DURING REFLOOD Flooding Injection Core Downcomer Total Accum Spill Time Temp Rate Carryover Height Height Flow Enthalpy (sec) (°F) (in/sec) Fraction (ft) (ft) Frac (lbm/sec) (Btu/lbm) 63.5 178.1 2.534 .741 4.48 15.61 .578 454.2 .0 .0 33.00 64.5 178.6 2.631 .744 4.53 15.49 .589 442.9 .0 .0 33.00 73.3 183.8 2.357 .741 5.00 14.63 .574 452.9 .0 .0 33.00 84.5 191.7 2.098 .738 5.54 13.96 .555 458.3 .0 .0 33.00 95.0 199.7 1.912 .737 6.00 13.65 .537 461.5 .0 .0 33.00 108.5 210.1 1.743 .736 6.54 13.57 .518 464.1 .0 .0 33.00 120.8 218.6 1.644 .736 7.00 13.73 .504 465.5 .0 .0 33.00 136.5 227.9 1.572 .738 7.55 14.10 .493 466.5 .0 .0 33.00 149.8 234.7 1.541 .741 8.00 14.51 .490 466.8 .0 .0 33.00 152.5 236.0 1.537 .742 8.09 14.60 .489 466.9 .0 .0 33.00 154.5 236.9 1.537 .742 8.16 14.67 .489 466.9 .0 .0 33.00 166.5 242.2 1.531 .745 8.55 15.04 .491 466.8 .0 .0 33.00 180.6 247.9 1.496 .748 9.00 15.37 .488 467.2 .0 .0 33.00 198.5 254.0 1.419 .750 9.55 15.57 .480 467.8 .0 .0 33.00 214.2 258.7 1.364 .752 10.00 15.61 .482 467.8 .0 .0 33.00 2 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14G DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 214.2 121.2 152.6 347.1 84.9 219.2 121.9 153.4 346.5 84.6 224.2 121.5 153.0 346.8 84.5 229.2 121.2 152.6 347.1 84.4 234.2 120.8 152.2 347.5 84.3 239.2 121.5 153.0 346.8 84.0 244.2 121.1 152.5 347.2 83.9 249.2 120.8 152.1 347.5 83.8 254.2 120.4 151.7 347.9 83.7 259.2 121.1 152.5 347.2 83.4 264.2 120.7 152.0 347.6 83.3 269.2 120.4 151.6 347.9 83.2 274.2 120.0 151.1 348.3 86.0 279.2 120.7 151.9 347.7 85.6 284.2 120.3 151.5 348.0 85.6 289.2 119.9 151.0 348.4 85.5 294.2 120.6 151.8 347.7 85.1 299.2 120.2 151.4 348.1 85.0 304.2 119.8 150.9 348.5 84.9 309.2 119.5 150.5 348.8 84.8 314.2 120.1 151.2 348.2 84.5 319.2 119.7 150.8 348.6 84.4 1 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14G (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 324.2 119.4 150.3 349.0 84.3 329.2 119.9 151.0 348.4 83.9 334.2 119.6 150.6 348.7 83.9 339.2 119.2 150.1 349.1 83.8 344.2 118.8 149.6 349.5 83.7 349.2 119.4 150.4 348.9 83.3 354.2 119.0 149.9 349.3 83.2 359.2 118.7 149.4 349.6 83.1 364.2 119.2 150.1 349.1 82.8 369.2 118.8 149.6 349.5 82.7 374.2 118.5 149.2 349.8 82.6 379.2 119.0 149.9 349.3 82.3 384.2 118.6 149.4 349.7 82.2 389.2 118.2 148.9 350.1 82.1 394.2 118.8 149.6 349.5 81.7 399.2 118.4 149.1 349.9 81.6 404.2 118.1 148.7 350.2 81.5 409.2 117.8 148.4 350.5 81.4 414.2 118.5 149.2 349.8 81.0 419.2 118.2 148.8 350.1 80.9 424.2 117.9 148.5 350.4 80.8 429.2 117.6 148.1 350.7 80.7 434.2 118.3 148.9 350.1 80.3 439.2 118.0 148.6 350.3 80.2 444.2 117.7 148.2 350.6 80.1 449.2 117.4 147.8 350.9 79.9 2 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14G (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 454.2 118.0 148.6 350.3 79.6 459.2 117.7 148.2 350.6 79.5 464.2 117.4 147.9 350.9 79.3 469.2 117.1 147.5 351.2 79.2 474.2 117.7 148.2 350.6 78.9 479.2 117.4 147.9 350.9 78.7 484.2 117.1 147.5 351.2 81.3 489.2 117.7 148.2 350.6 80.9 494.2 117.4 147.8 350.9 80.8 499.2 117.1 147.4 351.2 80.6 504.2 116.8 147.0 351.5 80.5 509.2 117.3 147.7 351.0 80.1 514.2 117.0 147.3 351.3 80.0 519.2 116.7 146.9 351.6 79.9 524.2 117.2 147.6 351.1 79.5 529.2 116.9 147.2 351.4 79.4 534.2 116.6 146.8 351.7 79.2 539.2 117.1 147.4 351.2 78.9 544.2 116.7 147.0 351.6 78.8 549.2 116.4 146.6 351.9 78.6 554.2 116.9 147.2 351.4 78.3 559.2 116.6 146.8 351.8 78.1 564.2 116.2 146.3 352.1 78.0 569.2 116.7 146.9 351.6 77.7 574.2 116.3 146.5 352.0 77.5 579.2 116.0 146.0 352.3 77.4 3 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14G (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 584.2 116.4 146.6 351.9 77.1 589.2 116.1 146.1 352.2 79.4 594.2 116.5 146.7 351.8 79.1 599.2 116.1 146.2 352.2 78.9 604.2 115.8 145.8 352.5 78.8 609.2 116.2 146.4 352.1 78.4 614.2 115.9 145.9 352.4 78.3 619.2 116.3 146.5 352.0 77.9 624.2 116.0 146.0 352.3 77.8 629.2 115.6 145.6 352.7 77.6 634.2 116.0 146.1 352.3 77.3 639.2 115.7 145.6 352.6 77.1 644.2 116.1 146.1 352.3 76.8 649.2 115.7 145.7 352.6 76.6 654.2 116.0 146.1 352.3 76.3 659.2 115.7 145.6 352.7 76.1 664.2 115.3 145.1 353.1 78.4 669.2 115.6 145.6 352.7 78.1 674.2 115.2 145.0 353.1 77.9 679.2 115.5 145.4 352.8 77.5 684.2 115.8 145.8 352.5 77.2 689.2 115.4 145.3 352.9 77.0 694.2 115.7 145.6 352.7 76.7 699.2 115.2 145.1 353.1 76.6 704.2 115.5 145.4 352.9 76.2 709.2 115.0 144.8 353.3 76.1 4 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14G (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 714.2 115.2 145.1 353.1 75.7 719.2 115.4 145.3 352.9 75.4 724.2 114.9 144.7 353.4 77.6 729.2 115.1 144.9 353.2 77.3 734.2 115.2 145.1 353.1 76.9 739.2 114.7 144.5 353.6 76.8 744.2 114.8 144.6 353.5 76.4 749.2 114.9 144.7 353.4 76.1 754.2 115.0 144.8 353.3 75.8 759.2 115.1 144.9 353.3 75.5 764.2 115.1 144.9 353.2 75.2 769.2 114.5 144.1 353.8 75.0 774.2 114.5 144.1 353.9 77.0 779.2 114.4 144.1 353.9 76.7 784.2 114.3 144.0 354.0 76.4 789.2 114.8 144.6 353.5 75.9 794.2 114.7 144.4 353.6 75.7 799.2 114.5 144.2 353.8 75.4 804.2 114.3 144.0 354.0 75.1 809.2 114.7 144.4 353.6 74.7 814.2 114.4 144.0 353.9 74.4 819.2 114.1 143.6 354.2 76.4 824.2 114.2 143.8 354.1 76.0 829.2 114.3 144.0 354.0 75.6 834.2 66.6 83.8 401.7 87.8 1089.1 66.6 83.8 401.7 87.8 5 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14G (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MINIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 1089.2 66.1 82.9 402.2 85.0 1089.2 66.1 82.9 402.2 85.0 1367.4 66.1 82.9 402.2 85.0 1367.5 57.6 66.3 410.7 13.6 2948.0 48.0 55.3 420.3 13.9 2948.1 50.5 58.1 453.9 40.0 3600.0 47.5 54.7 456.9 40.3

• Mass and Energy exiting the SG side of break

• • Mass and Energy exiting the pump side of break 6 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14H DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 21.4 .0 .0 .0 .0 21.9 .0 .0 .0 .0 22.1 .0 .0 .0 .0 22.2 .0 .0 .0 .0 22.3 .0 .0 .0 .0 22.4 .0 .0 .0 .0 22.4 .0 .0 .0 .0 22.5 57.4 67.5 .0 .0 22.6 20.7 24.4 .0 .0 22.7 18.2 21.4 .0 .0 22.9 21.6 25.4 .0 .0 23.0 31.8 37.4 .0 .0 23.1 37.0 43.6 .0 .0 23.2 43.0 50.6 .0 .0 23.3 48.3 56.8 .0 .0 23.4 53.7 63.3 .0 .0 23.5 58.3 68.6 .0 .0 23.6 61.7 72.7 .0 .0 23.7 65.1 76.7 .0 .0 23.8 66.8 78.7 .0 .0 23.8 68.4 80.6 .0 .0 23.9 71.6 84.3 .0 .0 24.0 74.7 87.9 .0 .0 1 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14H (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 24.1 77.6 91.4 .0 .0 24.2 80.5 94.8 .0 .0 24.3 83.3 98.1 .0 .0 24.4 86.0 101.3 .0 .0 25.4 109.9 129.5 .0 .0 26.4 129.7 152.9 .0 .0 27.4 160.1 188.8 742.5 60.5 28.4 387.0 458.1 3989.5 437.1 28.5 389.5 461.0 4008.0 441.8 29.5 389.1 460.6 3995.0 446.1 30.5 382.5 452.7 3926.3 440.4 31.5 375.6 444.5 3854.3 434.0 32.5 368.7 436.3 3782.2 427.5 33.0 365.3 432.3 3746.6 424.3 33.5 362.0 428.3 3711.3 421.0 34.5 355.5 420.6 3642.1 414.7 35.5 349.3 413.2 3574.8 408.5 36.5 343.3 406.0 3509.6 402.4 37.5 337.5 399.2 3446.5 396.5 38.5 332.0 392.6 3385.3 390.8 39.2 328.2 388.1 3343.6 386.9 39.5 326.7 386.2 3326.0 385.3 40.5 321.5 380.1 3268.6 379.9 41.5 316.6 374.3 3213.0 374.7 42.5 311.8 368.6 3159.1 369.6 2 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14H (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 43.5 307.3 363.2 3106.7 364.7 44.5 302.8 357.9 3055.8 359.9 45.5 298.6 352.8 3006.4 355.3 46.3 295.3 348.9 2967.8 351.6 46.5 294.4 347.9 2958.3 350.7 47.5 290.5 343.2 2911.5 346.3 48.5 286.6 338.6 2866.0 342.0 49.5 282.8 334.1 2821.5 337.8 50.5 279.2 329.8 2778.2 333.7 51.5 275.7 325.6 2736.0 329.7 52.5 272.2 321.6 2694.7 325.7 53.5 268.9 317.6 2654.4 321.9 54.2 266.6 314.9 2626.7 319.2 54.5 265.7 313.8 2615.0 318.1 55.5 262.5 310.0 2576.4 314.4 56.5 259.4 306.4 2538.7 310.8 57.5 256.4 302.8 2501.8 307.2 58.5 253.5 299.3 2465.6 303.8 59.5 250.7 295.9 2430.1 300.3 60.5 247.9 292.6 2395.3 297.0 61.5 245.1 289.4 2361.2 293.7 62.5 242.5 286.2 2327.7 290.4 63.5 239.9 283.2 2294.9 287.2 64.5 149.2 175.9 430.0 95.7 3 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14H (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 65.5 148.9 175.5 430.6 95.5 66.5 148.6 175.2 431.2 95.4 67.5 148.3 174.8 431.8 95.2 68.5 148.0 174.5 432.5 95.1 69.5 147.7 174.1 433.1 94.9 70.5 147.4 173.8 433.8 94.8 71.5 147.1 173.4 434.4 94.6 72.5 146.8 173.1 435.1 94.5 73.5 146.5 172.7 435.7 94.3 73.6 146.5 172.7 435.8 94.3 74.5 146.2 172.4 436.4 94.2 75.5 145.9 172.0 437.0 94.1 76.5 145.6 171.7 437.7 93.9 77.5 145.4 171.3 438.3 93.8 78.5 145.1 171.0 439.0 93.6 79.5 144.8 170.6 439.6 93.5 80.5 144.5 170.3 440.3 93.3 81.5 144.2 169.9 440.9 93.2 82.5 143.9 169.6 441.6 93.1 84.5 143.3 168.9 442.9 92.8 86.5 142.7 168.2 444.2 92.5 88.5 142.1 167.5 445.5 92.2 90.5 141.5 166.8 446.9 91.9 92.5 140.9 166.1 448.2 91.7 94.5 140.3 165.4 449.5 91.4 4 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14H (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 96.5 139.7 164.7 450.9 91.1 97.0 139.6 164.5 451.2 91.0 98.5 139.1 164.0 452.2 90.8 100.5 138.5 163.3 453.6 90.6 102.5 137.9 162.6 455.0 90.3 104.5 137.3 161.8 456.3 90.0 106.5 136.7 161.1 457.7 89.7 108.5 136.1 160.4 459.1 89.5 110.5 135.4 159.6 460.5 89.2 112.5 134.8 158.9 461.9 88.9 114.5 134.2 158.1 463.3 88.7 116.5 133.6 157.4 464.7 88.4 118.5 132.9 156.6 466.1 88.1 120.5 132.3 155.9 467.5 87.8 122.1 131.8 155.3 468.7 87.6 122.5 131.6 155.1 468.9 87.5 124.5 131.0 154.4 470.3 87.3 126.5 130.3 153.6 471.7 87.0 128.5 129.7 152.8 473.1 86.7 130.5 129.0 152.0 474.5 86.4 132.5 128.4 151.3 475.9 86.1 134.5 127.7 150.5 477.3 85.9 136.5 127.0 149.7 478.7 85.6 138.5 126.4 148.9 480.1 85.3 140.5 125.7 148.1 481.5 85.0 5 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14H (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 142.5 125.0 147.3 482.9 84.7 144.5 124.4 146.5 484.3 84.4 146.5 123.7 145.7 485.7 84.1 148.5 123.0 144.9 487.1 83.9 149.6 122.6 144.5 487.8 83.7 150.5 122.3 144.1 488.4 83.6 152.5 121.6 143.3 489.8 83.3 154.5 120.9 142.5 491.2 83.0 156.5 120.2 141.7 492.6 82.7 158.5 119.6 140.9 494.0 82.4 160.5 118.9 140.1 495.4 82.1 162.5 118.2 139.2 496.8 81.8 164.5 117.5 138.4 498.2 81.5 166.5 116.8 137.6 499.5 81.3 168.5 116.1 136.7 500.9 81.0 170.5 115.4 135.9 502.3 80.7 172.5 114.6 135.1 503.7 80.4 174.5 113.9 134.2 505.1 80.1 176.5 113.2 133.4 506.4 79.8 178.5 112.5 132.6 507.8 79.5 179.8 112.1 132.0 508.7 79.3 180.5 111.8 131.7 509.2 79.2 182.5 111.3 131.1 510.2 79.2 184.5 110.8 130.5 511.1 79.1 186.5 110.3 129.9 512.0 79.1 6 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14H (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

Time (Thousand (Thousand (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 188.5 109.8 129.3 512.9 79.1 190.5 109.3 128.7 513.8 79.0 192.5 108.8 128.1 514.7 79.0 194.5 108.3 127.6 515.6 78.9 196.5 107.8 127.0 516.5 78.9 198.5 107.3 126.4 517.3 78.8 200.5 106.8 125.8 518.2 78.8 202.5 106.3 125.3 519.1 78.7 204.5 105.8 124.7 520.0 78.7 206.5 105.3 124.1 520.8 78.6 208.5 104.9 123.5 521.7 78.6 210.5 104.4 123.0 522.6 78.5 212.5 103.9 122.4 523.4 78.4 213.4 103.7 122.2 523.8 78.4

• Mass and Energy exiting the SG side of the break

• • Mass and Energy exiting the pump side of the break 7 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14I PRINCIPAL PARAMETERS DURING REFLOOD DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS Flooding Injection Core Total Accum Spill Time Temp Rate Carryover Height Downcomer Flow Enthalpy (sec) (°F) (in/sec) Fraction (ft) Height (ft) Frac (lbm/sec) (Btu/lbm) 21.4 169.0 .000 .000 .00 .00 .333 .0 .0 .0 .00 22.2 166.4 22.694 .000 .67 1.40 .000 6117.9 6117.9 .0 74.50 22.4 165.1 24.146 .000 1.07 1.32 .000 6067.6 6067.6 .0 74.50 22.7 164.5 2.530 .113 1.32 1.96 .298 5952.2 5952.2 .0 74.50 22.9 164.4 2.633 .130 1.35 2.40 .310 5934.5 5934.5 .0 74.50 23.0 164.4 2.595 .163 1.37 2.69 .361 5887.9 5887.9 .0 74.50 23.2 164.4 2.604 .205 1.41 3.34 .390 5842.1 5842.1 .0 74.50 23.8 164.4 2.494 .302 1.50 4.99 .427 5710.0 5710.0 .0 74.50 24.4 164.5 2.425 .386 1.59 6.83 .442 5574.6 5574.6 .0 74.50 28.4 164.9 4.389 .635 2.01 15.61 .670 4980.9 4363.1 .0 69.35 29.5 165.0 4.190 .667 2.15 15.62 .667 4799.2 4184.3 .0 69.18 33.0 165.5 3.747 .710 2.50 15.62 .659 4434.4 3806.6 .0 68.63 39.2 167.3 3.345 .732 3.00 15.62 .644 3937.5 3290.5 .0 67.68 46.3 169.9 3.057 .740 3.50 15.62 .628 3498.4 2835.5 .0 66.64 54.2 173.4 2.826 .742 4.00 15.62 .611 3107.7 2432.0 .0 65.48 63.5 177.6 2.615 .743 4.54 15.62 .594 2731.6 2044.9 .0 64.07 64.5 178.1 2.030 .732 4.59 15.62 .483 712.1 .0 .0 33.00 1 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14I (Cont.)

PRINCIPAL PARAMETERS DURING REFLOOD DOUBLE-ENDED PUMP SUCTION BREAK - MAXIMUM SAFEGUARDS Flooding Injection Core Total Accum Spill Time Temp Rate Carryover Height Downcomer Flow Enthalpy (sec) (°F) (in/sec) Fraction (ft) Height (ft) Frac (lbm/sec) (Btu/lbm) 73.6 182.7 1.988 .733 5.00 15.62 .484 712.1 .0 .0 33.00 86.5 190.9 1.930 .735 5.56 15.62 .484 712.1 .0 .0 33.00 97.0 198.6 1.882 .736 6.00 15.62 .485 712.1 .0 .0 33.00 110.5 208.9 1.819 .739 6.55 15.62 .485 712.1 .0 .0 33.00 122.1 217.3 1.764 .740 7.00 15.62 .485 712.1 .0 .0 33.00 136.5 226.4 1.696 .743 7.54 15.62 .486 712.1 .0 .0 33.00 149.6 233.6 1.633 .744 8.00 15.62 .486 712.1 .0 .0 33.00 164.5 240.7 1.562 .746 8.51 15.62 .485 712.2 .0 .0 33.00 179.8 246.9 1.489 .747 9.00 15.62 .485 712.2 .0 .0 33.00 196.5 252.8 1.424 .749 9.51 15.62 .486 712.2 .0 .0 33.00 213.4 258.0 1.361 .751 10.00 15.62 .488 712.2 .0 .0 33.00 2 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14J DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 213.4 123.8 156.1 589.5 94.4 218.4 123.5 155.6 589.8 94.3 223.4 123.1 155.2 590.2 94.2 228.4 122.8 154.8 590.5 94.1 233.4 123.5 155.6 589.9 93.7 238.4 123.1 155.2 590.2 93.6 243.4 122.8 154.7 590.6 93.6 248.4 122.4 154.3 590.9 93.5 253.4 123.1 155.1 590.3 93.1 258.4 122.7 154.6 590.6 93.0 263.4 122.3 154.2 591.0 92.9 268.4 123.0 155.0 590.3 92.6 273.4 122.6 154.5 590.7 92.5 278.4 122.3 154.1 591.1 92.4 283.4 121.9 153.6 591.4 92.3 288.4 122.5 154.4 590.8 92.0 293.4 122.2 154.0 591.2 91.9 298.4 121.8 153.5 591.5 91.8 303.4 121.4 153.0 591.9 91.7 308.4 122.1 153.8 591.3 91.3 313.4 121.7 153.4 591.7 91.2 318.4 121.3 152.9 592.0 91.2 323.4 121.9 153.6 591.4 90.8 1 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14J (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 328.4 121.5 153.2 591.8 90.7 333.4 121.2 152.7 592.2 90.6 338.4 121.8 153.4 591.6 90.3 343.4 121.4 153.0 592.0 90.2 348.4 121.0 152.5 592.3 90.1 353.4 120.6 152.0 592.7 90.0 358.4 121.2 152.7 592.2 89.7 363.4 120.8 152.2 592.5 89.6 368.4 120.4 151.7 592.9 89.5 373.4 121.0 152.4 592.4 89.1 378.4 120.6 151.9 592.8 89.1 383.4 120.2 151.4 593.2 89.0 388.4 120.7 152.1 592.6 88.6 393.4 120.3 151.6 593.0 88.5 398.4 119.9 151.1 593.4 88.4 403.4 120.5 151.9 592.8 88.1 408.4 120.2 151.5 593.1 88.0 413.4 120.0 151.2 593.4 87.9 418.4 119.7 150.8 593.7 90.4 423.4 120.3 151.6 593.0 90.1 428.4 120.0 151.3 593.3 89.9 433.4 119.7 150.9 593.6 89.8 438.4 119.4 150.5 593.9 89.7 443.4 120.1 151.3 593.3 89.3 2 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14J (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 448.4 119.8 150.9 593.6 89.2 453.4 119.5 150.5 593.9 89.0 458.4 120.1 151.3 593.3 88.7 463.4 119.7 150.9 593.6 88.6 468.4 119.4 150.5 593.9 88.4 473.4 119.1 150.1 594.2 88.3 478.4 119.7 150.9 593.6 87.9 483.4 119.4 150.5 593.9 87.8 488.4 119.1 150.1 594.3 87.7 493.4 119.6 150.8 593.7 87.3 498.4 119.3 150.4 594.0 87.2 503.4 119.0 149.9 594.4 87.1 508.4 119.5 150.6 593.8 86.7 513.4 119.2 150.2 594.1 86.6 518.4 118.9 149.8 594.5 86.5 523.4 119.4 150.4 594.0 86.1 528.4 119.0 150.0 594.3 86.0 533.4 118.7 149.6 594.6 88.4 538.4 119.2 150.2 594.1 88.0 543.4 118.8 149.8 594.5 87.9 548.4 118.5 149.3 594.8 87.8 553.4 119.0 149.9 594.4 87.4 558.4 118.6 149.5 594.7 87.3 563.4 118.2 149.0 595.1 87.1 568.4 118.7 149.6 594.6 86.8 3 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14J (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 573.4 118.3 149.1 595.0 86.6 578.4 118.8 149.7 594.6 86.3 583.4 118.4 149.2 595.0 86.1 588.4 118.0 148.7 595.3 86.0 593.4 118.4 149.2 594.9 85.7 598.4 118.0 148.7 595.3 85.5 603.4 118.4 149.3 594.9 85.2 608.4 118.1 148.8 595.3 85.0 613.4 117.7 148.3 595.6 84.9 618.4 118.1 148.9 595.2 87.0 623.4 117.7 148.4 595.6 86.8 628.4 118.1 148.9 595.2 86.5 633.4 117.7 148.4 595.6 86.3 638.4 118.1 148.8 595.2 86.0 643.4 117.7 148.3 595.6 85.8 648.4 118.1 148.8 595.3 85.5 653.4 117.6 148.3 595.7 85.3 658.4 118.0 148.7 595.4 85.0 663.4 117.5 148.1 595.8 84.8 668.4 117.8 148.5 595.5 84.5 673.4 117.4 147.9 596.0 84.3 678.4 117.7 148.3 595.7 84.0 683.4 117.2 147.7 596.1 86.2 688.4 117.4 148.0 595.9 85.9 693.4 117.7 148.3 595.7 85.5 4 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14J (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 698.4 117.2 147.7 596.2 85.4 703.4 117.4 147.9 596.0 85.0 708.4 117.6 148.1 595.8 84.7 713.4 117.0 147.5 596.3 84.6 718.4 117.2 147.7 596.2 84.2 723.4 117.3 147.8 596.0 83.9 728.4 117.4 147.9 595.9 83.6 733.4 116.8 147.2 596.5 83.5 738.4 116.9 147.3 596.5 85.4 743.4 116.9 147.3 596.4 85.1 748.4 116.9 147.4 596.4 84.8 753.4 116.9 147.4 596.4 84.5 758.4 116.9 147.3 596.5 84.2 763.4 116.8 147.2 596.5 83.9 768.4 116.7 147.1 596.6 83.6 773.4 116.6 146.9 596.7 83.3 778.4 116.4 146.7 596.9 83.0 783.4 116.8 147.2 596.5 84.8 788.4 116.6 146.9 596.8 84.6 793.4 116.3 146.5 597.1 84.3 798.4 116.5 146.8 596.8 83.9 803.4 116.7 147.1 596.6 83.5 808.4 116.3 146.5 597.1 83.3 813.4 116.3 146.6 597.0 82.9 818.4 116.3 146.6 597.0 82.6 5 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14J (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 823.4 66.8 84.1 646.6 97.4 1079.6 66.8 84.1 646.6 97.4 1079.7 65.7 82.4 647.7 93.9 1083.4 65.6 82.4 647.7 94.4 1390.6 65.6 82.4 647.7 94.4 1390.7 55.7 64.0 657.7 21.7 2782.0 47.1 54.1 666.3 22.0 2782.1 49.5 57.0 454.9 40.1 3600.0 45.7 52.6 458.7 40.4

• Mass and Energy exiting the SG side of the break

• • Mass and Energy exiting the pump side of the break 6 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14K DOUBLE-ENDED HOT-LEG BREAK MASS BALANCE Time (Sec)

.00 19.20 19.20*

Mass (Thousand lbm)

Initial In RCS and ACC 626.47 626.47 626.47 Added Mass Pumped Injection .00 .00 .00 Total Added .00 .00 .00 Total Available 626.47 626.47 626.47 Distribution Reactor Coolant 416.98 49.96 49.96 Accumulator 209.49 173.99 173.99 Total Contents 626.47 223.96 223.96 Effluent Break Flow .00 402.50 402.50 ECCS Spill .00 .00 .00 Total Effluent .00 402.50 402.50 Total Accountable 626.47 626.45 626.45

• This time is the bottom of core recovery time, which is identical to the end of blowdown time due to the assumption of instantaneous refill.

1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14L DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MINIMUM SAFEGUARDS Time (Sec)

.00 21.40(1) 21.40(2) 214.19(3) 1089.17(4) 1367.37(5) 3600.0(6)

Mass (Thousand lbm)

Initial In RCS & 626.47 626.47 626.47 626.47 626.47 626.47 626.47 Accumulator Added Mass Pumped .00 .00 .00 85.24 495.00 625.28 1694.38 Injection Total Added .00 .00 .00 85.24 495.00 625.28 1694.38 Total Available 626.47 626.47 626.47 711.71 1121.47 1251.76 2320.85 Distribution Reactor 416.98 40.95 66.41 117.33 117.33 117.33 117.33 Coolant Accumulator 209.49 167.73 142.28 .00 .00 .00 .00 Total 626.47 208.68 208.68 117.33 117.33 117.33 117.33 Contents Effluent Break Flow .00 417.78 417.78 585.53 995.29 1125.57 2194.67 ECCS Spill .00 .00 .00 .00 .00 .00 .00 Total .00 417.78 417.78 585.53 995.29 1125.57 2194.67 Effluent Total Accountable 626.47 626.46 626.46 702.86 1112.62 1242.91 2312.00 Notes:

(1) End of Blowdown (2) Bottom of core recovery time, which is identical to the end of blowdown time due to the assumption of instantaneous refill.

(3) End of Reload (4) Time at which the Broken Loop SG equilibrates at the first intermediate pressure (5) Time at which the Intact Loop SG equilibrates at the second intermediate pressure (6) Time at which both SGs equilibrate to 14.7 psia 1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14M DOUBLE-ENDED PUMP SUCTION BREAK MASS BALANCE MAXIMUM SAFEGUARDS Time (Sec)

.00 21.40(1) 21.40(2) 213.36(3) 1079.74(4) 1390.62(5) 3600.00(6)

Mass (Thousand lbm)

Initial In RCS & 626.47 626.47 626.47 626.47 626.47 626.47 626.47 Accumulator Added Mass Pumped .00 .00 .00 130.69 748.68 970.44 2375.56 Injection Total Added .00 .00 .00 130.69 748.68 970.44 2375.56 Total Available 626.47 626.47 626.47 757.16 1375.15 1596.91 3002.04 Distribution Reactor 416.98 40.95 66.41 117.53 117.53 117.53 117.53 Coolant Accumulator 209.49 167.73 142.28 .00 .00 .00 .00 Total Contents 626.47 208.68 208.68 117.53 117.53 117.53 117.53 Effluent Break Flow .00 417.78 417.78 630.78 1248.77 1470.53 2875.67 ECCS Spill .00 .00 .00 .00 .00 .00 .00 Total Effluent .00 417.78 417.78 630.78 1248.77 1470.53 2875.67 Total Accountable 626.47 626.46 626.46 748.31 1366.30 1588.06 2993.20 Notes:

(1) End of Blowdown (2) Bottom of core recovery time, which is identical to the end of blowdown time due to the assumption of instantaneous refill (3) End of Reflood (4) Time at which the Broken Loop SG equilibrates at the first intermediate pressure.

(5) Time at which the Intact Loop SG equilibrates at the second intermediate pressure.

(6) Time at which both SGs equilibrate to 14.7 psia.

1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14N DOUBLE-ENDED HOT-LEG BREAK ENERGY BALANCE Time (Sec)

.00 19.20 19.20*

Energy (Million Btu)

Initial Energy In RCS, Acc, SG 675.94 675.94 675.94 Added Energy Pumped Injection .00 .00 .00 Decay Heat .00 5.77 5.77 Heat From Secondary .00 -.35 -.35 Total Added .00 5.42 5.42 Total Available 675.94 681.36 681.36 Distribution Reactor Coolant 245.25 11.99 11.99 Accumulator 15.62 12.97 12.97 Core Stored 22.87 9.19 9.19 Primary Metal 115.85 108.47 108.47 Secondary Metal 69.35 69.00 69.00 Steam Generator 207.00 205.72 205.72 Total Contents 675.94 417.34 417.34 Effluent Break Flow .00 263.52 263.52 ECCS Spill .00 .00 .00 Total Effluent .00 263.52 263.52 Total Accountable 675.94 680.86 680.86

• This time is the bottom of core recovery time, which is identical to the end of blowdown time due to the assumption of instantaneous refill.

1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14O DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE MINIMUM SAFEGUARDS Time (Sec)

.00 21.40(1) 21.40(2) 214.19 (3) 1089.17(4) 1367.37(5) 3600.00(6)

Energy (Million Btu)

Initial In RCS, Acc, SG 640.99 640.99 640.99 640.99 640.99 640.99 640.99 Energy Added Energy Pumped .00 .00 .00 2.81 16.33 20.63 74.05 Injection Decay Heat .00 5.79 5.79 24.82 85.54 101.78 209.46 Heat From .00 .43 .43 .43 9.95 10.04 10.04 Secondary Total Added .00 6.22 6.22 28.06 111.83 132.46 293.56 Total Available 640.99 647.21 647.21 669.04 752.81 773.45 934.54 Distribution Reactor Coolant 245.25 8.59 10.49 29.17 29.17 29.17 29.17 Accumulator 15.62 12.51 10.61 .00 .00 .00 .00 Core Stored 22.87 12.50 12.50 3.91 3.17 3.12 2.71 Primary Metal 115.85 109.61 109.61 88.92 50.61 45.22 40.05 Secondary Metal 34.40 34.82 34.82 31.86 19.12 16.34 14.61 Steam Generator 207.00 210.21 210.21 188.90 116.55 100.61 90.81 Total Contents 640.99 388.24 388.24 342.76 218.61 194.45 177.35 Effluent Break Flow .00 258.48 258.48 318.23 526.15 549.53 730.23 ECCS Spill .00 .00 .00 .00 .00 .00 .00 Total Effluent .00 258.48 258.48 318.23 526.15 549.53 730.23 Total Accountable 640.99 646.72 646.72 660.99 744.76 743.98 907.58 Notes:

(1) End of Blowdown (2) Bottom of core recovery time. This time is identical to the end of blowdown time due to the assumption of instantaneous refill.

(3) End of Reload (4) Time at which the Broken Loop SG equilibrates at the first intermediate pressure (5) Time at which the Intact Loop SG equilibrates at the second intermediate pressure (6) Time at which both SGs equilibrate to 14.7 psia 1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14P DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE - MAXIMUM SAFEGUARDS Time (sec)

.00 21.40(1) 21.40(2) 213.36(3) 1079.74(4) 1390.62(5) 3600.00(6)

Energy (Million Btu)

Initial Energy In RCS, Acc, SG 640.99 640.99 640.99 640.99 640.99 640.99 640.99 Added Energy Pumped Injection .00 .00 .00 4.31 24.71 32.02 101.15 Decay Heat .00 5.79 5.79 24.75 84.97 103.09 209.42 Heat From .00 .43 .43 .43 9.86 9.93 9.93 Secondary Total Added .00 6.22 6.22 29.49 119.54 145.04 320.50 Total Available 640.99 647.21 647.21 670.47 760.52 786.02 961.49 Distribution Reactor Coolant 245.25 8.59 10.49 29.17 29.17 29.17 29.17 Accumulator 15.62 12.51 10.61 .00 .00 .00 .00 Core Stored 22.87 12.50 12.50 3.91 3.10 3.05 2.71 Primary Metal 115.85 109.61 109.61 88.61 49.94 43.94 39.96 Secondary Metal 34.40 34.82 34.82 31.84 18.94 15.82 14.56 Steam Generator 207.00 210.21 210.21 188.76 115.44 97.55 90.42 Total Contents 640.99 388.24 388.24 342.31 216.59 189.53 176.82 1 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14P (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK ENERGY BALANCE - MAXIMUM SAFEGUARDS Time (sec)

.00 21.40(1) 21.40(2) 213.36(3) 1079.74(4) 1390.62(5) 3600.00(6)

Energy (Million Btu)

Effluent Break Flow .00 258.48 258.48 320.12 535.87 566.71 757.09 ECCS Spill .00 .00 .00 .00 .00 .00 .00 Total Effluent .00 258.48 258.48 320.12 535.87 566.71 757.09 Total Accountable 640.99 646.72 646.72 662.42 752.47 756.25 933.91 Notes:

(1) End of Blowdown (2) Bottom of core recovery time. This time is identical to the end of blowdown time due to the assumption of instantaneous refill.

(3) End of Reflood (4) Time at which the Broken Loop SG equilibrates at the first intermediate pressure.

(5) Time at which the Intact Loop SG equilibrates at the second intermediate pressure.

(6) Time at which both SGs equilibrate to 14.7 psia.

2 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14Q DOUBLE-ENDED PUMP SUCTION BREAK (SW FAILURE) - MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 21.4 .0 .0 .0 .0 21.9 .0 .0 .0 .0 22.1 .0 .0 .0 .0 22.2 .0 .0 .0 .0 22.3 .0 .0 .0 .0 22.4 .0 .0 .0 .0 22.4 .0 .0 .0 .0 22.5 57.4 67.5 .0 .0 22.6 20.7 24.4 .0 .0 22.7 18.2 21.4 .0 .0 22.9 21.6 25.4 .0 .0 23.0 31.8 37.4 .0 .0 23.1 37.0 43.6 .0 .0 23.2 43.0 50.6 .0 .0 23.3 48.3 56.8 .0 .0 23.4 53.7 63.3 .0 .0 23.5 58.3 68.6 .0 .0 23.6 61.7 72.7 .0 .0 23.7 65.1 76.7 .0 .0 23.8 66.8 78.7 .0 .0 23.8 68.4 80.6 .0 .0 23.9 71.6 84.3 .0 .0 24.0 74.7 87.9 .0 .0 24.1 77.6 91.4 .0 .0 24.2 80.5 94.8 .0 .0 1 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14Q (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK (SW FAILURE) - MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 24.3 83.3 98.1 .0 .0 24.4 86.0 101.3 .0 .0 25.4 109.9 129.5 .0 .0 26.4 129.7 152.9 .0 .0 27.4 160.1 188.8 742.5 60.5 28.4 387.0 458.1 3989.5 437.1 28.5 389.5 461.0 4008.0 441.8 29.5 389.1 460.6 3995.0 446.1 30.5 382.5 452.7 3926.3 440.4 31.5 375.6 444.5 3854.3 434.0 32.5 368.7 436.3 3782.2 427.5 33.0 365.3 432.3 3746.6 424.3 33.5 362.0 428.3 3711.3 421.0 34.5 355.5 420.6 3642.1 414.7 35.5 349.3 413.2 3574.8 408.5 36.5 343.3 406.0 3509.6 402.4 37.5 337.5 399.2 3446.5 396.5 38.5 332.0 392.6 3385.3 390.8 39.2 328.2 388.1 3343.6 386.9 39.5 326.7 386.2 3326.0 385.3 40.5 321.5 380.1 3268.6 379.9 41.5 316.6 374.3 3213.0 374.7 42.5 311.8 368.6 3159.1 369.6 2 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14Q (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK (SW FAILURE) - MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 43.5 307.3 363.2 3106.7 364.7 44.5 302.8 357.9 3055.8 359.9 45.5 298.6 352.8 3006.4 355.3 46.3 295.3 348.9 2967.8 351.6 46.5 294.4 347.9 2958.3 350.7 47.5 290.5 343.2 2911.5 346.3 48.5 286.6 338.6 2866.0 342.0 49.5 282.8 334.1 2821.5 337.8 50.5 279.2 329.8 2778.2 333.7 51.5 275.7 325.6 2736.0 329.7 52.5 272.2 321.6 2694.7 325.7 53.5 268.9 317.6 2654.4 321.9 54.2 266.6 314.9 2626.7 319.2 54.5 265.7 313.8 2615.0 318.1 55.5 262.5 310.0 2576.4 314.4 56.5 259.4 306.4 2538.7 310.8 57.5 256.4 302.8 2501.8 307.2 58.5 253.5 299.3 2465.6 303.8 59.5 250.7 295.9 2430.1 300.3 60.5 247.9 292.6 2395.3 297.0 61.5 245.1 289.4 2361.2 293.7 62.5 242.5 286.2 2327.7 290.4 63.5 239.9 283.2 2294.9 287.2 64.5 149.2 175.9 430.0 95.7 3 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14Q (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK (SW FAILURE) - MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 65.5 148.9 175.5 430.6 95.5 66.5 148.6 175.2 431.2 95.4 67.5 148.3 174.8 431.8 95.2 68.5 148.0 174.5 432.5 95.1 69.5 147.7 174.1 433.1 94.9 70.5 147.4 173.8 433.8 94.8 71.5 147.1 173.4 434.4 94.6 72.5 146.8 173.1 435.1 94.5 73.5 146.5 172.7 435.7 94.3 73.6 146.5 172.7 435.8 94.3 74.5 146.2 172.4 436.4 94.2 75.5 145.9 172.0 437.0 94.1 76.5 145.6 171.7 437.7 93.9 77.5 145.4 171.3 438.3 93.8 78.5 145.1 171.0 439.0 93.6 79.5 144.8 170.6 439.6 93.5 80.5 144.5 170.3 440.3 93.3 81.5 144.2 169.9 440.9 93.2 82.5 143.9 169.6 441.6 93.1 84.5 143.3 168.9 442.9 92.8 86.5 142.7 168.2 444.2 92.5 88.5 142.1 167.5 445.5 92.2 90.5 141.5 166.8 446.9 91.9 92.5 140.9 166.1 448.2 91.7 4 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14Q (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK (SW FAILURE) - MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 94.5 140.3 165.4 449.5 91.4 96.5 139.7 164.7 450.9 91.1 97.0 139.6 164.5 451.2 91.0 98.5 139.1 164.0 452.2 90.8 100.5 138.5 163.3 453.6 90.6 102.5 137.9 162.6 455.0 90.3 104.5 137.3 161.8 456.3 90.0 106.5 136.7 161.1 457.7 89.7 108.5 136.1 160.4 459.1 89.5 110.5 135.4 159.6 460.5 89.2 112.5 134.8 158.9 461.9 88.9 114.5 134.2 158.1 463.3 88.7 116.5 133.6 157.4 464.7 88.4 118.5 132.9 156.6 466.1 88.1 120.5 132.3 155.9 467.5 87.8 122.1 131.8 155.3 468.7 87.6 122.5 131.6 155.1 468.9 87.5 124.5 131.0 154.4 470.3 87.3 126.5 130.3 153.6 471.7 87.0 128.5 129.7 152.8 473.1 86.7 130.5 129.0 152.0 474.5 86.4 132.5 128.4 151.3 475.9 86.1 134.5 127.7 150.5 477.3 85.9 136.5 127.0 149.7 478.7 85.6 5 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14Q (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK (SW FAILURE) - MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 138.5 126.4 148.9 480.1 85.3 140.5 125.7 148.1 481.5 85.0 142.5 125.0 147.3 482.9 84.7 144.5 124.4 146.5 484.3 84.4 146.5 123.7 145.7 485.7 84.1 148.5 123.0 144.9 487.1 83.9 149.6 122.6 144.5 487.8 83.7 150.5 122.3 144.1 488.4 83.6 152.5 121.6 143.3 489.8 83.3 154.5 120.9 142.5 491.2 83.0 156.5 120.2 141.7 492.6 82.7 158.5 119.6 140.9 494.0 82.4 160.5 118.9 140.1 495.4 82.1 162.5 118.2 139.2 496.8 81.8 164.5 117.5 138.4 498.2 81.5 166.5 116.8 137.6 499.5 81.3 168.5 116.1 136.7 500.9 81.0 170.5 115.4 135.9 502.3 80.7 172.5 114.6 135.1 503.7 80.4 174.5 113.9 134.2 505.1 80.1 176.5 113.2 133.4 506.4 79.8 178.5 112.5 132.6 507.8 79.5 179.8 112.1 132.0 508.7 79.3 180.5 111.8 131.7 509.2 79.2 6 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14Q (Cont.)

DOUBLE-ENDED PUMP SUCTION BREAK (SW FAILURE) - MAXIMUM SAFEGUARDS REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 182.5 111.3 131.1 510.2 79.2 184.5 110.8 130.5 511.1 79.1 186.5 110.3 129.9 512.0 79.1 188.5 109.8 129.3 512.9 79.1 190.5 109.3 128.7 513.8 79.0 192.5 108.8 128.1 514.7 79.0 194.5 108.3 127.6 515.6 78.9 196.5 107.8 127.0 516.5 78.9 198.5 107.3 126.4 517.3 78.8 200.5 106.8 125.8 518.2 78.8 202.5 106.3 125.3 519.1 78.7 204.5 105.8 124.7 520.0 78.7 206.5 105.3 124.1 520.8 78.6 208.5 104.9 123.5 521.7 78.6 210.5 104.4 123.0 522.6 78.5 212.5 103.9 122.4 523.4 78.4 213.4 103.7 122.2 523.8 78.4

• Mass and Energy exiting the SG side of break

• • Mass and Energy exiting the pump side of break 7 of 7

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14R DOUBLE-ENDED PUMP SUCTION (SW FAILURE) - MAXIMUM SAFEGUARDS PRINCIPAL PARAMETERS DURING REFLOOD Flooding Injection Core Total Accum Spill Time Temp Rate Carryover Height Downcomer Flow Enthalpy (sec) (°F) (in/sec) Fraction (ft) Height (ft) Frac (lbm/sec) (Btu/lbm) 21.4 169.0 .000 .000 .00 .00 .333 .0 .0 .0 .00 22.2 166.4 22.694 .000 .67 1.40 .000 6117.9 6117.9 .0 74.50 22.4 165.1 24.146 .000 1.07 1.32 .000 6067.6 6067.6 .0 74.50 22.7 164.5 2.530 .113 1.32 1.96 .298 5952.2 5952.2 .0 74.50 22.9 164.4 2.633 .130 1.35 2.40 .310 5934.5 5934.5 .0 74.50 23.0 164.4 2.595 .163 1.37 2.69 .361 5887.9 5887.9 .0 74.50 23.2 164.4 2.604 .205 1.41 3.34 .390 5842.1 5842.1 .0 74.50 23.8 164.4 2.494 .302 1.50 4.99 .427 5710.0 5710.0 .0 74.50 24.4 164.5 2.425 .386 1.59 6.83 .442 5574.6 5574.6 .0 74.50 28.4 164.9 4.389 .635 2.01 15.61 .670 4980.9 4363.1 .0 69.35 29.5 165.0 4.190 .667 2.15 15.62 .667 4799.2 4184.3 .0 69.18 33.0 165.5 3.747 .710 2.50 15.62 .659 4434.4 3806.6 .0 68.63 39.2 167.3 3.345 .732 3.00 15.62 .644 3937.5 3290.5 .0 67.68 46.3 169.9 3.057 .740 3.50 15.62 .628 3498.4 2835.5 .0 66.64 54.2 173.4 2.826 .742 4.00 15.62 .611 3107.7 2432.0 .0 65.48 1 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14R (Cont.)

DOUBLE-ENDED PUMP SUCTION (SW FAILURE) - MAXIMUM SAFEGUARDS PRINCIPAL PARAMETERS DURING REFLOOD Flooding Injection Core Total Accum Spill Time Temp Rate Carryover Height Downcomer Flow Enthalpy (sec) (°F) (in/sec) Fraction (ft) Height (ft) Frac (lbm/sec) (Btu/lbm) 63.5 177.6 2.615 .743 4.54 15.62 .594 2731.6 2044.9 .0 64.07 64.5 178.1 2.030 .732 4.59 15.62 .483 712.1 .0 .0 33.00 73.6 182.7 1.988 .733 5.00 15.62 .484 712.1 .0 .0 33.00 86.5 190.9 1.930 .735 5.56 15.62 .484 712.1 .0 .0 33.00 97.0 198.6 1.882 .736 6.00 15.62 .485 712.1 .0 .0 33.00 110.5 208.9 1.819 .739 6.55 15.62 .485 712.1 .0 .0 33.00 122.1 217.3 1.764 .740 7.00 15.62 .485 712.1 .0 .0 33.00 136.5 226.4 1.696 .743 7.54 15.62 .486 712.1 .0 .0 33.00 149.6 233.6 1.633 .744 8.00 15.62 .486 712.1 .0 .0 33.00 164.5 240.7 1.562 .746 8.51 15.62 .485 712.2 .0 .0 33.00 179.8 246.9 1.489 .747 9.00 15.62 .485 712.2 .0 .0 33.00 213.4 258.0 1.361 .751 10.00 15.62 .488 712.2 .0 .0 33.00 2 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14S DOUBLE-ENDED PUMP SUCTION (SW FAILURE) MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 213.4 123.8 156.1 589.5 94.4 218.4 123.5 155.6 589.8 94.3 223.4 123.1 155.2 590.2 94.2 228.4 122.8 154.8 590.5 94.1 233.4 123.5 155.6 589.9 93.7 238.4 123.1 155.2 590.2 93.6 243.4 122.8 154.7 590.6 93.6 248.4 122.4 154.3 590.9 93.5 253.4 123.1 155.1 590.3 93.1 258.4 122.7 154.6 590.6 93.0 263.4 122.3 154.2 591.0 92.9 268.4 123.0 155.0 590.3 92.6 273.4 122.6 154.5 590.7 92.5 278.4 122.3 154.1 591.1 92.4 283.4 121.9 153.6 591.4 92.3 288.4 122.5 154.4 590.8 92.0 293.4 122.2 154.0 591.2 91.9 298.4 121.8 153.5 591.5 91.8 303.4 121.4 153.0 591.9 91.7 308.4 122.1 153.8 591.3 91.3 313.4 121.7 153.4 591.7 91.2 318.4 121.3 152.9 592.0 91.2 323.4 121.9 153.6 591.4 90.8 328.4 121.5 153.2 591.8 90.7 333.4 121.2 152.7 592.2 90.6 1 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14S (Cont.)

DOUBLE-ENDED PUMP SUCTION (SW FAILURE) MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 338.4 121.8 153.4 591.6 90.3 343.4 121.4 153.0 592.0 90.2 348.4 121.0 152.5 592.3 90.1 353.4 120.6 152.0 592.7 90.0 358.4 121.2 152.7 592.2 89.7 363.4 120.8 152.2 592.5 89.6 368.4 120.4 151.7 592.9 89.5 373.4 121.0 152.4 592.4 89.1 378.4 120.6 151.9 592.8 89.1 383.4 120.2 151.4 593.2 89.0 388.4 120.7 152.1 592.6 88.6 393.4 120.3 151.6 593.0 88.5 398.4 119.9 151.1 593.4 88.4 403.4 120.5 151.9 592.8 88.1 408.4 120.2 151.5 593.1 88.0 413.4 120.0 151.2 593.4 87.9 418.4 119.7 150.8 593.7 90.4 423.4 120.3 151.6 593.0 90.1 428.4 120.0 151.3 593.3 89.9 433.4 119.7 150.9 593.6 89.8 438.4 119.4 150.5 593.9 89.7 443.4 120.1 151.3 593.3 89.3 448.4 119.8 150.9 593.6 89.2 453.4 119.5 150.5 593.9 89.0 2 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14S (Cont.)

DOUBLE-ENDED PUMP SUCTION (SW FAILURE) MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 458.4 120.1 151.3 593.3 88.7 463.4 119.7 150.9 593.6 88.6 468.4 119.4 150.5 593.9 88.4 473.4 119.1 150.1 594.2 88.3 478.4 119.7 150.9 593.6 87.9 483.4 119.4 150.5 593.9 87.8 488.4 119.1 150.1 594.3 87.7 493.4 119.6 150.8 593.7 87.3 498.4 119.3 150.4 594.0 87.2 503.4 119.0 149.9 594.4 87.1 508.4 119.5 150.6 593.8 86.7 513.4 119.2 150.2 594.1 86.6 518.4 118.9 149.8 594.5 86.5 523.4 119.4 150.4 594.0 86.1 528.4 119.0 150.0 594.3 86.0 533.4 118.7 149.6 594.6 88.4 538.4 119.2 150.2 594.1 88.0 543.4 118.8 149.8 594.5 87.9 548.4 118.5 149.3 594.8 87.8 553.4 119.0 149.9 594.4 87.4 558.4 118.6 149.5 594.7 87.3 563.4 118.2 149.0 595.1 87.1 568.4 118.7 149.6 594.6 86.8 573.4 118.3 149.1 595.0 86.6 3 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14S (Cont.)

DOUBLE-ENDED PUMP SUCTION (SW FAILURE) MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 578.4 118.8 149.7 594.6 86.3 583.4 118.4 149.2 595.0 86.1 588.4 118.0 148.7 595.3 86.0 593.4 118.4 149.2 594.9 85.7 598.4 118.0 148.7 595.3 85.5 603.4 118.4 149.3 594.9 85.2 608.4 118.1 148.8 595.3 85.0 613.4 117.7 148.3 595.6 84.9 618.4 118.1 148.9 595.2 87.0 623.4 117.7 148.4 595.6 86.8 628.4 118.1 148.9 595.2 86.5 633.4 117.7 148.4 595.6 86.3 638.4 118.1 148.8 595.2 86.0 643.4 117.7 148.3 595.6 85.8 648.4 118.1 148.8 595.3 85.5 653.4 117.6 148.3 595.7 85.3 658.4 118.0 148.7 595.4 85.0 663.4 117.5 148.1 595.8 84.8 668.4 117.8 148.5 595.5 84.5 673.4 117.4 147.9 596.0 84.3 678.4 117.7 148.3 595.7 84.0 683.4 117.2 147.7 596.1 86.2 688.4 117.4 148.0 595.9 85.9 693.4 117.7 148.3 595.7 85.5 4 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14S (Cont.)

DOUBLE-ENDED PUMP SUCTION (SW FAILURE) MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 698.4 117.2 147.7 596.2 85.4 703.4 117.4 147.9 596.0 85.0 708.4 117.6 148.1 595.8 84.7 713.4 117.0 147.5 596.3 84.6 718.4 117.2 147.7 596.2 84.2 723.4 117.3 147.8 596.0 83.9 728.4 117.4 147.9 595.9 83.6 733.4 116.8 147.2 596.5 83.5 738.4 116.9 147.3 596.5 85.4 743.4 116.9 147.3 596.4 85.1 748.4 116.9 147.4 596.4 84.8 753.4 116.9 147.4 596.4 84.5 758.4 116.9 147.3 596.5 84.2 763.4 116.8 147.2 596.5 83.9 768.4 116.7 147.1 596.6 83.6 773.4 116.6 146.9 596.7 83.3 778.4 116.4 146.7 596.9 83.0 783.4 116.8 147.2 596.5 84.8 788.4 116.6 146.9 596.8 84.6 793.4 116.3 146.5 597.1 84.3 798.4 116.5 146.8 596.8 83.9 803.4 116.7 147.1 596.6 83.5 808.4 116.3 146.5 597.1 83.3 5 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14S (Cont.)

DOUBLE-ENDED PUMP SUCTION (SW FAILURE) MAXIMUM SAFEGUARDS POST-REFLOOD MASS AND ENERGY RELEASES Break Path No. 1 Flow* Break Path No. 2 Flow**

(Thousand (Thousand Time (sec) (lbm/sec) Btu/sec) (lbm/sec) Btu/sec) 813.4 116.3 146.6 597.0 82.9 818.4 116.3 146.6 597.0 82.6 823.4 66.8 84.1 646.6 97.4 1075.0 66.8 84.1 646.6 97.4 1075.1 64.8 81.5 648.5 93.4 1078.4 64.8 81.4 648.6 93.9 1403.5 64.8 81.4 648.6 93.9 1403.6 54.3 62.5 659.0 21.7 2128.0 48.7 56.0 664.6 21.9 2128.1 52.7 60.7 781.3 92.2 3600.0 45.7 52.6 788.3 93.0

• Mass and Energy exiting the SG side of break

• • Mass and Energy exiting the pump side of break 6 of 6

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14T DOUBLE-ENDED PUMP SUCTION (SW FAILURE) MASS BALANCE MAXIMUM SAFEGUARDS Time (Sec)

.00 21.40(1) 21.40(2) 213.36(3) 1075.07(4) 1403.54(5) 3600.00(6)

Mass (Thousand lbm)

Initial In RCS & 626.47 626.47 626.47 626.47 626.47 626.47 626.47 Accumulator Added Mass Pumped .00 .00 .00 130.69 745.35 979.66 2724.09 Injection Total Added .00 .00 .00 130.69 745.35 979.66 2724.09 Total Available 626.47 626.47 626.47 757.16 1371.83 1606.13 3350.56 Distribution Reactor 416.98 40.95 66.41 117.53 117.53 117.53 117.53 Coolant Accumulator 209.49 167.73 142.28 .00 .00 .00 .00 Total 626.47 208.68 208.68 117.53 117.53 117.53 117.53 Contents Effluent Break Flow .00 417.78 417.78 630.78 1245.45 1479.75 3224.18 ECCS Spill .00 .00 .00 .00 .00 .00 .00 Total .00 417.78 417.78 630.78 1245.45 1479.75 3224.18 Effluent Total Accountable 626.47 626.46 626.46 748.31 1362.97 1597.28 3341.71 Notes:

(1) End of Blowdown (2) Bottom of core recovery time. This time is identical to the end of blowdown time due to the assumption of instantaneous refill.

(3) End of Reflood (4) Time at which the Broken Loop SG equilibrates at the first intermediate pressure.

(5) Time at which the Intact Loop SG equilibrates at the second intermediate pressure.

(6) Time at which both SGs equilibrate to 14.7 psia.

1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14U DOUBLE-ENDED PUMP SUCTION (SW FAILURE) ENERGY BALANCE MAXIMUM SAFEGUARDS Time (Sec)

.00 21.40(1) 21.40(2) 213.36(3) 1075.07(4) 1403.54(5) 3600.00(6)

Energy (Million Btu)

Initial Energy In RCS, 640.99 640.99 640.99 640.99 640.99 640.99 640.99 Acc, SG Added Energy Pumped .00 .00 .00 4.31 24.60 32.33 194.25 Injection Decay Heat .00 5.79 5.79 24.75 84.69 103.81 209.42 Heat From .00 .43 .43 .43 9.81 9.93 9.93 Secondary Total Added .00 6.22 6.22 29.49 119.09 146.07 413.59 Total Available 640.99 647.21 647.21 670.47 760.08 787.05 1054.58 Distribution Reactor 245.25 8.59 10.49 29.17 29.17 29.17 29.17 Coolant Accumulator 15.62 12.51 10.61 .00 .00 .00 .00 Core Stored 22.87 12.50 12.50 3.91 2.98 2.95 2.71 Primary 115.85 109.61 109.61 88.61 49.14 42.92 39.96 Metal Secondary 34.40 34.82 34.82 31.84 18.80 15.53 14.56 Metal Steam 207.00 210.21 210.21 188.76 114.65 95.92 90.42 Generator Total 640.99 388.24 388.24 342.31 214.75 186.49 176.82 Contents Effluent Break Flow .00 258.48 258.48 320.12 537.28 568.50 846.95 ECCS Spill .00 .00 .00 .00 .00 .00 .00 Total .00 258.48 258.48 320.12 537.28 568.50 846.95 Effluent Total Accountable 640.99 646.72 646.72 662.42 752.03 754.99 1023.77 Notes:

(1) End of Blowdown (2) Bottom of core recovery time. This time is identical to the end of blowdown time due to the assumption of instantaneous refill.

(3) End of Reflood (4) Time at which the Broken Loop SG equilibrates at the first intermediate pressure.

(5) Time at which the Intact Loop SG equilibrates at the second intermediate pressure.

(6) Time at which both SGs equilibrate to 14.7 psia.

1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-14V LOCA MASS AND ENERGY RELEASE ANALYSIS ANS 1979 CORE DECAY HEAT POWER FRACTION Time (sec) ANS 1979 Decay Heat Fraction 10 0.053876 15 0.050401 20 0.048018 40 0.042401 60 0.039244 80 0.037065 100 0.035466 150 0.032724 200 0.030936 400 0.027078 600 0.024931 800 0.023389 1000 0.022156 1500 0.019921 2000 0.018315 4000 0.014781 6000 0.013040 8000 0.012000 10000 0.011262 15000 0.010097 20000 0.009350 40000 0.007778 60000 0.006958 80000 0.006424 100000 0.006021 150000 0.005323 400000 0.003770 600000 0.003201 800000 0.002834 1000000 0.002580 2592000 0.001745 1 of 1

BVPS-2 UFSAR Rev. 0 TABLE 6.2-21 PRESSURIZER SUBCOMPARTMENT VENT PATH DESCRIPTION Head Loss Coefficient K Vent From To Geom. Grating Path Node Node Area Inertia Expansion + Contraction* Loss Total No. No. No. ( ft-2 ) (ft-1) Fwd Rvs Factor Fwd Rvs 1 1 2 59.20 .146 .283 .270 .283 .270 2 2 3 120.00 .109 .135 .080 .135 .080 3 3 4 123.80 .059 .152 .089 .152 .089 4 4 1 55.40 .144 .201 .203 .201 .203 5 1 19 16.00 .115 2.450 2.064 2.450 2.064 6 4 19 16.00 .098 2.530 2.233 2.530 2.230 7 5 1 11.21 .523 .457 .457 .204 .661 .661 8 6 2 15.91 .362 .471 .471 .204 .675 .675 9 6 2 84.12 .068 .511 .511 .511 .511 10 7 3 39.59 .126 .580 .580 .204 .784 .784 11 7 3 41.54 .120 .620 .620 .620 .620 12 8 4 27.75 .205 .479 .479 .204 .683 .683 13 5 19 37.50 .066 2.331 1.684 2.331 1.684 14 6 19 17.03 .086 2.562 2.406 2.562 2.406 15 8 19 20.76 .075 2.543 2.253 .260 2.803 2.513 16 5 6 43.29 .109 .754 .688 .754 .688 17 6 7 125.64 .081 .271 .283 .271 .283 18 7 8 101.57 .044 .448 .505 .448 .505 19 5 8 40.76 .107 .661 .647 .661 .647 20 9 5 1.74 2.609 1.810 1.790 1.810 1.790 21 10 6 2.22 1.864 1.850 1.840 1.850 1.840 22 11 7 1.40 2.896 1.870 1.870 1.870 1.870 23 12 8 1.62 2.634 1.830 1.830 1.830 1.830 24 10 19 35.10 .100 1.906 1.385 .260 2.166 1.645 25 9 10 39.78 .081 .979 .903 .979 .903 26 10 11 190.86 .047 .361 .393 .361 .393 27 11 12 184.29 .025 .417 .468 .417 .468 28 12 9 70.43 .046 .690 .711 .690 .711 29 12 13 273.70 .041 .079 .052 .079 .052 30 13 14 156.60 .066 .398 .373 .398 .373 31 14 19 66.00 .086 2.400 1.971 2.400 1.971 32 14 19 25.91 .108 2.442 2.136 .260 2.702 2.396 33 20 18 8.89 .213 .445 .715 .445 .715 34 21 18 11.32 .149 .511 .942 .511 .942 35 22 18 7.16 .231 .525 .997 .525 .997 36 23 18 8.25 .212 .495 .887 .495 .887 1 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-21 (Cont)

Head Loss Coefficient K Vent From To Geom. Grating Path Node Node Area Inertia Expansion + Contraction* Loss Total No. No. No. ( ft-2 ) (ft-1) Fwd Rvs Factor Fwd Rvs 37 20 9 23.32 .285 .496 .496 .496 .496 38 21 10 119.60 .080 .219 .219 .219 .219 39 22 11 168.06 .069 .080 .080 .080 .080 40 23 12 66.07 .150 .190 .190 .190 .190 41 21 19 14.45 .309 1.780 1.371 .260 2.040 1.631 42 23 13 38.75 .291 .078 .138 .078 .138 43 15 17 10.62 .268 1.997 2.001 .170 2.167 2.171 44 15 17 10.62 .268 1.997 2.001 .170 2.167 2.171 45 15 16 55.42 .212 .485 .559 .485 .559 46 16 19 63.00 .101 1.629 1.108 1.629 1.108 47 16 19 20.99 .160 2.122 1.759 .260 2.382 2.019 48 15 19 44.85 .102 1.867 1.741 .260 2.127 2.001 49 20 21 16.71 .574 .180 .205 .180 .205 50 21 22 38.10 .333 .145 .100 .145 .101 51 22 23 37.17 .177 .189 .161 .189 .161 52 23 20 21.05 .326 .313 .360 .313 .360 53 15 20 8.02 .725 2.130 2.120 .400 2.530 2.520 54 15 21 24.18 .238 2.220 2.150 .400 2.620 2.550 55 15 22 31.43 .185 2.150 2.130 .400 2.550 2.530 56 15 23 6.96 .814 2.420 2.230 .400 2.820 2.630 57 15 13 62.95 .130 2.050 2.090 .400 2.450 2.490 58 13 19 70.59 .071 2.440 2.006 2.440 2.006 59 5 1 20.26 .289 .498 .498 .498 .498 60 8 4 11.61 .490 .520 .520 .520 .520 61 16 19 6.78 .397 1.698 1.570 1.698 1.570 NOTE:

• Includes wall friction.

2 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-22 PRESSURIZER SUBCOMPARTMENT NODAL DESCRIPTION Node Number Net Volume (ft3) 1 484 2 1,703 3 1,645 4 650 5 564 6 2,130 7 2,139 8 794 9 628 10 2,915 11 3,703 12 1,526 13 7,346 14 1,330 15 9,450 16 1,306 17 1,469 18 93 19 1,704,527 20 135 21 470 22 559 23 258 1 of 1

BVPS-2 UFSAR Rev. 0 TABLE 6.2-23 MASS AND ENERGY RELEASE RATES PRESSURIZER SPRAY LINE DOUBLE-ENDED RUPTURE Time Mass Flow Rate Energy Flow Rate (sec) (103 lb/sec) (106 Btu/sec) 0.0 0.0 0.0 1.010000D-03 4.027320 2.447550 2.000000D-03 4.544090 2.728269 3.020000D-03 4.646500 2.783795 4.020000D-03 4.712340 2.819180 5.000000D-03 4.741150 2.834352 6.000000D-03 4.743450 2.835017 1.304000D-02 4.678960 2.796182 2.406000D-02 5.051880 2.995971 3.106000D-02 5.198920 3.074445 4.108000D-02 5.059970 2.995857 4.517000D-02 5.017230 2.971817 5.810000D-02 5.109640 3.020708 6.401000D-02 5.128790 3.030622 7.000000D-02 5.144170 3.038526 8.115000D-02 5.000720 2.959195 8.908000D-02 4.888690 2.897659 9.714000D-02 4.962250 2.937598 1.080300D-01 5.226390 3.081829 1.121500D-01 5.246540 3.092737 1.302100D-01 4.994680 2.954570 1.361000D-01 4.994680 2.954499 1.400300D-01 5.000900 2.957848 1.620400D-01 4.907370 2.906511 1.720900D-01 4.934890 2.921379 1.881400D-01 4.957180 2.933351 2.051100D-01 4.858290 2.879170 2.151700D-01 4.855020 2.877278 2.351100D-01 4.916120 2.910353 2.451300D-01 4.909340 2.906563 2.650800D-01 4.946760 2.926725 2.950600D-01 4.896720 2.899199 3.300500D-01 4.825090 2.859695 3.500300D-01 4.850850 2.873396 3.750500D-01 4.846340 2.870517 4.251300D-01 4.897380 2.897620 4.600100D-01 4.848700 2.870605 4.900400D-01 4.854760 2.873406 5.201500D-01 4.837170 2.863345 5.700700D-01 4.874680 2.883823 6.400400D-01 4.843360 2.864974 7.200300D-01 4.846360 2.865421 7.900100D-01 4.824500 2.852574 8.700200D-01 4.816650 2.846887 1 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-23 (Cont)

Time Mass Flow Rate Energy Flow Rate (sec) (103 lb/sec) (106 Btu/sec) 9.500200D-01 4.794910 2.834357 1.030160D+00 4.779360 2.824859 1.130120D+00 4.747150 2.806104 1.500010D+00 4.656570 2.752474 2 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-24 MASS AND ENERGY RELEASE RATES PRESSURIZER SURGE LINE DOUBLE-ENDED RUPTURE Time Mass Flow Rate Energy Flow Rate (sec) (104 lb/sec) (Btu/sec) 0.0 1.025600 6.464500D+06 2.502000D-02 2.328200 1.529800D+07 5.005000D-02 2.343900 1.541100D+07 7.510000D-02 2.289400 1.507000D+07 1.001600D-01 2.138500 1.412500D+07 1.250500D-01 2.043400 1.355400D+07 1.500200D-01 2.002100 1.334100D+07 1.750200D-01 1.811100 1.216100D+07 2.000700D-01 1.602700 1.087800D+07 2.250100D-01 1.567700 1.074500D+07 2.500100D-01 1.511500 1.048000D+07 2.750300D-01 1.490300 1.039400D+07 3.000200D-01 1.449200 1.015200D+07 3.250100D-01 1.472700 1.029600D+07 3.500900D-01 1.503600 1.047400D+07 3.750100D-01 1.480200 1.034300D+07 4.000000D-01 1.472300 1.028200D+07 4.251500D-01 1.479700 1.031300D+07 4.500300D-01 1.476800 1.028800D+07 4.750300D-01 1.487800 1.033800D+07 5.000000D-01 1.499000 1.039200D+07 5.250900D-01 1.469700 1.022200D+07 5.500400D-01 1.459200 1.014500D+07 5.750500D-01 1.480200 1.025200D+07 6.000100D-01 1.478500 1.023800D+07 6.251900D-01 1.458600 1.011700D+07 6.501000D-01 1.454100 1.007900D+07 6.750500D-01 1.465900 1.013600D+07 7.000900D-01 1.469700 1.015000D+07 7.250700D-01 1.465400 1.011600D+07 7.500700D-01 1.466000 1.010700D+07 7.750200D-01 1.471400 1.012700D+07 8.000200D-01 1.473400 1.012800D+07 8.250600D-01 1.472700 1.011400D+07 8.500900D-01 1.474000 1.011000D+07 8.752100D-01 1.475900 1.011000D+07 9.000100D-01 1.476400 1.010200D+07 9.251100D-01 1.476600 1.009300D+07 9.501000D-01 1.477500 1.008700D+07 9.752000D-01 1.473900 1.004900D+07 1.000080D+00 1.480100 1.008200D+07 1.100050D+00 1.475200 1.000700D+07 1.200050D+00 1.473000 9.953400D+06 1 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-24 (Cont)

Time Mass Flow Rate Energy Flow Rate (sec) (104 lb/sec) (Btu/sec) 1.300040D+00 1.469600 9.893500D+06 1.375040D+00 1.467200 9.849600D+06 1.500140D+00 1.492700 9.961300D+06 1.600000D+00 1.549600 1.027700D+07 1.700090D+00 1.597700 1.053300D+07 1.800050D+00 1.628500 1.068300D+07 1.900090D+00 1.650400 1.077800D+07 1.975310D+00 1.660200 1.081100D+07 2.100080D+00 1.634900 1.063800D+07 2.200090D+00 1.601700 1.042500D+07 2.300080D+00 1.561300 1.016600D+07 2.400080D+00 1.521600 9.912100D+06 2.500070D+00 1.486600 9.687700D+06 2.600210D+00 1.457200 9.498900D+06 2.700220D+00 1.426400 9.299600D+06 2.800100D+00 1.418800 9.249000D+06 2.900130D+00 1.413000 9.210400D+06 3.000090D+00 1.405800 9.163600D+06 2 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-25

SUMMARY

OF SUBCOMPARTMENT PEAK CALCULATED PRESSURE DIFFERENTIAL USED IN STRUCTURAL ANALYSIS Peak Calculated Pressure Differential Break Blowdown P Time Subcompartment Considered Distribution (PSID) (sec)

Spray Pressurizer line cubicle DER Node 7 Node #

1 5.20 0.603 2 5.26 0.264 3 5.23 0.602 4 5.22 0.601 5 5.07 0.599 6 5.28 0.322 7 5.38 0.295 8 5.23 0.296 9 0.15 0.091 10 0.16 0.093 11 0.12 0.110 12 0.12 0.100 13 0.07 0.109 14 0.04 0.271 15 0.09 0.119 16 0.02 0.119 17 0.14 0.134 18 0.11 0.104 20 0.13 0.105 21 0.11 0.100 22 0.12 0.108 23 0.11 0.102 1 of 7

BVPS-2 UFSAR Rev. 0 TABLE 6.2-25 (Cont)

Peak Calculated Pressure Differential Break Blowdown P Time Subcompartment Considered Distribution (PSID) (sec)

Surge line DER Node 18 Node #

1 0.20 0.096 2 0.21 0.102 3 0.22 0.081 4 0.21 0.087 5 0.17 0.094 6 0.16 0.056 7 0.19 0.075 8 0.18 0.130 9 11.26 0.163 10 11.26 0.154 11 11.19 0.151 12 10.76 0.163 13 10.70 0.179 14 10.19 0.186 15 6.68 0.196 16 2.74 0.203 17 6.73 0.199 18 199.32* 0.053 20 18.28 0.056 21 11.48 0.154 22 11.18 0.157 23 11.09 0.154 NOTE:

• This P is across the pressurizer support skirt.

2 of 7

BVPS-2 UFSAR Rev. 0 TABLE 6.2-25 (Cont)

Peak Calculated Pressure Differential Break Blowdown P Time Subcompartment Considered Distribution (PSID) (sec)

Pressurizer Surge relief tank line cubicle DER Node 15 Node #

1 0.07 0.145 2 0.06 0.135 3 0.06 0.151 4 0.06 0.144 5 0.07 0.089 6 0.09 0.126 7 0.06 0.126 8 0.07 0.114 9 6.19 0.202 10 6.18 0.204 11 6.19 0.195 12 6.17 0.205 13 6.19 0.204 14 5.87 0.211 15 18.21 0.161 16 7.30 0.179 17 18.47 0.161 18 6.18 0.190 20 6.20 0.185 21 6.17 0.191 22 6.18 0.188 23 6.17 0.192 3 of 7

BVPS-2 UFSAR Rev. 0 TABLE 6.2-25 (Cont)

Peak Calculated Pressure Differential Break Blowdown P Time Subcompartment Considered Distribution (PSID) (sec) 320 in2 LDR Steam at the stm.

generator gen. outlet subcompartment nozzle 100% Node 32 Node#

1 6.25 0.028 2 5.60 0.111 3 5.86 0.114 4 5.80 0.037 5 6.89 0.029 6 6.69 0.025 7 5.83 0.079 8 5.65 0.106 9 5.78 0.106 10 5.05 0.312 11 6.04 0.098 12 6.14 0.102 13 4.98 0.157 14 5.03 0.162 15 5.08 0.097 16 4.76 0.312 17 4.90 0.264 18 4.96 0.147 19 4.77 0.283 20 4.72 0.298 21 4.72 0.300 22 4.72 0.250 23 4.76 0.152 24 4.76 0.274 25 0.82 0.131 26 0.74 0.354 27 0.60 0.372 28 0.61 0.131 29 3.62 0.313 31 12.59 0.102 33 12.66 0.099 4 of 7

BVPS-2 UFSAR Rev. 0 TABLE 6.2-25 (Cont)

Peak Calculated Pressure Differential Break Blowdown P Time Subcompartment Considered Distribution (PSID) (sec) 180 in2 at the reactor coolant pump outlet 50%-Node 8 nozzle 50%-Node 9 Node #

1 5.07 0.065 2 5.24 0.061 3 6.24 0.055 4 3.93 0.065 5 4.32 0.074 6 4.20 0.082

` 7 4.98 0.073 8 5.84 0.068 9 7.08 0.010 10 3.82 0.104 11 4.16 0.100 12 4.24 0.096 13 4.72 0.053 14 5.49 0.058 15 5.80 0.051 16 4.04 0.056 17 4.51 0.090 18 4.12 0.092 19 3.93 0.077 20 3.51 0.072 21 3.53 0.112 22 3.63 0.103 23 3.52 0.102 24 3.60 0.087 25 0.76 0.079 26 0.59 0.086 27 0.53 0.091 28 0.67 0.082 29 2.26 0.054 31 4.98 0.082 33 4.75 0.083 5 of 7

BVPS-2 UFSAR Rev. 0 TABLE 6.2-25 (Cont)

Peak Calculated Pressure Differential Break Blowdown P Time Subcompartment Considered Distribution (PSID) (sec) 707 in2 longitudinal split break at the stm. 50%-Node 32 gen. inlet 25%-Node 7 elbow 25%-Node 8 Node#

1 15.08 0.025 2 10.50 0.074 3 10.34 0.166 4 10.66 0.039 5 10.44 0.032 6 9.86 0.133

` 7 12.32 0.014 8 10.62 0.143 9 10.68 0.055 10 10.25 0.102 11 10.08 0.102 12 9.96 0.151 13 9.92 0.064 14 9.85 0.061 15 10.52 0.093 16 9.62 0.098 17 9.43 0.102 18 9.29 0.107 19 9.34 0.122 20 9.16 0.079 21 9.09 0.102 22 9.54 0.102 23 9.19 0.088 24 9.13 0.131 25 1.50 0.066 26 1.37 0.069 27 1.09 0.079 28 1.01 0.092 29 8.18 0.171 31 13.27 0.025 32 12.66 0.033 6 of 7

BVPS-2 UFSAR Rev. 0 TABLE 6.2-25 (Cont)

Peak Calculated Pressure Differential Break Blowdown P Time Subcompartment Considered Distribution (PSID) (sec) 50%-Node 32 707 in2 25% Node 7 longitudinal 25%-Node 8 Upper steam split break (1.0 Moody Flow generator at stm. gen. Correlation) cubicle inlet elbow Node #

25 1.82 .091 26 1.58 .085 27 1.28 .138 28 1.32 .131 7 of 7

BVPS-2 UFSAR Rev. 0 TABLE 6.2-27 STEAM GENERATOR SUBCOMPARTMENT VENT PATH DESCRIPTION Head Loss Coefficient K Vent From To Geom. Grating Path Node Node Area Inertia Expansion + Contraction Wall Loss Total No. No. No. ( ft2 ) (ft-1) Fwd Rvs Friction Factor Fwd Rvs 1 1 29 81.74 .042 .63 .47 .04 .67 .51 2 2 29 113.46 .028 .54 .51 .04 .55 .55 3 3 29 45.02 .039 1.01 .88 .04 1.05 .92 4 2 5 27.22 .285 .76 .67 .04 .80 .71 5 1 2 99.98 .125 .01 .02 .04 .05 .06 6 2 3 180.13 .042 .06 .08 .04 .10 .12 7 3 4 55.84 .252 .20 .23 .04 .24 .27 8 4 5 70.84 .129 .23 .22 .04 .27 .26 9 5 6 34.23 .248 .26 .27 .04 .30 .31 10 6 30 45.00 .142 1.64 1.64 .04 1.68 1.68 11 29 30 240.22 .011 1.01 1.01 .04 1.05 1.05 12 4 30 48.00 .150 1.67 1.67 .04 1.71 1.71 13 1 7 121.64 .079 .04 .04 .04 .08 .08 14 2 8 127.82 .05 .39 .39 .04 .79 .79 15 2 8 61.50 .05 .88 .88 .04 .224 1.14 1.14 16 3 9 138.80 .072 .02 .02 .04 .06 .06 17 4 10 111.74 .131 .007 .007 .04 .05 .05 18 5 11 180.33 .064 .004 .004 .04 .04 .04 19 6 12 72.80 .172 .003 .02 .04 .04 .06 20 7 12 12.81 .327 1.01 1.03 .04 1.05 1.07 21 8 11 51.00 .113 .77 .72 .04 .81 .76 22 7 8 80.09 .073 .54 .51 .04 .58 .55 23 8 9 178.20 .033 .34 .32 .04 .38 .37 24 9 10 30.08 .162 .97 .82 .04 1.01 .86 25 10 11 56.74 .093 .69 .68 .04 .73 .72 26 11 12 59.00 .111 .57 .57 .04 .61 .61 27 12 30 4.81 .658 1.55 1.55 .04 1.59 1.59 28 10 30 20.52 .242 1.50 1.50 .04 1.54 1.54 29 7 13 70.44 .111 .20 .20 .04 .224 .46 .46 30 8 14 173.70 .041 .24 .23 .04 .224 .50 .49 31 9 15 88.13 .088 .21 .22 .04 .224 .47 .48 32 10 16 84.46 .111 .08 .14 .04 .224 .34 .40 33 11 17 168.51 .041 .21 .21 .04 .224 .47 .47 34 12 18 98.49 .061 .21 .21 .04 .224 .47 .47 35 13 18 28.86 .213 .73 .73 .04 .77 .77 36 14 17 54.38 .106 .60 .57 .04 .64 .61 37 13 14 61.63 .102 .42 .37 .04 .46 .41 1 of 3

BVPS-2 UFSAR Rev. 0 TABLE 6.2-27 (Cont)

Head Loss Coefficient K Vent From To Geom. Grating Path Node Node Area Inertia Expansion + Contraction Wall Loss Total No. No. No. ( ft2 ) (ft-1) Fwd Rvs Friction Factor Fwd Rvs 38 14 15 119.66 .074 .10 .11 .04 .14 .15 39 15 16 25.38 .291 .72 .56 .04 .76 .60 40 16 17 40.48 .169 .57 .52 .04 .61 .56 41 17 18 44.39 .121 .63 .62 .04 .67 .66 42 18 30 34.56 .128 2.14 2.14 .04 2.18 2.18 43 16 30 37.50 .187 1.55 1.55 .04 1.59 1.59 44 13 19 87.72 .157 .00 .00 .04 .04 .04 45 14 20 123.77 .053 .53 .53 .04 .224 .79 .79 46 15 21 75.33 .116 .32 .32 .04 .224 .58 .58 47 16 22 105.18 .121 .05 .05 .04 .09 .09 48 17 23 209.86 .057 .09 .09 .04 .13 .13 49 18 24 105.57 .10 .17 .17 .04 .21 .21 50 19 24 65.76 .076 .75 .75 .04 .79 .79 51 20 23 194.51 .038 .42 .39 .04 .46 .43 52 19 20 146.26 .037 .55 .49 .04 .59 .53 53 20 21 248.74 .029 .33 .32 .04 .37 .36 54 21 22 128.31 .109 .22 .23 .04 .26 .27 55 22 23 202.41 .060 .10 .12 .04 .14 .16 56 23 24 143.44 .043 .53 .51 .04 .57 .55 57 24 30 48.00 .077 2.39 2.39 .04 2.43 2.43 58 20 30 48.62 .142 1.57 1.57 .04 .177 1.79 1.79 59 21 30 26.50 .236 1.49 1.49 .04 .177 1.71 1.71 60 22 30 14.00 .376 1.58 1.58 .04 1.62 1.62 61 22 30 53.07 .163 1.52 1.52 .04 .177 1.74 1.74 62 23 30 48.97 .124 1.68 1.68 .04 .177 1.90 1.90 63 19 25 29.19 .280 .99 .95 .04 1.03 .99 64 20 26 29.86 .190 1.25 1.41 .04 1.29 1.45 65 23 27 26.10 .202 1.31 1.40 .04 1.35 1.44 66 24 28 27.26 .232 1.25 1.24 .04 1.29 1.28 67 25 28 19.70 .140 1.01 1.01 .04 1.05 1.05 68 25 26 34.70 .093 .80 .80 .04 .84 .84 69 26 27 34.70 .121 .80 .80 .04 .84 .84 70 27 28 75.50 .065 .51 .51 .04 .55 .55 71 25 30 48.59 .051 1.25 1.25 .04 1.29 1.29 72 26 30 68.72 .039 1.23 1.23 .04 1.27 1.27 73 27 30 89.89 .033 1.21 1.21 .04 1.25 1.25 74 28 30 90.44 .035 1.18 1.18 .04 1.22 1.22 75 1 31 31.50 .169 1.42 1.22 .04 1.46 1.26 76 2 31 81.97 .131 .76 .62 .04 .80 .66 2 of 3

BVPS-2 UFSAR Rev. 0 TABLE 6.2-27 (Cont)

Head Loss Coefficient K Vent From To Geom. Grating Path Node Node Area Inertia Expansion + Contraction Wall Loss Total No. No. No. ( ft2 ) (ft-1) Fwd Rvs Friction Factor Fwd Rvs 77 5 31 73.79 .114 .90 .78 .04 .94 .82 78 6 31 35.41 .120 1.43 1.32 .04 1.47 1.36 79 31 32 99.46 .063 .76 .76 .04 .224 1.02 1.02 80 7 32 12.69 .214 1.73 1.77 .04 1.77 1.81 81 8 32 38.53 .146 1.32 1.45 .04 1.72 1.85 82 11 32 39.91 .132 1.29 1.46 .04 1.33 1.50 83 12 32 20.26 .158 1.52 1.65 .04 1.56 1.69 84 33 12 13.30 .336 1.08 .68 .04 1.12 .72 85 33 7 13.30 .352 1.01 .66 .04 1.05 .70 86 33 6 10.30 .348 1.10 .80 .04 1.14 .84 87 33 1 10.30 .391 1.06 .79 .04 1.10 .83 88 32 33 31.86 .163 1.69 1.22 .04 1.73 1.26 89 31 33 84.45 .084 1.06 .72 .04 1.10 .76 90 33 7 11.03 .452 1.09 .71 .04 1.13 .75 91 33 12 11.03 .519 .98 .68 .04 1.02 .72 3 of 3

BVPS-2 UFSAR Rev. 0 TABLE 6.2-28 STEAM GENERATOR SUBCOMPARTMENT NODAL DESCRIPTION Net Volume Node Number (ft3) 1 1,024 2 1,765 3 1,124 4 907 5 1,439 6 735 7 1,317 8 2,537 9 1,487 10 986 11 2,308 12 1,074 13 603 14 1,490 15 756 16 717 17 1,445 18 834 19 1,530 20 4,056 21 1,795 22 1,602 23 3,689 24 2,019 25 561 26 770 27 1,143 28 960 29 1,027 30 1,716,078 31 1,473 32 710 33 378 1 of 1

BVPS-2 UFSAR Rev. 0 TABLE 6.2-29 MASS AND ENERGY RELEASE RATES 360 IN2 LDR AT THE STEAM GENERATOR OUTLET NOZZLE Time Mass Flow Rate Energy Flow Rate (sec) (104 lb/sec) (107 Btu/sec) 0.0 0.0 0.0 1.0100D-03 1.46644 0.7301 2.0000D-03 1.89019 1.005 3.0200D-03 2.24415 1.193 4.0100D-03 2.70989 1.443 5.0000D-03 3.01268 1.649 6.0000D-03 3.22944 1.710 8.0000D-03 3.43589 1.825 1.0020D-02 3.39021 1.800 1.2030D-02 3.34821 1.777 1.5020D-02 3.51762 1.866 2.0020D-02 3.11841 1.654 2.3030D-02 2.95165 1.565 2.5000D-02 3.01085 1.597 2.9010D-02 2.80254 1.485 3.1010D-02 2.90247 1.539 3.4000D-02 2.75738 1.461 3.7020D-02 2.84539 1.507 3.0100D-02 2.79718 1.482 4.1020D-02 3.55730 1.387 4.2020D-02 3.53498 1.874 4.5010D-02 3.62314 1.921 5.0010D-02 3.94872 2.096 5.5010D-02 4.36179 2.315 6.0050D-02 4.38023 2.326 6.6050D-02 4.62400 2.455 7.1040D-02 4.44237 2.358 8.0020D-02 4.75770 2.527 8.6020D-02 4.80056 2.550 8.8030D-02 4.82318 2.561 1.0004D-01 4.59401 2.438 1.0807D-01 4.38502 2.326 1.2407D-01 4.56874 2.426 1.3008D-01 4.50811 2.393 1.4014D-01 4.46467 2.370 1.5005D-01 4.40963 2.340 1.6010D-01 4.44297 2.359 1.7203D-01 4.49341 2.386 1.7806D-01 4.47923 2.370 2.0502D-01 4.58427 2.436 2.2002D-01 4.52172 2.403 2.4003D-01 4.56181 2.406 1 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-29 (Cont)

Time Mass Flow Rate Energy Flow Rate (sec) (104 lb/sec) (107 Btu/sec) 3.00010D-01 4.55787 2.428 3.80010D-01 4.49416 2.401 4.05100D-01 4.52948 2.423 5.00260D-01 4.44197 2.387 6.00020D-01 4.32854 2.332 7.00010D-01 4.23482 2.299 8.00090D-01 4.13227 2.255 9.00040D-01 4.03755 2.215 1.00007D+00 3.97331 2.190 1.20002D+00 3.84359 2.139 1.40009D+00 3.78736 2.581 1.80004D+00 3.56703 2.381 1.80002D+00 3.44952 1.956 2.00005D+00 3.34527 1.904 2.20003D+00 3.22943 1.844 2.40012D+00 3.15043 1.805 2.60007D+00 3.62537 1.738 2.30017D+00 2.02537 1.603 3.00001D+00 2.76955 1.595 2 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-30 MASS AND ENERGY FLOW RATES 180 IN2 LDR AT THE REACTOR COOLANT PUMP OUTLET NOZZLE Time Mass Flow Rate Energy Flow Rate (sec) (104 lb/sec) (107 Btu/sec) 0.0 1.02564 0.5481 1.0000D-03 1.85908 0.9858 2.0100D-03 1.89412 1.007 3.0100D-03 2.04654 1.088 4.0100D-03 2.08563 1.109 5.0000D-03 2.11194 1.123 6.0100D-03 2.11314 1.124 7.0000D-03 2.10392 1.119 8.0300D-03 2.08910 1.111 1.0030D-02 2.05897 1.095 1.2020D-02 2.03615 1.082 1.4010D-02 2.02302 1.075 1.5000D-02 2.01992 1.074 1.6010D-02 2.02035 1.074 2.0000D-02 2.08074 1.107 2.2010D-02 2.13166 1.134 2.4000D-02 2.16966 1.155 2.6020D-02 2.18859 1.165 2.7010D-02 2.19191 1.166 2.8000D-02 2.30884 1.230 2.9000D-02 2.57437 1.369 3.0010D-02 2.95307 1.574 3.2000D-02 3.20083 1.705 3.4010D-02 3.34822 1.785 3.6020D-02 3.41168 1.819 3.8040D-02 3.42248 1.825 4.0020D-02 3.45159 1.841 4.2010D-02 3.49307 1.863 4.8030D-02 3.35292 1.787 5.7080D-02 3.49781 1.866 6.4010D-02 3.40959 1.818 6.6180D-02 3.40543 1.815 8.0030D-02 3.24940 1.731 9.5060D-02 2.99803 1.595 9.6120D-02 2.99717 1.594 9.7040D-02 2.99805 1.595 1.0605D-01 3.03307 1.613 1.2005D-01 2.99649 1.594 1.3412D-01 2.95979 1.574 1.5012D-01 2.99959 1.596 1.8207D-01 2.95729 1.573 1 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-30 (Cont)

Time Mass Flow Rate Energy Flow Rate (sec) (104 lb/sec) (107 Btu/sec) 1.98100D-01 2.98057 1.586 2.80090D-01 3.01695 1.605 3.50100D-01 2.91402 1.550 3.85020D-01 3.00367 1.598 4.10130D-01 2.90241 1.543 4.40010D-01 2.97684 1.584 4.70010D-01 2.91813 1.552 6.00030D-01 3.00351 1.598 6.90040D-01 2.95459 1.572 7.20060D-01 2.98387 1.588 8.20000D-01 2.97004 1.580 1.00006D+00 2.99995 1.597 1.20006D+00 3.01732 1.607 1.70006D+00 3.03020 1.614 1.56603D+00 3.04140 1.621 1.80012D+00 3.03994 1.622 2.00005D+00 3.03584 1.622 2.20002D+00 3.01368 1.622 2.40003D+00 2.98586 1.595 2.60019D+22 2.95262 1.582 2.80003D+00 2.82415 1.558 3.00007D+00 2.74881 1.530 2 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-31 MASS AND ENERGY RELEASE RATES 707 IN2 LONGITUDINAL INTRADOS SPLIT AT THE STEAM GENERATOR INLET ELBOW Time Mass Flow Rate Energy Flow Rate (sec) (104 lb/sec) (107 Btu/sec) 0.0 0.0 0.0 1.0000D-03 0.85935 0.541 2.0100D-03 1.21117 0.762 3.0200D-03 2.35905 1.483 4.0200D-03 3.43611 2.161 5.0200D-03 4.08434 2.568 6.0200D-03 4.45081 2.798 7.0000D-03 4.63897 2.915 8.0100D-03 4.72311 2.967 1.8010D-02 4.94518 3.103 2.3010D-02 5.01624 3.146 2.8010D-02 4.97659 3.121 2.9010D-02 4.97371 3.120 3.8050D-02 5.09590 3.198 4.8010D-02 5.19863 3.263 5.1010D-02 5.87217 3.690 5.4020D-02 5.43402 3.416 5.6020D-02 5.67902 3.570 5.8010D-02 5.62479 3.535 6.8000D-02 5.27236 3.316 7.8030D-02 5.15338 3.247 8.8000D-02 5.01559 3.167 9.8090D-02 4.86576 3.079 1.0805D-01 4.77536 3.027 1.1808D-01 4.73526 3.006 1.2805D-01 4.79491 3.048 1.4004D-01 4.83096 3.075 1.4810D-01 4.82055 3.071 1.6805D-01 4.75589 3.036 1.8804D-01 4.70265 3.006 2.0013D-01 4.65914 2.980 2.1512D-01 4.62556 2.961 2.5001D-01 4.66057 2.985 2.8004D-01 4.60615 2.951 3.0511D-01 4.54212 2.911 3.5012D-01 4.57662 2.932 4.0515D-01 4.48985 2.876 4.5516D-01 4.50214 2.884 5.0005D-01 4.43712 2.843 5.5011D-01 4.44384 2.847 6.0009D-01 4.38624 2.811 6.3011D-01 4.39090 2.814 1 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.2-31 (Cont)

`

Time Mass Flow Rate Energy Flow Rate (sec) (104 lb/sec) (107 Btu/sec) 6.5014D-01 4.37980 2.808 7.00010D-01 4.33078 2.778 7.50100D-01 4.31362 2.768 8.00090D-01 4.27366 2.745 8.50060D-01 4.24616 2.730 9.00100D-01 4.21213 2.711 9.50100D-01 4.17778 2.693 1.00001D+00 4.14529 2.676 1.05013D+00 4.10699 2.655 1.10018D+00 4.07349 2.637 1.15010D+00 4.03232 2.614 1.20040D+00 3.99611 2.594 1.25002D+00 3.95083 2.569 1.30003D+00 3.91050 2.546 1.35014D+00 3.86170 2.517 1.40004D+00 3.81716 2.491 1.45018D+00 3.76844 2.463 1.50013D+00 3.72217 2.436 1.55009D+00 3.67627 2.409 1.60004D+00 3.63134 2.383 1.65002D+00 3.58140 2.353 1.70007D+00 3.54292 2.330 1.75012D+00 3.50783 2.309 1.80021D+00 3.47465 2.288 1.90025D+00 3.41871 2.251 2.00002D+00 3.36742 2.215 2.10034D+00 3.31227 2.180 2.20026D+00 3.27017 2.146 2.30010D+00 3.22098 2.113 2.40001D+00 3.16928 2.081 2.50024D+00 3.11821 2.048 2.60022D+00 3.06548 2.015 2.70000D+00 3.01331 1.982 2.80002D+00 2.95994 1.948 2.90020D+00 2.90539 1.914 3.00019D+00 2.25279 1.881 2 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-50 Beaver Valley Power Station Unit 2 Initial Condition Assumptions MSLB Mass and Energy Releases Inside Containment NSSS Power, MWt 2910 Initial Conditions Power Level (%)

Parameter 100.6 70 30 0 RCS Average Temperature (°F) 588.5 578.6 565.4 547.0 RCS Flowrate (gpm) (Thermal 261,600 261,600 261,600 261,600 Design Flow)

RCS Pressure (psia) 2250 2250 2250 2250 Pressurizer Water Volume (ft3) 834.3 693.3 505.3 364.3 Feedwater Enthalpy (Btu/lbm) 436.0 385.4 305.3 70.7 SG Pressure (psia) 885 935 1011 1004 SG Water Level (% NRS) 51 51 51 51 1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-51 Beaver Valley Power Station Unit 2 Balance of Plant Assumptions MSLB Mass and Energy Releases Inside Containment Main Feedwater System Flowrate - DERs @ all powers Feedwater flow based on (until main feedwater isolation) system performance as a function of SG pressure.

Flowrate - split ruptures @ all powers Feedwater flow matches steam flow.

(until main feedwater isolation)

Unisolable volume from SG nozzle to MFIV 157 ft3 (faulted loop)

Unisolable volume from SG nozzle to FCV 264 ft3 assuming a single failure of the MFIV (faulted loop)

Total response time for feedwater 7.0 seconds isolation (instrument response, signal processing, and MFIV closure)

Auxiliary Feedwater System Flowrate to all steam generators Maximum flow to each SG is 310 gpm. The actual data used is a function of SG pressure.

Temperature (maximum value) 120°F Actuation delay time 0 seconds Assumed time of manual termination 30 minutes Main Steam System Total piping volume 7,192 ft3 Volume between the break and the nearest MSIV No failure of the faulted-loop MSIV 1,038 ft3 Failure of the faulted-loop MSIV 6,023 ft3 Total response time for steamline 7.0 seconds isolation (instrument response, signal processing, and MSIV closure) 1 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-51 (Cont)

Beaver Valley Power Station Unit 2 Balance of Plant Assumptions MSLB Mass and Energy Releases Inside Containment Isolation Setpoints For all double-ended ruptures, Feedwater isolation from a safety injection 1760 psia signal via low pressurizer pressure Feedwater isolation from a safety injection 460 psia signal via low steamline pressure in any loop dynamic compensation lead 50 seconds lag 5 seconds Steamline isolation from low steamline pressure 460 psia in any loop dynamic compensation lead 50 seconds lag 5 seconds Isolation Setpoints For all split ruptures, Feedwater isolation from a safety injection containment scope signal via Hi-1 containment pressure Steamline isolation from Hi-2 containment containment scope pressure 2 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-52 LBLOCA Mass and Energy Releases from BCL Vessel-Side Time (sec) Mass Flow (lbm/s) Energy Flow (Btu/s) 0 0 0 0.5 53899 28773060 1 47583 25544458 2 30986 17130028 3 25674 14228740 4 22701 12581345 5 20696 11479648 10 6987 5271856 15 4265 1963534 20 4464 879868 25 176 26980 30 93 0 35 32 0 40 6240 797967 45 1623 445206 50 204 142040 55 225 182306 60 156 47575 65 89 77304 70 33 25664 75 75 75761 80 37 26438 85 87 86789 90 224 74083 95 81 77320 100 302 109665 110 839 223422 120 944 240389 130 847 223149 1 of 2

BVPS-2 UFSAR Rev. 16 Table 6.2-52 (cont.)

LBLOCA Mass and Energy Releases from BCL Vessel-Side Time (sec) Mass Flow (lbm/s) Energy Flow (Btu/s) 140 418 119726 150 21 20756 160 21 22151 170 17 18681 180 35 37810 190 96 51695 200 812 205873 210 518 148513 220 741 189364 230 214 76856 240 33 34324 250 52 36638 260 73 45672 270 145 70845 280 79 54907 290 120 61814 300 172 82874 310 210 103847 320 815 210784 330 542 153176 340 283 89386 350 105 64664 360 56 46956 370 216 83446 379 243 89934 2 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-53 LBLOCA Mass and Energy Releases from BCL Vessel-Side Time (sec) Mass Flow (lbm/s) Energy Flow (Btu/s) 0 0 0 0.5 24639 13159514 1 24097 13037813 2 19067 10775565 3 14143 8275659 4 10118 6355736 5 7684 5246894 10 3168 2664892 15 771 821017 20 190 229993 25 -26 0 30 19 23663 35 20 26050 40 172 215294 45 114 139243 50 42 53484 55 62 78158 60 25 31981 65 39 49578 70 28 35314 75 42 53075 80 25 32428 85 49 61813 90 31 38924 95 43 54250 100 37 47537 110 41 52688 120 45 56783 130 38 48576 1 of 2

BVPS-2 UFSAR Rev. 16 Table 6.2-53 (cont.)

LBLOCA Mass and Energy Releases from BCL Loop-Side Time (sec) Mass Flow (lbm/s) Energy Flow (Btu/s) 140 32 40828 150 23 29517 160 22 28307 170 24 30994 180 31 38953 190 34 42830 200 39 49393 210 36 46134 220 36 45468 230 32 40281 240 31 38891 250 28 36135 260 34 42784 270 37 46983 280 38 47575 290 40 51167 300 43 54819 310 47 59217 320 47 58865 330 41 51856 340 38 47558 350 37 46814 360 35 43723 370 38 48498 379 41 51508 2 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-54 ACTIVE HEAT SINKS FOR MINIMUM CONTAINMENT PRESSURE ANALYSIS Containment Spray Parameters Number of pumps operating 2 Maximum spray flow 4450 gpm Fastest post-LOCA initiation of 37 sec spray pumps, assuming offsite power available Fastest post-LOCA initiation of spray pumps, assuming offsite power loss and no diesel failure 52 sec Containment Fan Coolers None 1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.2-55 Large Break LOCA Containment Wall Data Used for Calculation of Containment Pressure TAir Area Height TInit Thickness / Material Wall (oF) (ft2) (ft) (oF) (inches)

1. Painted Concrete 70 129,776 10 70 0.0035 / Paint 12.0 / Concrete

2. Stainless Steel Piping 70 8,400 10 70 0.28 / Stainless Steel

3. Galvanized Structural 70 21,783 10 70 0.0625 / Galvanized Steel Steel

4. Galvanized Ventilation 70 15,856 10 70 0.125 / Galvanized Ducts Steel

5. Carbon Structural Steel 70 6,406 10 70 0.0045 / Paint 0.0625 / Carbon Steel

6. Carbon Structural Steel 70 67,418 10 70 0.0045 / Paint 0.125 / Carbon Steel

7. Carbon Steel Liner / -20 42,469 100 70 0.0045 / Paint Concrete Shell 0.375 / Carbon Steel 54.0 / Concrete

8. Carbon Steel Liner / -20 19,638 100 70 0.0045 / Paint Concrete Dome 0.5 / Carbon Steel 30.0 / Concrete

9. Carbon Steel Liner / -20 9,036 100 70 0.0045 / Paint Concrete Dome Liner Plates 1.0 / Carbon Steel 30.0 / Concrete

10. Concrete / Carbon Steel 32 11,251 10 70 0.0035 / Paint Liner / Concrete Floor 24.0 / Concrete 0.25 / Carbon Steel 120.0 / Concrete

11. Carbon Structural Steel, 70 25,785 10 70 0.0045 / Paint Ducts, and Equipment 0.1879 / Carbon Steel

12. Carbon Structural Steel, 70 45,738 10 70 0.0045 / Paint Pipe Supports, and Piping 0.2565 / Carbon Steel

13. Carbon Structural Steel 70 17,720 10 70 0.0045 / Paint 0.3158 / Carbon Steel

14. Stainless Steel Refueling 70 13,394 10 70 0.25 / Stainless Steel Cavity Liner / Concrete 24.0 / Concrete

15. Stainless Steel Refueling 70 3,348 10 70 1.0 / Stainless Steel Cavity Liner / Concrete 24.0 / Concrete 1 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.2-55 (Cont)

Large Break LOCA Containment Wall Data Used for Calculation of Containment Pressure TAir Area Height TInit Thickness / Material Wall (oF) (ft2) (ft) (oF) (inches)

16. Carbon Steel Equipment 70 32,214 10 70 0.0045 / Paint 0.438 / Carbon Steel

17. Carbon Steel Pipe Rupture 70 11,489 10 70 0.0045 / Paint Restraints and Structural 0.6068 / Carbon Steel Steel

18. Carbon Steel Pipe Rupture 70 2,615 10 70 0.0045 / Paint Restraints 1.0312 / Carbon Steel

19. Carbon Steel Equipment and 70 3,843 10 70 0.0045 / Paint Structural Steel 1.4683 / Carbon Steel

20. Carbon Steel Equipment 70 7,648 10 70 0.0045 / Paint 4.593 / Carbon Steel 2 of 2

BVPS-2 UFSAR Rev. 26 TABLE 6.2-56 QUENCH SPRAY SYSTEM COMPONENT DESIGN DATA Components Design Parameters Refueling Water Storage Tank Quantity 1 Capacity (gal)* 900,000 Design pressure Atmospheric Design temperature (F) 150 Operating pressure Atmospheric Operating temperature (F) 45-50 Material(1) Type 304L Quench Pumps Quantity 2 Type Horizontal, Centrifugal Motor horsepower (hp) (each) 350 Seals Single mechanical Capacity (gal) 3,000 Head at rated capacity (ft) 330 Design pressure (psig) 250 Operating temperature (F) 50 Design temperature 150 Material(1)

Pump casing ASTM A351 Gr CF8 Shaft 17-4 PH Cond. H-1100 Impeller ASTM A351 Gr CF8

• Nominal volume of the straight sided portion of the tank 1 of 2

BVPS-2 UFSAR Rev. 26 TABLE 6.2-56 (Cont)

Components Design Parameters Material(1)

Pump casing SA 351-Type 304L Shaft Type 316 Stainless Steel Rotor Nitronic 60 Idler Nitronic 50 Quench Spray Headers Quantity 2 Elevation 871 ft 6 in and 846 ft-4 in Nozzles per header 39 and 120 Nozzle type SPRACO Model 1713A NOTES(1) Materials listed in this table may have been replaced with materials of equivalent design characteristics. The term equivalent is described in UFSAR Section 1.12, Equivalent Materials.

2 of 2

BVPS-2 UFSAR Rev. 7 TABLE 6.2-57 COMPONENT DESIGN DATA RECIRCULATION SPRAY SYSTEM Components Design Parameters Recirculation Pumps Quantity 4 Type Vertical, deep-well type turbine Motor power (hp) (each) 350 Seal type Tandem mechanical Capacity (gpm) 3,500 Total dynamic head at rated capacity (ft) 266 Design pressure (psig) 268 Design temperature (F) 280 Material(1)

Suction well casing SA-358 Type 304 Pump bowl ASTM A351 Gr. CF8 Shaft A-276 Type 304 Cond A Impeller ASTM A351 Gr. CF8 Recirculation Coolers Quantity 4 Design duty (Btu/hr) 1.43 x 108 Shell Tube Fluid Recirculated Ohio River spray water water (service water)

Design pressure (psig) 250 150 Design temperature (F) 250 250 Operating temperature In/Out (F) 196.4/114.5 86/143.3 Operating pressure 100 100 Material(1) Stainless Stainless Steel- Steel-Type 304 Type 304 Recirculation Spray Headers Quantity 2 Elevation 852 ft-5 in and 849 ft Nozzles per header 292 Nozzle type SPRACO Model 1713A NOTES(1) Materials listed in this table may have been replaced with materials of equivalent design characteristics. The term equivalent is described in UFSAR Section 1.12, Equivalent Materials.

1 of 1

BVPS-2 UFSAR Rev. 0 TABLE 6.2-58 CONTAINMENT VOLUMES COVERED BY QUENCH SPRAY Net Spray Volume Volume Percent Location ft3 ft3 Sprayed Dome 519,274 519,274 100.0 Operating floor to bend line (outside crane wall) 160,370 157,983 98.5 Operating floor to bend line (inside crane wall) 343,795 328,603 95.6 Refueling cavity 35,001 30,724 87.8 Reactor cavity 7,892 1,889 23.9 Annulus outside crane wall (738 ft 10 in to 767 ft 10 in) 98,885 74,945 75.8 Annulus outside crane wall (718 ft 6 in to 738 ft 10 in) 69,920 65,077 93.1 Basement 196,484 64,724 32.9 Steam generator cubicles - 135,371 79,153 58.5 total 48,772 29,006 59.5 steam generator cubicle 1 43,001 23,250 54.1 steam generator cubicle 2 43,598 26,897 61.7 steam generator cubicle 3 Pressurizer cubicle 23,819 2,546 10.7 Reactor head storage hatch (738 ft 10 in to 767 ft 10 in) 12,565 12,565 100.0 Incore instrumentation room 29,150 0 0 Pressurizer relief tank cubicle 16,286 0 0 Residual heat removal cubicle 28,667 0 0 Reactor head storage hatch (692 ft 11 in to 738 ft 10 in) 35,422 18,452 52.1 Incore instrumentation tunnel 3,732 0 0 Total 1,716,633 1,355,935 79.0 1 of 1

BVPS-2 UFSAR Rev. 26 TABLE 6.2-59 NET POSITIVE SUCTION HEAD FOR CONTAINMENT HEAT REMOVAL SYSTEM At Start of Recirculation Spray Elevation head (ft) 27.12 Pipe and sump losses (ft) 6.87 Available NPSH (ft) 20.25 Pump flow (gpm) 3,665 Required NPSH (ft) 18.3 Margin (ft) 2.3 1 of 1

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 CONTAINMENT ISOLATION FEATURES Distance Isol- from Outer-Cnmt ation/ Overpressure most Valve General Valve Position Isolation Safet Type C Leak Protection/ to Contain- Valve Closure Safety-Related Ref.

Penet Isolation Valves Design Valve Power Signal yFun Test Required Set point ment Valve Actuation Time (sec) (24) Power Source FSAR No. Line No. (1) Fluid Inside Outside Criteria Type Normal Shutdown Failure DBA (16) c-tion Inside Outside (psig) Wall Primary Secondary Inside Outside Inside Outside Sect.

(7)

System Name Chemical and X-28 2-CHS-002-2-2 Rx 2CHS*AOV200A 2CHS*AOV204 55 Globe Open Open Closed Closed CIA IO Yes Yes RV/600 24-3 Air None 10 60 BATT*2-1 BATT*2-2 9.3.4 volume control coolant 2CHS*AOV200B 55 Globe Open Open Closed Closed CIA IO Yes Air 10 BATT*2-1 system 2CHS*AOV200C 55 Globe Open Open Closed Closed CIA IO Yes Air 10 BATT*2-1 2CHS*HCV142 55 Globe Admin Open Closed Closed None IO Yes Air N/A BATT*2-1 Closed (19) Yes (19) None 2CHS*RV203 55 Relief - - - -

X-15 2-CHS-003-120-2 Rx (Weight Loaded) 2CHS*MOV289 55 Gate Open Open As is Closed SIS IO No No Inherent-2CHS* 15-4 Elect Manual None 10 None MCC*2-E05 9.3.4 coolant Check RV8144/2735 X-46 2-CHS-002-124-2 Borated (Weight Loaded) 2CHS*FCV160 55 Globe Admin Closed Closed Closed None IO No No Inherent-2CHS* 14-0 Air None None N/A None None 9.3.4 water Check Closed (19) RV160/2690 X-19 2-CHS-003-106-2 Rx 2CHS*MOV378 2CHS*MOV381 55 Gate Open Open As is Closed CIA IO Yes Yes Bypass to 28-4 Elect Manual 60 60 MCC*2-E05 MCC*2-E06 9.3.4 2-CHS-025-396-2 coolant (Weight Loaded) Yes - RV 382A/150 None Check X-35 2-CHS-002-95-2 Rx (Weight Loaded) 2CHS*MOV308A 55 Globe Open Open As is (9) (9) (9) No No Inherent-2CHS* 15-4 Elect Manual None N/A None MCC*2-E05 9.3.4 coolant Check RV260A/2690 X-36 2-CHS-002-94-2 Rx (Weight Loaded) 2CHS*MOV308B 55 Globe Open Open As is (9) (9) (9) No No Inherent-2CHS* 20-0 Elect Manual None N/A None MCC*2-E05 9.3.4 coolant Check RV260B/2690 X-37 2-CHS-002-93-2 Rx (Weight Loaded) 2CHS*MOV308C 55 Globe Open Open As is (9) (9) (9) No No Inherent-2CHS* 27-0 Elect Manual None N/A None MCC*2-E05 9.3.4 coolant Check RV260C/2690 Steam generator X-39 2-BDG-025-11-2 Demin None (4) 2BDG*AOV100A1 (20) Globe Open Closed Closed Closed (4) IO - No (10) 4-4 Air None 60 BATT*2-1 10.4.8 blowdown water system X-40 2-BDG-025-262-2 Demin None (4) 2BDG*AOV100B1 (20) Globe Open Closed Closed Closed (4) IO - No (10) 13-4 Air None 60 BATT*2-1 10.4.8 water X-41 2-BDG-025-264-2 Demin None (4) 2BDG*AOV100C1 (20) Globe Open Closed Closed Closed (4) IO - No (10) 11-4 Air None 60 BATT*2-1 10.4.8 water Component X-2 2-CCP-018-30-2 (13) 2CCP*MOV150-2 2CCP*MOV150-1 57 Btfy Open Open As is Closed CIB IO Yes Yes RV/150 6-0 Elect Manual 60 60 MCC*2-E05 MCC*2-E06 9.2.2 cooling water 2CCP*RV102 Relief None Yes system X-4 2-CCP-018-33-2 (13) 2CCP*MOV151-2 2CCP*MOV151-1 57 Btfy Open Open As is Closed CIB IO Yes Yes RV/150 3-8 Elect Manual 60 60 MCC*2-E06 MCC*2-E05 9.2.2 2CCP*RV103 Relief None Yes X-1 2-CCP-018-39-2 (13) 2CCP*MOV157-2 2CCP*MOV157-1 57 Btfy Open Open As is Closed CIB IO Yes Yes RV/150 5-8 Elect Manual 60 60 MCC*2-E06 MCC*2-E05 9.2.2 2CCP*RV105 Relief None Yes X-5 2CCP-018-36-2 (13) 2CCP*MOV156-2 2CCP*MOV156-1 57 Btfy Open Open As is Closed CIB IO Yes Yes RV/150 5-8 Elect Manual 60 60 MCC*2-E05 MCC*2-E06 9.2.2 2CCP*RV104 Relief None Yes 1 of 10

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 (Cont)

Isol- Distance ation/ from Outer-Cnmt Safety Overpressure most Valve General Valve Position Isolation Func- Type C Leak Protection/ to Contain- Valve Closure Safety-Related Ref.

Penet Isolation Valves Design Valve Power Signal tion Test Required Set point ment Valve Actuation Time (sec) (24) Power Source FSAR System Name No. Line No. (1) Fluid Inside Outside Criteria Type Normal Shutdown Failure DBA (16) (7) Inside Outside (psig) Wall Primary Secondary Inside Outside Inside Outside Sect.

Vent and drain X-38 2-DAS-002-19-2 Borated 2DAS*AOV100A 2DAS*AOV100B 56 Globe Open Open Closed Closed CIA IO Yes Yes RV/70 11-4 Air None 60 60 BATT*2-1 BATT*2-2 9.3.3 systems Water 2DAS*RV110 Relief - - - - None Yes X-48 2VRS-150-12-2 (14) 2VRS*AOV109A2 2VRS*AOV109A1 56 Globe Open Open Closed Closed CIA IO Yes Yes (10) 1-6 Air None 60 60 BATT*2-2 BATT*2-1 9.3.3 X-29 2-DGS-002-13-2 Borated 2DGS*AOV108A 2DGS*AOV108B 56 Globe Open Open Closed Closed CIA IO Yes Yes RV/125 5-8 Air None 60 60 BATT*2-1 BATT*2-1 9.3.3 water 2DGS*RV115 Relief - - - - None Yes Fuel pool cooling X-103 2-FNC-006-112-2 Dry Manual Manual 56 Ball Admin Closed Closed Closed None IO Yes Yes (10) 6-8 Manual None - - None None 9.1.3 and purification Closed system X-104 2-FNC-006-111-2 Dry Manual Manual 56 Ball Admin Closed Closed Closed None IO Yes Yes (10) 3-4 Manual None - - None None 9.1.3 Closed Reactor coolant X-45 2-RCS-003-151-2 Borated (Weight Loaded) 2RCS*AOV519 55 Globe Closed Closed Closed Closed CIA IO Yes Yes RV/150 16-4 Air None - 60 None BATT*2-2 5.1, system water Check 2RCS*RV100 Relief - - - - None Yes 9.2.8 X-49 2-RCS-750-120-2 Nitrogen (Weight Loaded) 2RCS*AOV101 55 Globe Closed Closed Closed Closed CIA IO Yes Yes (10) 12-0 Air None - 60 None BATT*2-2 5.1, Check 9.5.9 X-119a 2-RCS-750-301-2 (22) None (21) (21) (21) - - - - None S/M No No (10) (21) (21) (21) (21) (21) (21) (21) 6.2.4, thru 2-RCS-750-302-2 7.5.3.2 X-119f 2-RCS-750-303-2 (RVLIS)

Post-DBA X-55c 2-HCS-375-65-2 Cont atm 2HCS*SOV136A 2HCS*SOV136B 56 Globe Closed Closed Closed Open None S/M Yes Yes (10) (later) Elect None N/A N/A PNL*DC2-15 PNL*DC2-15 6.2.5 hydrogen control system X-57c 2-HCS-375-66-2 Cont atm 2HCS*SOV135A 2HCS*SOV135B 56 Globe Closed Closed Closed Open None S/M Yes Yes (10) (later) Elect None N/A N/A PNL*DC2-16 PNL*DC2-16 6.2.5 X-87 2-HCS-002-52-2 Cont atm None 2HCS*MOV117 56 Ball Closed Closed As is Closed None HC - Yes (10) 3-0 Elect Manual - N/A None MCC*2-E12 6.2.5 (12)

Manual Ball Admin Closed Closed Open - HC Yes (10) 9-6 Closed X-88 2-HCS-002-51-2 Cont atm None 2HCS*MOV116 56 Ball Closed Closed As is Closed None HC - Yes (10) 2-3 Elect Manual - N/A None MCC*2-E11 6.2.5 (12)

Manual Ball Admin Closed Closed Open - HC Yes (10) 9-6 Closed X-92 2-HCS-002-19-2 Cont atm None 2HCS*SOV114B (8) (11) Closed Closed Closed Closed None HC - Yes (10) 6-3 Elect None - N/A None PNL*AC2-E1 6.2.5 (12) 2HCS*SOV115B Closed Closed Closed Closed None HC Yes (10) 9-3 Elect None N/A PNL*AC2-E2 (12)

X-93 2-HCS-002-15-2 Cont atm None 2HCS*SOV114A (8) (11) Closed Closed Closed Closed None HC - Yes (10) 9-3 Elect None - N/A None PNL*AC2-E1 6.2.5 (12) 2HCS*SOV115A (11) Closed Closed Closed Closed None HC - Yes (10) 11-9 Elect None N/A PNL*AC2-E2 (12)

X-97b 2HCS-375-68-2 Cont atm 2HCS*SOV133B 2HCS*SOV134B 56 Globe Closed Closed Closed Open None S/M Yes Yes (10) (later) Elect None N/A N/A PNL*DC2-16 PNL-DC2-16 6.2.5 X-105b 2HCS-375-67-2 Cont atm 2HCS*SOV133A 2HCS*SOV134A 56 Globe Closed Closed Closed Open None S/M Yes Yes (10) (later) Elect None N/A N/A PNL*DC2-15 PNL-DC2-15 6.2.5 2 of 10

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 (Cont)

Distance Isol- from Outer-Cnmt ation/ Overpressure most Valve General Valve Position Isolation Safety Type C Leak Protection/ to Contain- Valve Closure Safety-Related Ref.

Pene Isolation Valves Design Valve Power Signal Func- Test Required Set point ment Valve Actuation Time (sec) (24) Power Source FSAR System Name t Line No. (1) Fluid Inside Outside Criteria Type Normal Shutdown Failure DBA (16) tion (7) Inside Outside (psig) Wall Primary Secondary Inside Outside Inside Outside Sect.

No.

Containment X-93 2-CVS-002-21-2 Cont atm None 2CVS*SOV151A (8) (11) Open Open Closed Closed CIA IO - Yes (10) 13-6 Elect None - 5 None PNL*DC2-07 9.5.10 vacuum and 2CVS*SOV152A Open Open Closed Closed CIA IO Yes (10) 16-6 Elect None <5 PNL*DC2-07 leakage monitoring systems X-92 2-CVS-002-22-2 Cont atm None 2CVS*SOV151B (8) (11) Open Open Closed Closed CIA IO - Yes (10) 16-0 Elect None - 5 None PNL*DC2-07 9.5.10 2CVS*SOV152B Open Open Closed Closed CIA IO Yes (10) 18-3 Elect None 5 PNL*DC2-06 X-94 2-CVS-008-28-2 Cont atm Manual Manual 56 Btfy Admin Closed Closed Closed None IO Yes Yes (10) 1-0 Manual None - - None None 9.5.10 Closed X-44 2-CVS-001-210-2 Cont atm 2CVS*SOV153B 2CVS*SOV153A 56 (11) Open Open Closed Closed CIA IO Yes Yes (10) 13-8 Elect None 60 60 PNL*DC2-06 PNL*DC2-07 9.5.10 X-43 2-CVS-001-211-2 Cont atm (Spring Loaded) 2CVS*SOV102 56 Globe Open Open Closed Closed CIA SM Yes Yes (10) 24-4 Elect None 60 BATT*2-2 9.5.10 Check (12)

X- 2-LMS-375-3-2 Cont atm None 2LMS*SOV953 (8) (11) Open Open Open Open None ESF - No (10) 5-0 Elect None - 60 None BATT*2-2 9.5.10 55b X- 2-LMS-375-19-2 Cont atm None 2LMS*SOV950 (8) (11) Open Open Open Open None ESF - No (10) 12-0 Elect None - 60 None BATT*2-1 9.5.10 57b X- 2-LMS-375-1-2 Cont atm None 2LMS*SOV952 (8) (11) Open Open Open Open None ESF - No (10) 11-0 Elect None - 60 None BATT*2-2 9.5.10 97a X- 2-LMS-375-2-2 Cont atm None 2LMS*SOV951 (8) (11) Open Open Open Open None ESF - No (10) 6-0 Elect None - 60 None BATT*2-1 9.5.10 105c 2LMS-375-30-2 Cont atm None Manual (8) Globe Admin Closed Closed Closed None IO - Yes (10) 1-0 Manual None - - None None X- Closed 105d Manual (8) Globe Admin Closed Closed Closed None IO Yes (10) 1-6 Manual None None Closed Main steam X-73 2-MSS-032-35-2 Steam None 2MSS*AOV101A (20) Globe Open Closed Closed Closed SLI IO - No (6) 20-2 Air None - 6 None (2) 10.3 system 2-MSS-002-920-2 Steam None 2MSS*AOV102A Globe Closed Closed Closed Closed SLI IO - No (6) 8-4 Air None - N/A None (2) 2-MSS-003-902-2 Steam None 2MSS*SOV105A (11) Closed Closed Open Open SIS ESF - No (6) 24-11 Elect None - N/A None PNL*DC2-07 2-MSS-375-244-2 Steam None 2MSS*SOV120 (20) Globe Closed Closed Closed Open SIS SM - No (6) 11-10 Elect None - N/A None PNL*DC2-03 10.3 2-MSS-006-158-2 Steam None 2MSS*SV101A (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 2-MSS-006-157-2 Steam None 2MSS*SV102A (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 2-MSS-006-156-2 Steam None 2MSS*SV103A (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 2-MSS-006-155-2 Steam None 2MSS*SV104A (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 2-MSS-006-154-2 Steam None 2MSS*SV105A (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 X-74 2-MSS-032-39-2 Steam None 2MSS*AOV101B (20) Globe Open Closed Closed Closed SLI IO - No (6) 18-8 Air None - 6 None (2) 10.3 2-MSS-002-921-2 Steam None 2MSS*AOV102B Globe Closed Closed Closed Closed SLI IO - No (6) 8-0 Air None - N/A None (2) 2-MSS-003-903-2 Steam None 2MSS*SOV105B (11) Closed Closed Open Open SIS ESF - No (6) 23-5 Elect None - N/A None PNL*DC2-07 2-MSS-375-244-2 Steam None 2MSS*SOV120 (20) Globe Closed Closed Closed Open SIS SM - No (6) 11-10 Elect None - N/A None PNL*DC2-03 10.3 2-MSS-006-163-2 Steam None 2MSS*SV101B (20) - - - - - None - - No (6) 11-5 - - - - None None 10.3 2-MSS-006-162-2 Steam None 2MSS*SV102B (20) - - - - - None - - No (6) 11-5 - - - - None None 10.3 2-MSS-006-161-2 Steam None 2MSS*SV103B (20) - - - - - None - - No (6) 11-5 - - - - None None 10.3 2-MSS-006-160-2 Steam None 2MSS*SV104B (20) - - - - - None - - No (6) 11-5 - - - - None None 10.3 2-MSS-006-159-2 Steam None 2MSS*SV105B (20) - - - - - None - - No (6) 11-5 - - - - None None 10.3 X-75 2-MSS-032-43-2 Steam None 2MSS*AOV101C (20) Globe Open Closed Closed Closed SLI IO - No (6) 20-2 Air None - 6 None (2) 10.3 2-MSS-002-922-2 Steam None 2MSS*AOV102C Globe Closed Closed Closed Closed SLI IO - No (6) 8-4 Air None - N/A None (2) 2-MSS-003-904-2 Steam None 2MSS*SOV105C (11) Closed Closed Open Open SIS ESF - No (6) 20-0 Elect None - N/A None PNL*DC2-07 2-MSS-375-244-2 Steam None 2MSS*SOV120 (20) Globe Closed Closed Closed Open SIS SM - No (6) 11-10 Elect None - N/A None PNL*DC2-03 10.3 2-MSS-006-168-2 Steam None 2MSS*SV101C (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 2-MSS-006-167-2 Steam None 2MSS*SV102C (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 2-MSS-006-166-2 Steam None 2MSS*SV103C (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 2-MSS-006-165-2 Steam None 2MSS*SV104C (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 2-MSS-006-164-2 Steam None 2MSS*SV105C (20) - - - - - None - - No (6) 14-5 - - - - None None 10.3 3 of 10

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 (Cont)

Distance Isol- Over- from Outer-Cnmt ation/ pressure most Valve General Valve Position Isolation Safety Type C Leak Protection/ to Contain- Valve Closure Safety-Related Ref.

System Penet Isolation Valves Design Valve Power Signal Func- Test Required Set point ment Valve Actuation Time (sec)(24) Power Source FSAR Name No. Line No. (1) Fluid Inside Outside Criteria Type Normal Shutdown Failure DBA (16) tion (7) Inside Outside (psig) Wall Primary Secondary Inside Outside Inside Outside Sect.

Steam drains X-73 2-SDS-150-76-2 Drains/ None 2SDS*AOV111A-1 (20) Globe Open Open Closed Closed SLI IO - No (6) 17-4 Air None - 60 None BATT*2-1 10.3 system 2-SDS-001-97-2 Steam None 2SDS*AOV129B (20) Globe Open Open Closed Closed SLI IO - No (6) 86 Air None - 60 None BATT*2-2 10.3 X-74 2-SDS-150-77-2 Drains/ None 2SDS*AOV111B-1 (20) Globe Open Open Closed Closed SLI IO - No (6) 15-4 Air None - 60 None BATT*2-1 10.3 2-SDS-001-97-2 Steam None 2SDS*AOV129B (20) Globe Open Open Closed Closed SLI IO - No (6) 86 Air None - 60 None BATT*2-2 10.3 X-75 2-SDS-150-78-2 Drains/ None 2SDS*AOV111C-1 (20) Globe Open Open Closed Closed SLI IO - No (6) 17-4 Air None - 60 None BATT*2-1 10.3 2-SDS-001-97-2 Steam None 2SDS*AOV129B (20) Globe Open Open Closed Closed SLI IO - No (6) 86 Air None - 60 None BATT*2-2 10.3 Steam vent X-73 2-SVS-010-170-2 Steam None 2SVS*PCV101A (20) Globe Closed Open Closed Open None IO/ESF - No (6) 25-6 Hyd Press Hand-Pump - N/A None MCC*2-E05 10.3 system 2-SVS-010-177-2 Steam None 2SVS*HCV104 (20) Globe Closed Open Closed Open None IO/ESF - No (6) 85 Hyd Press Hand-Pump - N/A None MCC*2-E14 10.3 X-74 2-SVS-010-172-2 Steam None 2SVS*PCV101B (20) Globe Closed Open Closed Open None IO/ESF - No (6) 26-6 Hyd Press Hand-Pump - N/A None MCC*2-E13 10.3 2-SVS-010-177-2 Steam None 2SVS*HCV104 (20) Globe Closed Open Closed Open None IO/ESF - No (6) 85 Hyd Press Hand-Pump - N/A None MCC*2-E14 10.3 X-75 2-SVS-010-174-2 Steam None 2SVS*PCV101C (20) Globe Closed Open Closed Open None IO/ESF - No (6) 27-0 Hyd Press Hand-Pump - N/A None MCC*2-E13 10.3 2-SVS-010-177-2 Steam None 2SVS*HCV104 (20) Globe Closed Open Closed Open None IO/ESF - No (6) 85 Hyd Press Hand-Pump - N/A None MCC*2-E14 10.3 Gas supply X-53 2-GNS-001-10-2 Nitrogen 2GNS*AOV101-2 2GNS*AOV101-1 55 Globe Open Open Closed Closed CIA IO Yes Yes (10) 15-4 Air None 10 60 BATT*2-2 BATT*2-1 9.5.9 system Containment X-90 2-HVR-042-1-2 Cont atm 2HVR*MOD23B 2HVR*MOD23A 56 Btfy Admin (15) Closed Closed RM IO Yes Yes (10) 2-9 Elect None 10 10 MCC*2-E14 MCC*2-E13 9.4.7.3 purge system Closed X-91 2-HVR-042-2-2 Cont atm 2HVR*MOD25B 2HVR*MOD25A 56 Btfy Admin (15) Closed Closed RM IO Yes Yes (10) 2-9 Elect None 10 10 MCC*2-E14 MCC*2-E13 9.4.7.3 Closed 2-HVR-008-3-2 Cont air - 2HVR*DMP206 56 Ball Admin (15) Closed Closed RM IO Yes (10) 2-9 Manual None - - None None 9.4.7.3 Closed Feedwater X-76 2-FWS-016-12-2 Demin None 2FWS*HYV157A (20) Gate Open Closed As is Closed FWI IO - No (6) 6-8 Hyd Press None - (26) None MCC*2-E09 10.4.7 and auxiliary water feedwater systems X-77 2-FWS-016-17-2 Demin None 2FWS*HYV157B (20) Gate Open Closed As is Closed FWI IO - No (6) 5-0 Hyd Press None - (26) None MCC*2-E09 10.4.7 water X-78 2-FWS-016-22-2 Demin None 2FWS*HYV157C (20) Gate Open Closed As is Closed FWI IO - No (6) 5-8 Hyd Press None - (26) None MCC*2-E09 10.4.7 water X-79 2FWE-003-18-2 Demin None 2FWE*HCV100E (20) Globe (3) Open Open As is Open None ESF No (6) 40-8 Elect/Hyd Hand-pump - N/A None MCC*2-E13 10.4.9 water 2FWE*HCV100F (20) Globe Open Open As is Open None ESF No (6) 50-9 Elect/Hyd Hand-pump - N/A MCC*2-E14 X-80 2-FWE-003-16-2 Demin None 2FWE*HCV100C (20) Globe (3) Open Open As is Open None ESF No (6) 42-0 Elect/Hyd Hand-pump - N/A None MCC*2-E13 10.4.9 water 2FWE*HCV100D Globe Open Open As is Open None ESF No (6) 22-8 Elect/Hyd Hand-pump - N/A MCC*2-E14 4 of 10

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 (Cont)

Distance Isol- Over- from Outer-Cnmt ation/ pressure most Valve General Valve Position Isolation Safety Type C Leak Protection/ to Contain- Valve Closure Safety-Related Ref.

Penet Isolation Valves Design Valve Power Signal Func- Test Required Set point ment Valve Actuation Time (sec)(24) Power Source FSAR System Name No. Line No. (1) Fluid Inside Outside Criteria Type Normal Shutdown Failure DBA (16) tion (7) Inside Outside (psig) Wall Primary Secondary Inside Outside Inside Outside Sect.

X-83 2-FWE-003-13-2 Demin None 2FWE*HCV100A (20) Globe (3) Open Open As is Open None ESF No No (6) 27-9 Elect/hyd Hand-pump - N/A None MCC*2-E13 10.4.9 water 2FWE*HCV100B Globe (3) Open Open As is Open None ESF No (6) 39-7 Elect/hyd Hand-pump - N/A MCC*2-E14 Residual heat X-24 2-RHS-006-14-2 Borated Manual Manual 55 Globe Admin Closed Closed Closed None IO Yes Yes RV/455 3-3 Manual None - - None None 5.4.7 removal water 2RHS*RV100 Relief Closed - - Yes - - - - - -

system Quench spray X-63 2-QSS-010-4-2 Borated (Weight 2QSS*MOV101A (8) Gate (5) Open Open As is Open CIB(17) CS Yes Yes (10) 9-0 Elect Manual - N/A (25) None MCC*2-E11 6.2.2 system water Loaded) 2QSS*RV101A Relief - - - - None - - Yes - - - - - - - None Check X-64 2-QSS-010-3-2 Borated (Weight 2QSS*MOV101B (8) Gate (5) Open Open As is Open CIB(17) CS Yes Yes (10) 10-0 Elect Manual - N/A (25) None MCC*2-E12 6.2.2 water Loaded) 2QSS*RV101B Relief - - - - None - - Yes - - - - - - - None Check Recirculation X-66 2-RSS-012-5-2 Borated None 2RSS*MOV155A (8) Btfy Open Open As is Open CIB(17) CS - No (10) 1-6 Elect Manual - N/A (25) - MCC*2-E11 6.2.2 spray system water -

X-67 2-RSS-012-6-2 Borated None 2RSS*MOV155C (8) Btfy Open Open As is Open CIB(17) CS/ESF - No (10) 1-6 Elect Manual - N/A (25) - MCC*2-E11 6.2.2 water -

X-68 2-RSS-012-7-2 Borated None 2RSS*MOV155D (8) Btfy Open Open As is Open CIB(17) CS/ESF - No (10) 1-6 Elect Manual - N/A (25) - MCC*2-E12 6.2.2 water -

X-69 2-RSS-012-8-2 Borated None 2RSS*MOV155B (8) Btfy Open Open As is Open CIB(17) CS - No (10) 1-6 Elect Manual - N/A (25) - MCC*2-E12 6.2.2 water -

X-70 2-RSS-012-3-2 Borated (Weight 2RSS*MOV156A 56 Gate(5) Open Open As is Open CIB(17) CS No No (10) 6-11 Elect Manual - N/A (25) None MCC*2-E11 6.2.2 water Loaded) 2RSS*RV156A - Relief - - - - None - - No - - - - - - - None Check X-71 2-RSS-012-4-2 Borated (Weight 2RSS*MOV156C 56 Gate(5) Open Open As is Open CIB(17) CS No No (10) 7-6 Elect Manual - 60 None MCC*2-E11 6.2.2 water Loaded) 2RSS*RV156C - Relief - - - - None - - No - - - - - - - None Check X-114 2-RSS-012-11-2 Borated (Weight 2RSS*MOV156D 56 Gate(5) Open Open As is Open CIB(17) CS No No (10) 7-5 Elect Manual - 60 None MCC*2-E12 6.2.2 water Loaded) 2RSS*RV156D - Relief - - - - None - - No - - - - - - - None Check X-115 2-RSS-012-12-2 Borated (Weight 2RSS*MOV156B 56 Gate(5) Open Open As is Open CIB(17) CS No No (10) 7-5 Elect Manual - N/A (25) None MCC*2-E12 6.2.2 water Loaded) 2RSS*RV156B - Relief - - - - None - - No - - - - - - - None Check Air supply X-11 2-IAC-004-31-2 Cont inst 2IAC*MOV133 2IAC*MOV134 56 Ball CLOSED CLOSED As is Closed CIA(28) IO Yes Yes (10) 25-6 Elect Manual N/A N/A MCC* MCC*2-E14 9.3.1 systems air 2-E13 X-59 2-IAC-003-49-2 Cont inst (Weight 2IAC*MOV130 56 Ball Open Open As is Closed CIA IO Yes Yes (10) 15-3 Elect Manual - 60 None MCC*2-E13 9.3.1 air Loaded)

Check 5 of 10

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 (Cont)

Distance Isol- from Outer-Cnmt ation/ Overpressure most Valve General Valve Position Isolation Safety Type C Leak Protection/ to Contain- Valve Closure Safety-Related Ref.

Penet Isolation Valves Design Valve Power Signal Func- Test Required Set point ment Valve Actuation Time (sec) (24) Power Source FSAR System Name No. Line No. (1) Fluid Inside Outside Criteria Type Normal Shutdown Failure DBA (16) tion (7) Inside Outside (psig) Wall Primary Secondary Inside Outside Inside Outside Sect.

X-42 2-SAS-002-66-2 Service Manual Manual 56 Globe Admin Closed Closed Closed None IO Yes Yes (10) 4-9 Manual None - - None None 9.3.1 air Closed Service water X-21 2-SWS-008-291-2 Service 2SWS*MOV155-2 2SWS*MOV155-1 57 Btfy Open Open As is Closed CIB IO Yes Yes RV/150 9-4 Elect Manual 60 60 MCC*2-E06 MCC*2-E05 9.2.1 system water 2SWS*RV155 Relief - - - - None - - Yes - - - - - - - None X-25 2-SWS-008-290-2 Service 2SWS*MOV154-2 2SWS*MOV154-1 57 Btfy Admin Closed As is Closed CIB (27) IO Yes Yes RV/150 9-4 Manual None N/A N/A None None 9.2.1 water 2SWS*RV154 Relief Closed - - - None - - Yes - - - - - - - None X-14 2-SWS-008-285-2 Service 2SWS*MOV153-2 2SWS*MOV153-1 57 Btfy Admin Closed As is Closed CIB (27) IO Yes Yes RV/150 9-4 Manual None N/A N/A None None 9.2.1 water 2SWS*RV153 Relief Closed - - - None - - Yes - - - - - - - None X-27 2SWS-008-284-2 Service 2SWS*MOV152-2 2SWS*MOV152-1 57 Btfy Open Open As is Closed CIB IO Yes Yes RV/150 9-4 Elect Manual 60 60 MCC*2-E06 MCC*2-E05 9.2.1 water 2SWS*RV152 Relief - - - - None - - Yes - - - - - - - None Safety injection X-61 2-SIS-010-9-2 Borated (Weight Loaded) 2SIS*MOV8889 (8) Gate Closed Closed As is Open None ESF No No Inherent 5-6 Elect Manual - N/A None MCC*2-E12 6.3 system water Check X-7 2-SIS-003-101-2 Borated (Weight Loaded) 2SIS*MOV869A (8) Gate Closed Closed As is Open None ESF No No Inherent 26-8 Elect Manual - N/A None MCC*2-E05 6.3 water Check X-17 2-SIS-003-88-2 Borated (Weight Loaded) 2SIS*MOV869B (8) Gate Closed Closed As is Open None ESF No No Inherent 35-0 Elect Manual - N/A None MCC*2-E06 6.3 water Check X-60 2-SIS-010-11-2 Borated (Weight Loaded) 2SIS*MOV8888B (8) Gate Open Open As is Open None ESF No No Inherent 13-0 Elect Manual - N/A None MCC*2-E12 6.3 water Check X-62 2-SIS-010-10-2 Borated (Weight Loaded) 2SIS*MOV8888A (8) Gate Open Open As is Open None ESF No No Inherent 6-0 Elect Manual - N/A None MCC*2-E11 6.3 water Check X-34 2-SIS-003-102-2 Borated (Weight Loaded) 2SIS*MOV836 (8) Gate Closed Closed As is Open None ESF No No Inherent 21-4 Elect Manual - N/A None MCC*2-E05 6.3 water Check 2-SIS-001-393-2 Borated 2SIS*MOV840 Closed Open Closed Closed None - - No Inherent 31-4 Elect None - N/A - PNL*AC2-E1 water X-113 2-SIS-003-96-2 Borated (Weight Loaded) 2SIS*MOV867C (8) Gate Closed Closed As is Open SIS(17) ESF No No Inherent 9-4 Elect Manual - 10 None MCC*2-E05 6.3 water Check 2SIS*MOV867D Gate Closed Closed As is Open SIS(17) - - No Inherent 13-8 Elect Manual - 10 - MCC*2-E06 X-20 2-SIS-001-32-2 Borated (Weight Loaded) Manual 55 Globe Admin Closed Closed Closed None IO Yes Yes RV/700 6-8 Manual None - - None None 6.3 water Check 2SIS*RV130 Closed

- - - - None - - Yes - - - - - - - None X-106 2-SIS-750-53-2 Borated 2SIS*MOV842 2SIS*AOV889 55 Globe Closed/ Open/ As is/ Closed/ CIA IO Yes Yes RV/700 9-0 Elect/ Manual/ 60 60 MCC*E-06 BATT*2-1 6.3 water 2SIS*RV175 Relief Closed Closed Closed Closed - - Yes - - Air None - - - -

6 of 10

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 (Cont)

Distance Isol- from Outer-Cnmt ation/ Overpressure most Valve General Valve Position Isolation Safety Type C Leak Protection/ to Contain- Valve Closure Safety-Related Ref.

Penet Isolation Valves Design Valve Power Signal Func- Test Required Set point ment Valve Actuation Time (sec) (24) Power Source FSAR System Name No. Line No. (1) Fluid Inside Outside Criteria Type Normal Shutdown Failure DBA (16) tion (7) Inside Outside (psig) Wall Primary Secondary Inside Outside Inside Outside Sect.

Fire protection X-99 2-FPW-006-10-2 Dry or (Weight Loaded) 2FPW*AOV206 56 Globe Closed Open Closed Closed CIA IO Yes Yes (10) 20-0 Air None None 60 None PNL*DC2-07 9.5.1 water system water & Check air X-101 2-FPW-004-13-2 Dry (Weight Loaded) 2FPW*AOV205 56 Globe Closed Open Closed Closed CIA IO Yes Yes (10) 9-3 Air None None 60 None PNL*DC2-07 9.5.1 Check Sample system X-55a 2-SSR-375-62-2 Nitrogen 2SSR*SOV130A1 2SSR*SOV130A2 55 Globe Closed Closed Closed Closed CIA S/M Yes Yes (10) 13-8 Elect None 60 60 BATT*2-1 BATT*2-2 9.3.2 X-55d 2-SSR-375-17-2 Borated 2SSR*AOV109A1 2SSR*AOV109A2 55 Globe Open Open Closed Closed CIA IO Yes Yes - 12-8 Air None <60 <60 BATT*2-1 BATT*2-2 -

water 2SSR*RV117 Relief - - - - Yes RV/700 None -

X-56c 2-SSR-375-225-2 Borated 2SSR*AOV102A1 2SSR*AOV102A2 55 Globe Closed Closed Closed Closed CIA IO Yes Yes 8-4 Air None 60 60 BATT*2-1 BATT*2-2 9.3.2 water 2SSR*RV118 Relief - - - - Yes RV/2485 None X-56b 2-SSR-375-224-2 Borated 2SSR*SOV128A1 2SSR*SOV128A2 55 Globe Closed Closed Closed Closed CIA S/M Yes Yes 15-8 Elect None 60 60 BATT*2-1 BATT*2-2 9.3.2 water 2SSR*RV120 Relief - - - - Yes RV/2485 None X-56d 2-SSR-375-221-2 Borated 2SSR*AOV100A1 2SSR*AOV100A2 55 Globe Closed Closed Closed Closed CIA IO Yes Yes 12-0 Air None 60 60 BATT*2-1 BATT*2-2 9.3.2 water 2SSR*RV119 Relief - - - - Yes RV/2485 None X-56a 2-SSR-750-65-2 Demin None 2SSR*AOV117A (8) Globe Open Open Closed Closed None IO - No (10) 9-4 Air None - 60 None BATT*2-1 9.3.2 water X-57d 2-SSR-750-76-2 Demin None 2SSR*AOV117B (8) Globe Open Open Closed Closed None IO - No (10) 15-8 Air None - 60 None BATT*2-1 9.3.2 water X-57a 2-SSR-375-54-2 Steam 2SSR*AOV112A1 2SSR*AOV112A2 55 Globe Closed Closed Closed Closed CIA IO Yes Yes RV/2485 6-8 Air None 60 60 BATT*2-1 BATT*2-2 2SSR*RV121 Relief - - - - Yes None X-97d 2-SSR-750-71-2 Demin None 2SSR*AOV117C (8) Globe Open Open Closed Closed None IO - No (10) 6-8 Air None - 60 None BATT*2-1 9.3.2 water X-97c 2-SSR-375-64-2 Borated 2SSR*SOV129A1 2SSR*SOV129A2 55 Globe Closed Closed Closed Closed CIA S/M Yes Yes 6-8 Elect None 60 60 BATT*2-1 BATT*2-2 water 2SSR*RV122 Relief - - - - - - Yes RV/600 None Post Accident X-105a 2-PAS-375-46-2 Cont 2PAS*SOV105A1 2PAS*SOV105A2 56 Globe Closed Closed Closed Closed CIA S/M Yes Yes (10) (Later) Elect None 60 60 PNL*DC2-03 PNL*DC2-10 Sampling atm Personnel air None None Air 2PHS*112 2PHS*110 56 Ball Admin Closed Closed Closed None IO (8) (8) (10) 10-0 Manual None - - - - 3.8.1 lock closed 2PHS*113 2PHS*111 56 Ball Admin Closed Closed Closed None IO (8) (8) (10) 10-0 Manual None - - - -

closed 2PHS*101 2PHS*100 56 Ball Admin Closed Closed Closed None IO (8) (8) (10) 10-0 Manual None - - None None closed Emergency air None None Air 2PHS*202 2PHS*201 56 Ball Admin Closed Closed Closed None IO (8) (8) (10) 3-0 Manual None - - None None 3.8.1 lock closed Fuel transfer X-65 None Air Bld. Flange Manual 56 Gate Admin Closed Closed Closed None IO (8) (23) (10) - Manual None - - None None tube closed 7 of 10

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 (Cont)

NOTES

1. The middle set of digits represents the line size. For example, for 2-CHS-002-2-2, 002 indicates a 2-inch line. All line sizes are represented by 3 digits, as indicated below:

Pipe Size Per Pipe Size, In. Line Designation 1/4 250 3/8 375 1/2 500 3/4 750 1 001 1 1/4 125 1 1/2 150 1 3/4 175 2 002 2 1/2 025 3 003 4 004 6 006 8 008 10 010 12 012 14 014 16 016 18 018 20 020 22 022 24 024 26 026 28 028 30 030 32 032 34 034 36 036 38 038 42 042 108 108

2. Power is available from either BATT*2-1 or BATT*2-2.

3. Valve is a Globe type.

4. Valves close automatically from control signals which indicate that the motor-driven auxiliary feedwater pumps have started, or that steam supplies have been initiated to the turbine-driven auxiliary feedwater pump, that steam generator sample radiation is high, or that blowdown tank level is high-high. Valves do not receive a CIA or CIB signal.

5. Relief valve is actually part of the isolation valve.

6. The ASME III, Class 2 lines in the main steam, steam generator blowdown, feedwater, and auxiliary feedwater systems have overpressure protection from the main steam safety valves. These valves range from a low set point of 1,075 psig to a high point of 1,125 psig.

8 of 10

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 (Cont)

NOTES (Cont)

7. Isolation/Safety Function - The listed function defines the isolation requirement or need for operability of the penetration after the containment isolation signal is generated. The systems identified as ESF, CS, and HC are considered essential systems.

IO - Isolation only ESF - Engineered safety features operations to mitigate consequences of accident CS - Containment depressurization HC - Post-accident H2 control S/M - Post-accident sampling/monitoring.

8. Section 6.2.4.2 discusses the exceptions to the General Design Criteria.

9. Valves remain open following an accident to ensure seal water supply to the reactor coolant pump and will be manually closed by the operator from the control room for long-term containment isolation.

10. Overpressure protection is provided by means other than noted, such as conservative pipe and valve design, post-DBA operating system, no fluid in lines, open to containment, credit for insulation, etc.

11. Valve is a Y pattern disc type.

12. Valve must be opened/reopened in the long term to fulfill its safety function.

13. Component cooling water composition - demineralized water and corrosion inhibitor.

14. Steam, noncondensibles, and water vapor.

15. Valve is opened only during cold shutdown or refueling.

16. Containment Isolation Signals:

CIA - Containment Isolation Phase A CIB - Containment Isolation Phase B SIS - Safety Injection Signal SLI - Steam Line Isolation Signal FWI - Feedwater Isolation Signal RM - Containment High Radiation Signal

17. Containment Isolation Signal opens the isolation valve to ensure availability for accident condition use.

18. Containment Isolation Phase B (CIB) Signal closes and extreme low-low RWST signal coincident with CIB opens valves.

19. Associated instrumentation for 2CHS*FCV160 and 2CHS*HCV142 is not seismically or environmentally qualified and does not receive Class 1E power. Containment isolation automatic signal is not required since valves are administratively closed during normal operation.

9 of 10

BVPS-2 UFSAR Rev. 26 TABLE 6.2-60 (Cont)

NOTES (Cont)

20. Main steam, steam generator blowdown, feedwater penetrations.

All isolation valves on these lines, including sampling lines for the steam generator blowdown (SGB) system, are located outside containment since these systems are considered closed inside containment and are designed to Safety Class 2, QA Category I, Seismic Category I inside containment. The isolation arrangements comply with GDC 57. The outside isolation valves of the main steam, feedwater, and SGB systems receive a closure signal from either main steam isolation, feedwater isolation or initiation of the auxiliary feedwater system, respectively. Branch lines connecting to the main steam or feedwater piping between the MSIVs and the containment wall include the atmospheric steam dump lines and the steam lines to the turbine-driven auxiliary feedwater pump. The steam supply lines to the turbine driven auxiliary feedwater pump contain two valves in series, the nearest to containment is designated the containment isolation valve; however, the redundant valve is equivalent and may be used in place of the designated containment isolation valve. All of these lines have at least one remotely-operated valve which can be manually closed by the operator. However, the steam lines to the turbine-driven auxiliary feedwater pump are used for safe-shutdown operation and therefore receive an auxiliary feedwater initiation signal to open. The branch lines to the main steam discharge radiation monitor (2MSS*RQI101) are not isolated, but instead, the isolation valve (2MSS*SOV120) receives a signal to open on safety injection.

21. Hydraulic isolation device.

22. Deaerated and deionized water.

23. This valve is not required to be type C leak tested due to the double barrier seal arrangement on the fuel transfer tube inside containment isolation flange.

24. The closure times shown are based on maximum limits set by offsite dose calculations. Many valves have faster closure times. These faster closure times, as established by preoperational or post-maintenance testing, will be used as the basis for ASME OM Code closure time testing.

25. No stroke time will be applicable to this valve. This valve is open during normal and shutdown conditions, Fails as-is, and receives a CIA or CIB signal to open.

26. Feedwater isolation for these penetrations is 7 seconds, composed of signal processing and valve stroke time.

27. No credit is taken for the CIB actuation. The valve is locked shut in Modes 1, 2, 3 & 4.

28. No credit is taken for the CIA actuation. The valve is locked shut in Modes 1,2,3 & 4.

10 of 10

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Beaver Valley Power Station Unit No. 2 Updated Final Safety Analysis Report

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Jn: Junction number (A): Junction flow area (ft. ) Z is elevation (ft.)

2 g: Junction with grating and/or stairs c: junction normally closed f: junction opens on set AP Figure 6.2-3 MAAP5 Node and Junction Arrangement Beaver Valley Power Station Unit No. 2 Updated Final Safety Analysis Report

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FIGURE 6.2-15 CONTAINMENT TEMPERATURE TIME -

HISTORY FOR THE LIMITING MSLB TEMPERATURE CASE BEAVER VALLY POWER STATION UNIT No.2 UPDATED FINAL SAFETY ANALYSIS REPORT PREPARED OH<;:::"g~/ CJE.DlJI 29-JL.N-2~11 ~9:27 M: \u2\UFS!IR\g6~2~~15.dg" THE CHSU Glllt~ SfSTEII 0 v L _______ or * *~--- - - - - - - - - - - - - - - - - ~ * * * **

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FLOW VOLUME j I I I I I I I I I Wj-1 Pi Wj Pi+ I Wj +I I I I IL ____ _ I I I I

- - - _ _ J_ _ _ _ _ _ ~-----.J NOTES:

DASHED LINES INDICATE NODE BOUNDARIES OR MASS AND ENERGY CONTROL VOLUMES SOLID LINES INDICATE INTERNAL JUNCTION OR MOMENTUM CONTROL VOLUMES FIGURE 6.2-18 STAGGERED MESH CONTROL VOLUME APPROXIMATION FOR THREED BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

SET INITIAL CONDITIONS FOR PROBLEM COMPUTE TIME STEP AND ADVANCE TIME COMPUTE JUNCTION MASS FLOW RATES COMPUTE NODAL PROPERTIES PRINT AS REQUESTED NO FIGURE 6.2-19 COMPUTATIONAL BLOCK DIAGRAM FOR THREED BEAVER VALLEY POWER STATION* UNIT 2 FINAL SAFETY ANALYSIS REPORT

EL.790'-o" EL.780~6'"

EL. 787'-ld' E L

• 763'-11" EL* 741-nl' 4

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4

• t" EL. 741-114 EL. 738'*10" EL. 734'-1d' 17 15 PRESSURIZER RELIEF TANK CUBICLE El. 718'*6" NOTES:

I) NODE 19 IS THE CONTAINMENT

2) SYMBOLS a) D CONTROL VOLUME (OR NODE) b) 0 JUNCTION NUMBER c)__. INDICATES VENT PATHS TO THE CONTAINMENT FIGURE 6. 2-20 PRESSURIZER SUBCOMPARTMENT NODALIZATION DIAGRAM BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

. . "*p **. :.*o*;*.*:*~ *.*.. **

NOTES:

I. NODES 5 THROUGH 8 ARE LOCATED DIRECTLY BELOW NODES I THROUGH 4 RESPECTIVELY

2. ~"' INDICATES PLATFORM GRATING EL. 780'-6" X X X INDICATES WIRE MESH DOOR

3. DASHED LINES INDICATE NODAL BOUNDARIES FIGURE 6.2-21 PRESSURIZER CUBICLE NODALIZATION- PLAN VIEW 1 11 E 23 NODE MODEL BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

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t. NODES 20 THROUGH 23 ARE LOCATED DIRECTLY BELOW NODES 9 THROUGH 12 RESPECTIVELY FIGURE 6.2-ZZ

2. X X X INDICATES WIRE MESH DOORS PRESSURIZER CUBICLE

3. DASHED LINES INDICATE NODAL NODALIZATION 1

BOUNDARIES PLAN VIEW EL. 738 -10" 2 3 NODE MODEL BEAVER VALLEY POWER STATION-uNITZ FINAL SAFETY ANALYSIS REPORT

t..

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NOTES:

I. X X X INDICATES WIRE MESH DOOR

2. DASH LINE INDICATES NODAL BOUNDARY FIGURE 6. 2- 23 PRESSURIZER RELIEF TANK CUBICLE NODALIZATION 1 11 PLAN VIEW EL. 718 -6 2 3 NODE MODEL BEAVER VALLEY POWER STATION -UNIT2 FINAL SAFETY ANALYSIS REPORT

EL. 792' -0"

I 9

, NODE 2 NODE I i

(!

j_

t. .

• .") ~--..J:t<~X::....XQ...oQX...:X:....::rX..::....::X....:X:...XQo.~XI..,::X~X::...,X~I"X~X;::,q I El. 780'- 6" I

I v I NODE 6  : NODE 5

~*

I I

I I

~* ~-------,  !----*--- -;-* -~ ... : i.: *\..,

EL. 767'-10"

/)

.. v "1: I .. ~

1----~---

I 1

I NODE 10 I NODE 9 4* I I

I

• 6 I

I I

~ . I I

I I

l 1

. ~ ~- - - - - .....1~....-_ _'T.'T'_ ___,J-t II EL. 741-11/4 I

NODE 21 I NODE 20 4 I EL. 738'*10"

'----------,---I---....--~

• ~*. A: .... ~ ~ ~-.***:+*

IJ *A

,------~---'-

.j.

NODE 15

. . .A:

SECTION 1-1 NOTES:

I) NODE 18 IS LOCATED INSIDE THE PRESSURIZER SUPPORT SKIRT

2) SYMBOLS FIGURE 6.2-24

~ INDICATES GRATING PRESSURIZER CUBICLE ELEVATION VIEW SECTION 1-1

- - - NODAL BOUNDARIES BEAVER VALLEY POWER STATION- UNIT 2 FINAL SAFETY ANALYSIS REPORT

M

.--NODE 3 NODE 1-p-

r'

..q-

,C\

a:

(/)

en (L ......

w IJ a:::

J

~

J

(/) N

(/) ......

w a:::

(L 0

0.1 0.2 0.3 0.4 o.s 0.6 0.7 0.8 o.g 1.0 TIME {SEC)

FIGURE 6.2-25 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SPRAY LINE DER NODES 1 . 3 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

...... r (r

lJ)

...... NODE 2-, NODE 4

..q r~ boD a:

J" (f) a_ (Y')

~ ~v

(

w 0::::

(f) C\J (f) ......

w 0::::

a_

0 J

"b. 0 0 *1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 o.g 1.0 TIME (SEC)

FIGURE 6.2-26 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SPRAY LINE DER NODES 2 E. 4 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

(.1) lJ)

NODE 7~ -

/__,

{\/') :::._....---

v

..q ~NODE 5 a:

(J) a_

(Y)

~~

w 0:::

IV

l (J) N (J) .......

w 0:::

a_

~

0

)

0 b .o 0. 1 0.2 0.3 0.4 o.s 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6.2-27 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SPRAY LINE DER NODES 5 E. 7 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lD ---

II

M lJ)

NODE

~ NODE 8 v -

-..q {1, a:

(f)

('(') J 0.... ...... I w

a:::

J ~

(f) N (f) w a:::

I 0....

...... v 0

~ .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2-28 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SPRAY LINE DER NODES 6 E. 8 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

(1"')

N 0

NODES 9,10,11& 12 0')

CI

( f) a.... (0 w

0:::: r-

=>

(f)

(f) w w

0::::

a.... L/)

v (1"')

N

0. 1 0.2 0-3 0.4 0.5 0.6 0.7 o.s o.g 1.0 TIME (SEC)

FIGURE 6.2-29 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SPRAY LINE DER NODES 9, 10,11 E. 12 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

(Y")

N 0

NODES 13, 14,15 E.16 a: (7)

(/)

(L CD w

0:::: r-

J

(/)

(/) U) w '

0::::.

(L lf) v (Y")

N 0 *1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 .0 TIME (SEC)

FIGURE 6.2-29A PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SPRAY LINE DER NODES 13, 14, 15 E. 16 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

(Y'}

N I

...... I 0 -*

NODES 17,18,19 . 20 a: (1) o-i (f) ll.. CD w

Q:: ['

=>

(f)

(f) w w

Q::

ll.. lJ)

(Y'}

N

0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2-298 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SPRAY LINE DER NODES 17, 18,19 E. 20 BEAVER VALLEY POWER STATION'-UNIT 2 FINAL SAFETY ANALYSIS REPORT

('f")

N 0

NODES 21,22 e. 23 a: 0>

U) a_ ([)

w a::: r-U)

U) w w

a:::

a_ U)

'¢

('f")

N o.o 0. 1 0.2 0.3 0.4 O.G 0.6 Q.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2-29C PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SPRAY LINE DER NODES 21,22 E. 23 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

N 0

NODES 1,2,3.4 m

a: CD (f) a... r--

w 0::: CD

~

(f)

(f) lJ) w 0:::

a... ~

(T)

N 0

0.00 0.05 o.to o.tE 0.20 0.25 0.30 0.35 0.40 0.45 0.50 TIME (SEC)

'FIGURE 6.2-30 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER AT THE PRESSURIZER NOZZLE NODES 1,2,3 E. 4 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

N 0

m NODES 5,6,7E.8 a: (X)

>--1 (f)

Q_ c-w

~ lD (f)

(f) LJ) w

~

Q_

~

en N

9J.oo o.o5 0.10 o.15 0.20 0.25 o.3o 0.35 0.40 0.45 0.50 TIME (SEC)

FIGURE 6.2-31 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER AT THE PRESSURIZER NOZZLE NODES 5,6, 7 £. 8 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0 (I')

v NODES 9,10,11&12

./

- r-0 a: N -c J~

( I) 0..

w c:t: -

ll)

~f (I)

(I) w c:t: 0 0.. ~

ll)

~.oo o.o5 o.to o.t5 o.2o 0.25 o.3o o.35 o.4o o.45 o.5o TIME (SEC)

FIGURE 6.2-32 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER AT THE PRESSURIZER NOZZLE NODES 9,10,11 f.t 12 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0 en I

NODE 13 rJ I ------

a: 0

...... N

~

U) 0...

j/

t w

rr ll) 0::::

=>

U) NODE 15 U) w 0:::: 0 0...

~.oo o.o5 o.to o.t5 0.20 0.25 o.3o o.35 o.4o o.45 o.5o TIME (SEC)

FIGURE 6.2-33 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER AT THE PRESSURIZER NOZZLE NODES 13 E. 15 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0 (r)

U')

N

,..... ~DE14 a: 0 , --....._

/

...... N

(/) NODE17

-~

a_

w v

U')

L~

0:: NOD~I6

J

(/)

(/)

w 0:: 0 J

a_

NODE19

~.oo o.o5 o.to o.t5 o.2o 0.25 o.3o 0.35 o.4o o.45 o.5o TIME (SEC}

FIGURE 6.2-34 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER AT THE PRESSURIZER NOZZLE NODES 14,16,17 E. 19 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0 l/)

N 0

0 /

I'--

N 18

~

a:

......... 0 l/)

I (f) a....

w a:::

J (f) 0 (f) 0 w

a:::

a....

0 l/)

0 o.oo o.o5 o.to o.t5 o.2o 0.25 o.3o o.35 o.4o 0.45 o.5o TIME (SEC}

FIGURE 6.2-35 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER AT THE PRESSURIZER NOZZLE NODE 18 BEAVER VALLEY POWER STAT ION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0 Vz....--

(I')

NODE 20

, ( ~-< ~

v ---

II)

N

-~ -

0 a: N

~ (1- ~

( /)

0... NJOES 21,22 1 E. 23 w

- tU 1

II) 0:::

U)

~

U) w 0::: 0 0...

II)

~.oo o.o5 o.to o.t5 o.2o 0.25 o.3o 0.35 o.4o o.45 o.5o TIME (SEC)

FIGURE 6.2-36 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER AT THE PRESSURIZER NOZZLE NODES 20,21, 22 E. 23 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

N NODES 1,2,3&4 0

...... ~

m a: (X)

U)

Q_ ('

w a:::: w U)

U) l/)

w a::::

Q_

'<t" (T')

N 0 *1 0.2 0.3 0.4 0.5 0.6 0.7 O.B 0-9 1 *0 TIME (SEC)

FIGURE 6.2-37 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER IN THE PRESSURIZER RELIEF TANK CUBICLE NODES 1,2,3 E. 4 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0 (J)

---

- NODES 5,6,7:.8 I

a: 00 (f) a_ (""--

w a::: w (f)

(f) IJ) w a:::

a_ -.;t (Y')

N

't .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 *0 TIME {SEC)

FIGURE 6.2-38 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER IN THE PRESSURIZER RELIEF TANK CUBICLE NODES 5,6, 7 t. 8 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0

('I')

an N

0 a: N

( /)

Q_

,.--NODES 9,10,116.12 w an

..... r

/

a:::

(/)

(/)

w a::: 0 Q_

l/)

0. 1 0.2 0.3 0.4 o.s 0.6 0.7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6.2-39 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER IN THE PRESSURIZER RELIEF TANK CUBICLE NODES 9, 10, 11 E. 12 BEAVER VALLEY POWER STAT ION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

0

('I'}

lJ)

C\1 0

a: C\1

( /)

0...

NODEJ3 w ,....

lJ) v

/

0::

(/)

(/)

w 0::

0... ....

0

,.../

lJ) 9J .o 0. 1 0-2 0.3 0-4 o.s 0.6 0.7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6.2-40 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER IN THE PRESSURIZER RELIEF TANK CUBICLE NODE 13 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

/ r\ ~

I ll)

N ~* 1--.

NODE IS 0

a:

I N

NODE 14

( f.)

~

0...

~

'!/

ll)

I.JJ .....

~

a::

(f.)

(f.)

I""\ NODE 16 w

a:: 0

..... '.L/

0...

ll)

~.o 0. 1 0.3 0.4 0.5 o.s 0.7 o.a o.s 1 *0 TIME (SEC)

FIGURE 6.2-41 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER IN THE PRESSURIZER RELIEF TANK CUBICLE NODES 14, 15 E. 16 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0 (Y) I NODE 17 Ul if ~ r---....

I N

I 0

a: N NODE 18

(

( f) a...

II w Ul ~

0::: /

J NODEI9

(

(f)

(f) w 0::: 0 a...

Ul

0. 1 0-2 0-3 0.4 o.s 0.6 0-7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6.2-42 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER IN THE PRESSURIZER RELIEF TANK CUBICLE NODES 17, 18 E. 19 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0

(\")

1./)

N a:

( /)

a_

0 N

NODES 21, 2 & 23 w

/

~

..... r 1./)

0::::

(/)

(/)

w 0:::: 0 a_

1./)

9:Lo 0. 1 0.2 0.3 0.4 o.s o.s 0.7 o.e 0.9 1 *0 TIME (SEC)

FIGURE 6.2-43 PRESSURIZER CUBICLE AVERAGE PRESSURE VERSUS TIME SURGE LINE DER IN THE PRESSURIZER RELIEF TANK CUBICLE NODES 21,22 . 23 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

NOTES FIGURE 6.2-44

1. NODE ('ffi IS THE CONTAINMENT STEAM GENERATOR SUBCOMPARTMENT

2. SYMBOLS NODALIZATION DIAGRAM a) D CONTROL VOLUME (OR NODE) BEAVER VALLEY POWER STATION-UNIT 2 b) 0 JUNCTION NUMBER FINAL SAFETY ANALYSIS REPORT

26"DUCT Q; ...

NODE 29/

I

~-

I'

.~

• .fl NODE 31 I NODE 3 I

I ..

NODE 6 31" I.D.

RC PIPE I

NODE 5 I NODE 4

\

.. ."p*

• _J 1

FIGURE 6.2-45 STEAM GENERATOR SUBCOMPARTMENT NODALIZATION PLAN VIEW ELEVATION 718'*6"T0727'-2

BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

. (?

• 0

• p.

q,

• v*

NODE 10 d38"DUCT FIGURE 6.2-46 STEAM GENERATOR SUBCOMPARTMENT NODALIZATION 11 PLAN VIEW EL. 727._0.. TO 740'-3 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

1 I

NODE 13 (19) PLATF.

EL.747'-6" .

. v v...-'

.. v NODE 18 (24)

NODE 17 (23)

NODE 16 (22)

_j 1

NOTE:

NODES 19 THROUGH 24 ARE LOCATED Dl RECTLY ABOVE NODES 13 THROUGH 18, RESPECTIVELY FIGURE 6.2-47 STEAM GENERATOR SUBCOMPARTMENT NODALIZATION 1 11 11 PLAN VIEW EL. 740 -3 TO 767!.10 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

Fl G U R E 6. 2- 48 STEAM GENERATOR SUBCOMPARTMENT NODALIZATION PLAN VIEW EL. 767'-10" BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

NODE 30 (CONTAII'I.t.1ENT)

.. EL. 735

• s" J~ NODE II

.* EL. 728'6" ~

SG28 .:NODE II' 5

EL.718'6" :*: ~~~~~~~~~

EL. 777' 10"

• ~**.I .I>* * : .: ~"' *. ** :- : : ~ . ci. *: NODE 2

~

• . t>*

.~ NODE 27 \ - I" ~~ 26 SECTION 2-2 EL.767'10" 'r---i

~. *. ~: ... .; . *.

0 * ., 01 I o\

• 0 "'

!'7' NODE 23 EL. 747'6" II>.'

EL. 740' 3"

.. EL. 732' 3~"

EL. 727' o" ".*..

• fl EL.718'6" EL.714'2" SECTION 1-1 FIGURE 6.2-49 STEAM GENERATOR SUBCOMPARTMENT NODALIZATION ELEVATION VIEWS SECTIONS 1-1 AND 2-2 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

1/)

N 0

N (NODE 5 a:

r

(/) 1/)

a... /"'"

~ NODE 1 w

0:::

J

(/)

(/)

0 w ~

0:::

a...

1/)

i

0. 1 0.2 0.3 0.4 0.5

~

0.6 0.7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6.2-50 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES I E. 5 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N 0

N r_NODE 6 a:

~

CJ) lJ)

~ r ..........

~~

Q_

\.NODE 2 w

a:::

CJ) 0 CJ) ......

w ~

a:::

Q_

lJ) 9J .o 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2-51 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 2 E. 6 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

1./)

N 0

N a: NODE 3 t...

vf U) 1./)

CL

" ~\1' w NODE 4 0:::

U) v U) ..... 1-J 0

w 0:::

CL 1./)

9J.o

• o.1 0.2 0.3 0.4 0.5 0.6 0.7 o.a o.s 1 .0 TIME (SEC)

FIGURE 6.2 -52 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 3 E. 4 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

LJ)

N I

0 N

.{NODE II a:

( /)

a_

LJ)

...... -r ~ -NODE 7 w

0:::

J

(/)

0

(/) ......

w 0:::

a_

\

LJ)

0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 .0 TIME (SEC)

FIGURE 6.2-53 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 7 E. 11 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

Ul N

0 N

a:

J~DE 12 rfY

( f) Ul J~ /'\.../ ~

Q_

" \

--NODE 8 w

a:::

=>

(f)

(f) 0 w

a:::

Q_

Ul 9J .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 o.a 0.9 1.0 TIME (SEC)

FIGURE 6.2-54 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 8 E. 12 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

L/)

C\1 0

C\1 a:

~ It NODE 9 fV

(/)

L/)

a... ~ /

w

~NODE 10 0::::

(/)

(/) .....

0 J

w 0::::

a...

L/)

9J .o 0. 1 0.2 0.3 0.4 o.s o.s 0.7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6.2-55 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 9 E. 10 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

1.1)

N 0

N a:

(NODE 17 (J) 1.1) fl t~~E Q_

w 13 0::::

(J) 0 (J) ..... J w

0::::

Q_

1.1)

0. 1 0.2 0.3 0.4 0.5 0.6 0.7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6.2-56 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 13 E. 17 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

ll)

N D

N a: (NODE 14 II' (f) ll)

Q_ .- J --

w a:::

\'N: 18

J (f)

D I (f) w ~ f.J a:::

Q_

ll) 9J .o 0. 1 0.2 0.3 0.4 0.5 o.s 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6. 2-57 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 14 E. 18 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N I

0 N

a: (NODE 15 lJ) tl CJ) a_ ......

w 0::::

~ODE 16 CJ) 0 CJ) ......

w ~

0::::

a_

lJ) 9J .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 o.a o.s 1.0 TIME (SEC)

FIGURE 6.2-58 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SO. IN. BREAK NODES 15 E. 16 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N 0

N CI:

........ ,...NODE 19 lJ) r U) a_

w 0:::

~ODE 23 U)

U) .....

0 J

w 0:::

a_

lJ)

~.o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6.2-59 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 19 E. 23 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

~--------------------------------------------------------------------------*--------,--------

U)

N 0

N a:

...... /NODE 24 f

(/) U) v v-a_

w Q:::

~NODE 20

(/)

(/)

0 J

w Q:::

a_

U) 9J .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 o.a 0.9 1 .0 TIME (SEC)

FIGURE 6.2-60 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 20 E. 24 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REFORT

~------------------------------------------------*-----------------------------------------~

l/)

N 0

N a:

UJ a.... .....

l/)

r-NODE 21 w

a::::

> -NODE 22 UJ 0 UJ ..... ~

w a::::

a....

l/)

0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 o.s 1.0 TIME (SEC)

FIGURE 6. 2-61 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 21 . 22 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

L/)

N 0

N a:

(/) L/)

0....

w a:::: {_NODE 25

J

(/)

0

(/)

w  !\_NODE 27 a::::

0....

L/)

't .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6.2-62 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 25 . 2 7 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N I \

0 N

a:

( J)

Q_

lJ) w a::::

> ..-NODE 26 (J) 0 /

(J) ......

w a:::: r--.-NODE 28 Q_

lJ)

0. 1 0.2 0.3 0.4 0.5 o.s 0.7 o.a 0.9 1.0 TIME (SEC)

FIGURE 6.2-63 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 26 E. 28 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

L/)

N 0

N a:

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(L

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r w

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w a::: (NODE 30 (L

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FIGURE 6.2-64 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 29 E. 30 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

0 CD

(

0 l/)

,-NODE 32

~

0 a: ~

(J)

~I (L

w 0 (Y')

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J (J)

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a::: 0 N

// ....

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'b .o 0. 1 0.2 0.3 0-4 0.5 0.6 0.7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6. 2 -65 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 320 SQ. IN. BREAK NODES 31,32 E. 33 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

Ll)

I N

a N

a:

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(NODE 1 w

a::::

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a::::

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Ll) 9J .c 0.1 0.2 0.3 0.4 0.5 0.6 0.7 o.e 0.9 1.0 TIME (SEC)

FIGURE 6.2-66 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 1 e 5 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

U)

N 0

N a:

CJ) U)

(L i(NODE 2 r---

w -...._..-

a::: (NODE 6 CJ)

CJ) 0

'J w

a:::

(L U) 9J.o

• o.1 0-2 0.3 0.4 o.s 0.6 0.7 o.a 0.9 1 .0 TIME (SEC)

FIGURE 6.2-67 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 2 E. 6 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

L/)

N I

0 N

a:

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1\

(f) L/)

_£NODE 3

~r CL w

0:::

J (f)

~,~ /t?o:E 4

(f) .....

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CL

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0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 .0 TIME (SEC)

FIGURE 6.2-68 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 3 E. 4 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N 0

N a:

CJ)

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(NODE 7 w

a:::

CJ)

CJ) w -

0

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lJ) 9J .o 0. 1 0.2 0.3 0.4 0.5 o.s 0.7 0.8 o.s 1 *0 TIME (SEC)

FIGURE 6.2-69 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 7 E. 11 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

IJ)

N 0

N rJ--\.

a:

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(J) IJ) a_ ..... 1\

r...

w a::: ~ (NODE 12 (J)

(J) .....

0 J

w a:::

a_

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FIGURE 6.2-70 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 8 E. 12 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ) 0

"" ~

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Q_

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J CJ) 0 CJ) ...... J w

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FIGURE 6.2-71 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 9 E. 10 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

Lf)

N 0

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a::::

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CJ) 0

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FIGURE 6.2-72 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 13 E. 17 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N I

0 N

a:

( f) a...

lJ)

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0:::

=>

(f)

(NODE 18 0

(f) ...... lj w

0:::

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lJ) 9J .o 0.1 0.2 0.3 0.4 o.s 0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2-73 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 14 E. 18 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N 0

N a:

( f) a_ .....

lJ) ..

r\

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\

n w

Q:::

tf kV"'-

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(f) .....

0 IJ' I

w Q:::

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lJ) 9J .o 0.1 0.2 0.3 0.4 o.s 0.6 0.7 o.s 0.9 1.0 TIME (SEC)

FIGURE 6.2-74 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 15 E. 16 BEAVER VALLEY ~OWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

C\1 0

C\1 a:

( /)

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a::::

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w a::::

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FIGURE 6.2-75 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 19 £. 23 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

IJ)

N 0

N a:

( f)

(L 1./)

.-NODE 20 w

£ 0:::

l rf/'-*--...__NODE 24 (f) 0 U)

..D w

0:::

(L IJ)

0. 1 0.2 0.3 0.4 o.s Q.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2-76 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 1BO SQ. IN. BREAK NODES 20 . 24 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N 0

N a:

(J) lJ) ur \:~DE a_

~NODE 22 w

~

(J) 21 0

(J) ...... ~

w

~

a_

lJ) 9l.o 0 *1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2-77 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 21 £. 22 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

(\J 0

(\J f a:

U) lJ) a.... ......

w a::

> {NODE 25 U) 0 U) ......

w a:: lNODE 27 a....

lJ) 9J .o 0. 1 0.2 0.3 0.4 0.5 Q.6 0.7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6.2-78 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 25 £. 27 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

l/)

N a

N a:

(f) l/)

a_ ......

w a::::

J /NODE 26 (f) a (f) ......

w ~

a::::

a_

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~.o I

0. 1 0.2 0.3 0.4 0.5 Q.6 0.7 o.e o.g 1.0 TIME (SEC)

FIGURE 6. 2-79 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 26 £. 28 BEAVER VALLEY POWER STATION- UNIT 2 FINAL SAFETY ANALYSIS REPORT

1./)

N 0

N a:

rf-(J) 1./)

a_ ......

31

~DE w

0:::

J (J) 0 (NODE 29 (J) ...... .I w

0:::

a_

1./)

9J .o 0. 1 0.2 0.3 0.4 o.s 0.6 0.7 o.a o.s 1.0 TIME (SEC)

FIGURE 6. 2-80 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK*

NODES 29 E. 31 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

a

(.£)

a LJ) a: a

'¢

(/)

a_

w a

('t) a::::

l

(/)

(/)

w a:::: a a_ N

,...NODE 33 a ~ 32 NODE 30 0.1 o.z 0.3. 0.4 0.5 0.6 0.7 o.a o.s 1 .0 TIME (SEC)

FIGURE 6.2-81 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 180 SQ. IN. BREAK NODES 30,32 E. 33 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

Ul N

r--- ... NODE 1

~ f"'/"" -

[t'f'/

0 N

NODE 5 a:

(f) Ul a... ......

w 0::::

1 (f) 0 (f) ......

w 0::::

a...

Ul 9J .o 0. 1 0.2 0.3 0.4 o.s 0.6 0.7 o.a 0.9 1 *0 TIME (SEC)

FIGURE 6.2-82 BV2-33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME ,

707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 1 E. 5 9EAVE~ VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

UJ N  :

I I

--NODE 2

./

rt V' 0  !---

N NODE 6 a:

(./) UJ ---- --------- --- -- -- -

a_ ...... *~-

w 0:::

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0

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w 0:::

a_

UJ 9J .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2-83 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 2 E. 6 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

l/)

N

. / NODE 3 rrt a -

~ ........ I-

--~--

N NODE 4 a:

(J) l/)

Q_

w 0:::

l (J)

(J) a w ~ I 0::: I I

I Q_

I!

l/)

0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2 -84 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 3 £. 4 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N AN( rV' NODE 7 VJ_

0 N

'-/

NODE 11 a:

(J) lJ)

(L w

0::::

J (J)

(J) .....

0 w

0::::

(L lJ) 9J .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

F1GURE 6.2-85 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 7 . 11 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

1./)

N

1. " I-NODE 8

[?

I 0 ..-./

N

~V-NODE12 a:

(J) 1./)

(L ...... I w

0::

J (J) 0 (J) ......

w 0::

(L 1./)

0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6.2-86 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 8 £. 12 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

lJ)

N NODE 9 v

0 L

/rj N

~-

NODE 10 a:

l

(/) lJ) a_ __.

w a::::

(/)

(/) .....

0 w

a::::

a_

lJ) 9J .o 0. 1 0.2 0.3 0.4 o.s 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6.2-87 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 9 E. 10 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

Ul N

0

/NODE 13

()' rx;;O~E N I 17 a:

Ul U) a...

w a:::

U)

U) .....

0 w

a:::

a...

Ul

't .o 0.1 0.2 0.3 0.4 0.5 .0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE 6. 2-88 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 13 £. 17 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

LJ)

N r;

jNOOE 14 0

N ~

R ; E 18 a:

J rv CJ) LJ)

Q_ ......

w 0:::

J CJ) 0 CJ) ......

w 0:::

Q_

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0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6. 2-89 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE.PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 14 E. 18 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

L/)

N

./NODE 15 0

N

"'~\) ~~ {NODE 16 a:

(f) a_

L/)

v

~

w 0:::

(f)

(f) 0

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w 0:::

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9J .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 TIME (SEC)

FIGURE G. 2-90 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE. PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 15 E. 16 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

l/)

C\1

( NODE 19 I

0

,-- I f

C\1 t_NODE 23 a:

(f) l/)

(L ...... I w

0:::

J (f) j 0

(f) ...... ~

w 0:::

(L l/)

9J .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6.2-91 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE.PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 19 E. 23 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

Lf)

N 0

NODE 24 r!

N I p-NODE 20 a:

(f) Lf) a_ __.

w a::::

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(f) 0 J

w ~

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0. 1 0.2 0.4 0.5 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6.2-92 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE" PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 20 E. 2 4 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

IJ)

N I

0 fNODE 22 t

N a:

~E 21

- r if) IJ) a...

w a::::

J if) if) 0 J

.JI w '

a::::

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~-0 0. 1 0.2 0.3 0.4 o.s 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6.2-93 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 21 . 22 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

l/)

C\1 0

C\1 a:

(f) l/)

(L w (NODE 25 a::::

> \NODE 27 (f) 0 ~ 1---

(f) w ~

a::::

(L l/)

0. 1 0.2 0.3 0.4 0.5 0.6 0.7 O.B 0.9 1 *0 TIME (SEC)

FIGURE 6. 2-94 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE .PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 25 £. 27 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

~-------------------------------------------------------------------------------------------~

l/)

N 0

N a:

(/)

a_

l/)

w (NODE 26 a:::

J \NODE 28

(/)

0 ~-

(/) ...... f..-'

w a:::

a_

l/)

'11 .o 0. 1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6.2-95 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 26 . 28 BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

l/)

N 31 JNO~

0 N

~{\

v

/

/

{.NODE 29 a:

I

(/) l/)

a... ......

w a::

J

(/)

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w a::

a...

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~.o 0 *1 0.2 0.3 0.4 o.s 0.6 0.7 o.B 0.9 1.0 TIME (SEC)

FIGURE 6.2-96 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 29 E. 31 BEAVER VALLEY POWEP STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

~----------------------------------------------------------------------------------------~

0 w

0 lJ) 0 a: "<t (f)

(NODE 32

~

a.....

_c-.

w 0 a:: ('I")

J (f)

(f) (NODE 33 w

a:: 0 1\~ ~

a..... N (NODE 30 0

0.1 0.2 0.3 0-4 o.s 0.6 0.7 0.8 0.9 1 *0 TIME (SEC)

FIGURE 6.2-97 BV-2 33 NODE STEAM GENERATOR CUBICLE ABSOLUTE PRESSURE VERSUS TIME 707 SQ. IN. STM. GEN. INLET ELBOW SPLIT NODES 30,32 e. 33 BEAVER VALLEY POWER STATION- UNIT 2 FINAL SAFETY ANALYSIS REPORT

FLOW +-VOLUME C - ~VOLUMED RESTRICTOR "'"""'\.. ,_-----~~)----~

(TYPICAL) '

MSIV VSLI =VOLUMES A+ 8 +C +D A WITHOUT MSIV FAILURE VSL 2 : { A+ D WITH MSIV FAILURE VOLUMED~

-VOLUME a-MSIV

+-VOLUME A -

BREAK-.-----------~~r---------~

&-------------------t:X}-- AUXILIARY FEEDWATER PIPE MAIN FEEDWATER PIPE WHERE:

MSIV : MAIN STEAM ISOLATION VALVE SGMASS : INITIAL STEAM GENERATOR INVENTORY VSLI : TOTAL STEAM PIPING VOLUME VSL2 : VOLUME BETWEEN BREAK AND NEAREST FUNCTIONING MSIV FWIV : FEEDWATER ISOLATION VALVE FWCV : FEEDWATER CONTROL VALVE FIGURE 6.2-118 SCHEMATIC DIAGRAM OF SECONDARY SYSTEM BEAVER VALLEY POWER STATION- UNIT 2 FINAL SAFETY ANALYSIS REPORT

Rev. 16 CONTAINMENT PRES URE USED FOR BEAVER VALLEY UNIT 2 40 35

<::2 30

• (f) o_

Q)

\._

(f)

(f)

~ 25 o_

20 15 0 100 200 300 400 Time After Break (s)

Figure 6.2-119 Lower Bound Containment Pressure Beaver Valley Power Station Unit No. 2 Updated Final Safety Analysis Report

14 TK21 REFUEUNG 30 WATER STORAGE TANK

~

w

E II 2 101A z

~

8 1 w 0

~

P21B QUENCH SPRAY PUMP CHEMICAL ADDmON SYSTEM IS DISABLED ALL VALVE AND EQUIPMENT IDENTIFICATION NUMBERS ON THIS FIGURE ARE PRECEDED BY THE SYSTEM DESIGNATOR "2Qss*

UNLESS OTHERWISE INDICATED FIGURE 6.2-121 QUENCH SPRAY SYSTEM BEAVER VALLEY POWER STATION - UNIT 2 UPDATED FINAL SAFETY ANALYSIS REPORT

___ )~:~~R:?~t~J~=?! ___________ :~u~~Y~§~~~~~0~~!-~~~~--------------------------------------------------------------------------------------------------------------------------------

REV. 12 RECIRCULATION SPRAY HEADERS TO SAFETY INJECTION RECIRCULATION SPRAY COOLERS I-z w

z

<(

I-z 0

u RECIRCULATION w 0

SPRAY PIT PIT PUMPS Vl z

ALL VALVE AND EQUIPMENT IDENTIFICATION NUMBERS ON THIS FIGURE ARE PRECEDED BY THE SYSTEM DESIGNATOR "2RSSN UNLESS OTHERWISE INDICATED.

FIGURE 6.2-122 INSIDE CONTAINMENT RECIRCULATION SPRAY SYSTEM

REFERENCE:

STATION DRAWING OM 13-1 BEAVER VALLEY POWER STATION UNIT NO. 2 UPDATED FINAL SAFETY ANALYSIS REPORT

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ONTAINMENT QUENCH AND OEi"ALi" DE.T ".t.F" ~ECIRCULATING SPRAY PIPING RP*F9D AL,.E.ANI.lf. ~1: OETS. .. AI..C *"t:*

41S.V A OR 8 WOIUZONTA.L NOZnE ASSY AOA&IS*Pl,.CHED NOZZLf PART PLAN 8-RECIRC !>PRAY HEII.DERS 0.,.1 BEAVER VALLEY POWER STATION-UNIT 2

{5UNOU29)

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2. SPATIAL DROPLET SIZE DISTRIBUTION OF SPRACO 1713A NOZZLE APPLYING SURFACE AREA CORRECTION AND SPRAYING WATER AT FIGURE 6.2-125 40 PSIG UNDER LABORATORY CONDITIONS SPATIAL DROPLET SIZE DISTRIBUTION BEAVER VALLEY POWER STATION- UNIT 2 FINAL SAFETY ANALYSIS REPORT

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FIGURE 6.2-126 QUENCH SPRAY COVERAGE ON 1 11 OPERATING FLOOR, EL. 767 -10 BEAVER VALLEY POWER STATION- UNIT2 FINAL SAFETY ANALYSIS REPORT

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I EL.846'-4" BEND LINE EL.767~1d'~~~~~~~~~~~~~~~~~~~~~~OPERATING 10 20 30 40 50 60 63 FLOOR RADIUS FROM { OF REACTOR -FEET FIGURE 6.2-127 QUENCH SPRAY COVERAGE IN CONTAINMENT DOME BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

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DEPRESSURIZATION CASE PSOER (MIN ESF l BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT OWN. BY* A. NULPH 0/CHK.* ?*-? 2SAR130 DATE* 03-14-90

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STEAM GENERATOR SHIELD WALL ABOVE OPERATING FLOOR NOT FIGURE 6. 2-138 SHOWN FOR CLARITY CONTAINMENT INTERNAL STRUCTURE SEAVER VALLEY POWER STATION- UNIT 2 FINAL SAFETY ANALYSIS REPORT

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BVPS-2 UFSAR Rev. 26 6.3 EMERGENCY CORE COOLING SYSTEM 6.3.1 Design Bases Summary Description The primary function of the emergency core cooling system (ECCS) following a loss-of-coolant accident (LOCA) is to remove the stored and fission product decay heat from the reactor core such that fuel rod damage, to the extent that it would impair effective cooling of the core, is prevented.

The ECCS consists of the high head safety injection (HHSI)/charging pumps, the refueling water storage tank (RWST), low head safety injection (LHSI) pumps, recirculation spray pumps, and the safety injection (SI) accumulators with the associated valves, instrumentation, and piping.

Plants listed in Section 1.3 have essentially similar systems as Beaver Valley Power Station - Unit 2. The ECCS is designed to cool the reactor core as well as to provide additional shutdown capability following initiation of the following accident conditions:

1. A LOCA, including a pipe break or a spurious relief or safety valve opening in the reactor coolant system (RCS), which would result in a discharge larger than that which could be made up by the normal make-up system.

2. A rupture of a control rod drive mechanism causing a rod cluster control assembly (RCCA) ejection accident.

3. A steam or feedwater system break accident, including a pipe break or a spurious relief or safety valve opening in the secondary steam system which would result in an uncontrolled steam release or a loss of feedwater.

4. A steam generator tube rupture.

The acceptance criteria for the consequences of each of these accidents are described in Chapter 15 in the respective accident analyses sections.

The bases used in design and for selection of ECCS functional requirements are derived from 10 CFR 50 Appendix K limits for fuel cladding temperature, etc following any of the preceding accidents as delineated in 10 CFR 50.46. The subsystem functional parameters are selected so that, when integrated, the 10 CFR 50 Appendix K requirements are met over the range of anticipated accidents and single failure assumptions.

The reliability of the ECCS has been considered in selection of the functional requirements, selection of the particular components, and location of components and connected piping. Redundant components are provided where the loss of one component would impair reliability. Valves are provided in series where isolation is desired and in parallel when flow paths are to be established for ECCS performance. Redundant sources of the SI actuation signal are available so that the proper and timely operation of the ECCS will be assured. Sufficient instrumentation is 6.3-1

BVPS-2 UFSAR Rev. 26 available so that a failure of an instrument will not impair readiness of the system. The active components of the ECCS are powered from separate buses that are normally energized from the unit station service transformers. In addition, redundant sources of onsite power are available through the use of the emergency diesel generators to assure adequate power for all ECCS requirements. Each emergency diesel generator is capable of driving all pumps, valves, and necessary instruments associated with one train of the ECCS.

All motor-operated valves (MOVs) required to be moved from one position to another following the injection phase are located to prevent vulnerability to flooding. Spurious repositioning of valves due to the actuation of its positioning device coincident with a LOCA has been analyzed and is not considered credible. However, those valves whose spurious repositioning could result in a loss of the ECCS function have their power removed.

The environmental qualification of active ECCS equipment is discussed in Section 3.11.

Protection of the ECCS from missiles is discussed in Section 3.5.

Protection of the ECCS against dynamic effects associated with ruptures of piping is described in Section 3.6.

The elevated temperature of the sump solution during recirculation is well within the design temperature of all ECCS components. In addition, consideration has been given to the potential for corrosion of various types of metals exposed to the fluid conditions prevalent immediately after the accident or during long term recirculation operations (Section 6.1.1).

6.3.2 System Design The ECCS components are designed such that adequate core cooling in the event of a design basis LOCA is provided: 1) in the injection phase by a minimum of two accumulators, one HHSI/charging pump and one LHSI pump, and

2) in the recirculation phase by a minimum of one charging pump and one recirculation spray pump with their associated valves and piping. The redundant onsite emergency diesel generators assure adequate emergency power to all required electrically-operated components in the event that a loss-of-offsite power (LOOP) occurs simultaneously with a LOCA, even assuming a single failure in the emergency power system.

6.3.2.1 Schematic Piping and Instrumentation Diagrams The ECCS is shown on Figures 6.3-1 and 6.3-2.

The components of the ECCS are interlocked as listed:

1. The SI signal is interlocked with the following components and initiates the indicated actions:

a. The HHSI/charging pumps start,

b. The LHSI pumps, start, 6.3-2

BVPS-2 UFSAR Rev. 26

c. The HHSI/charging pump discharge to cold legs isolation valves 2SIS*MOV867A,B and 2SIS*MOV867C,D open,

d. The RWST to HHSI/charging pump valves 2CHS*LCV115B,D open,

e. The normal charging line isolation valves close,

f. The volume control tank (VCT) to HHSI/charging pump suction isolation valves 2CHS*LCV115C,E close, and

g. The accumulator isolation valves open, if closed (these valves are normally open and have their power removed).

2. The following valves close on a containment isolation Phase A (CIA) signal:

a. Isolation valves in the nitrogen (N2) supply line to the accumulators,

b. Isolation valves in the check valve test lines, and

c. Isolation valves in the sampling lines from the accumulators, pressurizer, and hot and cold legs.

3. The recirculation spray pump suction isolation valves from the containment sump are interlocked to open upon receipt of a containment isolation Phase B (CIB) signal.

4. On an extreme low RWST level concurrent with an SI signal the following actions occur:

a. The LHSI pump discharge crossover isolation valves 2SIS*MOV8887A,B close,

b. The recirculation spray pump discharge valves 2SIS*MOV8811A,B to the LHSI discharge lines into the RCS open,

c. The recirculation spray header isolation valves 2RSS*MOV156C,D are closed,

d. The LHSI pumps stop on the limit switch signals open from valves 2SIS*MOV8811A,B,

e. The HHSI/charging pump suction isolation valves from the LHSI header 2SIS*MOV863A,B are opened on limit switch signal open from valves 2SIS*MOV8811A,B,

f. The HHSI/charging pump suction isolation valves from the RWST 2CHS*LCV115B,D are closed on limit switch signa1 open from valves 2SIS*MOV863A,B, and 6.3-3

BVPS-2 UFSAR Rev. 26

g. The LHSI pump suction isolation valves from the RWST 2SIS*MOV8809A,B are closed on stop signal from LHSI pumps.

The LHSI pump miniflow valves open on low flow and close on high flow coincident with the LHSI pump operation. The valves close when the LHSI pumps are stopped.

6.3.2.2 Equipment and Component Descriptions The component design and operating conditions are specified to the most severe conditions to which each respective component is exposed during either normal plant operation, or during operation of the ECCS. For each component, these conditions are considered in relation to the code to which it is designed. By designing the components in accordance with applicable codes, and with due consideration for the design and operating conditions, the fundamental assurance of structural integrity of the ECCS components is maintained. Components of the ECCS are designed to withstand the appropriate seismic loadings in accordance with their safety class as given in Table 3.2-1. Specific equipment parameters are shown in Table 6.3-1. Sections 3.7N and 3.9N describe the seismic and mechanical component designs.

The major mechanical components of the ECCS are:

Accumulators The accumulators are pressure vessels partially filled with borated water and pressurized with nitrogen gas. During normal operation each accumulator is isolated from the RCS by two check valves in series.

Should the RCS pressure fall below the accumulator pressure, the check valves open and borated water is forced into the RCS. One accumulator is attached to each of the cold legs of the RCS.

Mechanical operation of the swing-disc check valves is the only action required to open the injection path from the accumulators to the core via the cold leg. Connections are provided for remotely adjusting the level and boron concentration of the borated water in each accumulator during normal plant operation as required. Accumulator water level may be adjusted either by draining to the Primary Drains Transfer Tank or by pumping borated water from the RWST to the accumulator. Samples of the solution in the accumulators are taken periodically to check boron concentration. Reduction of accumulator water level is detected by level indicators and low level alarms to allow the operator to take prompt action to maintain plant operation within the requirements of the Technical Specifications covering accumulator operability.

Accumulator pressure is provided by a supply of nitrogen gas and can be adjusted as required during normal plant operation; however, the accumulators are normally isolated from this nitrogen supply. Gas relief valves on the accumulators protect them from pressures in excess of design pressure.

6.3-4

BVPS-2 UFSAR Rev. 26 The accumulators are located within the containment but outside of the secondary shield wall which protects them from missiles. Since the accumulators are located within the containment, a release of the nitrogen gas in the accumulators would cause an increase in normal containment pressure. Containment pressure increase following release of the gas from all accumulators has been calculated and is well below the containment pressure set point for ECCS actuation. Depressurization of the accumulators due to a loss of the nitrogen gas would be detected by the accumulator pressure indicators and alarms. Thus, the operator could take action promptly to maintain plant operation within the requirements of the Technical Specification covering accumulator operability.

Isolation and/or depressurization of the accumulators to permit cold shutdown operations is discussed in Section 5.4.7.

Low Head Safety Injection Pumps The LHSI pumps start automatically on receipt of an SI signal and deliver water to the RCS from the RWST during the injection phase. The pumps are horizontal centrifugal type, with a self-contained mechanical seal.

A minimum flow bypass line is provided for the pumps to recirculate and return the pump discharge fluid to the RWST should these pumps be started with their normal flow paths blocked. Once flow greater than 1,000 gpm is established to the RCS, the bypass line is automatically closed. This line prevents dead-heading of the pumps and permits pump testing during normal operation.

During the switchover from injection to recirculation, these pumps are stopped and the LHSI function is provided by two of the four recirculation spray pumps.

The low head pump performance curve is given on Figure 6.3-4.

High Head Safety Injection/Charging Pumps The HHSI/charging pumps are started automatically on receipt of a SI signal and are automatically aligned to take suction from the RWST during injection. During recirculation, suction is provided from the containment sump via the recirculation spray pumps and portions of the LHSI piping.

The charging pumps deliver flow to the RCS at the prevailing RCS pressure.

Each centrifugal charging pump is a multistage diffuser design, barrel-type casing with vertical suction and discharge nozzles.

A minimum flow recirculation path is provided to protect the two operable charging pumps by preventing the pumps from reaching a deadhead condition.

That is, if the reactor coolant pressure is too high to allow any operating charging pump to deliver flow through the injection lines, sufficient flow to prevent pump damage will be recirculated through the seal water heat exchanger and back to the pump suction header.

6.3-5

BVPS-2 UFSAR Rev. 26 During normal plant operation, at least one charging pump is continuously in service. The other charging pumps may be tested during power operation via the minimum flow bypass lines.

The high head pump performance curve is given on Figure 6.3-5.

Available and required net positive suction head (NPSH) for ECCS pumps are shown in Table 6.3-1. The safety intent of Regulatory Guide 1.1 is met by the design of the ECCS such that adequate NPSH is provided to system pumps.

Positive Displacement Hydrostatic Test Pump This pump serves two functions, neither of which is safety-related.

Permanent connections are provided to the accumulators to allow addition of borated water from the RWST. Temporary connections permit the use of this pump in hydro-testing the plant piping systems.

Recirculation Spray Pumps The recirculation spray pumps provide containment spray during the injection phase as well as maintaining long term ECCS recirculation during the recirculation phase. Two of the four recirculation spray pumps are automatically realigned to LHSI during the transfer from injection to recirculation. In the recirculation mode, these two pumps provide water directly to the RCS as well as to the suction of the HHSI/charging pumps.

Section 6.2.2 discusses component design and NPSH considerations for these pumps.

Valves Design features employed to minimize valve leakage include:

1. Where possible, packless valves are used.

2. Valves with packed stuffing boxes are provided with intermediate stem leakoff connections when possible.

3. Valves that are not packless which are normally open, except check valves and those which perform a control function, are provided with backseats to limit stem leakage.

4. Normally closed globe valves are installed with recirculation fluid pressure under the seat to prevent stem leakage of recirculated (radioactive) water.

5. Relief valves are enclosed, that is, they are provided with a closed bonnet.

Motor-operated valves which must function against system pressure are designed such that they function with a pressure drop equal to full system pressure across the valve disc.

6.3-6

BVPS-2 UFSAR Rev. 26 Weight-loaded, swing-type check valves are used as the inside containment check valve on the SI lines. This type of weight-loaded check valve is also used on each line from the recirculation spray pump discharges to the LHSI headers.

Carbon steel manual valves are employed to pass non-radioactive fluids only.

During normal operation the accumulator check valves are in the closed position with a nominal differential pressure across the disc of approximately 1,650 psi. Since the valves remain in this position, except for testing or when called upon to open following an accident, and are therefore not subject to the abuse of flowing operation or impact loads caused by sudden flow reversal and seating, they do not experience significant wear of the moving parts and are expected to function with minimal back-leakage. This back-leakage can be checked via the test connection as described in Section 6.3.4.

Relief valves are installed in various sections of the ECCS to protect lines which have a lower design pressure than the RCS. Table 6.3-2 lists the ECCS relief valves with their capacities and set points.

Accumulator Motor-Operated Valve Controls As part of the plant normal shutdown administrative procedures, the operator is required to energize and close these valves. This prevents a loss of accumulator water inventory to the RCS and is done shortly after the RCS has been depressurized below the SI unblock set point. The redundant pressure and level alarms on each accumulator would remind the operator to close these valves, if any were inadvertently left open.

Power is disconnected after the valves are closed.

During plant start-up, the operator is instructed, via procedures, to energize and open these valves when the RCS pressure reaches the SI set point. Monitor lights in conjunction with an audible alarm will alert the operator should any of these valves be left inadvertently closed once the RCS pressure increases beyond the SI unblock set point. In addition, these valves will automatically receive a signal to open when the RCS pressure exceeds the unblock pressure.

The accumulator isolation valves are not required to move during power operation or in a post-accident situation. For a discussion of limiting conditions for operation and surveillance requirements of these valves, refer to the Technical Specifications (Chapter 16).

The accumulator isolation valves receive a SI signal to ensure that they are open in the event of an accident which initiates SI. Instrumentation associated with these valves is discussed in Sections 6.3.5, 7.3.1.1.2, and 7.6.4.

Motor-Operated Valves and Controls Remotely-operated valves for the injection mode (that is, valves which normally are in their ready position and do not require a SI signal) have 6.3-7

BVPS-2 UFSAR Rev. 26 their positions indicated on a common portion of the control board. If a component is out of its proper position, its monitor light will indicate this on the control board. At any time during operation when one of these valves is not in the ready position for injection, this condition is shown visually on the control board, and an audible alarm is sounded in the main control room.

Inadvertent mispositioning of MOVs due to failures in the control circuitry has been examined and found not to be a credible concern.

However, those valves which are not required to change position during the injection and automatic transfer to recirculation phases have received additional review. For these valves, power has been removed from the valves for which spurious repositioning would affect the ECCS function.

The ECCS valves which have power removed are:

1. Accumulator discharge isolation valves (2SIS*MOV865A, B, C),

2. The HHSI/charging pump discharge to the cold legs (2SIS*MOV841),

3. The HHSI/charging pump discharge to the hot legs (2SIS*MOV869A/B),

4. The LHSI pump discharge to the hot legs (2SIS*MOV8889),

5. The HHSI/charging pump discharge header cross-connect valves (2CHS*MOV8132A,B and 2CHS*MOV8133A,B), and

6. The HHSI/charging pump minimum flow discharge header isolation valve (2CHS*MOV373).

During normal plant operation the accumulator discharge isolation valves (2SIS*MOV865A,B, & C) can have their power removed to prevent spurious operation (see Note 3 on Figure 7.3-83). This is accomplished by a banana plug disconnect on the main control board. This similar power removal scheme is available for other previously mentioned ECCS valves with the exception of 2CHS*MOV373 which has its power isolated at the motor control center.

Table 6.3-3 provides information on various MOVs such as valve position indication, valve interlocks, and alarms.

6.3.2.3 Applicable Codes and Classifications Section 3.2 discusses the applicable codes and standards which apply to individual ECCS components.

6.3.2.4 Materials Specifications and Compatibility Materials employed for components of the ECCS are given in Table 6.3-4.

Materials are selected to meet the applicable material requirements of the codes in Table 3.2-1 and the following additional requirements:

6.3-8

BVPS-2 UFSAR Rev. 26

1. All parts of components in contact with borated water are fabricated of or clad with austenitic stainless steel or equivalent corrosion resistant material.

2. All parts of components in contact (internal) with sump solution during recirculation are fabricated of austenitic stainless steel or equivalent corrosion resistant material.

3. Valve seating surfaces are hard-faced with Stellite Number 6 or equivalent to prevent galling and to reduce wear.

4. Valve stem materials are selected for their corrosion resistance, high tensile properties, and resistance to surface scoring by the packing.

The elevated temperature of the sump solution during recirculation is well within the design temperature of all ECCS components. In addition, consideration has been given to the potential for corrosion of various types of metals exposed to the fluid conditions prevalent immediately after the accident or during long term recirculation operations (Section 6.1.1).

Environmental testing of ECCS equipment inside the containment is discussed in Section 3.11. This equipment is required to operate following a LOCA.

6.3.2.5 System Reliability Reliability of the ECCS is considered in all aspects of the system from initial design to periodic testing of the components during plant operation. The ECCS is a two train, fully redundant, standby safeguard feature. The system has been designed and proven by analysis to withstand any single credible active failure during injection, or any single active or passive failure during recirculation, and maintain the performance objectives discussed in Section 6.3.1. Separate series-aligned trains of pumps and flow paths are provided for redundancy as only one train and flow path is needed to satisfy the performance requirements. The initiating signals for the ECCS are derived from independent sources as measured from process (for example, low pressurizer pressure) or environmental variables (for example, containment pressure). Redundant as well as functionally independent variables are measured to initiate the safeguards signal. Each train is physically separated and protected where necessary so that a single event cannot initiate a common failure. Power sources for the ECCS are divided into two independent trains supplied from separate emergency buses powered from offsite power. Sufficient diesel generating capacity is maintained onsite to provide required power to each train should offsite power be unavailable. The diesel generators and their auxiliary systems are completely independent and a diesel generator is dedicated to each one of the two ECCS trains.

The preoperational testing program assures that the systems, as designed and constructed, will meet the functional requirements as calculated in the design. The ECCS is designed with the ability for on-line testing of most components so the availability and operational status can be readily 6.3-9

BVPS-2 UFSAR Rev. 26 determined. In addition, the integrity of the ECCS is assumed through examination of critical components during the routine in-service inspection (ISI).

The reliability program further extends to the procurement of ECCS components such that only designs which have been proven by past use in similar applications are acceptable for use. The quality assurance program, as described in Chapter 17, assures receipt of components only after manufacture and test to the applicable codes and standards.

Definitions of Terms Period of Recovery: The time necessary to bring the plant to a cold shutdown and regain access to failed equipment. The recovery period is the sum of the short and long term periods as further defined.

Incident: Any natural or accidental event of infrequent occurrence and its related consequences which affect plant operation and require the use of engineered safeguards systems. Such events, which are analyzed independently and are not assumed to occur simultaneously, include the LOCA, steam line ruptures, steam generator tube ruptures, etc. A system blackout may be an isolated occurrence or may be concurrent with any event requiring engineered safeguards systems use.

Short Term: The time immediately following the incident during which automatic actions are performed, system responses are checked, type of incident is identified, and preparations for long term recovery operation are made. The short term for the LOCA, for example, is the injection phase.

Long Term: The remainder of the recovery period following the short term.

In comparison with the short term where the main concern is to prevent or limit site release, the long term period of operation involves bringing the plant to cold shutdown conditions where access to the containment can be gained and repair effected. The long term for the LOCA, for example, is the recirculation phase.

Active Failure: The failure of a powered component such as a piece of mechanical equipment, a component of the electrical supply system, or instrumentation and control equipment to act on command to perform its design function. Examples include the failure of a MOV to move to its correct position, the failure of an electrical breaker or relay to respond, the failure of a pump, fan, or diesel generator to start, etc.

Passive Failure: The structural failure of a static component which limits the component's effectiveness in carrying out its design function.

Examples include cracks in pipes, sprung flanges, valve packing leaks, or pump seal failures.

Active Failure Criteria The ECCS is designed to accept a single failure following the incident without loss of its protective function. The system design will tolerate the failure of any single active component in the ECCS itself or in the 6.3-10

BVPS-2 UFSAR Rev. 26 necessary associated service systems at any time during the period of required system operations following the incident.

A single active failure analysis is presented in Table 6.3-5, and demonstrates that the ECCS can sustain the failure of any single active component in either the short or long term and still meet the level of performance for core cooling.

Since the operation of the active components of the ECCS following a steam line rupture is identical to that following a LOCA, the same analysis is applicable and the ECCS can sustain the failure of any single active component and still meet the level of performance for the addition of shutdown reactivity.

Passive Failure Criteria The following philosophy provides for necessary redundancy in component and system arrangement to meet the intent of the general design criteria on single failure as it specifically applies to failure of passive components in the ECCS. Thus, for the long term, the system design is based on accepting either a passive or an active failure, assuming no failures in the short term.

Redundancy of Flow Paths and Components for Long Term Emergency Core Cooling The design of the ECCS utilizes the following criteria:

1. During the long term cooling period following a LOCA, the emergency core cooling flow paths shall be separable into two subsystems, either of which can provide minimum core cooling functions and return spilled water from the floor of the containment back to the RCS.

2. Either of the two subsystems can be isolated and removed from service in the event of a leak outside the containment.

3. Adequate redundancy of check valves is provided to tolerate failure of a check valve during the long term as a passive component.

4. Should one of these two subsystems be isolated in this long term period, the other subsystems remain operable.

5. Provisions are also made in the design to detect leakage from components outside the containment, collect this leakage, and to provide the maintenance of the affected equipment.

Thus, for the long term emergency core cooling function, adequate core cooling capacity exists with one flow path removed from service.

6.3-11

BVPS-2 UFSAR Rev. 26 Subsequent Leakage From Components in Safeguards Systems The features described in this section meet the intent of NUREG-0737, Item III.D.1.1 (USNRC 1980), for the ECCS.

With respect to piping and mechanical equipment outside the containment, considering the provisions for visual inspection and leak detection, leaks will be detected before they propagate to major proportions. A review of the equipment in the system indicates that the largest sudden leak potential would be the sudden failure of a pump shaft seal. Evaluation of leak rate, assuming only the presence of a seal retention ring around the pump shaft, showed that flows less than 50 gpm would result. Piping leaks, valve packing leaks, or flange gasket leaks have been of a nature to build up slowly with time and are considered less severe than the pump seal failure.

Larger leaks in the ECCS are prevented by the following:

1. The piping is classified in accordance with ANS Safety Class 2 and receives the ASME Section III Class 2 quality assurance program associated with this safety class.

2. The piping, equipment and supports are designed to ANS Safety Class 2 seismic classification permitting no loss of function for the design basis earthquake.

3. The system piping is located within a controlled area on the plant site.

4. The piping system receives periodic pressure tests and is accessible for periodic visual inspection.

5. The piping is austenitic stainless steel which, due to its ductility, can withstand severe distortion without failure.

Therefore, the design of the auxiliary building and related equipment is based upon handling of leaks up to a maximum of 50 gpm. Means are also provided to detect and isolate such leaks in the emergency core cooling flow path within 30 minutes. With these design ground rules, continued function of the ECCS will meet minimum core cooling requirements.

A single passive failure analysis is presented in Table 6.3-6. It demonstrates that the ECCS can sustain a single passive failure during the long term phase and still retain an intact flow path to the core to supply sufficient flow to maintain the core covered and effect the removal of decay heat. The procedure followed to establish the alternate flow path also isolates the component which failed.

Potential Boron Precipitation Boron precipitation in the reactor vessel can be prevented by a backflush of cooling water through the core.

6.3-12

BVPS-2 UFSAR Rev. 26 During the long term cooling phase of ECCS operation, recirculation flow will be redirected from the cold legs to the hot legs. Each of the two charging pumps has a separate flow path through hot leg connections to provide backflushing through the core. In addition, the recirculation spray pumps can be aligned to deliver flow to the hot legs through the common cross connection. Hot leg recirculation, in addition to preventing boron precipitation by backflushing the core, provides subcooled water to terminate boil off.

Lag Times Lag time for initiation and operation of the ECCS is limited by pump start-up time and consequential loading sequence of the motor onto the emergency buses. Most valves are normally in the position conducive to safety, therefore valve opening time is not considered for these valves.

In the case of a blackout, a 10 second delay is assumed for diesel start-up. The HHSI/charging pumps and all valves are then applied to the buses and the LHSI pumps will start 5 seconds later. If there is no LOOP the same starting sequence is followed without delay, the first load being started upon receipt of the SI signal.

6.3.2.6 Protection Provisions The provisions taken to protect the system from damage that might result from dynamic effects of pipe rupture are discussed in Section 3.6. The provisions taken to protect the system from missiles are discussed in Section 3.5. The provisions taken to protect the system from seismic damage are discussed in Sections 3.7, 3.9, and 3.10. Thermal stresses on the RCS are discussed in Section 3.9.

6.3.2.7 Provisions for Performance Testing Provisions to facilitate performance testing are discussed in Section 6.3.4.

6.3.2.8 Manual Actions No manual actions are required during the short term injection phase.

Transfer from injection to cold leg recirculation is initiated automatically and only limited operator actions are required to complete the transfer. The manual actions required of the operator to complete the transfer are:

1. The isolation valve in the redundant HHSI flow path to the cold legs is opened (2SIS*MOV836), and

2. The redundant HHSI flow paths are separated by closing the isolation valves in both the common suction and discharge headers of the HHSI/charging pumps (2CHS*MOV8130A, B; 2CHS*MOV8131A, B; 2CHS*MOV8132A, B; and 2CHS*MOV8133A, B).

3. If HHSI pump C is operating, then on the pump suction side either 2CHS*MOV8130A (2CHS*MOV8130B-analogous) or 2CHS*MOV8131A (2CHS*MOV8131B-analogous) is closed, but not both. Similarly, 6.3-13

BVPS-2 UFSAR Rev. 26 on the HHSI pump C discharge side, either 2CHS*MOV8132A (2CHS*MOV8132B-analogous) or 2CHS*MOV8133A (2CHS*MOV8133B-analogous) is closed, but not both. This is to prevent isolation of HHSI pump C should it be operating to supply one of two SI trains.

The switchover to hot leg recirculation for long term cooling requires further manual actions as described in Table 6.3-7. The operator terminates recirculation spray pump flow to the RCS cold legs and establishes flow to the RCS hot legs. Similar actions need to be taken to switch HHSI/charging pump flow to provide hot leg recirculation.

Those valves in the ECCS and recirculation spray system, which are essential to ECCS operation, have been evaluated for accessibility in the event these valves must be manually positioned during the long term. All critical valves are either accessible or have been provided with reach rods for manual operation.

6.3.3 Performance Evaluation The accidents discussed in Chapter 15 which result in ECCS operation are as listed:

1. Accidental depressurization of the main steam system (MSS),

2. Small LOCAs,

3. Major rupture of reactor coolant pipe,

4. Major steam system pipe rupture,

5. Steam generator tube rupture,

6. Feedwater system pipe break,

7. Inadvertent RCS depressurization, or

8. A RCCA ejection accident.

Simplified functional flow diagrams are shown on Figure 6.3-6. The time sequence for ECCS component actuation is discussed in Chapter 15 with the appropriate accident analysis.

Accidental Depressurization of the Main Steam System The most severe core conditions resulting from an accidental depressurization of the MSS are associated with an inadvertent opening of a single steam dump, relief, or safety valve.

Safety injection is actuated from any of the following:

1. Two out of three low pressurizer pressure signals,

2. Two out of three Hi-1 containment pressure signals, 6.3-14

BVPS-2 UFSAR Rev. 26

3. Two out of three low steam line pressure signals in any one loop, or

4. Manual initiation of the safety injection signal from main control room.

The SI signal (and other actuation signals resulting from a SI signal) will close all feedwater control valves, trip the main feedwater pumps, and close the main feedwater isolation valves. Following the SI signal, the HHSI/charging pump suction is switched from the VCT to the RWST. At this time, isolation valves 2SIS*MOV867A,B,C, and D open and the HHSI/charging pumps discharge water into the RCS cold legs. At the same time, the LHSI pumps are started but they only provide recirculation flow back to the RWST since the RCS pressure remains above the shutoff head of the pumps. The accumulators do not discharge since the RCS pressure also remains above accumulator pressure.

Results and Conclusions of Accidental Depressurization of Main Steam System The assumed steam release is typical of the capacity of any single steam dump relief or safety valve. The boron solution provides sufficient negative reactivity to meet the departure from nucleate boiling (DNB) design basis. The cooldown for this case is more rapid than the case of steam release from all steam generators through one steam dump, relief, or safety valve. The transient is quite conservative with respect to cooldown, because no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the steam generators. Since the transient occurs over a period of about 5 minutes, the neglected stored energy is likely to have a significant effect in slowing the cooldown. The analysis shows that there will be no violation of the DNB design basis after reactor trip assuming a stuck RCCA, with offsite power available, and assuming a single failure in the ECCS. Therefore, a departure from nucleate boiling ratio less than 1.30 does not exist.

Loss of Reactor Coolant From Small Ruptured Pipes or From Cracks In Large Pipes Which Actuates the Emergency Core Cooling System A LOCA is defined as a rupture of the RCS piping or of any line connected to the system. Ruptures of small cross section will cause expulsion of the coolant at a rate which can be accommodated by the HHSI/charging pumps, which would maintain an operational water level in the pressurizer permitting the operator to execute an orderly shutdown.

The maximum break size for which the normal makeup system can maintain the pressurizer level is obtained by comparing the calculated flow from the RCS through the postulated break against the charging pump makeup flow at normal RCS pressure, that is 2,250 psia. A makeup flow rate from one centrifugal charging pump is typically adequate to sustain pressurizer level and pressure for a break through a 3/8-inch diameter hole. As part of the normal makeup system, a second charging pump is available to provide additional makeup flow to help maintain pressurizer level and 6.3-15

BVPS-2 UFSAR Rev. 26 pressure. A small break, as considered in this section, is then defined as a rupture of the RCS with a total cross-sectional area less than 1.0 ft2 and for which makeup via normal charging flow is not sufficient.

The ECCS operation following a small break LOCA is similar to that for a large break LOCA except that the time periods for ECCS component actuation are substantially longer. Additionally, in the event the small break LOCA does not result in the containment pressure initiating the CIB signal, the operator would be required to manually start the recirculation spray pumps when the RWST reaches low level.

Results and Conclusions from Analysis for Small Break LOCA The analysis of this break has shown that the high head portion of the ECCS, together with the accumulator, provides sufficient core flooding to keep the calculated peak clad temperature below required limits of 10 CFR 50.46. Hence, adequate protection is afforded by the ECCS in the event of a small break LOCA.

Major Reactor Coolant System Pipe Ruptures (Loss-of-Coolant Accident)

Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. Reactor trip signal occurs when the pressurizer low pressure trip set point is reached. A SI signal is actuated when the appropriate set point is reached. These provisions limit the consequences of the accident in two ways:

1. Reactor trip and borated water injection provide additional negative reactivity insertion to supplement void fraction in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

2. Injection of borated water provides heat transfer from the core and prevents excessive clad temperature.

The operation of the ECCS is as described in Section 6.3.2.1 Results and Conclusions for Major Reactor Coolant System Pipe Rupture Conclusions - Thermal Analysis For breaks up to and including the double-ended severance of a reactor coolant pipe, the ECCS will meet the acceptance criteria as presented in 10 CFR 50.46. That is:

1. The calculated peak fuel element clad temperature provides margin to the requirement of 2,200F.

2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.

6.3-16

BVPS-2 UFSAR Rev. 26

3. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation limits of 17 percent are not exceeded during or after quenching.

4. The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long-lived radioactivity remaining in the core.

Major Secondary System Pipe Rupture The steam release arising from a main steam line break would result in an initial increase in steam flow, which decreases during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure. The cooldown results in a reduction of core shutdown margin when the moderator temperature coefficient is negative. If it is assumed that the most reactive RCCA is stuck in the fully withdrawn position after reactor trip, there is an increased probability that the core will return to power. This return to power is a potential problem because of the high peaking factors which exist, assuming the most reactive RCCA to be stuck in its fully withdrawn position.

Safety injection is actuated from any of the following:

1. Two out of three low pressurizer pressure signals,

2. Two out of three Hi-1 containment pressure signals,

3. Two out of three low steam line pressure signals in any one loop, or

4. Manual initiation of the safety injection signal from the main control room.

The HHSI/charging pumps begin pumping RWST water into the RCS. This pumping action occurs as the RCS pressure is decreasing and restores the RCS to the correct volume.

Results and Conclusions of Major Secondary System Pipe Rupture The analysis has shown that even assuming a stuck RCCA with or without offsite power, and assuming a single failure in the engineered safeguards, the core remains in place and intact. Radiation doses will not exceed 10 CFR 100 guidelines.

Although DNB and possible clad perforation following a Main Steam Line Break are not necessarily unacceptable and not precluded in the criterion, the preceding analysis shows that no DNB occurs for any rupture assuming the most reactive RCCA stuck in its fully withdrawn position.

6.3-17

BVPS-2 UFSAR Rev. 26 Steam Generator Tube Rupture The postulated accident is the complete severance of a steam generator tube occurring at power. The higher pressure reactor coolant causes inflow to the affected steam generator and loss of volume and pressure in the pressurizer. The steam side experiences a steam flow-feed flow mismatch in one steam generator. The reactor trips on low pressurizer pressure or overtemperature T from the continued loss of RCS inventory and SI is also initiated at the same time.

Since the initial signals available to the operator are also indicative of other accident types, the operator must determine that a tube rupture has occurred. Steam generator water level rising more rapidly in one steam generator than the others is a unique indication of this accident.

Using the recovery procedure outlined in Chapter 15 the operator identifies the affected steam generator, and after reduction of the RCS to the proper set points, initiates operation of the residual heat removal (RHR) system.

Results and Conclusions of Steam Generator Tube Rupture A steam generator tube rupture will cause no subsequent damage to the RCS or the reactor core. An orderly recovery from the accident can be completed even assuming simultaneous LOOP.

Existing Criteria Used to Judge the Adequacy of the ECCS In order to assure the performance of the ECCS in the accident analysis, minimum performance levels are designated. To assure that these performance levels are maintained, Technical Specifications have been established and documented in Chapter 16.

The following criteria are used to judge the adequacy of the ECCS performance:

1. Criteria from 10 CFR 50.46

a. Peak cladding temperature calculated shall not exceed 2,200F.

b. The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

c. 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 amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding around the plenum, were to react.

d. Calculated changes in core geometry shall be such that the core remains amenable to cooling.

6.3-18

BVPS-2 UFSAR Rev. 26

e. After any calculated successful initial operation 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 long-lived radioactivity remaining in the core.

2. In the case of the accidental depressurization of the MSS, an additional criteria for the adequacy of the ECCS is: assuming a stuck RCCA, with or without offsite power, and assuming a single failure in the engineered safety features (ESF), there will be no violation of the DNB design basis after reactor trip for a steam release equivalent to the accidental opening, with failure to close, of the largest of any single steam dump, relief, or safety valve.

3. For a major secondary system pipe rupture, the additional criterion is: assuming a stuck RCCA with or without offsite power, and assuming a single failure in the ESF, the core remains in place and intact.

Shared Function Components During normal operation, some components of the ECCS may be used to support power operation. In this case, these components will automatically realign to the safeguards mode upon receipt of a SI signal.

These components are as listed:

1. The HHSI/charging pumps are normally aligned to provide seal water to the reactor coolant pumps and for replacement of letdown. During ECCS operation these pumps inject refueling water into the cold legs. In emergency operation, the charging pump suction is automatically switched to the RWST.

2. The RWST and LHSI pumps are used to fill the refueling canal for refueling operations. For other periods of operation the RWST and the LHSI pumps are aligned for injection.

An evaluation of all components required for operation of the ECCS demonstrates that either:

1. The component is not shared with other systems.

2. If the component is shared with other systems, it is either aligned during normal plant operation to perform its accident function or if not aligned to its accident function, two valves in parallel are provided to align the system for injection, and two valves in series are provided to isolate portions of the system not utilized for injection. (These valves are automatically actuated by the SI signal.)

3. The component shared with another system is used for a different ESF in the short term then subsequently realigned for the ECCS operation in the recirculation phase.

6.3-19

BVPS-2 UFSAR Rev. 26 Table 6.3-8 indicates the alignment of components during normal operation and the realignment required to perform the accident function. In all cases of component operation, safeguards operation has the priority usage such that a safeguard signal will override all other signals and start or align systems for their safeguards function.

Limits on System Parameters The analyses show that the design basis performance characteristic of the ECCS is adequate to meet the requirements for cooling following an accident with the minimum ESF equipment operating. In order to ensure this capability in the event of the simultaneous failure to operate any single active component, Technical Specifications are established for reactor operation (Chapter 16).

Normal operating status of ECCS components is given in Table 6.3-9.

The ECCS components are available whenever the coolant energy is high and the reactor is critical. During low temperature physics tests, there is a negligible amount of stored energy in the coolant and low decay heat.

Therefore, an accident comparable in severity to accidents occurring at operating conditions is not possible and ECCS components are not required.

The principal system parameters and the number of components which may be out of operation in test, quantities and concentrations of coolant available, and allowable time in a degraded status, are addressed in Chapter 16. If efforts to repair the faulty component are not successful, the plant is placed into a lower operational status, that is, hot standby to hot shutdown, hot shutdown to cold shutdown, etc.

6.3.4 Inspection and Testing Requirements 6.3.4.1 Emergency Core Cooling System Performance Tests Pre-operational testing of the ECCS is conducted during the hot functional testing of the RCS following flushing and hydrostatic testing to demonstrate the capability of HHSI and accumulator injection at operating temperature and pressure.

After HFT with the system cold and the reactor vessel head removed, pre-operational testing of the recirculation spray pumps in the recirculation mode will be conducted with water drawn from the containment sump and delivered to the suction of the HHSI/charging pumps. During this test, the HHSI/charging pumps are not started and water is not injected into the RCS.

Separate flow tests of the LHSI and HHSI/charging pumps are conducted during pre-operational testing (with the reactor vessel head off) to check operation. Each pump would be aligned to take suction from the RWST and to discharge into the reactor vessel through the injection lines. At this time, the throttle valves in the HHSI lines are positioned to limit pump runout and equalize injection flow to all RCS loops. Data will be taken to determine pump head and flow at this time.

6.3-20

BVPS-2 UFSAR Rev. 26 Pumps are also run with only the miniflow circuits open and data taken to determine a second point on the head/flow characteristic curve.

Each accumulator is filled with water from the RWST and pressurized with nitrogen, with the isolation valve in the discharge line closed. The valve is opened and the accumulator allowed to discharge into the reactor vessel as part of pre-operational testing with the reactor cold and vessel head off.

The SI block switch is reset and the breakers on the lines supplying offsite power are tripped manually so that operation of the emergency diesel generators is tested in conjunction with the ECCS. During the test in the HHSI/charging and LHSI pumps would inject into the reactor vessel, via the RCS cold legs, with the overflow from the reactor vessel spilling into the refueling canal.

Section 14.2.12 discusses the testing to be performed on these systems.

6.3.4.2 Reliability Tests and Inspections Description of Test Planned.

Where possible, without interruption of service, routine periodic testing of the ECCS components and all necessary support systems at power is planned. Valves which must operate as part of the system safety function are operated through a complete cycle, and pumps are operated individually recirculating through their miniflow lines with the exception of the charging pumps, which are tested by their normal charging function. If such testing indicates a need for corrective maintenance, the redundancy of equipment in these system permits such maintenance to be performed without shutting down or reducing load under certain conditions. These conditions include consideration such as a period within which the component should be restored to service and the capability of the remaining equipment to provide the minimum required level of performance during such a period.

The operation of the remote isolation valve and the upstream check valve in each accumulator tank discharge line may be tested by opening the remote test line valves just downstream of the isolation valve and check valve, respectively. Flow through the test line can be observed on instruments, and the opening and closing of the discharge line isolation valve can be monitored by position indication lights in the main control room.

Where series pairs of check valves form the high pressure to low pressure isolation barrier between the RCS and ECCS piping outside the reactor containment, periodic testing of these check valves must be performed to provide assurance that certain postulated failure modes will not result in a loss-of-coolant from the low pressure system outside containment with a simultaneous loss of SI pumping capacity.

A manual testing procedure is used for determination of the integrity of the pressure boundary formed by series check valves. The tests performed verify that each of the series check valves can independently sustain differential pressure across its disc, and also verify that the valve is 6.3-21

BVPS-2 UFSAR Rev. 26 in its closed position. The required periodic tests are to be performed after each refueling just prior to plant start-up, after the RCS has been pressurized.

Lines in which the series check valves are to be tested are the LHSI pump cold leg injection lines. To implement the periodic component testing requirements, Technical Specifications (Chapter 16) are established.

During periodic system testing, a visual inspection of pump seals, valve packings, flanged connection, and relief valves is made to detect leakage.

An ISI provides further confirmation that no significant deterioration is occurring in the ECCS fluid boundary.

Design measures have been taken to assure that the following testing can be performed:

1. Active components may be tested periodically for operability (pumps on miniflow, certain valves, etc).

2. An integrated system actuation test (details of the testing of the sensors and logic circuits associated with the generation of a SI signal together with the application of this signal to the operation of each active component are given in Section 7.2) can be performed when the plant is cooled down and the RHR system is in operation. The ECCS will be arranged so that no flow will be introduced into the RCS for this test.

3. An initial flow test of the full operational sequence can be performed.

The design features which assure this test capability are specifically:

1. Power sources are provided to permit individual actuation of each active component of the ECCS.

2. The LHSI pumps are tested periodically when the plant is at power, using the miniflow recirculation lines.

3. The HHSI/charging pumps are either normally in use for charging service or can be tested periodically using miniflow recirculation lines.

4. Remotely-operated valves can be exercised during routine plant maintenance.

5. Level and pressure instrumentation is provided for each accumulator tank for continuous monitoring of these parameters during plant operation.

6. Flow from each accumulator tank can be directed at any time through a test line to determine check valve leakage and to demonstrate operation of the accumulator MOVs.

7. A flow indicator is provided in the LHSI pump headers.

Pressure instrumentation is also provided in these lines.

6.3-22

BVPS-2 UFSAR Rev. 26

8. An integrated system test can be performed when the plant is cooled down and the RHR system is in operation. This test demonstrates the operation of the valve, pump circuit breakers, and automatic circuitry, including diesel starting and the automatic loading of ECCS components on the diesels (by simulating a LOOP to the Class lE electrical buses).

Chapter 16 discusses the selection of test frequency, acceptability of testing, and measured parameters. A description of the ISI program is also included in Section 6.6. The ECCS and its components are designed to meet the intent of ASME Section XI for ISI.

On January 11, 2008, the NRC issued Generic Letter 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems. Generic Letter 2008-01 requested licensees to evaluate the licensing basis, design, testing, and corrective action programs for the emergency core cooling, decay heat removal, and containment spray systems.

As a result, the company performed evaluations that included the review of gas susceptible piping locations, the development of activities to monitor various piping locations as appropriate based on industry experience and plant specific experience, and acceptance of some generic locations that normally accumulate voids that do not adversely affect the design function(s) of the system.

The company established a gas accumulation prevention and management program to ensure that gas accumulation is reasonably prevented or maintained less than the amount that challenges the functionality of the applicable systems and that appropriate action is taken when conditions adverse to quality are identified.

6.3.5 Instrumentation Requirements.

6.3.5.1 Actuation Signal The actuation signal which initiates SI is referred to as the SI signal.

The SI signal would primarily be used to automatically start the LHSI and HHSI/charging pumps.

Two separate and redundant actuation trains are provided. Each actuation train is assigned to a corresponding electrical power train to ensure that, in the event of a single failure in the actuation logic, at least one emergency diesel generator, one LHSI, and one HHSI/charging pump would receive an actuation signal.

Each actuation train is driven by a separate protection logic train to ensure that, in the event of a single failure in the protection logic, the minimum safeguards equipment would still receive an actuation signal.

The signals which are generated by the protection logic and used to initiate the SI signal are the following:

1. Two out of three low pressurizer pressure signals, 6.3-23

BVPS-2 UFSAR Rev. 26

2. Hi-1 containment pressure trip signal produced by two out of three Hi-1 containment pressure signals,

3. Low steam line pressure signal produced by two out of three low steam line pressure signals in one line, or

4. Manual SI actuation from the main control board.

In addition to the SI signal, the protection logic initiates the following actuation signals which are not specifically related to the operation of the ECCS (specifically excluded is a description of the protection logic signals which are generated to initiate steam line isolation):

The CIA Signal - The actuation signal which would initiate CIA and containment ventilation isolation is referred to as the CIA signal. The CIA signal is initiated from the same protection logic signals which produced the SI signal, with the exception that a separate manual actuation switch is provided on the main control board which would permit the operator to initiate the CIA actuation without initiating SI actuation.

The CIB Signal - The actuation signal which would initiate spray actuation and CIB is referred to as the CIB signal. The signals which are generated by the protection logic and used to initiate the CIB signal are the following:

1. Hi-3 containment pressure trip signal produced by two out of four Hi-3 containment pressure signals, or

2. Manual actuation from the main control board.

6.3.5.2 Pressure Indication Low Head Safety Injection Pump Suction and Discharge Pressure The LHSI pump suction and discharge pressure are indicated locally.

Accumulator Pressure Duplicate pressure channels are installed on each accumulator. Pressure indication in the main control room, and high and low pressure alarms, are provided by each channel.

Test Line Pressure A local pressure indicator used to check for proper seating of the accumulator check valves between the injection lines and the RCS is installed on the leakage test line.

Hydrotest Pump Discharge Local pressure indication is provided to monitor hydro-test pump discharge pressure.

6.3-24

BVPS-2 UFSAR Rev. 26 6.3.5.3 Flow Indication Charging Pump Injection Flow Charging pump injection and recirculation header flow are indicated in the main control room.

Low Head Safety Injection Pump Minimum Flow A flow meter installed in each LHSI pump discharge header provides control for the valves located in the pump minimum flow line.

Low Head Safety Injection Flow Low head safety injection flow during injection and recirculation is indicated in the main control room.

Test Line Flow Local indication of the leakage test line flow is provided to check for proper seating of the accumulator check valves between the injection lines and the RCS.

6.3.5.4 Level Indication Refueling Water Storage Tank Level The RWST is provided with eight water level transmitters. Two transmitters are provided for wide range level measurement, with indicators provided in the main control room. Two transmitters are provided for narrow range level indication, with indicators provided in the main control room.

Level alarms are provided to protect against possible overflow of the RWST to assure that a sufficient volume of water is always available in the RWST in conformance with the Technical Specifications, and to indicate that the useable volume of the RWST has been exhausted.

Four transmitters (four channels) are provided to automatically actuate the ECCS and containment spray system switchover from injection mode to recirculation mode following an accident.

Accumulator Water Level Duplicate water level channels are provided for each accumulator. Both channels provide indication in the main control room and actuate high and low water level alarms.

6.3.5.5 Valve Position Indication Valve positions are indicated on the main control board by red (open) and green (closed) lights. When both lights are on, they indicate a valve in an intermediate position.

6.3-25

BVPS-2 UFSAR Rev. 26 Accumulator Isolation Valve Position Indication The accumulator isolation MOVs are provided with two sets of red (open) and green (closed) position indicating lights located at the main control board switch for each valve. One set of these lights is powered by the valve control power and actuated by valve motor operator limit switches.

This set of indication lights is not deenergized when MOV control power is removed during normal power operation in accordance with Table 8.3-5. The other set of lights is powered by a separate 120 V ac supply and are actuated by stem-mounted limit switches on the valve.

An alarm annunciator point is activated by a valve motor operator limit switch whenever an accumulator valve is not fully open for any reason with the system at pressure (the pressure at which the SI block is unblocked is approximately 1,900 psig). This alarm will be recycled at approximately l-hour intervals to remind the operator of the improper valve lineup (Section 7.6.3).

6.3.6 Reference for Section 6.3 U.S. Nuclear Regulatory Commission 1980. Clarification of TMI Action Plan Requirements. NUREG-0737.

6.3-26

BVPS-2 UFSAR Tables for Section 6.3

BVPS-2 UFSAR Rev. 16 TABLE 6.3-1 EMERGENCY CORE COOLING SYSTEM COMPONENT PARAMETERS Accumulators Quantity 3 Design pressure (psig) 700 Design temperature (F) 300 Operating temperature (F) 70 to 150 Nominal operating pressure (psig) 640 Total volume (ft3) (each) 1,450 Nominal water volume (ft3) (each) 925 Nominal volume N2 Gas (ft3) (each) 525 Boric acid concentration (ppm) 2,300 to 2,600 Relief valve set point (psig) 700 HHSI/Charging Pumps Quantity 3 Design pressure (psig) 2,800 Design temperature (F) 300 Design flow rate (gpm)* 150 Design head at design flow rate (ft) 5,800 (minimum)

Maximum flow rate (gpm) 580 runout ****

Design head at maximum flow rate (ft) 1,250-1,650 Design discharge head at shutoff (ft) 6,000-6,200 Motor rating (bhp)** 600 NPSH, available at 580 gpm maximum 47 flow rate (ft)

NPSH, maximum required at 580 gpm 40 maximum flow rate(ft)

Low Head Safety Injection Pumps Quantity 2 Design pressure (psig) 240 Design temperature (F) 200 Design flow rate (gpm) 3,000 Design head (ft) 225 Maximum flow rate (gpm) 5,000 Design discharge head of shutoff (ft) 350 NPSH, available (ft) 38 NPSH, maximum required (ft) 18 1 of 2

BVPS-2 UFSAR Rev. 16 TABLE 6.3-1 (Cont)

Hydrotest Pump Quantity 1 Design pressure (psig) 3,125 Design temperature (F) 250 Normal operating temperature (F) 130 Normal flow rate (gpm) 15 Design flow rate (gpm) 26.5 Developed head at design flow rate (psig) 3,125 Motor rating (bhp) 60 Recirculating Pumps (Section 6.2.2 - design parameters)

Motor-Operated Valves Maximum opening or closing time Up to and including 8 inches 10 to 15 sec***

Over 8 inches 15 to 20 sec***

NOTES:

• Includes miniflow

• • 1.15 service factor not included

• • • Times vary depending upon size, valve type, and type of actuator.

• • • • Maximum runout flow rate is 585 gpm during recirculation mode of ECCS (580 gpm is injection phase of ECCS).

2 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.3-2 EMERGENCY CORE COOLING SYSTEM RELIEF VALVE DATA Back Fluid Fluid Inlet Set Pressure Back Dis- Temperature, Pressure Constant Developed Description charged Normal (F) (psig) (psig) (psig) Capacity N2 supply to accumu-lators N2 Gas Ambient 700 0 0 1,500 scfm Low head cold leg safety in-jection lines Water 100 220 3 50 25 gpm Accumulator to contain-ment N2 Gas 120 700 0 0 1,500 scfm Hydrotest pump dis-charge Water 100 700 0 0 30 gpm Low head hot leg safety injection lines Water 100 220 3 50 25 gpm 1 of 1

BVPS-2 UFSAR Rev. 26 TABLE 6.3-3 MOTOR-OPERATED ISOLATION VALVES IN EMERGENCY CORE COOLING SYSTEMS Automatic Position Location Features Indication Alarms on Comments HHSI/charging pump Opens on SI signal MCB Incorrect discharge to cold legs position (2SIS*MOV867A, B, C, D)

Accumulator discharge 1. Open (if closed) MCB Incorrect These valves are (2SIS*MOV865A, B, C) on SI signal position normally open during

2. Open (if closed) power operation and on RCS pressure have their power above SI unblock removed.

pressure LHSI pump suction Close on LHSI pump MCB Incorrect These valves are from RWST stop signal position normally open during (2SIS*MOV8809A, B) coincident with SI power operation.

signal Recirculation spray Open on RWST level MCB Incorrect pump discharge to extreme low signal position LHSI paths to cold coincident with SI legs (2SIS*MOV8811A,B) signal Recirculation spray Open a stem position MCB Incorrect pump discharge to signal (open) from position HHSI/charging pump valves suction (2SIS*MOV863A, B) 2SIS*MOV8811A, B HHSI/charging pump None MCB Incorrect These valves are discharge to hot legs position normally closed and (2SIS*MOV869A, B) have their power removed. They are opened by the operator during the transfer from cold leg to hot leg recirculation.

1 of 4

BVPS-2 UFSAR Rev. 26 TABLE 6.3-3 (Cont)

Automatic Position Location Features Indication Alarms on Comments HHSI/charging pump None MCB Incorrect This valve is normally discharge to cold legs position closed and has its power (2SIS*MOV836) removed. The valve is opened by the operator to provide redundant HHSI subsystems during the transfer from injection to cold leg recirculation.

LHSI pump discharge Close on RWST MCB Incorrect cross-connect extreme low level position (2SIS*MOV8887A, B) signal coincident with a SI signal LHSI pump discharge None MCB Incorrect These valves are to cold legs position normally open and are (2SIS*MOV8888A, B) closed by the operator during the transfer from cold leg to hot leg recirculation.

LHSI pump discharge None MCB Incorrect This valve is normally to hot legs position closed and has its power (2SIS*MOV8889) removed. The valve is opened by the operator during the transfer from cold leg to hot leg recirculation.

Recirculation spray pump Open on CIB MCB Incorrect These valves are suction from containment signal position normally open and sump (2RSS*MOV155C, D) receive a CIB signal to open (if closed).

Recirculation spray Close on RWST MCB Incorrect pump discharge to extreme low level position containment spray header signal coincident (2RSS*MOV156C, D) with a SI signal 2 of 4

BVPS-2 UFSAR Rev. 26 TABLE 6.3-3 (Cont)

Automatic Position Location Features Indication Alarms on Comments HHSI/charging pump 1.Close on SI signal MCB Incorrect These valves are suction from VCT 2.Close on VCT low- position interlocked with (2CHS*LCV115C, E) low level signal HHSI/charging pump suction valves 2CHS*LCV15B, D from RWST. The valves will close after the RWST valves open.

HHSI/charging pump 1. Open on SI signal MCB Incorrect suction from RWST 2. Open on VCT low- position (2CHS*LCV115B, D) low level signal HHSI/charging pump None MCB Incorrect These valves are miniflow position normally open and have (2CHS*MOV373 their power removed.

and 2CHS*MOV275B)

HHSI/charging pump None MCB Incorrect miniflow position (2CHS*MOV275A, C)

HHSI/charging pump Close on SI signal MCB None discharge to normal charging path (2CHS*MOV289 and 2CHS*MOV310)

HHSI/charging pump None MCB Incorrect These valves are suction header position normally open and are (2CHS*MOV8130 A, B closed by the operator and 2CHS*MOV8131 A, during the transfer from B) injection to cold leg recirculation.

3 of 4

BVPS-2 UFSAR Rev. 26 TABLE 6.3-3 (Cont)

Automatic Position Location Features Indication Alarms on Comments HHSI/charging pump None MCB Incorrect These valves are discharge header position normally open and have (2CHS*MOV8132 A, B their power removed.

and 2CHS*MOV8133 A, B) They are closed by the operator during the transfer from injection to cold leg recirculation.

HHSI/charging pump None MCB Incorrect This valve is normally discharge to cold position open and has its power legs (2SIS*MOV841) removed. This valve is part of the cold shutdown design (Section 5.4.7).

4 of 4

BVPS-2 UFSAR Rev. 7 TABLE 6.3-4 MATERIAL(1) EMPLOYED FOR EMERGENCY CORE COOLING SYSTEM COMPONENTS Component Material(1)

Accumulators Carbon steel, clad with austenitic stainless steel Pumps HHSI/Charging Austenitic stainless steel Low head safety injection Austenitic stainless steel Valves Motor-operated valves containing radioactive fluids Pressure containing parts Austenitic stainles s steel or equivalent Body-to-bonnet Low alloy steel bolting and nuts Seating and surfaces Stellite Number 6 or equivalent Stems Austenitic stainless steel or 17-4 pH stainless steel Diaphram Valves Austenitic stainless steel Accumulator Check Valves Parts contacting borated Austenitic stainless steel water Clapper arm shaft 17-4 pH stainless steel Relief Valves Stainless steel bodies Stainless steel Carbon steel bodies Carbon steel All nozzles, discs, Austenitic stainless steel spindles, and guides 1 of 2

BVPS-2 UFSAR Rev. 7 TABLE 6.3-4 (Cont)

Component Material(1)

Bonnets for stainless Stainless steel or steel plated valves without a balancing carbon steel bellows All other bonnets Carbon steel Piping All piping in contact with Austenitic stainless steel borated water NOTES(1) Materials listed in this table may have been replaced with materials of equivalent design characteristics. The term equivalent is described in UFSAR Section 1.12, Equivalent Materials.

2 of 2

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 FAILURE MODES AND EFFECTS ANALYSIS EMERGENCY CORE COOLING SYSTEM - ACTIVE COMPONENTS Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

1. Motor-operated Fails to close on Provides HHSI/ Failure reduces Valve position indication Valve is interlocked with gate valve demand. charging pump redundancy of (open to closed position HHSI/charging pump 2CHS*LCV115C suction isolation from providing VCT change) at MCB. Valve suction valve (2CHS*LCV115E VCT. isolation. No effect close position monitor light 2CHS*LCV115B analogous) on system operation for group monitoring of (2CHS*LCV115D) from as redundant isolation valve components at MCB. RWST. Valve closes on 2CHS*LCV115E actuation by SI signal (2CHS*LCV115C) after valve provides tank 2CHS*LCV115B (2CHS-discharge isolation. LCV115D) reaches the full open position.

2. Motor-operated a. Fails to open Provides HHSI/ a. Failure reduces Valve position indication a. Valve is interlocked gate valve on demand. charging pump redundancy of (closed to open position with VCT level 2CHS*LCV115B suction isolation from providing a flow path change) at MCB. Valve instrumentation.

(2CHS*LCV115D RWST. from RWST to suction open position monitor light Valve opens upon analogous) of HHSI/charging for group monitoring of actuation by a SI pumps. Additionally, components at MCB. signal or by a VCT permissive for VCT low-low level signal.

isolation valve 2CHS*LCV115C (2CHS*LCV115E) to close will not be generated. No effect on system operation as a redundant RWST to HHSI/charging pump suction path will be provided by valve 2CHS*LCV115D (2CHS*LCV115B).

VCT isolation will be provided by re-1 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks dundant isolation valve 2CHS*LCV115E (2CHS*LCV115C).

See item 1.

b. Once open, b. Failure reduces b. Valve is automatically fails to close on redundancy of isolating closed during the demand. the RWST during sump switchover from recirculation following a injection to LOCA. No effect on recirculation following a safety for system LOCA.

operation. Backflow of radioactive fluid into the RWST is precluded by a check valve.

3. HHSI/charging Fails to deliver Provides high Failure reduces HHSI/charging pump One HHSI/charging pump pump 2CHS*P21A working fluid. pressure injection redundancy of providing discharge header flow is used for normal charging (pumps flow during injection and fluid flow to RCS. (FI-943) at MCB. Open of RCS during plant 2CHS*P21B, C recirculation phases. Redundant HHSI/charging pump switchgear circuit operation. Pump circuit analogous) pump will provide minimum breaker indication at MCB. breaker is aligned to close flow requirements. Circuit breaker close on actuation by a SI signal.

position monitor light for group monitoring of components at MCB.

2 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

4. Intentionally Deleted

5. Motor-operated Fails to close on Provides isolation of Failure reduces Same methods of Valve is aligned to close gate valve demand. normal charging line redundancy of isolating detection as stated in upon actuation by a SI 2CHS*MOV289 during the injection and HHSI/charging pump item 1. signal.

(2CHS*MOV310 recirculation phases. discharge to normal analogous) charging line. No effect on safety for system operation. Isolation will be provided by alternate isolation valve 2CHS*MOV310 (2CHS*MOV289).

6. Motor-operated a. Fails to open Provides HHSI flow a. Failure reduces Same methods of a. Valve is aligned to gate valve on demand. path to RCS for both redundancy of detection as those stated open upon actuation by 2SIS*MOV867A HHSI/charging pumps providing HHSI flow for Item 2. In addition, a SI signal.

(2SIS*MOV867B during injection and for paths. No effect on valve open position alarm analogous) HHSI/charging pump safety for system for group monitoring of during cold leg operation. Alternate components at MCB.

recirculation. path will be provided by valve 2SIS*MOV867B (2SIS*MOV867A).

3 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

b. Once open, b. Failure reduces b. Valve is closed by the fails to close on redundancy of isolating operator during the demand. HHSI flow to cold legs switchover from cold during switchover to leg to hot leg hot leg recirculation. recirculation.

No effect for system operation. Redundant valves 2SIS*MOV867C,D will provide isolation capability. Should failure be caused by loss of a Class 1E bus, core flushing would be accomplished by the operable RSS pump and HHSI/charging pump.

7. Motor-operated Same as item 6. Same as item 6. Same effect as item 6 Same methods of gate valve except for valve numbers. detection as those stated 2SIS*MOV867C for item 6.

(2SIS*MOV867D analogous)

8. Motor-operated a. Fails to close Controls LHSI pump a. Failure reduces fluid Valve position indication Valve is regulated by a gate valve on demand. miniflow during injection flow delivered to RCS (open to closed position signal from a flow 2SIS*MOV8890A phase. from LHSI pump change) at MCB. The transmitter located in the (2SISMOV8890B 2SIS*P21A LHSI pump discharge flow pump discharge header.

analogous) (2SIS*P21B). indication, FI-946 (FI-945) The control valve opens Minimum flow at MCB. when the LHSI pump requirements for LHSI discharge flow is less than will be met by the a low flow set point and redundant LHSI pump. closes when the flow 4 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

b. Fails to open b. Failure results in exceeds a high flow set on demand. insufficient LHSI pump point coincident with LHSI flow for a small LOCA pump running.

or steam line break, resulting in possible pump damage. Should the pump become inoperable, adequate LHSI flow will be provided by the redundant LHSI pump.

9. LHSI pump Fails to deliver Provides LHSI flow Failure reduces LHSI pump discharge Pump circuit breaker is 2SIS*P21A working fluid. during injection phase. redundancy of providing flow, FI-946 (FI-945) at aligned to close on (2SIS*P21B emergency coolant to the MCB. Open pump actuation by a SI signal.

analogous) RCS from the RWST at low switchgear circuit breaker pressure. Minimum flow indication at MCB. Circuit requirements will be met by breaker close position the redundant LHSI pump. monitor light for group monitoring of components at MCB. Common breaker trip alarm at MCB.

10. Motor-operated Fails to open on Provides LHSI flow path Failure reduces Same methods of Valve is aligned to open on gate valve demand. from RSS pump redundancy of providing detection as those stated actuation by RWST 2SIS*MOV8811A 2RSS*P21C emergency coolant during for item 6. In addition, extreme low level signal (2SIS*MOV8811B (2RSS*P21D) to RCS recirculation. Minimum failure may be detected coincident with a SI signal.

analogous) during recirculation flow requirements will be through monitoring of phase. provided by the redundant LHSI pump discharge RSS pump 2RSS*P21D flow, FI-946 (FI-945) at (2RSS*P21C) through MCB.

valve 2SIS*MOV8811B (2SIS*MOV8811A).

5 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

11. Motor-operated Fails to open on Provides flow from RSS Failure reduces Same methods of Valve is aligned to gate valve demand. pump 2RSS*P21C redundancy of providing detection as those stated open on actuation 2SIS*MOV863A (2RSS*P21D) to suction flow from RSS pump to for item 6. by position signal (2SIS*MOV863B of HHSI/charging pump suction of HHSI/charging open from valve analogous) 2CHS*P21A pumps. No effect on safety 2SIS*MOV8811A (2CHS*P21B) during for system operation. (2SIS*MOV8811B).

recirculation. Redundant valve 2SIS*MOV863B (2SIS*MOV863A) opens to provide flow path to suction of HHSI/charging pumps.

12. Motor-operated a. Fails to close Provides LHSI train a. Failure reduces Valve position indication a. Valve is gate valve on demand. separation during cold redundancy of (open to closed position aligned to close 2SIS*MOV8887A leg recirculation. providing LHSI train change) at MCB. on actuation by (2SIS*MOV8887B separation. No effect a RWST analogous) on safety for system extreme low operation. Redundant level signal valve 2SIS*MOV8887B coincident with (2SIS*MOV8887A) a SI signal.

closes to provide separation.

b. Once closed, b. Failure reduces b. Valve is opened fails to open on redundancy of by the operator demand. providing LHSI flow to during the RCS hot legs for core switchover from flushing. No effect on cold leg to hot safety for system leg recirculation.

operation. The LHSI flow to RCS hot legs will be provided through redundant valve 2SIS*MOV8887B (2SIS*MOV8887A).

6 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

13. Motor-operated Fails to close on Provides isolation of Failure reduces Same methods of Valve is automatically gate valve demand. LHSI pump 2SIS*P21A redundancy of isolating the detection as stated for closed during the 2SIS*MOV8809A (2SIS*P21B) from RWST during the item 4. switchover from injection to (2SIS*MOV8809B RWST during recirculation phase recirculation following a analogous) recirculation. following a LOCA. No LOCA.

effect on safety for system operation. Back flow of radioactive fluid into the RWST is precluded by a check valve.

14. Motor-operated Fails to close on Provides HHSI train Failure reduces Same methods of Valve is closed by the gate valve demand. separation during redundancy of separating detection as stated for operator during the 2CHS*MOV8130A recirculation. HHSI trains. No effect on item 4. switchover from injection to (2CHS*MOV8130B safety for system operation. cold leg recirculation analogous) Train separation will be following a LOCA. If HHSI provided by redundant pump C is operating, then isolation valve either 2CHS*MOV8130A 2CHS*MOV8130B (2CHS*MOV8130B) or (2CHS*MOV8130A) 2CHS*MOV8131A (2CHS*MOV8131B) will be closed, but not both.

15. Motor-operated Fails to close on Same as item 14. Same as item 14. Same methods of Same as item 14.

gate valve demand. detection as stated for 2CHS*MOV8131A item 4.

(2CHS*MOV8131B analogous) 7 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

16. Motor-operated Fails to close on Same as item 14. Same as item 14. Same methods of Valve is normally open with gate valve demand. detection as stated for power removed. Valve is 2CHS*MOV8132A item 4. closed by the operator (2CHS*MOV8132B during the switchover from analogous) injection to cold leg recirculation following a LOCA. If HHSI pump C is operating, then either 2CHS*MOV8132A (2CHS*MOV8132B) or 2CHS*MOV8133A (2CHS*MOV8133B) will be closed, but not both.

17. Motor-operated Fails to close on Same as item 14. Same as item 14. Same methods of Same as item 16.

gate valve demand. detection as stated for 2CHS*MOV8133A item 4.

(2CHS*MOV8133B analogous)

18. Motor-operated a. Fails to open Provides redundant a. Failure prevents Same methods of a. Valve is normally gate valve on demand. HHSI path to RCS cold separation of HHSI detection as stated for closed with power 2CHS*MOV836 legs during flow into two separate item 6. removed. Valve is recirculation. trains. The HHSI flow opened by the operator will be provided by during the switchover alternate path through from injection to cold valves leg recirculation.

2SIS*MOV867A,B, and 2SIS*MOV867C,D.

8 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

b. Once open, b. Failure reduces b. Valve is closed by the fails to close on redundancy of operator during the demand. providing HHSI flow to switchover from cold RCS hot legs for core leg to hot leg flushing. No effect on recirculation.

safety for system operation. The HHSI flow to RCS hot legs will be provided by HHSI/charging pump 2CHS*P21C through valve 2SIS*MOV869B.

19. Recirculation spray Fails to deliver Provides LHSI during Failure reduces Same as item 9 except for Pump is part of Pump 2RSS*P21C working fluid. recirculation phase. redundancy of providing flow indication FI-157C RSS and is aligned (2RSS*P21D LHSI flow directly to RCS (FI-157D). automatically during analogous) and to suction of switchover from HHSI/charging pump injection to 2CHS*P21A (2CHS*P21B). recirculation The LHSI flow to the RCS (Section 6.2.2).

and to the suction of both HHSI/charging pumps will be provided by the redundant pump 2RSS*P21D (2RSS*P21C).

20. Motor-operated Fails to open on Provides LHSI flow path Failure prevents use of Valve position indication Valve is normally gate valve demand. to RCS hot legs. LHSI flow for core flushing (closed to open position closed with power 2SIS*MOV8889 during hot leg recirculation. change) at MCB. Valve removed. Valve is Core flushing will be closed position monitor opened by the provided by HHSI/charging light and alarm for group operator during the pumps through valves monitoring of components switchover from 2SIS*MOV869A,B and by at MCB. cold leg to hot leg RSS pumps through valves recirculation.

2SIS*MOV8888A,B.

9 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

21. Motor-operated Fails to open on Provides HHSI flow Failure prevents use of Same methods of Same as item 20.

gate valve demand. path to RCS hot legs. HHSI/charging pump detection as stated for 2SIS*MOV869A 2CHS*P21A (2CHS*P21B) item 20.

(2SIS*MOV869B for core flushing during hot analogous) leg recirculation. Core flushing will be provided by the redundant HHSI/charging pump and LHSI flow from RSS pumps 2RSS*P21C,D.

22. Motor-operated Fails to close on Provides LHSI flow path Failure reduces Same methods of Valve is closed by gate valve demand. to RCS cold legs. redundancy of using LHSI detection as stated for the operator during 2SIS*MOV8888A flow for core flushing during item 4. the switchover from (2SIS*MOV8888B hot leg recirculation. Core cold leg to hot leg analogous) flushing will be provided by recirculation.

redundant LHSI path through valve 2SIS*MOV8887B (2SIS*MOV8887A) and HHSI flow through 2SIS*MOV869A and B.

10 of 11

BVPS-2 UFSAR Rev. 26 TABLE 6.3-5 (Cont)

Effect On Failure Detection Component Failure Mode Function System Operation Method Remarks

23. Motor-operated Fails to close on Provides RSS pump Failure reduces LHSI flow Same method of detection Valve closes on gate valve demand. flow path to RSS to RCS cold legs and to as stated for item 4. actuation by RWST 2RSS*MOV156C header. suction of HHSI/charging extreme low level (2RSS*MOV156D pump 2CHS*P21A coincident with an analogous) (2CHS*P21B). Minimum SI signal.

flow requirements will be provided by redundant LHSI train.

List of Abbreviations and Acronyms HHSI - High head safety injection RWST - Refueling water storage tank LHSI - Low head safety injection SI - Safety Injection LOCA - Loss-of-coolant accident VCT - Volume control tank MCB - Main control board RCS - Reactor coolant system RSS - Recirculation spray system 11 of 11

BVPS-2 UFSAR Rev. 0 TABLE 6.3-6 EMERGENCY CORE COOLING SYSTEM RECIRCULATION PIPING PASSIVE FAILURE ANALYSIS LONG TERM PHASE Indication of Flow Path Loss of Flow Path Alternate Flow Path Low Head Recirculation From recir- Accumulation of Via the independent, culation spray water in a recir- identical low head flow to low head culation spray path utilizing the se-injection pump compartment cond recirculation spray header or auxiliary pump building sump High Head Recirculation From con- Accumulation of water From containment sump to tainment in a recirculation the high head injection sump to the spray pump compart- headers via alternate high head ment, charging pump recirculation spray injection compartment, or the header auxiliary building sump 1 of 1

BVPS-2 UFSAR Rev. 26 TABLE 6.3-7 SEQUENCE OF SWITCHOVER OPERATION FROM INJECTION TO RECIRCULATION The injection mode of ECCS operation is initiated automatically and requires no operator action. During injection the LHSI and HHSI/charging pumps take suction from the RWST and deliver borated water to the cold legs. The accumulators discharge into the cold legs as RCS pressure drops below their set pressure.

Upon a CIB signal coincident with an RWST low level signal (as defined in Section 6.2.2.2.2), the recirculation spray pumps start automatically and provide recirculation of the containment sump water through the recirculation coolers to the recirculation spray headers.

During the switchover from injection to cold leg recirculation, two of the four recirculation spray pumps are automatically realigned to provide the LHSI pump function and to provide flow to the suction of the HHSI/charging pumps.

Switchover from injection to cold leg recirculation occurs automatically as the level in the RWST drops to the extreme low level set point. On RWST extreme low level coincident with a SI signal, the following actions occur automatically:

1. The LHSI crossover isolation valves close (2SIS*MOV8887A, B),

2. The recirculation spray pump discharge valves to the LHSI discharge lines into the RCS open (2SIS*MOV8811A, B),

3. The recirculation spray header isolation valves close (2RSS*MOV156C, D),

4. The HHSI/charging pump suction isolation valves from the LHSI pumps (2SIS*MOV863A, B) open on limit switch signal open from valves 2SIS*MOV8811A, B,

5. The LHSI pumps stop on limit switch signal open from valves 2SIS*MOV8811A, B,

6. The HHSI/charging pump suction isolation valves from the RWST (2CHS*LCV115B, D) close on limit switch signal open from valves 2SIS*MOV863A, B, and

7. The LHSI pump suction valves from the RWST (2SIS*MOV8809A, B) close on signal from LHSI pump stop.

Following these automatic actions, the operator performs the following manual actions:

1 of 2

BVPS-2 UFSAR Rev. 26 TABLE 6.3-7 (Cont)

1. Opens the alternate HHSI path isolation valve (2SIS*MOV836),

and

2. Separates the two HHSI/charging subsystems by closing the appropriate isolation valves in the HHSI/charging pump suction header (either 2CHS*MOV8130A, B or 2CHS*MOV8131A, B) and the HHSI/charging pump discharge header (either 2CHS*MOV8132A, B or 2CHS*MOV8133A, B).

The ECCS is now aligned for cold leg recirculation. Two of the four recirculation spray pumps provide LHSI into the cold legs while also providing suction to the HHSI/charging pumps. Two separate subsystems are provided, each consisting of a recirculation spray pump and a HHSI/charging pump.

At approximately 6.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> following the LOCA, the operator would realign the ECCS for hot leg recirculation by the following actions:

1. Closes the LHSI discharge isolation valves to the cold legs (2SIS*MOV8888A, B),

2. Opens the LHSI crossover isolation valves (2SIS*MOV8887A, B),

3. Opens the isolation valve in the common LHSI discharge line to the hot legs (2SIS*MOV8889) (during the switchover of the HHSI/charging pumps to hot leg recirculation, the HHSI/charging pumps are stopped individually while respective isolation valves are realigned),

4. After HHSI/charging pump A is stopped, the isolation valve to the cold legs is closed (2SIS*MOV836),

5. The corresponding isolation valve to the hot legs is opened (2SIS*MOV869A), and

6. The HHSI/charging pump is restarted.

The other HHSI/charging pump is then realigned in the same manner.

Both cold leg isolation valves (either 2SIS*MOV867A, B or 2SIS*MOV867C, D) are closed and hot leg isolation valve (2SIS*MOV869B) is opened.

The ECCS is now aligned for hot leg recirculation with two recirculation spray pumps and two HHSI/charging pumps providing flow to the hot legs.

2 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.3-8 ECCS SHARED FUNCTIONS EVALUATION Normal Operating Accident Component Arrangement Arrangement Refueling water Lined up to suction of Lined up to suction of storage tank LHSI pumps HHSI/charging and LHSI pumps HHSI/charging Lined up for charging Line up for HHSI.

pumps service Valves for realignment meet single failure criteria Recirculation Lined up to provide Lined up to the spray pumps flow from the recirculation spray containment sump to headers in the short the recirculation term (two of the four spray headers pumps are aligned to the LHSI header) and to the HHSI/charging pump suction in the recirculation phase Low head safety Lined up to cold legs Lined up to cold legs injection pumps of reactor coolant of reactor coolant piping piping 1 of 1

BVPS-2 UFSAR Rev. 13 TABLE 6.3-9 NORMAL OPERATING STATUS OF EMERGENCY CORE COOLING SYSTEM COMPONENTS FOR CORE COOLING Number of LHSI pumps operable 2 Number of HHSI charging pumps operable 2 Refueling water storage tank minimum useable volume (gal) 859,248 Boron concentration in RWST (minimum ppm) 2,400 Boron concentration in accumulators (nominal ppm) 1,900 Number of accumulators 3 Minimum accumulator pressure(psig) 600 Nominal accumulator water volume (ft 3) 925 System valves, interlocks, and piping required All for the above components which are operable 1 of 1

TO RCS REV. 12 FILL HEADER OUTS IDE INSIDE 1---_.,--l TO CHARGING CONTAINMENT CONTAINMENT HEADER TO PZR 2CHS AUXILIARY TK21 ( A/8 l SPRAY FROM RCP EAL LEAKOFF REGENERATIVE HEAT EXCHANGER TANK 2CHS-E23 REFUEL! NG 20SS WATER TK21 STORAGE 123 LO

"-=y--'

FROM RHR AND SI ACCUMULATORS ALL DAMPER, VALVE AND EQUIPMENT IDENTIFICATION NUMBERS ON THIS FlGURE ARE PRECEDED BY THE SYSTEM DESIGNATOR u2SISn UNLESS OTHERWISE INDICATED.

P218 LOW HEAD SAFETY FROM RECIRCULATION INJECTION PUMPS SPRAY SYSTEM FIGURE 6.3-1 SAFETY INJECTION SYSTEM

REFERENCE:

STATION DRAWINGS OM 6-1.

7-1A. 7-1B. 7-2. 7-3, 11-1 AND 13-2 BEAVER VALLEY POWER STATION UNIT NO. 2 UPDATED FfNAL SAFETY ANALYSIS REPORT

REV. 12 FROM FROM 1---~--1 NITROGEN RWST SYSTEM INSIDE CONTAINMENT VENT TO CONTAINMENT ATMOSPHERE TO TO TO FLOOR FLOOR FLOOR DRAINS DRAINS DRAINS SAFETY SAFETY SAFETY INJECTION INJECTION INJECTION ACCUMULATOR ACCUMULATOR ACCUMULATOR A B c TO VENTS TO TO VENTS TO TO VENTS TO AND DRAINS SAMPLE AND DRAINS SAMPLE AND DRAINS SAMPLE SYSTEM SYSTEM SYSTEM SYSTEM SYSTEM SYSTEM RHR RHR RETURN RETURN TO RCS TO RCS TO RCS LOOP A LOOP B LOOP C COLD LEG COLD LEG COLO LEG TO RWST ALL VALVE AND EQUIPMENT IDENTIFICATION NUMBERS ON THIS FIGURE ARE PRECEDED BY THE SYSTEM DESIGNATOR "2SIS" UNLESS

<(

OTHERWISE INDICATED.

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u w ~ FIGURE 6.3-2 SAFETY INJECTION ACCUMULATORS Cl 1./) RCS REACTOR COOLANT SYSTEM z RWST REFUELING WATER STORAGE REFERENCEI STATION DRAWING OM 11-2 TANK BEAVER VALLEY POWER STATION UNIT NO. 2 UPDATED FINAL SAFETY ANALYSIS REPORT

1000 100 400 EFFICIENCY 900 90 360 800 80 320 HEAO-CAPACITY 700 70 280 a:

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100 10 40 0 0 0 0 1000 2000 3000 4000 5000 6000 GALLONS PER MINUTE FIGURE 6.3-4 PUMP HEAD CHARACTERISTIC CURVE- LOW HEAD-SAFETY INJECT ION PUMPS BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

Figure 6.3-5 (Sheet 1 of 3) Rev. 16 Pump Head Characteristic Curve High Head-Centrifugal Charging Pump Beaver Valley Power Station-Unit 2 Final Safety Analysis Report 7000 70 6000 60 D.

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Figure 6.3-5 (Sheet 2 of 3) Rev. 16 Pump Head Characteristic Curve High Head-Centrifugal Charging Pump Beaver Valley Power Station-Unit 2 Final Safety Analysis Report aooo~--~~--~~--~--~----~--~~~~~~~--~----~--~----~--~----~--~ao 7000 70 6000 60 a..

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Figure 6.3-5 (Sheet 3 of 3) Rev. 16 Pump Head Characteristic Curve High Head-Centrifugal Charging Pump Beaver Valley Power Station-Unit 2 Final Safety Analysis Repor1

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Rev. 12 IIC aJI.D LS UllP I LHSI PlW j UtSI PI.M' z IRC FIGURE 6.3-6 { SH. t OF 9 )

PROCESS FLOW DIAGRAM SAFETY INJECTION SYSTEM BEAVER VALLEY POWER STAT JON -UNIT 2 FINAL SAFETY ANALYSIS REPORT

Rev 12 fl.: LQlP I O.l..D LEG

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PROCESS FLOW DIAGRAM SAFETY iNJECTION SYSTEM BEAVER VALLEY POWER STATION -UNIT 2 FINAL SAFETY ANALYSIS REPORT

VOH Rev. 12 ACC.

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PROCESS FLOW DIAGRAM SAFETY INJECTION SYSTEM BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANAL YSlS REPORT

BVPS-2 UFSAR Rev. 12 NOTES TO FIGURE 6.3-6 The following general assumptions were utilized to develop the system modes of operation:

1. The process flow diagrams are provided for illustrative purposes only and are not intended to represent the flow rates used in the various accident analyses. Flow rates to the RCS are provided in Chapter 15, where appropriate. The process flow diagrams are developed assuming typical pump curves and balanced system resistances. The flow rates for FSAR accident analyses are developed using either maximum design pump curves enhanced by 10 percent or minimum design pump curves degraded by 5 percent (whichever is appropriate) and worst case assumptions pertaining to design system resistances (for example, maximum allowable resistances in lines connected to unbroken loops and m~n~mum allowable resistances in lines connected to the broken loop).

2. The system operating conditions presented for the injection and recirculation modes are all based on the assumption of the RCS being fully depressurized and in equilibrium with the containment at 0 psig.

3. The accumulator delivery is considered as an independent mode of operation and the process conditions presented are based on the assumption that the accumulators are fully discharged and depressurized at 0 psig.

4 of 9

BVPS-2 UFSAR Rev. 12 NOTES TO FIGURE 6.3-6 (Cont)

Modes of Operation Mode A: Normal Standby This mode p~esents the process conditions for the case of normal ECCS standby.

Mode B: Injection/Maximum Safeguards This mod~ repr-esents the c.s.se of way__irn1Jll'l s~f'~gu.arric:. t.1here all safeguards pumps operate following accumulator delivery. Two LHSI pumps and two HHSI/charging pumps operate, taking suction from the RWST and delivering to the reactor through three cold leg connections.

Mode C: Injection/Minimum Safeguards - Train A Operating This mode represents the process conditions for the case of minimum safeguards with LHSI pump C and HHSI/charging pump A taking suction from the RWST and delivering to the reactor through three cold leg connections.

Mode D: Injection/Minimum Safeguards - Train B Operating This mode represents the case of minimum safeguards with LHSI pump B and HHSI/charging pump C taking suction from the RWST and delivering to the reactor through three cold leg connections.

This mode of operation is similar to that of Mode C.

Mode E: Cold Leg Recirculation/Maximum Safeguards This mode represents the case of cold leg recirculation with recirculation spray pumps C and D, and HHSI/charging pumps A and C operating.

In this mode the safeguards pumps operate in series, with only the recirculation spray pumps capable of taking suction from the containment sump. The recirculated coolant is then delivered by the recirculation spray pumps through the recirculation spray coolers to the HHSI/charginq pumps, which deliver to the reactor through three cold leg connections.

The recirculation spray pumps also deliver flow directly to the reactor through the same three cold leg connections.

Mode F: Cold Leg Recirculation/Minimum Safeguards - Train A Operating.

This mode represents the case of cold leg recirculation with recirculation spray pump C and HHSI/charging pump A operating.

5 of 9

BVPS-2 UFSAR Rev. 12 NOTES TO FIGURE 6.3~6 (Cent)

In this mode the safeguards pumps operate in series, with only the recirculation spray pump capable of taking suction from the containment sump. The recirculated coolant is then delivered by the recirculation spray pump through a recirculation spray cooler to the HHSI/charging pump, which delivers to the reactor through three cold leg connections. ~

The recirculation spray pump also delivers flow directly to the reactor through the same three cold leg connections.

Mode G: Cold Leg Recirculation/Minimum Safeguards - Train B Operating This mode represents the case of cold leg recirculation with recirculation spray pump D and HHSI/charging pump C operating. As such, this mode of operation is similar to that of Mode F.

Mode H; Hot Leg Recirculation/Maximum Safeguards This mode represents the case of hot leg recirculation with recirculation spray pumps c and D and HHSI/charging pumps A and C operating.

In this mode, the safeguards pumps again operate in series with the recirculation spray pumps taking suction from the containment sump. The recirculated coolant is then delivered by the recirculation spray pumps through the recirculation spray coolers to the HHSI/charging pumps and which deliver to the reactor through three hot leg connections. The recirculation spray pumps also deliver directly to the reactor through two hot leg connections.

Mode I: Hot Leg Recirculation/Minimum Safeguards

• Train A Operating This mode represents the case of hot leg recirculation with recirculation spray pump C and HHSI/charging pump A operating.

In this mode, the safeguards pumps again operate in series with only the recirculation spray pump taking suction from the containment sump. The recirculated coolant is then delivered by the recirculation spray pump through a recirculation spray cooler to the HHSI/charging pump, which delivers to the reactor through three hot leg connections. The recirculation spray pump also delivers directly to the reactor through two hot leg connections.

Mode J: Hot Leg Recirculation/Minimum Safeguards - Train B Operating This mode represents the case of hot leg recirculation with recirculation spray pump D and HHSI/charging pump C operating. As such, this mode of operation is similar to that of Mode I.

6 of 9

- I BVPS-2 UFSAR Rev. 12 NOTES TO FIGURE 6.3-6 (Cont)

Valve Alignment Chart Valve No. c

-- -A B

- D E F G H I

-J lA O* 0 0 0 c c 0 c c 0 lB 0 0 0 0 c 0 c c 0 c 2A C** c c c c c c c c c 2B c c c c c c c c c c 3A c c c c 0 0 c 0 0 c 3B c c c c 0 c 0 0 c 0 4A c c c c 0 0 c 0 0 c 4B c c c c 0 c 0 0 c 0 SA 0 0 0 0 c c 0 0 0 0 SB 0 0 0 0 c 0 c 0 0 0 6A 0 0 0 0 0 0 0 c c c 6B 0 0 0 0 0 0 0 c c c 7 c c c c c c c 0 0 0 SA 0 c c 0 c c 0 c c 0 88 0 c 0 c c 0 c c 0 c 9A c 0 0 c c c c c c c 9B c 0 c 0 c c c c c c lOA 0 0 0 0 c 0 0 c 0 0 lOB 0 0 0 0 c 0 0 c 0 0 llA 0 0 0 0 c 0 0 c 0 0 llB 0 0 0 0 c 0 0 c 0 0 12A 0 c c 0 c c 0 c c 0 12B 0 c c 0 c c 0 c c 0 12C 0 c c 0 c c 0 c c 0 13 0 c 0 c c 0 c c 0 c 14A 0 c c 0 c c 0 c c 0 148 0 c 0 c c 0 c c 0 c lSA 0 0 0 0 c 0 0 c 0 0 158 0 0 0 0 c 0 0 c 0 0 l6A 0 0 0 0 c 0 0 c 0 0 16B 0 0 0 0 c 0 0 c 0 0 17A c 0 0 c 0 0 c c c c 17B c 0 c 0 0 c 0 c c c 18A c 0 0 c 0 0 c c c c 18B c 0 c 0 0 c 0 c c c 19 c c c c c c c 0 c 0 20A c c c c 0 c c c c c 20B c c c c c c c 0 0 c 21A 0 0 0 0 0 0 0 0 0 0 21B 0 0 0 0 0 0 0 0 0 0 21C 0 0 0 0 0 0 0 0 0 0 7 of 9

BVPS-2 UFSAR Rev. 12 NOTES TO FIGURE 6.3~6 (Cent)

HODE C: Injection Phase 1 Minimum Safeguards ~ Train A Operating (Runout Conditions Following Accumulator Delivery)

Pressure Temperature Flow Location Fluid (psig) (OF) (gpm)*** (lb/hr)**** Volume 1 Refueling Water atm. 55 - 50,000 gal.

2 II 25 55 4,223 2.11 3 It 55 535 0.268 4 II 55 3,688 1.85 SA 55 3,688 1.85 It SB II 55 0 0 6A If 91 55 0 0 68 55 II 0 0 7 II 55 0 0 91 55 3,688 1.85 SA II 88 II 55 0 Q_

9A II 91 55 0 0 9B II 54 55 0 0 lOA It 54 55 0 0 lOB II 54 55 0 0 llA II 52 55 1,982 0.992 0.854 llB

. 52 55 1,706 II 12 54 55 0 0 13A If 0 55 1,272 0.637 13B II 0 55 1,404 0.750 13C IJ 0 55 1,498 0.703 l4A Refueling 23 55 0 0 Water-148 II 29 55 0 0 14C II 31 55 0 0 lSA II 23 55 162 0.081 lSB II 29 55 161 0.081 lSC II 31 55 162 0.081 16A II 0 55 0 0 16C II 0 55 0 0 17A It 0 55 0 0 l7C II 0 55 0 0 18A II 0 55 0 0 18C It 0 55 0 0 19A II 0 55 535 0.268 198 II 55 0 0 20 II 55 0 0 21A II 0 55 535 0.268 218 II 55 0 0 21C II 55 0 0 22 II 55 0 0 8 of 9

BVPS~ 2 UFSAR Rev. 12 NOTES TO FIGURE 6.3-6 (Cent)

Pressure Temperature Flow Location Fluid (psig} (OF) {gpm)*** (lb/hr)**** Vol tune

-~

23A II 841 55 535 0-268 23B Refueling 838 55 0 0 Water 23C II 833 55 0 0 24 II 841 55 0 0 25 II 833 55 so 0.025 26! II 0 55 n 0 26B II 0 55 0 0 27 II 833 55 485 0.243 900 gal.

28A Refueling 0 55 0 0 Water 288 II 617 55 485 0.243 29A N2 0 100 0 0 1,450 ft 3 298 N2 0 100 0 0 1,450 ft 3 29C Nz 0 100 0 0 1,450 ft3 30A N2 0 100 0 0 308 Nz 0 100 0 0

":!""'* - N2 0 100 0 0 NOTES:

• O = Open

• • C = Closed

• • • At reference conditions 55°F and 0 psig

• • • • x 10 5 9 of 9

BVPS-2 UFSAR Rev. 26 6.4 HABITABILITY SYSTEMS The habitability system for the control room envelope encompasses equipment and supplies to ensure that the main control room operators are able to remain in the area and take action to operate Beaver Valley Power Station - Unit 2 (BVPS-2) safely under normal conditions, as well as during all postulated design basis accidents (DBA). The habitability systems include radiation shielding, redundant radiation monitors with automatic control room isolation capability, redundant air supply and filtration systems, redundant air-conditioning systems, fire protection, personnel protective and first aid equipment, food and water storage, emergency lighting, utility and sanitary facilities.

For detailed descriptions of the individual systems, refer to specific sections of this Final Safety Analysis Report.

6.4.1 Design Bases The habitability system of the control room is designed in accordance with the following criteria:

1. General Design Criterion 2, as it relates to the ability of structures housing the facility and the facility components to withstand the effects of natural phenomena such as earthquakes, hurricanes, and floods, as established in Chapters 2 and 3.

2. General Design Criterion 3, as it relates to protection against fire hazards.

3. General Design Criterion 4, with respect to structures housing the facility and the facility components being capable of withstanding the effects of external missiles and internally-generated missiles, pipe whip, and jet impingement forces associated with pipe breaks, such that safety functions will not be precluded.

4. General Design Criterion 5, as it relates to shared systems and components important to safety being capable of performing required safety functions.

5. General Design Criterion 19, and 10 CFR 50.67 as it relates to providing adequate radiation protection during LOCA, CREA, and FHA accident conditions.

6. Regulatory Guide 1.26, as it relates to the quality group classification of system components.

7. Regulatory Guide 1.29, as it relates to the seismic design classification of systems and components.

8. Regulatory Guide 1.52, regarding air filtration equipment requirements.

6.4-1

BVPS-2 UFSAR Rev. 26

9. Regulatory Guide 1.76, regarding design basis tornado for nuclear power plants.

10. Regulatory Guide 1.78, regarding assumptions for evaluating the habitability of the main control room following postulated chemical releases.

11. Regulatory Guide 1.95, regarding protection of main control room operators from a postulated chlorine release.

12. Regulatory Guide 1.117, regarding tornado design classification.

13. Meets the intent of NUREG-0737, Action Item III.D.3.4 (USNRC 1980) as it relates to control room habitability.

6.4.2 System Design 6.4.2.1 Definition of Control Room Envelope The Beaver Valley Power Station is served by a single control room that supports both units. Each unit has a separate control area in the control room. The two control areas are contained in a single control room envelope, and are modeled as a single region. The control room envelope is defined in the Control Room Envelope Habitability Program. Table 6.4-1 compares the ventilation parameters between BVPS-1 and BVPS-2.

Operators must occupy the main control room continually and the computer room frequently. The mechanical equipment rooms are seldom occupied but, due to their configuration and function, they are also part of the envelope.

6.4.2.2 Ventilation System Design The control room air-conditioning system is safety-related and maintains the main control room ambient air temperature, under normal conditions, at 75F. Detailed system and component information is provided in Section 9.4.1. Figure 3.8-25 shows the layout of the control room.

The emergency diesel generators supply power to emergency lighting and electrically-powered motors and controls associated with redundant air-conditioning and filtration systems in the event of a loss of normal power. In addition, the emergency 125 V dc system (Section 8.3.2) provides power for emergency lighting in the main control room. All spaces served by these two air-conditioning and filtration units form the control room envelope and are shown on Figure 9.4-1.

Fresh air is supplied to the control room envelope during normal plant operation. Two redundant emergency supply filtration units are provided to filter the fresh air for breathing and pressurization after an accident. Each of the emergency supply filtration units has a capacity of 1,000 cfm.

6.4-2

BVPS-2 UFSAR Rev. 26 The design, testing, and maintenance of the control room emergency supply filtration unit is in accordance with Regulatory Guide 1.52, with the exception of those items discussed in Section 1.8. When the main control room must be isolated manually (for example, due to smoke) or automatically (for example, due to a loss-of-coolant accident (LOCA), or high radiation levels in the control room), the supply air isolation butterfly valves in the outdoor air intake ducts close. The locations of the BVPS-l and BVPS-2 control room air intakes are shown on Table 6.4-3 and Figure 6.4-5. Hazardous materials released from accidents offsite are identified in Section 2.2.3. The intent of NUREG-0737, Action Item III.D.3.4, is met in accordance with the control room habitability study for BVPS-1 and BVPS-2 for hazardous chemical releases.

Following a containment isolation Phase B signal or a high radiation signal, about 1,000 cfm of outdoor air will be taken through the air intake and supplied to the control room envelope through one of the redundant emergency supply filtration units. Introduction of CO 2 from the CO2 storage facility is not possible since the facility is not installed within the control room envelope. Sufficient ventilation is provided to prevent a buildup of noxious gases from batteries installed within the control room envelope.

6.4.2.3 Leaktightness The control room emergency ventilation system is designed to provide a positive pressure in the control room envelope using filtered outside air to minimize unfiltered inleakage during emergency operation through doors, ducts, pipes, and cable penetrations that could be caused by wind effects and pressure variations. Special construction features are provided to maintain the leaktightness of the control room envelope including compression seals for access doors and equipment removal hatches, penetration seals for pipes, ducts, and electrical penetrations, and water trap seals for sanitary piping.

Control room envelope unfiltered air inleakage testing is performed in accordance with the Control Room Envelope Habitability Program.

In addition to meeting the requirements related to radiation, the following design features are provided:

1. Centrally located redundant Category I area radiation monitors.

2. Automatic main control room isolation on radiation detection.

3. Construction details to control main control room leakage agree with recommended practices (Atomics International 1965).

Concrete and concrete block surfaces are coated with a surface treatment to reduce leakage due to porosity, cracks, and construction joints.

6.4.2.4 Interaction with Other Zones and Pressure-Containing Equipment 6.4-3

BVPS-2 UFSAR Rev. 26 The air-conditioning units for the control room envelope do not supply air to any other area. The return air in this system is only from the control room envelope area.

With the exception of the portable fire extinguishers, service water piping, refrigerant piping, and potable water system, there are no pressurized tanks, steam or hot water lines, or other pressurized pipes in the control room envelope. The seismically designed hose stations serving the control room envelope are located in the stairways on the next floor below and outside the envelope. There are no carbon dioxide supply pipes to the control room envelope. A 24-ton CO2 storage tank is located outside the west wall of the Auxiliary Building approximately 100 ft. from the BV-2 Control Room Air Intake. Failure of this tank has been analyzed and it has been determined that control room habitability will be maintained.

6.4.2.5 Shielding Design The design of the control room envelope includes adequate radiation shielding and ventilation control to maintain acceptable radiation levels in the main control room under accident conditions, as discussed in Section 12.3.2.

In accordance with General Design Criterion 19, personnel exposure is limited, for the duration of any accident postulated in Chapter 15, to 5 Rem TEDE, per 10 CFR 50.67 and R.G. 1.183.

The postulated accident radioactivity sources inside and outside the control room envelope are stated in Chapter 15.

The effects of the LOCA and all other design basis accidents are evaluated to determine the doses which the main control room personnel might receive at BVPS-2. The LOCA analysis is based on:

1. Major reactor coolant system (RCS) pipe rupture (LOCA) at BVPS-2 or a

2. Major RCS pipe rupture (LOCA) at BVPS-1.

For purposes of analysis, it is assumed that each accident occurs with and without a loss of offsite power. Accidents are not postulated to occur simultaneously.

The main control room personnel are potentially exposed to sources from several locations following the LOCA. The sources considered for the design of control room shielding include: 1) the containment building (direct and sky shine dose), 2) the external cloud (from containment and emergency core cooling system (ECCS) leakage), 3) sources in adjacent buildings, and 4) iodine collection on the main control room intake filter, and 5) the refueling water storage tank (RWST).

The containment building is considered as one of the sources of radiation used for main control room shielding design due to its location and the large amount of activity contained within its bounds. A significant fraction of the containment free air volume is located above grade and its 6.4-4

BVPS-2 UFSAR Rev. 26 dose contribution is evaluated to determine the main control room 30 day dose. The containment shine source is based on the time-dependent airborne activity inside containment released during the fuel gap and reactor in-vessel release periods, as described in Section 15.6.5.4.

The external cloud is due to containment leakage during the LOCA plus the ECCS and RWST leakage over 30 days. The containment leakage contribution to the cloud source is a function of the containment airborne inventory available for leakage and the containment leak rate of 0.10 percent of the containment air weight per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 0.05 percent per day for the remaining duration of the accident. The airborne inventory available for leakage and the leakage pathways are described in Section 15.6.5.4.

The ECCS leakage and RWST back-leakage contribution to the control room dose is based on containment sump water containing all non-gaseous activity released during the fuel gap and reactor in-vessel release phases, and at an assumed leakage rate of twice the maximum allowable ECCS leakage.

Two adjacent areas that potentially contain airborne sources due to leakage from the containment, the ECCS, or the RWST are considered contributors to the main control room dose due to shine. These source locations are the cable spreading area below the BVPS-2 control room, and the cable tray mezzanine below the BVPS-1 control room. The amount of radioactive iodine collected on the intake filters due to these leakage sources is considered in determining the direct shine dose in the control room. The direct shine dose from the BVPS-2 RWST is negligible, because it is shielded by the Unit 2 Containment building.

The main control room walls and adjacent structures provide shielding required to minimize exposure to operating personnel, as shown on Figures 6.4-2, 6.4-3 and 6.4-4, while a description of the shielding can be found in Section 12.3.2.9.

Figure 6.4-2 shows a cross-sectional view of the main control room and the relative distances to the BVPS-1 and BVPS-2 containments. Also shown are shield wall and floor thicknesses. Figures 6.4-3 and 6.4-4 show cross-sectional views from the BVPS-2 main control room facing BVPS-1 and BVPS-2 containments, respectively. Table 6.4-3 and Figure 6.4-5 identify the radiation release points for both BVPS-1 and BVPS-2.

6.4.3 System Operational Procedures The air-conditioning systems for the control room are in continuous operation and under automatic control. The operation of the units are controlled by thermostats which, by varying parameters of the refrigeration unit and operating the room reheat coils, maintain the main control room at 75F.

Except for the air distribution ductwork within each area, the air-conditioning system for the control room is segregated into two 100 percent capacity trains. The redundant standby equipment starts with the loss of static pressure across the operating unit.

6.4-5

BVPS-2 UFSAR Rev. 26 If static pressure is not lost and the temperature in the control room continues to rise above the normally maintained 75F (indicating loss of refrigerant supply to air-conditioning unit), a manual transfer of air-conditioning units is required.

If smoke is detected, isolation of control room ventilation is under administrative control. (Refer to Section 9.4.1.)

6.4.4 Design Evaluation The main control room air-conditioning system maintains a suitable environment for personnel occupancy and equipment operation during normal and emergency conditions. Two service water cooling coils in the return air stream are also provided as an additional back-up method of cooling the main control room area if required. Components of the air-conditioning systems are seismically designed and are housed in the Seismic Category I control building. Control room envelope ventilation parameters are shown in Table 6.4-1.

6.4.4.1 Radiological Protection Upon receipt of a containment isolation Phase B signal, or a high radiation signal from the control room area monitors, the normal outside air supply dampers automatically close, thus isolating the control room envelope.

This signal also initiates the control room emergency ventilation system (CREVS). The system is capable of maintaining the control room envelope ambient pressure slightly above atmospheric pressure, thereby limiting inleakage for an indefinite period of time. Periodic control room envelope unfiltered air inleakage tests are performed to confirm that the control room envelope is operable.

To provide operational margin, it is assumed that the joint BVPS-1 & BVPS-2 unfiltered intake plus inleakage during normal plant operation is a maximum of 1250 cfm.

Following the accident, the control room envelope is maintained at a positive pressure for an indefinite period of time due to the operation of the redundant emergency supply systems. Each system can draw outside air through an emergency supply filtration unit, which consists of a HEPA filter and carbon adsorber, with assumed effective iodine removal efficiencies of 99% for particulate aerosols and 98% for radioiodines.

These emergency supply filtration units and associated air handling equipment are designed to Seismic Category I and Safety Class 3 requirements.

Filtration of the Control Room ventilation system recirculation flow during all modes of operation by particulate air filters (intended for dust removal) is not credited in radiological dose consequence analyses.

The control room ventilation system recirculation flow may remain in service during accident conditions to maintain the control room within design temperature limits.

6.4-6

BVPS-2 UFSAR Rev. 26 Control room ventilation design parameters used for the LOCA, CREA, and MSLB analyses are presented in Table 6.4-1a. The atmospheric dispersion factors were calculated using ARCON96 for these accidents, and the factors are presented in Tables 15.0-14 and 15.0-15.

The information and data required to develop the radiological consequences for the main control room are presented in the respective sections describing the design basis accident analysis. Radiation doses to a control room operator due to the various postulated DBAs are summarized in Table 15.0-13. Exposure from inhalation is principally attributable to airborne radioactivity in the main control room envelope due to:

1. Intake prior to main control room isolation,

2. Inleakage during main control room isolation, prior to pressurization,

3. Post-pressurization ventilation filtered intake and unfiltered inleakage.

The CIB signal isolates the control room envelope almost immediately after a LOCA in either Unit. For CREA and MSLB, manual operator action by t=30 minutes post-accident is assumed when necessary to maintain habitability.

The analyses consider a conservative selection of parameters to calculate the accident dose. Ventilation intake prior to control room envelope isolation, and unfiltered inleakage are the main contributors to the dose.

The allowance for inleakage is based on the results of tracer gas testing and includes 10 cfm for ingress and egress.

The maximum normal ventilation intake rate (for both BVPS-1 and BVPS-2 intakes) prior to isolation and an appropriate clean up rate post-isolation are used to maximize the dose estimate. The CREVS filtered intake flow varies between 800 and 1000 cfm, including allowance for uncertainties. Sensitivity studies have shown that assuming the minimum intake flow is more limiting.

The analysis also assumes coincident loss of offsite power. Considering the time delay for startup and load sequencing on the emergency diesel generator and CREVS fan logic relay delays, a total auto start delay of 137 seconds was assumed in the analysis.

The main control room doses presented in Table 15.0-13 have been calculated to be less than the limit specified in General Design Criterion 19 and less than 5.0 Rem TEDE per 10 CFR 50.67. The main control room may, therefore, be safely occupied during any condition of operation.

6.4.4.2 Toxic Gas Protection The main control room design provides protection of the personnel in the main control room from any toxic effects from spills of chemicals stored onsite. The effects of spills of chemicals along transportation routes are evaluated in Section 2.2.3.

6.4-7

BVPS-2 UFSAR Rev. 26 Self-contained breathing apparatus units and sufficient reserve air cylinders are available to support the minimum control room shift composition for at least six hours. This satisfies Regulatory Guide 1.78 and 1.95. Sufficient additional units are provided to support the members of the emergency squad stationed outside the control room for one hour, after which these personnel would move away from the area affected by the toxic release. Air cylinders brought from off-site locations may be used to extend capacity beyond six hours.

The storage areas of toxic gases and chemicals that could produce toxic gases are shown in Table 6.4-3 and on Figure 6.4-5.

6.4.5 Inspection and Testing Requirements The major items of equipment that maintain the habitability of the main control room are the emergency supply filtration units, their fans, mechanical refrigeration units, air-conditioning units, and their control systems. The system is inspected, tested, and air balanced periodically.

Portions of the system are in continuous operation. Periodic operation of the standby equipment, in conjunction with routine observation and maintenance during normal operation, ensure system availability.

6.4.6 Instrumentation Requirements The instrumentation and controls included for main control room habitability are addressed in Section 9.4.1.5.

6.4.7 References for Section 6.4 American Nuclear Society (ANS) 1977. American National Standard Neutron and Gamma-Ray Flux-to-Dose-Rate Factors. ANSI/ANS-6.l.l-1977 (N666).

Atomics International 1965. Application data in: Conventional Buildings for Reactor Containment,Section IV. NAA-SR-10100.

DiNunno, J.J.; Anderson, F.D.; Baker, R.E.; Waterfield, R.L. 1962.

Calculation of Distance Factors for Power and Test Reactor Sites, U.S.

Atomic Energy Commission Technical Information Document TID-14844.

U.S. Nuclear Regulatory Commission (USNRC) 1980. Clarification of TMI Action Plan Requirements. NUREG-0737.

U.S. Nuclear Regulatory Commission 1982. Control Room Habitability Study for BVPS-1 and BVPS-2. Personal Communication between J.J. Carey, VP, DLC, and D.A. Chaney, USNRC, Project Manager, Operating Reactor Branch No. 1, Division of Licensing. Letter dated February 9, 1982.

U.S. Nuclear Regulatory Commission (USNRC) 1995. Voltage-based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, Generic Letter 95-05.

6.4-8

BVPS-2 UFSAR Rev. 26 Regulatory Guide 1.183, Revision 0, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors,"

July 2000.

6.4-9

BVPS-2 UFSAR Tables for Section 6.4

BVPS-2 UFSAR Rev. 26 TABLE 6.4-1 CONTROL ROOM ENVELOPE VENTILATION DESIGN PARAMETERS BVPS-1 BVPS-2 Parameter Value Value Gross volume (ft3) 151,000 79,400 Net free volume (ft3) 114,000 59,000 Post-accident filtered air makeup 1,000 1,000 rate (cfm)

Air changes per hour --- 0.35*

Recirculation air flow rate (cfm) 33,500 19,800 Control room emergency supply filtration unit characteristics Type Moisture separator,

• • electric heating coil, HEPA, charcoal, HEPA Filter bed depth 2 inches 2 inches charcoal charcoal Iodine filter efficiency Methyl iodide (at 30C and 99.5% 99.5%

70% RH)

Recirculation air filter characteristics Type Roll and Roll and high high efficiency efficiency Redundant automatic detection equipment Area radiation monitors setpoint 0.470 0.476 (mr/hr)

NOTE:

• Based on 1,000 cfm air makeup rate and combined net free volume.

• • Unit 2 only 1 of 1

BVPS-2 UFSAR Rev. 26 TABLE 6.4-1a CONTROL ROOM ENVELOPE VENTILATION SYSTEM PARAMETERS USED FOR LOCA, CREA, AND MSLB ANALYSES Control Room Parameters Free Volume 173,000 ft3 Unfiltered Normal Operation Intake and 1250 cfm (Notes 1 & 2)

Inleakage Isolation Mode Inleakage 450 cfm (Note 1)

Pressurization Mode Filtered Intake 800 to 1000 cfm Pressurization Mode Recirculation Not Credited CREVS Intake Filter Efficiency 99% (aerosols) 98% (elemental/organic iodine)

Pressurization Mode Unfiltered Inleakage 165 cfm (Note 1)

Occupancy Factors 0 to 24 hr (1.0) 1 to 4 d (0.6) 4 to 30 d (0.4)

Operator Breathing Rate 0 to 30 d (3.5E-04 m3/sec)

Delay in Initiation of Control Room Emergency Ventilation System due to LOOP Auto-Start on receipt of CIB (Note 3)

CR in isolation mode T=77 seconds (Note 4)

CR in emergency pressurization mode using T=137 seconds (Note 5)

BVPS-2 CREVS Manual CR in emergency pressurization mode T=30 minutes Notes:

1. Upper bound analytical value includes test measurement uncertainties and a 10 cfm allowance for ingress/egress.

2. To provide operational margin, the radiological dose consequence analyses assume that the unfiltered intake plus inleakage into the joint BVPS-1 & BVPS-2 control room is a maximum of 1250 cfm during normal plant operation.

3. High radiation signal is not credited in any analyses.

4. Credited in LOCA analysis only; time includes Emergency Diesel Generator start and EDG load sequencer delays.

5. Automatic start of BVPS-2 CREVS is not credited in any analyses.

1 of 1

BVPS-2 UFSAR Rev. 12 TABLE 6.4-3 LOCATIONS OF CONTROL ROOM AIR INTAKE, TOXIC GAS STORAGE, AND RADIATION RELEASE POINTS*

Station Coordinate N (ft) E (ft) Elevation Radiation Release Points D. BVPS-1 Elevated Release 3730 7550 885 E. BVPS-1 Ventilation Vent 3946.5 7753 815 F. BVPS-1 Turbine Building Vent 4014 7637.5 805 G. BVPS-1 and BVPS-2 Process Vent 4290 8395 1210 H. BVPS-2 Elevated Release 3907 8121 8906" I. BVPS-2 Ventilation Vent 3843 8025 820 J. BVPS-2 Turbine Building Vent 3676 8100 847 Toxic Gas Storage K. Ammonium Hydroxide/Hydrazine 3740 7960 7526" L. Nitrogen 3570 7820 7356" M. Carbon Dioxide 1 ton tank (Unit 1) 3677 7717 7356" 1 ton tank (Unit 1) 4056 7735 7356" 2 ton tank (Unit 2) 3890 7950 7356" 1 ton tank (Unit 2) 3830 7909 7356" 1 - 7.5-ton tank (Unit 2) 3623 8263 7306" N. Hydrogen 3570 7840 7356" Q. Morpholine/Sulfuric Acid 4020 7400 7136" Control Room Air Intakes R. BVPS-1 Control Room Intake 3870 7770 7356" S. BVPS-2 Control Room Intake 3890 7820 74610" NOTE:

• Refer to Figure 6.4-5.

1 of 1

1 1

UNIT 1 2..J L.. 3 UNIT 2

.KEY PLAN SCALE 1"* 400'-0" REACTOR REACTOR CONTAINMEMT CONTAINMENT UNIT 2 UNIT 1 I

L ______________ _,

WASTE I I COMP FAC I I I I I AUX BLDG I I I I I

WAREHOUSE

~.:+/-:=============================~-- __ ...J 0 40 eo CONT. RM. EXTN. CONTROL UNIT 1 SCALE FEET SECTION LOOKING SOUTH FIGURE 6.4-2 SECTION 1-1 UNIT 1 & 2 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

r 1

UNIT 1 KEY PLAN SCALE 1"'* 400'-0" REACTOR CONTAINMENT UNIT 1 I

II 'I I

SECTION LOOKING WEST UNIT 1 40 eo SCALE FEET Fl GURE 6.4-3 SECTION 2-2 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

r 1

UNIT 1 UNIT 2 KEY PLAN SCALE 1"*400'*0" REACTOR CONTAINMENT UNIT 2 0

1nl CONNECTING 1!dJ TUNNEL SECTION LOOKING EAST eo 40 UNIT 2 SCALE FEET FIGURE 6.4-4 SECTION 3-3 BEAVER VALLEY POWER STATION-UNIT 2 FINAL SAFETY ANALYSIS REPORT

Removed in Accordance with RIS 2015-17 FIGURE 6.4-!I CONTROL ROOM AIR INTAKES TOXIC GAS S TORAGE AND RADIATION RELEASE POINTS BEAVER VALLEY POWER STATION*UNIT 2 UPDATED FINAL SAFETY ANALYSIS REPORT

BVPS-2 UFSAR Rev. 26 6.5 FISSION PRODUCT REMOVAL AND CONTROL SYSTEMS 6.5.1 Engineered Safety Feature Filter Systems The main control room area emergency supply filtration system, as described in Section 9.4.1, is an Engineered Safety Feature (ESF) filter system that is used to mitigate the consequences of accidents.

The supplementary leak collection and release system (SLCRS) filters are not credited for accident mitigation. The system is described in Section 6.5.3.2.

6.5.1.1 Design Bases The ESF filter systems are designed in accordance with the following criteria:

1. General Design Criterion 2, as it relates to the system being capable of withstanding the effects of natural phenomena, as established in Chapters 2 and 3.

2. General Design Criterion 5, as it relates to shared systems.

No portion of the system is shared. However, dampers in all control room envelope normal intakes must close in order to isolate the common control room envelope.

3. General Design Criterion 19, as it relates to systems designed for habitability of the main control room for all accident conditions, with the exception of the Loss of Coolant Accident (LOCA), the Control Rod Ejection Accident (CREA) and the Fuel Handling Accident (FHA) which follow the criteria provided in 10 CFR 50.67.

4. General Design Criterion 41, as it relates to the design of systems to be used for containment atmosphere cleanup following postulated accidents and to control releases to the environment.

5. General Design Criteria 42 and 43, as they relate to the design of systems to permit the inspection and testing of containment ESF atmosphere cleanup systems.

6. General Design Criterion 60, as it relates to the control of the release of radioactive materials to the environment.

7. General Design Criterion 61, as it relates to the design of systems for radioactivity control under normal and postulated accident conditions.

8. General Design Criterion 64, as it relates to monitoring radioactive releases under normal, anticipated operational occurrences, and postulated accident conditions from ESF atmosphere cleanup systems.

6.5-1

BVPS-2 UFSAR Rev. 26

9. Regulatory Guide 1.26, as it relates to the quality group classification of systems and components, as discussed in Section 1.8.

10. Regulatory Guide 1.29, as it relates to the seismic design classification of system components, as discussed in Section 1.8.

11. Regulatory Guide 1.52, as it relates to system design requirements, maximum system flow requirements, and system functional requirements, as discussed in Section 1.8.

12. Regulatory Guide 1.183, as it relates to use of the Alternative Source Term design criteria in 10 CFR 50.67 for LOCA, CREA, and FHA.

6.5.1.2 System Design The ESF filter systems are designed in accordance with Regulatory Guide 1.52, except as stated in Section 1.8. Table 6.5-1 compares design features of the ESF filter systems with Regulatory Guide 1.52.

6.5.1.3 Safety Evaluation The ESF filter systems are evaluated in Sections 15.6.5 and 15.7.4 to show that they adequately remove airborne radioactive material during the postulated accident. The system design features ensure that the systems operate before and after the postulated accidents, as described in Sections 6.5.3.2.3 and 9.4.1.3.

6.5.1.4 Inspection and Testing Requirements The ESF filter systems are inspected after installation to ensure that the equipment is properly installed and operates correctly. The systems are tested and balanced after installation. Preliminary tests are performed as described in Section 14.2.12.

6.5.1.5 Instrumentation Requirements The following instrumentation is supplied on ESF systems, as required by operating conditions:

1. High differential pressure drop indication across moisture separators, prefilters, high efficiency particulate air filters (HEPA), and charcoal adsorber banks. All the indicators have a high pressure drop alarm,

2. High and high-high temperature alarms on charcoal adsorbers,

3. Humidity indication upstream and downstream of the heater, with high humidity alarm downstream of heater,

4. Temperature indication upstream and downstream of the heater, 6.5-2

BVPS-2 UFSAR Rev. 26

5. Open and closed indicators on motor-operated damper, and

6. Low airflow to heater alarm.

Instrumentation for the main control room emergency supply filtration system and for the SLCRS is described in Sections 9.4.1.5 and 6.5.3.2.5, respectively.

6.5.1.6 Materials The ESF filter systems utilize the following materials:

1. Ductwork, filter housings, filter component mounting frames, and water drains are fabricated from corrosion-resistant or painted carbon steel. Paints used for carbon steel can withstand all expected radiolytic and environmental conditions.

2. Prefilters meet Underwriters Laboratory (UL) Class 1 requirements in conformance with UL 900-1977 (UL 1977) and are listed in the current UL Building Materials List. The HEPA filters are qualified in accordance with UL 586-1977 and MIL-F-51068E (1981).

3. The adsorbent is steam-activated, virgin coconut shell carbon with physical properties in accordance with Table 2 of Regulatory Guide 1.52.

4. Elastomeric materials are capable of withstanding all expected radiolytic and environmental conditions.

6.5.2 Containment Spray as a Fission Product Cleanup System The quench spray system (QSS) and recirculation spray system (RSS), as described in Section 6.2.2, are safety-related systems that provide water spray to the containment during the unlikely event of a LOCA. Both systems function to depressurize the containment and to minimize the release of radioactive iodine and other aerosols to the environment. The analysis of the radiological consequences of the LOCA is given in Section 15.6.

6.5.2.1 Design Bases The following are the design bases of the QSS and RSS for removing iodine from the containment atmosphere:

1. General Design Criterion 41, as it relates to the design which permits containment atmosphere cleanup.

2. General Design Criterion 42, as it relates to the design which permits inspection of containment atmosphere cleanup systems.

3. General Design Criterion 43, as it relates to the design which permits testing of containment atmosphere cleanup systems.

6.5-3

BVPS-2 UFSAR Rev. 26

4. The system is capable of functioning effectively with the single failure of an active component in the spray system, any of its subsystems, or any of its support systems.

5. The amount of radioactive iodine and other aerosols in the containment atmosphere following a DBA is reduced so that the outleakage will result in a Total Effective Dose Equivalent (TEDE) below the recommended limits of 10 CFR 50.67.

6. The spray system is designed to obtain adequate coverage of the containment volume in order to limit the site boundary dose following a DBA to a value less than that established in 10 CFR 50.67.

7. The spray nozzles are designed to minimize the possibility of clogging while producing droplet sizes effective for iodine absorption.

8. The QSS and RSS remove elemental iodines and particulates from the containment atmosphere. Iodine removal coefficients used for dose calculations are discussed in Section 6.5.2.3.1.

9. The long-term pH of the sump water using sodium tetraborate as a buffer is expected to remain above 7.0, considering radiolysis of the sump water and electric cable jacketing materials.

10. The QSS is designed to initiate automatically by an appropriate accident signal and is capable of continuous operation until the refueling water storage tank is emptied (Section 6.2.2).

6.5.2.2 System Design The QSS consists of two parallel flow paths. Each flow path consists of one spray pump and associated piping and valves. Both flow paths provide quench spray to opposite sides of the two spray headers. The QSS design is discussed in detail in Section 6.2.2, and component data are given in Table 6.2-56.

The quench spray nozzles are manufactured by Spray Engineering Company (SPRACO) and are Model 1713A. Section 6.2.2 discusses the quench spray header design and the regions of the containment that are sprayed.

The QSS is capable of operating continuously until RWST is emptied. The system meets the redundancy requirements of an ESF and will satisfy the system performance requirements despite the most limiting single active failure in the short term, or the most limiting single active or passive failure in the long term.

The QSS becomes effective in less than 90 seconds after the postulated event. The chronology of events for system operation is discussed in Section 6.2.2.

6.5-4

BVPS-2 UFSAR Rev. 26 6.5.2.3 Design Evaluation 6.5.2.3.1 Fission Product Cleanup In the effectively sprayed region, fission product cleanup is actively accomplished by the quench and recirculation spray systems and passively by transport of particulates to the spray droplets and heat sink surfaces as a result of steam condensation on these surfaces (diffusiophoresis). In the unsprayed region, only passive gravitational settling promotes particulate removal.

The bounding fission product cleanup calculation for Units 1 and 2 is performed with the containment atmospheric conditions and with power uprating conditions. Since the bounding plant parameters applied for the fission product cleanup calculation such as the spray flow rates, spray droplet size, RWST temperature are conservative for the Unit 2 operation with the containment at sub-atmospheric conditions, all fission product cleanup results are applicable whether the plant is operating at sub-atmospheric or atmospheric conditions.

Removal of Particulates by Sprays The particulates are effectively removed from the containment atmosphere by the quench and recirculation spray systems. The particulate removal rate is calculated with Stone & Websters proprietary SWNAUA Computer Program (Lischer 1993). The SWNAUA Program is a derivative of the NAUA/MOD4 Computer Program (Bunz 1982) which has been modified for DBA calculations to include a conservative model for aerosol removal by sprays.

The model correlations that were implemented into SWNAUA tend to underestimate the spray removal coefficient. The spray model was originally described in Elia, 1993. For the effectively sprayed region of the containment, SWNAUA employs only the conservatively developed spray removal model and conservative condensation rates for the diffusiophoresis calculation when performing DBA calculations. While agglomeration is considered, its impact on the resulting particulate removal rates is negligible. In summary, the aerosol removal rates calculated by SWNAUA are conservative lower bound estimates.

There are several aerosol mechanics phenomena that promote the depletion of aerosols from the containment atmosphere. These include the natural phenomena of agglomeration, gravitational settling, diffusional plate-out, and diffusiophoresis; and removal by fluid mechanical interaction with the falling droplets that enter the containment atmosphere through the spray system nozzles. The particulate removal calculation for the effectively sprayed region takes credit for the removal effectiveness of only diffusiophoresis and sprays. Agglomeration of the aerosol is considered.

If gravitational settling and diffusional plate-out were considered, the spray removal coefficients would be slightly reduced but the total removal effectiveness by all removal mechanisms would increase. In the unsprayed region, only gravitational settling of aerosols is credited.

6.5-5

BVPS-2 UFSAR Rev. 26 The spray model in SWNAUA evaluates the particulate removal efficiency for each particle size in the aerosol by the following mechanisms: inertial impaction, interception, and Brownian diffusion. The aerosol removal constant due to spray is presented in NUREG-0772 as:

3 Fm h vspray - vsed spray = x 4 R sp w V vspray where spray = Particulate removal constant for spray Fm = Spray mass flow rate h = Spray fall height

= Collision efficiency Rsp = Spray droplet radius w = Density of the spray droplet V = Effectively sprayed volume of containment vspray = Velocity of the spray droplets vsed = Aerosol sedimentation velocity The collision efficiency is divided into three contributing mechanisms as described in BMI-2104:

i r d where i = Efficiency due to inertial impaction r = Efficiency due to interception d = Efficiency due to Brownian diffusion For viscous flow around the spray droplet, the inertial impaction efficiency is given in NUREG-0772:

1 i = 2 0.75 ln (2 Stk )

1 + Stk - 1.214 6.5-6

BVPS-2 UFSAR Rev. 26 The critical Stokes number, Stk, for viscous flow is 1.214; for Stk below this value, the model assumes the efficiency of inertial impaction is 0.

The Stk is calculated from BMI-2104:

2 p r 2 Cc ( vspray - vsed )

Stk =

9 R sp where r = Aerosol particle radius p = Aerosol density Cc = Cunningham slip correction factor

= Gas viscosity For droplet sizes typical of nuclear plant spray systems, the data of Walton and Woolcock show that the inertial impaction efficiency will be closer to that predicted for potential flow around the droplet. Calvert fitted this data to the expression:

2 Stk i =

Stk + 0.7 The inertial impaction efficiency predicted by this equation is always higher than that predicted by the viscous flow expression given above.

Calvert's fit is employed in this calculation.

For the remaining constituents of the collision efficiency, the spray model employs an interception efficiency of the form:

2 3 r 1 r r x 1-2 R sp 3 R sp which is a conservative approximation of the expression given by BMI-2104.

The efficiency due to Brownian motion is also taken from this report:

d = 3.5 Pe-2/3 where Pe = Peclet number

= 2vsprayRsp/DB DB = Aerosol diffusion coefficient 6.5-7

BVPS-2 UFSAR Rev. 26

= kBoltzTB (Fuchs 1964) kBoltz = Boltzmann constant

= 1.3804 X 10-16 erg/K T = Temperature, K Fuchs gives the aerosol mobility, B:

Cc B=

6r In most cases, the collision efficiency is dominated by inertial impaction, but for small aerosols, Brownian diffusion may become dominant.

The inertial impaction efficiency increases as aerosol size is increased, whereas the Brownian diffusion efficiency increases as aerosol size decreases.

The model can handle a distribution of up to 20 droplet radii with the spray removal efficiency being determined for each aerosol size bin.

However, the droplet diameter distribution can be accurately represented by a single diameter equal to the mass mean diameter (Elia 1993).

The bounding plant parameters for Units 1 and 2 are listed below.

Bounding Plant Parameters for Fission Product Cleanup Calculations Parameter Value Sprayed Containment Volume 3.062 x 1010 cm3 Fall Height 2,403 cm Spray Flow Rate 1,821 gpm (120 to 2080 sec) 2,910 gpm (2080 to 3855 sec) 7,871 gpm (3855 to 4227 sec) 6,113 gpm (4227 to 10158 sec) 4,740 gpm (10158 to 11545 sec) 3,178 gpm (11545 to 345600 sec)

Spray droplet radius 500 microns 6.5-8

BVPS-2 UFSAR Rev. 26 The containment pressure, temperature, relative humidity, and steam condensing rate transients following the NUREG-1465 style DBA are presented in Table 6.5-2.

Description of Aerosol The chemical composition of the aerosol is important only as it relates to the aerosol density utilized in the development of spray lambdas. The chemical composition during the gap release phase is assumed to be predominantly CsOH. The chemical composition during the early in-vessel release phase is assumed to be 20 percent CsOH, 20 percent indium, and 60 percent silver. These compositions are based on a review of the SASCHA experimental results (Albrecht 1984). The aerosol input data for SWNAUA are provided below.

Description of Aerosol Minimum Aerosol Radius 1.0E-07 cm Maximum Aerosol Radius 1.0E-02 cm Maximum Number of Aerosol Size Bins 100 From 30 sec to 1830.0 sec Aerosol Injection Rate 9.74 gm/sec Mean Geometric Radius 7.5E-06 cm Geometric Standard Deviation 1.56 Aerosol Density 3.7 gm/cc From 1830 sec to 6510.0 sec Aerosol Injection Rate 92.94 gm/sec Mean Geometric Radius 4.0E-05 cm Geometric Standard Deviation 1.46 Aerosol Density 4.6 gm/cc Removal of Particulates by Diffusiophoresis Diffusiophoresis entrains particulate matter in steam as it flows toward condensation surfaces. In this calculation, steam is assumed to condense on the spray droplets, on particulate matter, and on heat sinks. The diffusiophoresis model in the SWNAUA computer code is the same as that in the NAUA/MOD4 computer code.

The steam condensation rates used by SWNAUA are calculated by the LOCTIC computer code (Cho 1993). The LOCTIC code calculates conservative DBA containment pressure and temperature responses. Because LOCTIC predicts a conservatively high containment pressure transient, the rate of steam condensation from the containment atmosphere is minimized. The steam 6.5-9

BVPS-2 UFSAR Rev. 26 condensation rates that are input to the diffusiophoresis calculation of SWNAUA are taken as the steam removal rates from the containment atmosphere determined by LOCTIC.

The coefficient for removal of particulates from the effectively sprayed and unsprayed regions of the containment are plotted versus time in Figures 6.5-3 and 6.5-4, respectively. For the effectively sprayed region, the aerosol removal is due to sprays and diffusiophoresis. The particulate removal coefficient in the unsprayed region is due to gravitational settling only.

Removal of Elemental Iodine The calculated removal rate for elemental iodine in the vapor phase by sprays always exceeds 20 hr-1, the maximum value permitted by NUREG-0800, Standard Review Plan Section 6.5.2. Therefore, elemental iodine is conservatively assumed to be removed by sprays at either the same rate as the aerosol particles when the aerosol removal rate is lower than 20 hr -1 or at 20 hr-1 when the aerosol removal rate is calculated to be higher than the NRC limit.

A plateout removal coefficient for elemental iodine is calculated with the model provided in NUREG-0800, Standard Review Plan Section 6.5.2. In the effectively sprayed region, plate-out coefficients of 4.1075 hr -1 and 0.5358 hr-1 are calculated for the period before initiation of sprays and after initiation of sprays, respectively.

No credit is taken for elemental iodine removal in the unsprayed region.

Effectively Sprayed Containment Volume Fraction The sprayed volume fraction of the containment is determined by superimposing spray patterns for various spray nozzle orientations onto containment arrangement drawings. The sprayed volume is the volume of unblocked spray patterns. The spray patterns are based on the nozzle manufacturers laboratory tests at atmospheric conditions. The patterns have been compressed to account for the higher density atmosphere that exists during a DBA. The effectively sprayed volume is calculated by combining highly mixed unsprayed regions with directly sprayed regions.

The effective spray coverage fraction is determined to be 60.0 percent of the containment free volume. The concentration of fission products is expected to be uniform in the containment volume above the operating floor since this volume is open with very few obstructions to mixing. The sprayed volume is taken as the free volume above the operating floor plus the volume below the operating floor that is covered by recirculation spray. Actually, the whole containment is expected to be uniform in fission product concentration based on the discussion in Section 6.2.5, but the sprayed volume fraction has been limited to 60.0 percent.

6.5-10

BVPS-2 UFSAR Rev. 26 Containment Mixing The mixing rate between the effectively sprayed volume and the unsprayed volume of the containment is assumed to be 2 hr-1, the rate permitted by NUREG-0800, the Standard Review Plan Section 6.5.2.

6.5.2.3.2 Range of Spray pH Quench spray consists of a boric acid solution with a spray pH as low as 4.6. As indicated in Standard Review Plan (SRP), Section 6.5.2, Rev 2, Containment Spray as A Fission Product Cleanup System, fresh sprays (sprays with no dissolved iodine) are effective at scrubbing elemental iodine and thus a spray additive is unnecessary during the initial injection phase when the spray solution is being drawn from the Refueling Water Storage Tank (RWST). As described in the SRP, research has shown that elemental iodine can be scrubbed from the atmosphere with borated water, even at low pH. The SRP provides an equation for calculating a first order removal coefficient that is not dependent on pH.

The conditions assumed in calculating the minimum expected spray pH for the system are given in Table 6.5-4. The spray pH will remain at or above the value given in the table for all operating modes of the system. The values of the parameters used in calculating the limiting pHs are those Technical Specification limits which tend to minimize pH as appropriate.

6.5.2.3.3 Ultimate Sump pH The minimum expected ultimate sump pH is given in Table 6.5-5 along with the boric acid and NaTB sources considered in the analysis. The values of the parameters listed in this table are consistent with the appropriate Technical Specification limits which minimize the pH.

6.5.2.4 Inspection and Testing Requirements The inspection and testing of the quench spray system is described in Section 6.2.2.4.

6.5.2.5 Instrumentation Requirements The instrumentation required by the QSS is given in Section 6.2.2.5.

6.5.2.6 Materials The boric acid solution shows little change at high temperatures (130C) with or without radiation (Eggleton 1967; Fittel and Row 1971; Greiss and Bacarella 1969). The solution is not susceptible to significant radiolytic or pyrolytic decomposition under conditions found in nuclear power plant containments.

6.5.3 Fission Product Control Systems 6.5.3.1 Primary Containment 6.5.3.1.1 Design Bases The containment is a steel-lined, reinforced concrete construction. A complete description of the primary containment is presented in Sections 3.8.1, 3.8.2, and 6.2.1.

6.5.3.1.2 Deleted 6.5-11

BVPS-2 UFSAR Rev. 26 6.5.3.1.3 Containment Purge System Section 9.4.7.3 provides a description of the containment purge system.

6.5.3.1.4 Primary Containment Leakage Although the primary containment is designed to be leaktight, a design leakage rate is established and periodically verified at 0.1 percent of containment mass per day or less (Table 6.5-6). During a DBA, this leakage value is assumed for the initial 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period, after which the reduction in containment pressure to subatmospheric levels precludes any further leakage. Section 6.2.6 provides a description of the containment leakage testing program and monitoring system, and Section 15.6.5 discusses the fission product release via the containment during a DBA.

Information on the fission product removal system (containment spray system) is provided in Section 6.5.2.

6.5.3.2 Supplementary Leak Collection and Release System The function of the SLCRS is to collect potential containment leakage to the cable vault and rod control area, charging pump cubicles, component cooling water (CCW) pumps area, safeguards area, auxiliary building, and fuel building. The air is processed and filtered before release to the atmosphere at an elevated point. The system is designated as Safety Class 3. The SLCRS is shown on Figure 6.5-2. Design parameters of the SLCRS are listed in Table 6.5-7.

6.5.3.2.1 Design Bases The SLCRS is designed to the following criteria:

1. The SLCRS provides safety related cooling to the cable vault and rod control area by maintaining temperature at or below 120F during loss of offsite power, containment isolation phase B, or loss of chilled water cooling to the area. SLCRS is not required for cooling this area in the event of a tornado.

2. General Design Criterion 2, as it relates to the system being capable of withstanding the effects of natural phenomena, as established in Chapters 2 and 3.

3. General Design Criterion 4, as it relates to the portion of structures housing the system, and the portion of the system itself necessary for safe shutdown being capable of withstanding the effects of external and internally-generated missiles, pipe whip, and jet impingement forces associated with pipe breaks. The ventilation equipment room of the SLCRS, located on the top of the auxiliary building, is not required to be protected against tornadoes, hurricanes, or missiles because these natural phenomena are assumed not to occur within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before or after a DBA.

4. General Design Criterion 5, as it relates to shared systems and components important to safety.

5. General Design Criterion 17, as it relates to assuring proper functioning of the essential electric power system.

6.5-12

BVPS-2 UFSAR Rev. 26

6. General Design Criterion 41, as it relates to the containment atmosphere cleanup system being designed to control fission product releases to the environment following postulated accidents.

7. General Design Criterion 42, as it relates to the containment atmosphere cleanup system being designed to permit periodic inspection.

8. General Design Criterion 43, as it relates to the containment atmosphere cleanup system being designed to permit appropriate functional testing.

9. General Design Criterion 60, as it relates to the control of the release of radioactive materials to the environment.

10. Regulatory Guide 1.26, as it relates to the quality group classification of systems and components. This system is classified as QA Category I, Safety Class 3.

11. Regulatory Guide 1.29, as it relates to the seismic design classification of system components. This system is classified as Seismic Category I.

12. Regulatory Guide 1.52, with exceptions as indicated in Section 1.8, as it relates to system design requirements, maximum system flow requirements, design provisions for radiation detection, and isolation provisions for air filtration and adsorption unit.

13. Regulatory Guide 1.140, as it relates to the design, testing, and maintenance criteria for atmospheric cleanup systems.

14. Branch Technical Positions ASB 3-1 and MEB 3-1, as they relate to breaks in high and moderate energy piping system outside containment.

Other design bases include:

1. The maintenance of negative pressure in areas contiguous to the containment (except the main steam and feedwater valve area) and in the fuel building.

2. Filtration by impregnated charcoal absorber banks for radioactive iodine removal utilizing gasketless type module design, thus eliminating elastomer seals. Charcoal beds are seal-welded to the assembly and have a filling device to allow filling to a minimum density of 30 lb/ft3.

3. The provision of continual monitoring for radioactive particulate, iodine, and gaseous nuclides in the exhaust air being discharged at an elevated release point.

6.5-13

BVPS-2 UFSAR Rev. 26

4. The use of redundant demister assemblies, each with an electric heating coil, filter banks, and exhaust fans operable on emergency power.

5. The system is designated as nuclear safety-related.

6 Equipment and system capability of withstanding the design basis earthquake without loss of function.

7. Continuous exhaust through the HEPA filters and charcoal filtration units from the auxiliary building, charging pump cubicles, fuel building, and solid waste handling building.

8. Safety-related equipment in the Charging Pump Cubicles, Component Cooling Pump Area, Auxiliary Personnel Airlock, and the Cable Vault Areas Elev. 735' and 755' requires SLCRS flow for cooling. An Emergency Ventilation System, discussed in Section 9.4.3.2, as available for cooling those areas in the rare circumstances where SLCRS is unavailable.

6.5.3.2.2 System Description The components and operation of the SLCRS are shown on Figure 6.5-2 (design basis analysis (DBA) flow rates are also provided) and in Tables 6.5-7 and 6.5-8. The primary function of the SLCRS is to ensure that radioactive leakage from the primary containment following a DBA is collected and filtered for iodine removal prior to discharge to the atmosphere at an elevated release point through a ventilation vent. This ventilation vent also discharges the exhaust from the gland seal steam exhaust system as described in Section 9.4.15.

The SLCRS consists of: 1) one 10,500 cfm and one 29,000 cfm leak collection normal exhaust fans powered from the normal buses, 2) two 34,000 cfm leak collection filter exhaust fans powered from the emergency buses, 3) four 28,500 cfm filter banks, and 4) two 13,000 cfm emergency charging pump cubicle exhaust fans.

Air is exhausted from the fuel building, solid waste handling building, auxiliary building, charging pump cubicles, CCW pump area, post accident sampling system panel and personnel sampling area, and from the areas contiguous to the reactor containment except the main steam and feedwater valve area. The areas contiguous to the reactor containment are the personnel access hatch area, equipment hatch enclosure, purge duct area, main steam and feedwater valve area, cable vault and rod control building at el 735 ft-6 in and el 755 ft-6 in, pipe tunnel, and safeguards areas.

The capacity of each leak collection filter exhaust fan is in excess of the estimated air inleakage to the containment contiguous areas and the other buildings delineated in the previous paragraph. The excess capacity of the fan ensures a negative pressure in the areas being exhausted.

Tables 6.5-7 and 6.5-8 list nominal air flow rates required to ensure the negative pressure.

6.5-14

BVPS-2 UFSAR Rev. 26 In order to limit air leakage into these structures to less than the design capacity of a leak collection exhaust fan, all penetration pipes, ducts, and cables are sealed at or near the point where they pass from the contiguous structure to some other structure, such as for example, the auxiliary building. A flexible sealing compound is used between electrical cables and sleeves. Doors are either locked or self-closing, and are under administrative control.

The modes of operation of the SLCRS and nominal air flow rates are shown in Table 6.5-8. One of the leak collection normal exhaust fans is used to exhaust the various areas listed previously. The other fan is used as a standby. After plant shutdown, the standby fan is used for purging the reactor containment.

During normal plant operation, the inlet vanes of "A" Train normal exhaust fan is manually set to exhaust unfiltered air at the nominal rate of 10,500 cfm from the SLCRS areas, except for the solid waste handling building, auxiliary building, charging pump cubicles, CCW pump area, and the fuel building. The exhaust air from the excepted SLCRS areas is demisted, filtered, and exhausted by the leak collection filter exhaust fans. The other demister assemblies and main filter banks are used as standby.

On a containment isolation Phase A signal, or on a high radiation signal from monitors in the ventilation exhaust from the areas contiguous to the containment, the air that is normally exhausted by the leak collection normal exhaust fans (Figure 6.5-2) is diverted so that it first flows through one of the two parallel demister assemblies and then through the aligned main filter banks before flowing to the leak collection filter exhaust fans. Each demister assembly consists of a moisture separator and an electric heating coil. The moisture separator removes entrained water droplets while the electric heating coil reduces the relative humidity to less than 70 percent before the exhaust air stream passes through the main filter banks. Moisture separators are effective for the removal of at least 99 percent by weight of the entrained moisture in an air stream containing 0.005 lb of entrained moisture per ft3, and at least 99 percent by count of the 1 to 10 micrometer diameter droplets without visible carryover when operating at rated capacity of 57,000 cfm. Each main filter bank consists of HEPA filter, charcoal, and a second set of HEPA filter.

The charcoal filters are effective for radioactive iodine removal and the second HEPA filters remove particulates and charcoal fines at a rated efficiency of 99.97 percent when tested with 0.30 micron dioctyl phthalate. The charcoal filters are rated for 0.25 second residence time.

The charcoal filters are shown on Figure 6.5-2. They are of flat parallel bed design, containing approximately 500 pounds of charcoal per bed. The media is new, impregnated, activated, coconut shell charcoal. The qualification of impregnated carbon will be in accordance with Table 5-1 of ANSI N509-1980. Charcoal cells are leak-tested according to ANSI 510.

A heat detection alarm and manual water spray system is provided to prevent ignition in the event of decay heat buildup. Overtemperature conditions are alarmed locally and in the main control room.

6.5-15

BVPS-2 UFSAR Rev. 26 The leak collection filter exhaust fans discharge through a duct to an elevated release point 150 feet above grade. This elevated release point is located on the top of the containment structure, which is 144 feet above grade. The duct and supporting structure is designed to accommodate seismic forces.

The leak collection normal exhaust and filter exhaust fans are also used for reactor containment purging after plant shutdown to remove radioactivity from containment atmosphere. Section 9.4.7, Containment Ventilation System, provides additional information about the purge system.

Should the SLCRS suffer a loss of function, the emergency exhaust fan system, as shown on Figure 9.4-4, may be started manually from the main control room to remove the heat generated by the charging pumps and the CCW pumps. The emergency exhaust fan system consists of ducting, motor-operated dampers, and two axial flow fans with back draft dampers located within the tornado missile-protected portion of the auxiliary building.

The fans are powered from the emergency buses.

6.5.3.2.3 Safety Evaluation The SLCRS incorporates redundant 100 percent capacity leak collection exhaust fans, demister assemblies, and main filter banks. In addition, there are redundant dampers where required. The redundant fans, electric heating coils, and dampers are connected to redundant emergency buses, which are capable of being supplied either from normal 4,160 V buses 2A and 2D or emergency diesel generators 2-1 and 2-2 (Figure 8.3-1). Thus, there is sufficient redundancy in the system to ensure system reliability.

The SLCRS collects, filters, and releases at an elevated point, the leakage from the containment following a DBA. Essentially, all the leakage from the containment following a DBA flows into containment contiguous areas. These areas house the various containment penetrations, ESF equipment circulating radioactive water, and equipment used for plant shutdowns. The SLCRS, with the exception of the ESF portion of the system, is not tornado missile-protected.

The elevated release point in the SLCRS is located above the top of the containment and has a discharge flow rate of about 57,000 cfm. The contiguous area exhaust is normally exhausted directly to atmosphere, but the exhaust is automatically diverted through one of the demister assemblies and main filter banks on an accident signal and is discharged at this elevated release point. Upon failure of both hydrogen recombiners, the hydrogen control system purge blower will take suction from either recombiner suction line. The discharge of the blower is connected directly into the SLCRS contiguous area exhaust ductwork (see Section 6.2.5).

A FMEA to determine if the I&C and electrical portions meet the single failure criterion, and to demonstrate and verify how the GDC and IEEE Standard 279-1971 requirements are satisfied, has been performed on the supplementary leak collection and release system. The FMEA methodology is discussed in Section 7.3.2. The results of this analysis can be found in the separate FMEA document (Section 1.7).

6.5-16

BVPS-2 UFSAR Rev. 26 6.5.3.2.4 Inspection and Testing Requirements The system is inspected after installation to ensure that the equipment is properly installed and operates correctly. The system is tested and balanced after installation. Preliminary tests are performed as described in Section 14.2.12.

6.5.3.2.5 Instrumentation Requirements The fans and dampers for the SLCRS have manual controls and indicating lights in the main control room.

During normal operation, one leak collection normal exhaust fan is manually started. Upon failure of this fan, the standby normal exhaust fan is manually started. After plant shutdown, the normal exhaust fan is manually started for reactor containment purging; however, if high radiation is detected in the containment, the normal exhaust fan is automatically stopped and the air is diverted through the leak collection filters. During normal operation, both leak collection filter exhaust fans are manually started. Isolation dampers for both demister assemblies are manually opened. The outlet isolation dampers of the inservice filter bank are manually opened. The total flow through the leak collection filter exhaust fans is maintained constant by the automatic modulation of the variable inlet vanes mounted on each fan.

Following an accident signal, or when high radiation is detected in the areas contiguous to the containment, the leak collection normal exhaust fans are automatically isolated and the standby demister assembly and filter bank are automatically started so that all flow is routed through the leak collection filter exhaust fans. Each leak collection exhaust fan can be manually started or stopped from the main control room to supply emergency ventilation to the charging pump cubicles and the CCW pumps.

The following indications of system operation are provided in the main control room:

1. Auto trip of each leak collection normal exhaust fan is annunciated,

2. Auto trip of each leak collection filter exhaust fan is annunciated,

3. Auto trip of each emergency exhaust fan is annunciated,

4. High radiation in the containment and the exhausted air are annunciated, and

5. High differential pressure across each filter and low flow through a heater are annunciated.

6.5-17

BVPS-2 UFSAR Rev. 26 6.5.4 References for Section 6.5 ANSI/ANS Standard 56.5. 1979. PWR and BWR Containment Spray System Design Criteria.

Eggleton, A.E.J. 1967. A Theoretical Examination of Iodine-Water Partition Coefficients. UKAEA, AERE-R4887.

Fittel, H.E. and Row, T. H. 1971. Radiation and Thermal Stability of Spray Solutions. Nuclear Technology, p 442.

Griess, J. C. and Bacarella, A. A. 1969. Design Considerations of Reactor Containment Spray System - Part III, The Corrosion of Materials in Spray Solutions. ORNL-TM-2412, Part III, p 15.

Underwriter Laboratories 1977. Test Performance of Air Filter Units. UL 586-1977.

Underwriter Laboratories 1977. Test Performance of High Efficiency Particulate Air Filter Units. UL 900-1977.

U.S. Department of Defence 1981. Filter, Particulate High Efficiency, Fire Resistant. MIL-F-51068E.

Lischer, D.J., Users Manual, Aerosol Behavior in a Condensing Atmosphere (SWNAUA), June 1993, (Stone & Webster Proprietary)

Bunz, H., Kayro, M., Schck, W., 1982, NAUA/Mod4 - A Code for Calculating Aerosol Behaviour in LWR Core Melt Accidents, Code Description and User Manual, KfK.

Elia, Frank A. Jr. and Lischer, D. Jeffrey, Advanced Method for Calculating the Removal of Airborne Particles with Sprays, 1993, ASME paper no. 93-WA/SERA-5.

NUREG-0772, 1981, Technical Bases for Estimating Fission Product Behavior During LWR Accidents.

Battelle Columbus Laboratories, BMI-2104, Vol. III, draft report, 1984, Radionuclide Release Under Specific LWR Accident Conditions.

Walton, W. H., and Woolcock, A., 1960, The Suppression of Airborne Dust by Water Spray, Interm. J. Air Pollution 3, 129-153.

Calvert, S., 1970, Venturi and Other Atomizing Scrubbers Efficiency and Pressure Drop, AIChE Journal 16, 392-396.

Fuchs, N.A., 1964, The Mechanics of Aerosols, revised and enlarged edition, Dover Publications, Inc.

Albrecht, H. and H. Wild, Review of the Main Results of the SASCHA Program on Fission Product Release Under Core Melting Conditions," ANS Meeting on Fission Product Behavior and Source Term Research, Snowbird, Utah, 15-19 July 1984.

6.5-18

BVPS-2 UFSAR Rev. 26 Cho, J. H., Users Manual, Loss of Coolant Transient Inside Containment (LOCTIC), January 1993, (Stone & Webster Proprietary)

NUREG-0800, 1988, Standard Review Plan, Containment Spray as a Fission Product Cleanup System, Section 6.5.2, Revision 2.

6.5-19

BVPS-2 UFSAR Tables for Section 6.5

BVPS-2 UFSAR Rev. 0 TABLE 6.5-1 COMPARISON OF ENGINEERED SAFETY FEATURES FILTER SYSTEM DESIGN FEATURES WITH REGULATORY GUIDE 1.52 Regulatory Control Room Area Supplementary Guide 1.52 Pressurization Leak Collection Paragraph Filtration System System C.1.a A A C.1.b A A C.1.c A A C.1.d A A C.1.e A A C.2.a A A C.2.b A B C.2.c A A C.2.d A A C.2.e A A C.2.f A B C.2.g B B C.2.h B B C.2.i A A C.2.j B B C.2.k A A C.2.l B B C.3.a B B C.3.b A A C.3.c A A C.3.d A A C.3.e B B C.3.f A B C.3.g B B C.3.h A A C.3.i B B C.3.j A A C.3.k A A C.3.l B B C.3.m A A C.3.n B B C.3.o A A C.3.p B B C.4.a B B C.4.b A A C.4.c A A C.4.d B B C.4.e A A 1 of 2

BVPS-2 UFSAR Rev. 0 TABLE 6.5-1 (Cont)

Regulatory Control Room Area Supplementary Guide 1.52 Pressurization Leak Collection Paragraph Filtration System System C.5.a B B C.5.b B B C.5.c B B C.5.d B B C.6.a A A C.6.b B B NOTES:

A - Designed in accordance with Regulatory Guide 1.52.

B - Section 1.8 discusses design exceptions and justifications.

2 of 2

BVPS-2 UFSAR Rev. 26 TABLE 6.5-2 CONTAINMENT THERMODYNAMIC DATA - LOSS OF COOLANT ACCIDENT Steam Condensing Relative Time Pressure Time Temp Time Rates Time Humidity (sec) (psla) (sec) (°F) (sec) (gm/sec) (sec) Fraction 0.0 14.2 0.0 108.0 0.0 0.0 0.0 0.500 0.001 14.2 0.001 108.1 40.1 279591 5.0 0.500 40.1 52.3 40.1 258.5 80.1 120655 10.0 1.000 80.2 51.2 80.2 256.0 120.2 101211 345600.0 1.000 120.6 49.6 120.6 253.2 160.2 84363 160.2 49.0 160.2 251.3 200.2 74004 200.2 48.3 200.2 250.0 240.2 66345 240.2 47.8 240.2 249.4 280.7 61909 280.7 47.6 280.7 248.9 321.0 58435 321.0 47.4 321.0 248.5 361.0 55660 361.0 47.4 361.0 248.4 401.0 53361 401.0 47.3 401.0 248.3 441.0 51477 441.0 47.3 441.0 248.3 481.0 49834 481.0 47.4 481.0 248.3 522.1 48459 522.1 47.5 522.1 248.4 562.1 47186 562.1 47.6 562.1 248.6 602.1 47106 602.1 47.7 602.1 248.9 642.1 46143 642.1 47.9 642.1 249.2 682.1 45339 682.1 48.1 682.1 249.5 722.1 44671 722.1 48.2 722.1 249.8 762.1 44046 762.1 48.4 762.1 250.1 802.1 43450 802.1 48.6 802.1 250.5 842.1 40987 1 of 4

BVPS-2 UFSAR Rev. 26 TABLE 6.5-2 (Cont)

Containment Thermodynamic Data - Loss of Coolant Accident Steam Condensing Time Pressure Time Temp Time Rates Time Relative (sec) (psla) (sec) (°F) (sec) (gm/sec) (sec) Humidity 842.1 48.9 842.1 250.9 882.1 39002 882.1 48.6 882.1 250.4 922.1 36625 922.1 48.3 922.1 249.8 962.1 35512 962.1 48.0 962.1 249.2 1002.1 34577 1002.1 47.7 1002.1 248.6 1042.1 33760 1042.1 47.4 1042.1 248.1 1082.1 33032 1082.1 47.2 1082.1 247.7 1122.1 32377 1122.1 47.0 1122.1 247.3 1162.1 31885 1162.1 46.8 1162.1 246.9 1202.1 31344 1202.1 46.6 1202.1 246.6 1242.1 32339 1242.1 46.5 1242.1 246.3 1282.1 31822 1282.1 46.3 1282.1 246.0 1322.1 31350 1322.1 46.2 1322.1 245.7 1362.1 30936 1362.1 46.1 1362.1 245.4 1402.1 30555 1402.1 45.9 1402.1 245.2 1442.1 30201 1442.1 45.8 1442.1 244.9 1482.1 29871 1482.1 45.7 1482.1 244.7 1522.1 29564 1522.1 45.6 1522.1 244.5 1562.1 29183 1562.1 45.4 1562.1 244.1 1602.1 28593 1602.1 45.2 1602.1 243.6 1642.1 28095 1642.1 45.0 1642.1 243.2 1682.1 27661 1682.1 44.8 1682.1 242.8 1722.1 27262 1722.1 44.6 1722.1 242.4 1800.1 26465 1762.1 44.4 1762.1 242.0 2004.0 25498 1804.0 44.2 1804.0 241.6 2151.1 23937 1844.0 44.0 1844.0 241.2 2164.0 28256 1884.0 43.8 1884.0 240.8 2844.0 25766 2 of 4

BVPS-2 UFSAR Rev. 26 TABLE 6.5-2 (Cont)

Containment Thermodynamic Data - Loss of Coolant Accident Steam Condensing Time Pressure Time Temp Time Rates Time Relative (sec) (psla) (sec) (°F) (sec) (gm/sec) (sec) Humidity 1924.0 43.6 1924.0 240.5 3066.0 23153 1964.0 43.5 1964.0 240.1 3248.4 19907 2004.0 43.3 2004.0 239.8 3448.4 14942 2044.0 43.1 2044.0 239.4 3548.4 10768 2078.0 43.0 2078.0 239.1 4049.5 13086 2093.0 43.0 2080.0 239.1 4359.5 13219 2140.0 42.8 2081.0 239.3 4759.5 8477 2170.0 42.1 2100.0 239.1 4959.5 10098 2212.0 41.0 2152.4 238.2 5059.5 10485 2252.0 40.1 2200.0 234.3 5159.5 10929 2350.0 37.9 2252.0 231.3 5259.5 11132 2404.0 36.3 2964.0 216.9 5363.8 5849 2564.0 35.5 3605.0 209.0 5463.8 8500 2964.0 33.8 3877.5 204.6 5563.8 9584 3084.0 33.3 3957.0 196.5 5663.8 10344 3404.0 32.1 4749.5 180.1 5763.8 10871 3605.0 31.4 4849.5 183.7 5863.8 11165 3957.0 29.0 5749.5 191.2 6526.9 11867 4749.5 24.9 6749.5 191.4 7329.6 11391 4849.5 25.3 7263.9 192.6 10144.7 10108 5949.5 27.2 18074.7 188.9 10274.1 6472 6749.5 27.2 35998.0 180.5 11662.9 5552 7263.9 27.4 72198.0 170.9 12362.9 8920 18074.7 26.8 144020.0 162.6 20062.9 8936 35998.0 24.9 259020.0 155.3 30080.0 8293 72198.0 23.1 346020.0 150.4 49997.6 7255 144020.0 21.7 59997.6 6910 3 of 4

BVPS-2 UFSAR Rev. 26 TABLE 6.5-2 (Cont)

Containment Thermodynamic Data - Loss of Coolant Accident Steam Condensing Time Pressure Time Temp Time Rates Time Relative (sec) (psla) (sec) (°F) (sec) (gm/sec) (sec) Humidity 259020.0 20.8 69997.6 6689 346020.0 20.2 79997.6 6368 90000.0 6020 100000.0 5643 200020.0 5044 259020.0 4685 300020.0 4455 346028.5 4189 4 of 4

BVPS-2 UFSAR Rev. 18 TABLE 6.5-4 PARAMETERS FOR CALCULATING MINIMUM SPRAY PH DURING QUENCH SPRAY OPERATION Minimum*

pH = 4.6 Quench spray flow rate (gpm) 4,450 Boron concentration in the RWST (ppm) 2,600 NOTES:

• Minimum quench spray pH is calculated utilizing the following:

1. Maximum quench spray flow rates

2. Maximum boron concentration in the RWST 1 of 1

BVPS-2 UFSAR Rev. 18 TABLE 6.5-5 PARAMETERS FOR MINIMUM ULTIMATE SUMP PH CALCULATION RWST volume (gal) 866,592 Boron concentration in RWST (ppm) 2,600 Reactor coolant system (RCS) volume (gal) 66,386 Boron concentration in RCS (ppm) 2,400 Safety injection (SI) accumulator volume (gal) 8,020 Boron concentration in accumulators (ppm) 2,600 Weight of NaTB in Containment Sump pH Control (lbs.) 13,980 (Minimum)

Resulting minimum ultimate sump pH* 7.0 NOTE:

• The minimum ultimate sump pH is calculated utilizing the following:

1. Maximum boron concentration in the RWST, RCS, and SI accumulators,

2. Maximum volume of the RWST, RCS, and SI accumulators, and

3. Minimum weight of NaTB in the containment sump pH control system.

1 of 1

BVPS-2 UFSAR Rev. 16 TABLE 6.5-6 PRIMARY CONTAINMENT INFORMATION Data Description Parameter Value Type of structure Steel-lined reinforced concrete Primary containment design leak rate 0.1% per day Primary containment operation 13.5 psia Primary containment internal fission product removal systems Ice condenser Not applicable Spray system (accident) 3,000 gpm each pump Filter system (normal operation) 2 at 10,000 cfm each H2 purge mode (direct; to recirculation systems; to annulus)

Purge initiation time Long term backup Purge rate 50 cfm Primary containment purge Normal plant operation Containment not purged At cold shutdown 29,000 cfm max*

Valve arrangement Figure 6.5-2

• As Built Parameter 1 of 1

BVPS-2 UFSAR Rev. 1 TABLE 6.5-7 SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM PRINCIPAL COMPONENTS AND DESIGN PARAMETERS Components Design Parameters Leak collection normal exhaust fan Quantity 1* 1*

Capacity, (cfm) each 29,000* 17,350*

Static pressure (in WG) 6.6* 9.4*

Motor (hp) each 50 50 Leak collection filter exhaust fan Quantity 2 Capacity (cfm) each 34,000*

Static pressure (in WG) 21.1 Motor (hp) each 200 Emergency charging pump cubicle exhaust fan 2

Quantity 13,000 Capacity (cfm) each 6.0 Static pressure (in WG) 30 Motor (hp) each Filter house assembly Quantity 4 Capacity (cfm) each 28,500 Charcoal filter pressure drop (in WG) 0.9 Elemental iodine removal efficiency 99.5%

(at 30C and 95% RH)

Upstream HEPA pressure drop (in WG)

New 1.3 @ 1500 cfm per cell Dirty (requires replacement) 3 HEPA efficiency 99.97%

Downstream HEPA pressure drop (in WG)

New 1.3 @ 1500 cfm per cell Dirty (requires replacement) 2 HEPA efficiency 99.97%

• As Built Parameters 1 of 1

BVPS-2 UFSAR Rev. 13 TABLE 6.5-8 SUPPLEMENTARY LEAK COLLECTION AND RELEASE SYSTEM AIR FLOW RATES Modes of Plant Operation Normal Normal Normal Reactor Trip Purge Purge Refueling Refueling Rad. Signal in RC Loss of Filter DBA or Loss of High Activity in No Activity in No Activity in Fuel High Activity in Fuel Incident condition None None Contg. Areas Exhaust Fan Normal Power Reactor Cntmnt Reactor Cntmnt Bldg. Bldg.

Filtration modes Filtered Unfilt Filtered Unfilt Filtered Unfilt Filtered Unfilt Filtered Unfilt Filtered Unfilt Filtered Unfilt Filtered Unfilt Air Flow (cfm) From:

Aux bldg and Solid Waste 39,250 0 26,750 0 17,000 0 0 0 20,500 0 17,000 0 20,500 0 20,500 0 Handling Building Aux bldg charging and 14,750 0 14,550 0 15,000 0 16,750 0 12,500 0 15,000 0 12,500 0 12,500 0 component cooling pumps area Main steam valve area** 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 Fuel bldg 3,000 0 3,000 0 2,000 0 3,500 0 2,500 0 2,000 0 2,500 0 2,500 0 RC contiguous area 0 10,500 11,100 0 0 10,500 15,000 0 0 10,500 0 10,500 0 10,500 0 10,500 RC purge 0 0 0 0 0 0 0 0 9,400*** 0 0 29,000 7,500 0 7,500 0 Total flow (cfm): 57,000 10,500 55,400 0 34,000 10,500 35,250 0 44,900 10,500 34,000 39,500 43,000 10,500 43,000 10,500 Exhausted by: Filter** Normal*** Filter Normal Filter Normal Filter Normal Filter Normal Filter Normal Filter Normal Filter Normal exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan exh fan Flow per fan (cfm):

Fan A 28,500 10,500 27,700 0 34,000 10,500 35,250 0 44,900 10,500 34,000 10,500 43,000 10,500 43,000 10,500 Fan B 28,500 Standby 27,700 Standby Lost Standby 0 (single 0 0 Standby 0 29,000 Standby Standby failure)

Filter Capacity (cfm):

Main filter - Bank A 59,000 -- 59,000 0 59,000 -- 59,000 -- 59,000 -- 59,000 -- 59,000 -- 59,000 --

Main filter - Bank B Standby -- Standby -- Standby -- Standby -- Standby -- Standby -- Standby -- Standby --

NOTES:

• • Flow path is not used - the duct is blanked off.

• • • Flow path isolates on High High Activity in Reator Containment, flow is zero at that time. Air flow limitation is not required for the purpose of controlling a fuel handling accident radiological release.

1 of 1

MOISTURE SEPARATOR NO. 1 REV. 12 FL TA NOTE: FILTERS 204(A/Bl. 206<AIBl, 250A CH219A 207<A1Bl AND 209<A/8l ARE HEPA FILTERS.

ELEVATED RELEASE TO ATMOSPHERE PURGE EXHAUST FROM CONTAINMENT VENTILATION SYSTEM TYPICAL FOR ALL FOUR FILTER TO SEAL TANK\ _ HOUSE 2DBS-TK21 ASSEMBLIES TYPICAL FOR MOISTURE SEPARATORS NOS. 1 AND 2 2108 MOISTURE SEPARATOR NO. 2 FLTA 250B CH219B 2188 FROM ~

INSTRUMENT II AIR FROM BLANKED CHARGING PUMP CUBICLE ST~AM AND COMPONENT COOLING VENT TO PUMP AREA ATMOSPHERE MAIN AND FEEDWATER AREA COMBINED FROM 20018 AUXILIARY BUILDING AND SOLID WASTE HANDLING BUILDING RQ 1101 ALL DAMPER, VALVE AND EQUIPMENT FROM IDENTIFICATION NUMBERS ON THIS FIGURE SCREEN/ GLAND STEAM ARE PRECEDED BY THE SYSTEM DESIGNATOR OPENING EXHAUST SYSTEM "2HVSu UNLESS OTHERWISE INDICATED.

NOTE:

FROM THE EXHAUST FLOW VALUES OF 476 CFM F' IGURE 6. 5-2 PURGE EXHAUSir---~--~ NORMAL EXHAUST THROUGH DAMPER 203A. 514 CFM THROUGH SYSTEM FANS DAMPER 205 AND 1.306 CFM THROUGH DAMPER SUPPLEMENTARY LEAK COLLECTION 202 ARE THE MINIMUM REQUIRED AIRFLOWS FOR AND RELEASE SYSTEM EQUIPMENT IN THE CABLE VAULT AND ROD

REFERENCE:

STATION DRAWINGS OM 16-1 AND 2 CONTROL AREA IN THE DBA ALIGNMENT !THOSE BEAVER VALLEY POWER STATION UNIT NO. 2 DAMPERS NOT SHOWN ON THIS FIGURE>. UPDATED FINAL SAFETY ANALYSIS REPORT

BVPS-2 UFSAR Rev. 26 6.6 IN-SERVICE INSPECTION OF ASME CODE CLASS 2 AND CLASS 3 COMPONENTS This section addresses the in-service inspection (ISI) of ASME Boiler and Pressure Vessel Code,Section III, Class 2 and Class 3 components, as required by the 1983 edition of the ASME Boiler and Pressure Vessel Code,Section XI, with addenda through Summer 1983.

After Beaver Valley Power Station - Unit 2 (BVPS-2) becomes operable, the ISI program shall be periodically updated and subsequent ISIs shall be performed, to the extent practical, to meet the latest applicable addendum to ASME XI that has been endorsed by the U.S. Nuclear Regulatory Commission (USNRC) in accordance with 10 CFR 50.55a(g).

6.6.1 Components Subject to Examination Code Class 2 systems and components will be in-service inspected in accordance with and to the extent required by ASME XI, Subsection IWC and Table IWC-2500-1. Code Class 3 systems and components will be in-service inspected in accordance with and to the extent required by ASME XI, Subsection IWD and Table IWD-2500-1.

All components in Class 2 and Class 3 fluid systems comprising the ISI program will be examined at regular intervals during BVPS-2 service lifetime by nondestructive examinations, pressure tests, operational surveillance, or combinations of these inspections, as set forth in the BVPS-2 ISI program. The general requirements of ASME XI, Subsection IWA, are used as a guideline for the BVPS-2 ISI program examinations, procedures, and records. Certain exceptions may be taken whenever specific written relief is granted by the USNRC in accordance with 10 CFR 50.55a(g)(6)(i).

6.6.2 Accessibility The design and arrangement of Code Class 2 systems and components provide access for all ISIs required by ASME XI, Subsection IWC. Adequate access is provided for performance of required volumetric and surface examinations specified in ASME XI, Table IWC-2500-1.

The design and arrangement of Code Class 3 systems and components provide for performance of all visual inspections and surveys to meet the requirements of ASME XI, Subsection IWD, Table IWD-2500-1. Special design considerations are given to those systems that are intended to be examined during normal plant operation.

Code Class 2 welds which receive ultrasonic examination are contoured and finished to permit meaningful examination of the welds by eliminating irregular or rough surfaces and sharp discontinuities which would interrupt the flow of sound between the transducer and the weld.

Irregular weld geometry that would tend to scatter ultrasonic beams has been avoided. Removable insulation is provided on those piping systems requiring volumetric and surface inspection.

In addition, the placement of pipe hangers and supports with respect to those welds requiring inspection has been reviewed and modified, where necessary, to permit adequate access to these areas for inspection.

6.6-1

BVPS-2 UFSAR Rev. 26 A verification review will be performed to assure compliance with ASME XI code requirements. The review and documentation will consist of the following tasks:

1. Listing of all ASME III Class 2 and 3 lines requiring examination.

2. Developing a set of piping isometric drawings showing the location of welds requiring inspection.

3. Tabulating all candidates for weld examination and identifying the type of inspection required.

4. Review that all welds have adequate clearance and accessibility for inspection personnel and equipment. This review will be performed on the 1/16 scale engineering model and on installed components.

5. Documentation of the clearance and access study.

This documentation will be reviewed periodically during BVPS-2 construction to ensure its accuracy and the information will be updated and maintained.

6.6.3 Examination Techniques and Procedures In-service inspection examination techniques for Code Class 2 systems and components are volumetric, surface, and visual to meet the requirements of Table IWC-2500-1 of ASME XI. Ultrasonic techniques are generally employed where volumetric examination is required, and either liquid penetrant or magnetic particle techniques are employed where surface examination is required. Visual examinations are conducted in accordance with the requirements of ASME XI, Table IWC- 2500-1 and Subarticle IWA-2210.

Code Class 3 systems and components are given a visual survey examination during system in-service tests, component functional tests, or system pressure tests to detect evidence of component leakage, structural distress, or corrosion, in accordance with the requirements of Table IWD-2500-1 and Subarticle IWA-2210 of ASMS XI.

6.6.4 Inspection Intervals For Code Class 2 and Class 3 systems and components, the ISIs will be conducted in accordance with ASME XI Subarticle IWA-2400. The inspection schedule for Class 2 systems and components will be per Program B, Table IWC-2412-1. In-service inspections for Code Class 3 systems and components are conducted when systems are undergoing either a system in-service test, component functional test, or system pressure test, as specified by ASME XI, Subarticle IWD-2400 and Table IWD-2500-1.

To the extent practicable, it is intended that ISIs be performed during normal plant outages, such as refueling shutdowns or maintenance shutdowns occurring during the inspection interval.

6.6-2

BVPS-2 UFSAR Rev. 26 6.6.5 Examination Categories and Requirements The ISI categories and requirements for Code Class 2 systems and components are in agreement with, and are designed to permit ISI required by, ASME XI, Table IWC-2500-1.

In-service inspection categories and requirements for Code Class 3 systems and components are in agreement with, and are designed to permit ISI required by, ASME XI, Table IWD-2500-1.

All welds requiring volumetric or surface examination must be identified by number for inspection. For identification and tracking of welds (requiring volumetric and surface examination) during the design and fabrication stages, the pipe fabricators weld joint numbering system is utilized.

6.6.6 Evaluation of Examination Results Evaluation of Code Class 2 and Class 3 component examination results will be made in accordance with the requirements of ASME XI, Article IWB-3000.

Repair procedures for Code Class 2 components will comply with the repair rules of ASME XI, Article IWC-4000. Repair procedures for Class 3 components will comply with the repair rules of ASME XI, Article IWD-4000.

When Section XI promulgates rules for flaw evaluation under Articles IWC-3000 and IWD-3000, and if the associated addenda are referenced by 10 CFR 50.55a, these rules will be incorporated into this section, if applicable. Components found to contain unacceptable indications will be repaired under the rules of Articles IWC-4000 for Class 2 and IWD-4000 for Class 3, or will be replaced under the rules of Article IWA-4000 and Articles IWC-4000 for Class 2 and IWD-4000 for Class 3.

6.6.7 System Pressure Tests System pressure tests on Code Class 2 systems and components will be conducted to comply with the criteria established in ASME XI, Article IWC-5000. System pressure tests on Code Class 3 systems and components will be conducted to comply with the criteria established in ASME XI, Article IWD-5000. The general requirements of Article IWA-5000 apply to pressure testing of both classes of components.

6.6.8 Augmented In-Service Inspection to Protect Against Postulated Piping Failures The augmented ISI program provides for examination of high energy piping that penetrates the primary containment. The examination requirement for piping welds will be defined by Table 4.1-1 of WCAP-14572, Revision 1-NP-A.

Welds located between the containment wall outside the containment, and the main steam valve house wall (as defined by FSAR Figures 3.6B-13 and 3.6-14), consisting of only the main steam and feedwater lines in the main steam valve house; all high energy branch lines within the main steam valve house up to the first relief or non-manual isolation valve; and 6.6-3

BVPS-2 UFSAR Rev. 26 those lines identified as "extended" break exclusion zone piping (reference Table 210.5-1 in response to NRC Question 210.5) are examined using volumetric techniques for circumferential butt or longitudinal welds at locations selected with the risk-informed methodology described in WCAP-14572, Revision 1-NP-A, and WCAP-14572, Revision 1-NP-A, Addendum 1-A. During each inspection interval, these welds will be volumetrically examined.

6.6-4

BVPS-2 UFSAR Rev. 26 Appendix 6A Generic Letter 2004-02 Containment Sump Evaluation TABLE OF CONTENTS Section Title Page 6A.1 CONTAINMENT SUMP DESCRIPTION ............................... 6A-4 6A.1.1 General Plant System Description ........................... 6A-4 6A.1.2 Description of Containment Sump Strainers .................. 6A-5 6A.2

SUMMARY

OF GL 2004-02 EVALUATIONS PERFORMED ................ 6A-6 6A.2.1 Pipe Break Characterization ................................ 6A-6 6A.2.2 Debris Generation .......................................... 6A-6 6A.2.3 Latent Debris and Foreign Materials ........................ 6A-7 6A.2.4 Debris Transport to the Sump Strainers ..................... 6A-8 6A.2.5 Sump Strainer Evaluations .................................. 6A-8 6A.2.6 Debris Source Term Reduction ............................... 6A-10 6A.2.7 Upstream Effects of Debris Accumulation .................... 6A-10 6A.2.8 Downstream Effects - Components and Systems ................ 6A-10 6A.2.9 Downstream Effects - Fuel and Vessel ....................... 6A-11 6A.2.10 Chemical Effects ........................................... 6A-12 6A.3 ANALYZED DEBRIS LIMITS FOR CONTAINMENT ..................... 6A-13 References for Appendix 6A ......................................... 6A-14 6A-1

BVPS-2 UFSAR Rev. 26 LIST OF TABLES Table Title 6A-1 Bounding Break Locations in Each Area of Containment 6A-2 In-Vessel Debris Effects Key Parameter Evaluation 6A-3 Analyzed Debris Limits for Containment -

Applicable to All LOCA Scenarios 6A-4 Analyzed Debris Limits for Containment -

Applicable to Breaks Inside the Reactor Cavity 6A-5 Analyzed Debris Limits for Containment -

Applicable to Breaks Outside the Reactor Cavity 6A-2

BVPS-2 UFSAR Rev. 26 LIST OF FIGURES Figure Title 6A-1 Containment Sump and Strainer Assembly 6A-2 Containment Sump Top-Hat Strainer Module 6A-3

BVPS-2 UFSAR Rev. 26 APPENDIX 6A GENERIC LETTER 2004-02 CONTAINMENT SUMP EVALUATION This Appendix provides a discussion of the conformance of BVPS-2 to the requirements of NRC Generic Letter 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors (Reference 1). The content of the information provided in this Appendix is that prescribed by Revised Content Guide for Generic Letter 2004-02 Supplemental Responses, dated November 2007 (Reference 2) and PWROG-16073-P, TSTF-567 Implementation Guidance, Evaluation of In-Vessel Debris Effects, Submittal Template for Final Response to Generic Letter 2004-02 and FSAR Changes, Revision 0 (Reference 3).

6A.1 CONTAINMENT SUMP DESCRIPTION 6A.1.1 General Plant System Description The containment recirculation sump is a collecting reservoir designed to provide an adequate supply of water, with a minimum amount of debris, to the Emergency Core Cooling System (ECCS) and Recirculation Spray System (RSS). Downstream components are protected from debris by a strainer assembly immediately upstream of the containment sump. The containment sump performance meets the following NRC acceptance criteria:

General Design Criteria 35 through 43.

10 CFR 50.46(b)(5)

Regulatory Guide 1.82, Revision 3 Following a LOCA, the injection mode of ECCS operation is initiated automatically and requires no operator action. During injection, the Low Head Safety Injection (LHSI) and High Head Safety Injection (HHSI) pumps take suction from the Refueling Water Storage Tank (RWST) and deliver borated water to the cold legs.

Each of the two Quench Spray (QS) pumps takes suction from the RWST and discharge into separate 360-degree spray rings near the top of containment. When the RWST level reaches the low level setpoint, Recirculation Spray Pumps start automatically and provide recirculation of the containment sump water through the Recirculation Spray Coolers to the spray headers. Each 360-degree spray ring header is fed by two risers, where each riser originates from one of the recirculation coolers.

Switchover from injection to cold leg recirculation occurs automatically as the level in the RWST drops to the extreme low level set point. The LHSI pumps are stopped, while the discharge flow paths for two RSS pumps are automatically realigned to the RCS cold legs and the HHSI pump suctions.

6A-4

BVPS-2 UFSAR Rev. 26 After the transfer to recirculation takes place and approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the LOCA occurs, the operators initiate hot leg recirculation, at which time all ECCS flow paths are re-aligned from the cold legs to the hot legs. The alignment is switched between hot leg and cold leg recirculation every 9.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the initial transfer to hot leg recirculation.

6A.1.2 Description of Containment Sump Strainers The four recirculation spray pumps take suction from the containment sump, which is enclosed by a protective screen assembly. The surface area of the strainer assembly is sufficient to ensure that the specific BVPS-2 debris strainer loading does not impair ECCS and RSS performance and approach flow velocities are low enough to prevent entrainment of most small particles. A passive single failure is not assumed to be credible for the strainer assembly; therefore, the entire strainer area is credited for both trains.

The BVPS-2 containment sump strainer assembly provides 3,396 ft² of strainer area. The strainer assembly consists of 113 top-hat strainers, placed in a rectangular grid arrangement, mounted to the floor near the outside wall of containment. The top-hats are constructed of a series of perforated plate tubes.

The new strainer arrangement for BVPS-2 consists of three segments, A, B, and C, with connectors between segments. Segment A is located over the existing sump trench. Each segment has vertically orientated, cylindrical top-hat style strainer assemblies supported on structural frames. Each top-hat is approximately 3 feet long and consists of four perforated plate tubes of different diameters stacked one inside the other. The perforated plates are made from 14 gage stainless steel plates with 3/32 inch diameter holes. A bypass eliminator material made of woven stainless-steel wire is sandwiched between the tubes. Top-hats have a square flange at the bottom for attachment to the supporting frames. A cruciform near the flange acts as a vortex suppressor. Additionally, in segment A, vortex suppression grating is installed between the top-hats and the RSS pump inlets. There are water boxes below each of the three separate segments to collect and channel recirculated containment water to the sump trench.

The layout of strainer assembly is shown in Figure 6A-1. Individual strainer modules are depicted in Figure 6A-2.

The design of the containment sump is in accordance with Regulatory Guide 1.82, Revision 3, with the following exceptions and justification for each:

1. The recirculation spray pumps take a suction from a single sump. A passive failure of the containment sump screen has been determined to be not credible and is therefore not assumed.

2. A portion of the containment floor slopes down toward the sump, but a raised lip is provided, which directs normal floor drainage to the 6A-5

BVPS-2 UFSAR Rev. 26 segmented section of the containment sump and will prevent small debris from being swept directly into the sump.

3. The BVPS-2 net positive suction head (NPSH) calculation methodology credits containment overpressure. This approach was deemed conservative and acceptable by the NRC in License Amendment No. 167.

6A.2

SUMMARY

OF GL 2004-02 EVALUATIONS PERFORMED Analysis has been performed to determine the susceptibility of the ECCS and Containment Heat Removal recirculation functions for BVPS-2 to the adverse effects of post-accident debris blockage and operation with debris-laden fluids. These analyses conform to the greatest extent practicable to the methodology of NEI 04-07, Pressurized Water Reactor Sump Performance Evaluation Methodology (Reference 4) as approved by the NRC Safety Evaluation Report dated December 6, 2004 (Reference 5). The following is a summary description of the analysis areas performed.

6A.2.1 Pipe Break Characterization The piping considered for breaks is limited to RCS and connecting piping 2 inches in diameter or greater within the LOCA accident boundary. The LOCA accident boundary is defined in Section 3.6N.2.3.2 and shown in Figure 3.6N-2.

Regulatory Position 1.3.2.3 of Regulatory Guide 1.82, Water Sources for Long-Term Recirculation Cooling Following a Loss-of-Coolant Accident, Revision 3 (Reference 6), was used to select the spectrum of breaks for evaluation. A summary of bounding break locations in each area of containment is provided in Table 6A-1.

6A.2.2 Debris Generation The debris generation analysis determined the quantities and types of debris generated by the spectrum of LOCA scenarios based on guidance provided in References 4 and 5. The analysis determined the Zone of Influence (ZOI) for various types of insulation and coatings found in containment and the quantities of debris within the ZOI for each LOCA scenario listed in Table 6A-1.

The inventory of insulation in BVPS-2 containment was established through a series of containment walkdowns and evaluation of plant drawings and documentation.

Most insulation in the vicinity of high-energy RCS piping is Reflective Metallic Insulation (RMI). This insulation consists of concentric layers of stainless steel foils encased within stainless-steel jacketing. Debris formed from RMI will sink in the recirculation pool and has a negligible impact on sump strainer performance, as the strainers are raised above the containment floor.

6A-6

BVPS-2 UFSAR Rev. 26 Insulation installed in BVPS-2 containment that impacts sump strainer performance includes the following:

Calcium Silicate Microporous insulation - includes Min-K insulation installed on piping and Microtherm insulation installed on the Reactor Vessel.

Temp-Mat high-density fiberglass insulation Thermal Wrap and Thermal Insulating Wool low-density fiberglass insulations.

These insulating materials are included in the list of restricted materials in containment.

Calcium silicate and microporous insulation are assumed to fail as fine particulates that remain suspended in the recirculation fluid.

Fiberglass insulations are assumed to fail as a combination of fine debris, small pieces, and/or large pieces. Fine debris consists of individual fibers suspended in the recirculation fluid. Pieces of insulation 4 inches and larger are considered large pieces. Pieces of insulation smaller than 4 inches are considered small pieces. The distribution of fine, small, and large piece fibrous insulation debris is conservative with respect to that recommended by References 4 and 5 and is aligned with the results of industry testing.

Analyzed limits for insulation within a LOCA ZOI are provided in Tables 6A-4 and 6A-5.

Most of the coatings used inside containment are qualified for design basis accident environments and will remain intact following a LOCA.

Qualified coatings installed within the ZOI of a LOCA are dislodged by the hydraulic forces of the break and fail as fine particulate debris.

Coatings not qualified for the post-DBA environment are assumed to delaminate and transport to the containment sump strainers. Unqualified coatings include both commercial grade coatings not tested to DBA conditions and degraded qualified coatings. An unqualified coatings inventory was established through a series of containment walkdowns and evaluation of plant documentation. The analyzed limits for unqualified coatings in containment are provided in Table 6A-3.

6A.2.3 Latent Debris and Foreign Materials Latent debris consists of unintended dirt, dust, lint, etc., that are already present in containment prior to the LOCA. This debris was quantified through debris samples taken from surfaces in various locations and orientations in containment. The average debris density of the samples was extrapolated over all surfaces in containment to obtain the total latent debris load. Latent debris is considered 85% particulate and 15%

fibrous in accordance with References 4 and 5. The analyzed limit for latent debris in containment is provided in Table 6A-3.

Thin impermeable materials, such as tags, labels, and tape that are not qualified for the containment environment are assumed to delaminate and 6A-7

BVPS-2 UFSAR Rev. 26 transport to the containment sump, where they will block a portion of the strainer thereby reducing its effective surface area. Containment walkdowns were performed to determine the surface area of tags, labels, and tape in containment, with a 30% margin included to account for those not visible from the accessible areas of the walkdown. Plant documentation was used to justify the environmental qualification of various types of tags and labels identified in the walkdown. The analyzed limit for thin impermeable materials is provided in Table 6A-3.

6A.2.4 Debris Transport to the Sump Strainers 100% of accident-generated and latent debris sources that affect strainer performance are assumed to transport to the containment sump strainers with the exception of small and large piece fibrous insulation debris.

This is conservative as all types of debris are expected, to some extent, to be held up from upstream obstacles such as floor grating and curbs and to settle in inactive regions of the recirculation pool.

Large piece debris does not transport to the recirculation pool as it is too large to pass through floor grating. The debris transport analysis concludes that approximately 20% of small piece fibrous debris generated from an RCS loop break will be held up on floor grating and not reach the sump strainers. The analyses of all other break scenarios conservatively assumes that 100% of small piece fibrous debris is transported to the sump strainers.

A separate debris transport analysis of the RMI foil distribution throughout containment was performed to evaluate the potential for RMI foils to accumulate at choke points upstream of the sump strainers.

Section 6A.2.7 discusses the impact of the RMI accumulation at these locations.

6A.2.5 Sump Strainer Evaluations Strainer Head Loss and NPSH The sump strainers are sized to prevent the formation of a contiguous debris bed of sufficient thickness to induce excessive head loss that challenges the NPSH requirements of the downstream pumps. This was verified through laboratory testing of a set of prototype top-hat strainer modules. The strainer modules were installed in a test tank through which debris-laden water was recirculated. The scaled quantities and combinations of debris added to the tank for two of these tests bounds the calculated debris loads for the spectrum of BVPS-2 LOCA scenarios. One test was performed to bound the debris loads resulting from a break of the RCS loop piping inside the reactor cavity, while a second test bounds the debris loads of all other LOCA scenarios. For this reason, a separate set of debris limits are established for breaks located in the reactor cavity.

During the test, head loss was measured across the debris bed at varying flow rates. The maximum tested flow rate is equivalent to 13,700 gpm of containment sump flow. Test data was used to develop a correlation of head 6A-8

BVPS-2 UFSAR Rev. 26 loss versus flow rate and temperature, which is utilized by the MAAP-DBA program to determine strainer head loss and pump NPSH at any given time during the LOCA scenario. NPSH analysis of the Recirculation Spray Pumps is described in Section 6.2.2.3.2 and results are provided in Table 6.2-59.

Vortex Analysis and Minimum Submergence Testing was performed to verify no vortex formation will occur for the strainer modules closest to the pump suction. These modules will experience disproportionately high flow rates prior to formation of a debris bed and are more susceptible to vortex formation during this time.

Test observations confirmed that no vortex formation occurs at twice the maximum screen approach velocity when the water level is at the top of the strainers.

The minimum required strainer submergence during operation of the Recirculation Spray Pumps is 6 inches above the top of the top-hat strainers. The containment water level is tracked by the MAAP-DBA program as part of the LOCA containment analysis. The minimum strainer submergence is 8.87 inches for a 2-inch break located at the top of the pressurizer and therefore meets minimum submergence requirements.

Void Formation Analysis A void formation analysis was performed for the flow path between the sump strainers and pump impellers. Head losses from the debris bed and downstream flow path may result in void formation due to deaeration of dissolved gases or flashing of the recirculation fluid to steam.

Regulatory Guide 1.82 (Reference 6) requires the void fraction to be less than two percent and a correction factor for NPSH is required be applied for void fractions between zero and two percent. The MAAP-DBA analysis uses a void fraction of 0.3%, resulting in a 15% increase in NPSH required. This increase is reflected in the NPSH required values provided in Section 6.2.2.3.2. The void formation analysis concludes that the void fraction remains less than 0.3% at all points in the flow path between the sump strainers and pump impellers for all LOCA scenarios.

Structural Analysis The code used to design the BVPS-2 containment sump strainer assembly is the AISC Specification for the Design, Fabrication, and Erection of Structural Steel - Seventh Edition. The AISC code does not provide reduction in strength due to elevated temperatures. Therefore, the material property values used at elevated temperatures are from ASME Section III, 1971 and 1974 Editions. Stud material properties for the top-hats are from ASME Section III, 1984.

The load combinations representing accident conditions account for loads from strainer differential pressure, weight of debris, and hydrodynamic loads due to sloshing of the recirculation pool during an earthquake. The strainers are isolated from high energy systems by major structural features and are protected from dynamic effects such as pipe whip, jet impingement, and missile impact due to their location.

6A-9

BVPS-2 UFSAR Rev. 26 6A.2.6 Debris Source Term Reduction The Containment Coatings Inspection and Assessment Program inspects one half of the accessible areas of the containment building during each refueling outage to identify and repair degraded coatings.

The Containment Cleaning Program cleans one fourth of the containment building during each refueling outage. This ensures the quantity of latent debris in containment remains less than that assessed during the containment walkdown performed prior to the start of the program.

The Plant Labeling and Tagging Program restricts the use of tags and labels to those qualified for the post-LOCA containment environment.

Administrative control of restricted materials in containment covers several materials that adversely affect sump strainer performance. These include aluminum, unqualified coatings, insulation, and Benelex neutron shielding. All modifications that introduce these materials into containment require engineering approval through the Engineering Change Process. The quantities of each of these materials is tracked by engineering.

A containment closeout inspection is performed prior to startup from a refueling outage. The inspection identifies and removes foreign material in containment that remains after completion of outage activities.

6A.2.7 Upstream Effects of Debris Accumulation Debris accumulation at choke points in the upstream flow path to the sump can potentially reduce the volume of water that reaches the sump due to water holdup, reducing strainer submergence. The potential for upstream debris accumulation was evaluated inside the reactor cavity. Debris blockage of both the gap between the reactor vessel and neutron shield tank and the gaps in the RCS loop piping penetrations would need to occur to produce significant water holdup. The analysis concludes no debris induced water holdup occurs in the reactor cavity beyond that assumed in the containment analysis.

6A.2.8 Downstream Effects - Components and Systems The downstream impact of containment sump debris on the performance of the ECCS and RSS flow path components was evaluated using the guidance of Westinghouse WCAP-16406-P-A, Evaluation of Downstream Effects in Support of GSI-191, Revision 1 (Reference 7) and its associated SER, dated December 20, 2007 (Reference 8). The effects of debris ingested through the containment sump strainers during recirculation mode include erosive wear of stationary surfaces, abrasive wear of rotating surfaces, and potential blockage of flow paths.

6A-10

BVPS-2 UFSAR Rev. 26 The smallest clearance in the downstream flow path is 0.128 inches. No flow path blockage is expected with a screen perforation diameter of 3/32 (0.0938) inches.

Erosive wear of valves, spray nozzles, orifices, heat exchangers, and piping was evaluated and found to meet the acceptance criteria of Reference 5 for the mission time of 30 days.

The high head safety injection pumps were evaluated for hydraulic performance, mechanical seal performance, and a rotor dynamic (vibration) analysis based on debris-induced wear of the wear rings and pressure reducing sleeves. Hydraulic and mechanical seal performance were acceptable per WCAP-16406-P-A requirements. The results of the rotor dynamic analysis meet the acceptance criteria in American Petroleum Institute Standard API 610, as referred to in Reference 7, for the mission time of 30 days.

A wear analysis of the RSS pumps was performed in accordance with Reference 7. The wear results for all components evaluated were found to be within the acceptance criteria for the mission time of 30 days. With relatively low wear, hydraulic performance of the pumps was evaluated as acceptable. The seals for these pumps are not exposed to debris-laden fluid.

6A.2.9 Downstream Effects - Fuel and Vessel Methods and results contained in WCAP-17788-P, Volume 1, Revision 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090) (Reference 9) were used to evaluate the accumulation of fiber inside the reactor vessel. During the post-LOCA sump recirculation phase, debris ingested by the ECCS could accumulate at the reactor core inlet or inside the heated core, potentially challenging long-term core cooling.

The quantity of fiber accumulation inside the reactor vessel was calculated for the worst-case hot leg break scenario and compared to the in-vessel debris limits defined in Reference 9.

Prototype strainer bypass testing was performed to determine the fraction of fibrous debris that will pass through the containment sump strainers.

Test results indicate that approximately 5.8% of fine fibrous debris that reaches the sump strainers will pass through to downstream systems. This bypass fraction is credited in the calculation of in-vessel debris loading with the results adjusted to compensate for the tested strainer flow rate.

The calculated quantity of fiber that accumulates in the reactor vessel meets the limits defined by Reference 9 thus ensuring long-term core cooling will not be challenged by debris-induced blockage of the ECCS flow path within the reactor vessel.

The NRC has not generically approved WCAP-17788-P, Revision 1 for use. An evaluation was performed by Energy Harbor Nuclear Corp., using the guidance provided in Reference 3, to demonstrate applicability of the methods and results to BVPS-2. The applicability evaluation compares the values of key parameters assumed in the WCAP-17788-P analysis to BVPS-2 6A-11

BVPS-2 UFSAR Rev. 26 specific values. The key parameter comparison is summarized in Table 6A-2.

The evaluation concludes that the WCAP-17788-P, Revision 1 methods and results are applicable to BVPS-2.

The effects of in-vessel downstream chemical effects are discussed in Section 6A.2.10.

6A.2.10 Chemical Effects The containment sump strainers have been sized to account for an increase in head loss due to interactions of the sump water with various materials in containment, most notably metallic aluminum. In the early stages of recirculation, the high temperature/high pH water corrodes aluminum submerged in the sump recirculation pool or exposed to containment spray.

As the sump recirculation pool cools over time, dissolved aluminum precipitates out of solution, forming chemical products that may accumulate on the sump strainers or in the fuel assembly bottom nozzles.

Quantification of chemical precipitates at the sump strainers was performed using the methodology provided by WCAP-16530-NP, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191, Revision 0 (Reference 10), as modified by NRC SER dated December 21, 2007 (Reference 11). The key inputs to this evaluation include the following:

Surface area of aluminum submerged in the sump recirculation pool (submerged aluminum)

Surface area of aluminum exposed to recirculation spray (unsubmerged aluminum)

Sump/Spray pH as a function of time Sump/Spray temperature as a function of time Results of this evaluation were used to determine the quantity of chemical precipitates to be added to the debris mixture used during prototype strainer head loss testing. Tested chemical quantities are based on a sodium hydroxide buffering agent; the sodium tetraborate buffering agent installed at BVPS-2 results in a lower maximum pH and reduced chemical precipitates. The MAAP-DBA containment analysis applies head loss due to chemical effects when the sump temperature decreases to 140°F.

Autoclave chemical effects testing documented in WCAP-17788-P Volume 5, Revision 1 (Reference 12) investigated post-accident corrosion, dissolution, and precipitation reactions to determine the earliest time that chemical products are expected to be generated inside containment.

The testing has demonstrated that the generation of chemical products will be sufficiently delayed by comparing the BVPS-2 post-LOCA plant conditions to the autoclave test conditions. The comparison concluded that the generation of chemical products will be delayed until after the time that complete core inlet blockage can be tolerated and the time of transfer to hot leg recirculation.

6A-12

BVPS-2 UFSAR Rev. 26 Aluminum is considered a restricted material in containment. The inventory of submerged and unsubmerged aluminum is tracked by Engineering. Submerged aluminum is that which is at or below the maximum post-LOCA containment water level, while unsubmerged aluminum is located above this level. The maximum containment water level is elevation 703 ft 4 in. Analyzed limits for submerged and unsubmerged aluminum in containment are provided in Table 6A-3.

6A.3 CONTAINMENT ANALYZED DEBRIS LIMITS Containment accident-generated and transported debris is defined as the quantity of debris calculated to arrive at the containment sump strainers following a LOCA. The analyzed debris limits are the design basis debris loads used as input for the sump strainer head loss, downstream component wear, and core blockage evaluations.

Containment accident-generated and transported debris loads shall remain less than the analyzed debris limits. Maintaining these containment debris loads less than the analyzed debris limits ensures that RSS Pump NPSH requirements, downstream component wear limits, and core cooling flow path availability are not challenged by the limiting LOCA scenarios.

The analyzed debris limits are provided in Tables 6A-3, 6A-4, and 6A-5.

Table 6A-3 provides debris limits that apply to all LOCA scenarios. The additional limits in Table 6A-4 apply only to those LOCAs located in the reactor cavity and correspond to the debris loads used in the strainer head loss testing performed specifically for the reactor vessel nozzle break scenario. The additional limits provided by Table 6A-5 apply to LOCAs located outside the reactor cavity and correspond to the debris loads used in the strainer head loss testing that bounds all LOCA scenarios except for the reactor vessel nozzle break.

6A-13

BVPS-2 UFSAR Rev. 26 References for Appendix 6A

1. NRC Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents for Pressurized-Water Reactors," September 13, 2004. (ADAMS Accession No. ML042360586)

2. NRC Document, Revised Content Guide for Generic Letter 2004-02 Supplemental Responses, dated November 2007 (ADAMS Accession No. ML073110389)

3. Pressurized Water Reactor Owners Group Report PWROG-16073-P, TSTF-567 Implementation Guidance, Evaluation of In-Vessel Debris Effects, Submittal Template for Final Response to Generic Letter 2004-02 and FSAR Changes, Revision 0

4. Nuclear Energy Institute (NEI) document NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology," Revision 0

5. NRC SER dated December 6, 2004, Safety Evaluation by the Office of Nuclear Reactor Regulation Related to NRC Generic Letter 2004-02, Nuclear Energy Institute Guidance Report (Proposed Document Number NEI 04-07), Pressurized Water Reactor Sump Performance Evaluation Methodology " (ADAMS Accession No. ML043280007)

6. Regulatory Guide 1.82, "Water Sources for Long Term Recirculation Cooling Following a Loss of Coolant Accident," Revision 3 (ADAMS Accession No. ML033140347)

7. WCAP-16406-P, Evaluation of Downstream Sump Debris Effects in Support of GSI-191, Revision 1.

8. NRC SER dated December 20, 2007, Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report (TR) WCAP-16406-P, Revision 1, Evaluation of Downstream Sump Debris Effects in Support of GSI-191, Pressurized Water Reactor Owners Group. (ADAMS Accession No. ML073520295)

9. WCAP-17788-P, Volume 1, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090), Revision 1

10. WCAP-16530-NP-A, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191., dated March 2008, (ADAMS Accession No. ML081150379)

11. NRC SER dated December 21, 2007, Final Safety Evaluation by the Office of Nuclear Reactor Regulation Topical Report WCAP-16530-NP, Evaluation of Post-Accident Chemical Effects in Containment Sump Fluids to Support GSI-191. (ADAMS Accession No. ML073520891)

12. WCAP-17788-P, Volume 5, Comprehensive Analysis and Test Program for GSI-191 Closure (PA-SEE-1090)Autoclave Chemical Effects Testing for GSI-191 Long-Term Cooling, Revision 1 6A-14

BVPS-2 UFSAR Rev. 26 Table 6A-1 BOUNDING BREAK LOCATIONS IN EACH AREA OF CONTAINMENT LOCA Scenario Location Break Size Maximum Zone of Influence Radius (1)

RCS Loop Piping RCS Loop Cubicles 31 inches 74 ft (Entire Cubicle)

RCS Loop Piping Reactor Cavity 29 inches 70 ft vertically 90° around vessel on both sides of break (RV acts as a robust barrier)

PZR Surge Line Lower PZR Cubicle 14 inches 34 ft (Entire Cubicle)

PZR PORV inlet Upper PZR Cubicle 6 inches 15 ft piping and PZR Spray Line PZR Spray Line Lower Containment 4 inches 10 ft (1) The maximum ZOI radius for any material determined through industry testing is equal to 28.6 pipe diameters. The ZOI radius for each specific material is provided in the debris generation analysis. The ZOI is typically limited to the cubicle in which the break occurs, as walls and floors act as robust barriers.

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BVPS-2 UFSAR Rev. 26 Table 6A-2 IN-VESSEL DEBRIS EFFECTS KEY PARAMETER EVALUATION Parameter WCAP-17788 Value BVPS-2 Value Evaluation Minimum Sump 20 41.3 Later switchover Switchover Time time results in a (min) lower decay heat at the time of debris arrival, reducing the potential for debris induced core uncovery and heatup.

Maximum Hot Leg 24 6.5 Latest hot leg Switchover Time switchover occurs (hr) well before earliest potential chemical product generation Rated Thermal 3658 2900 Lower thermal power Power (RTP) results in lower decay heat.

Maximum Alternate WCAP-17788, Volume 4, WCAP-17788 Volume 4, When adjusted for Flow Path (AFP) Table 6-1 Table RAI-4.2-24 RTP, AFP resistance Resistance is less than the analyzed value, which increases the effectiveness of the AFP.

Minimum ECCS 8 40 Maximum debris bed Recirculation resistance at the Flow (gpm/FA) core inlet occurs at lower flow rates.

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BVPS-2 UFSAR Rev. 26 Table 6A-3 ANALYZED DEBRIS LIMITS FOR CONTAINMENT - APPLICABLE TO ALL BREAK SCENARIOS Debris Type Units Analyzed Limit Latent Debris pounds 200 Thin Impermeable Materials square feet 500 Exposed Unqualified Coatings(1) pounds 247.8 Exposed Unqualified Coatings(1) cubic feet 1.978 Submerged Aluminum square feet 179 Unsubmerged Aluminum square feet 1,360 (1) Does not include unqualified coatings installed under insulation Page 1 of 1

BVPS-2 UFSAR Rev. 26 Table 6A-4 ANALYZED DEBRIS LIMITS FOR CONTAINMENT -

APPLICABLE TO BREAKS INSIDE THE REACTOR CAVITY Debris Type Units Analyzed Limit Microporous Insulation pounds 340.5 Fibrous Debris(1) pounds 30.0 All Coatings(2) cubic feet 2.622 (1) Fibrous Debris = Fibrous Insulation + (0.15) * (Latent Debris)

(2) Includes the total of qualified and unqualified coatings.

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BVPS-2 UFSAR Rev. 26 Table 6A-5 ANALYZED DEBRIS LIMITS FOR CONTAINMENT -

APPLICABLE TO BREAKS OUTSIDE THE REACTOR CAVITY Debris Type Units Analyzed Limit Calcium Silicate Insulation pounds 96.0 (1)

Fibrous Debris pounds 66.4 Fibrous Debris (1) pounds 56.0 (Upper Pressurizer Cubicle only)

Microporous Insulation pounds 16.0 (1) Fibrous Debris = Fibrous Insulation + (0.15) * (Latent Debris)

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