ML11356A157
| ML11356A157 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 12/22/2011 |
| From: | Entergy Nuclear Operations |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| Shared Package | |
| ML11356A150 | List: |
| References | |
| RAS 21610, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 | |
| Download: ML11356A157 (188) | |
Text
NYSR0014A Revised: December 22, 2011 Indian Point 2 UFSAR Revision 20 OAGI0000215_0001
IP2 UFSAR UPDATE Instructions and Key to Unit 2 UFSAR for Revision 20 Current Changes - Revision 20:
The 2006 revision to the Unit 2 UFSAR is Revision 20. The current changes made in this update are highlighted in a gr~Y§h§t~@~~§t9Kgr9qn~.
Historical Information:
The UFSAR has some parts marked as "Historical Information". Information that is highlighted with a green background is designated as "Historical Information" and is not updated.
The definition of Historical Information means that the information meets the following criteria:
- Information relating to initial plant licensing and start-up that was included in the original FSAR to meet the requirements of 10CFRSO.34(b).
- Information that was accurate at the time the plant was originally built, but is not intended or expected to be updated for the life of the plant (unless required by the Commission).
- Information that is not affected by changes to the plant or its operation.
- Information the does not change with time.
Deleted Information and Figures:
Deleted information that can be shown to have been deleted, will have the word "Deleted" in yellow highlighting.
NRC Orders and License Conditions:
Information relocated to the UFSAR from the Technical Specifications or included by NRC Order, become a Licensing Condition and cannot be removed from the UFSAR without NRC approval. Where these are identified, there will be a [Note.] with bold text and igr~:!i 6.Hi.6.UI6.dril*
Fission Product Barrier Design Basis Limits for IP2:
Information contained in the Unit 2 UFSAR that represent fission produc~PCirriE:!T<:lE:!~i9n basis limits for the plant are highlighted with an aqua shaded background or ~q4~I~n~riQg, if the change is contained in a gray shaded background. These are fundamental to the fission product barrier, located in the UFSAR, numerical and cannot be exceeded, changed or altered without NRC approval.
UFSAR Figures and Plant Drawings:
In an effort to improve the accuracy of the figures in the UFSAR and to reduce redundant work, Revision 19 replaced UFSAR Figures, where possible, with references to the current plant drawings located in Merlin.
Revision 20 has "snapshots" of the referenced plant drawings for use by the NRC and for user convenience. However, if you are doing any plant technical work, you should refer to the most current revision of the official plant drawing in Merlin.
Searching the UFSAR:
The Unit 2 UFSAR is viewed and searched in Adobe Acrobat. The best way to find the information you are looking for is to open the left hand tab named "Bookmarks". This will unfold 1 of 2 Revision 20, 2006 OAG1000021S_0002
Indian Point 2 UFSAR & Plant Drawing Cross Reference 1.2 FIGURES Figure No. Title Plant Drawing Rev Figure 1.2-1 Indian Point Nuclear Generating Units 1 & 2 [Historical]
Figure 1.2-2 Deleted Figure 1.2-3 Deleted (Plot Plan) 9321-1002 8 Figure 1.2-4 Cross Section of Plant [Historical]
Figure 1.2-5 Deleted (Primary Auxiliary Bldg. General Arrangement 9321-2510 7 Plans)
Figure 1.2-6 Deleted (Primary Auxiliary Bldg. General Arrangement 9321-2511 29 Sections)
Figure 1.2-7 Deleted (General Arrangement of Equipment -Control 209812 (Sh.1) Room Unit #1 & #2)
Figure 1.2-7 Deleted (Control Building General Plan) 9321-3052 37 (Sh.2)
Figure 1.2-8 Deleted (Spent Fuel Pit Building General Arrangement) 9321-2514 15 Figure 1.2-9 Deleted (Holdup Tank Building General Arrangement) 9321-2517 12 1.7 FIGURES ure No. Title Plant Drawin ure 1.7-1 Functional Relationshi 1.8 FIGURES Title Plant Drawin Organization Chart WEDCO Reliability Group [Historical]
1.11 FIGURES Figure No. Title Plant Drawing Rev Figure 1.11-1 Ten Percent of Gravity Response Spectra o Figure 1.11-2 Fifteen Percent of Gravity
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Response Spectra o Figure 1.11-3 Fuel Storage Building o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference North-South Model [Historical]
Figure 1.11-4 Fuel Storage Building East-West Model [Historical]
Figure 1.11-5 Indian Point Unit 1 Superheater Building North-South Section Figure 1.11-6 Indian Point Unit 1 Superheater Building East-West Section Figure 1.11-7 Column Line "G" Figure 1.11-8 Representation of Lumped Mass Model of Superheater Building Used in Dynamic Analysis 2.2 FIGURES Figure No. Title Plant Drawing Rev Figure 2.2-1 Aerial Photo of Indian Point Site and Surrounding Area
[Historical]
Figure 2.2-2 Indian Point Building Identification [Historical]
2.3 FIGURES Figure No. Title Plant Drawing Rev Figure 2.3-1 Topographical Map of Indian Point and Surrounding Area [Historical]
2.4 FIGURES Figure No. Title Plant Drawing Rev Figure 2.4-1 Schematic Sector/Zone Diagram Figure 2.4-2 Indian Point Station Ten and Fifty Mile Radius Map Figure 2.4-3 Five Mile Sector/Zone Diagram o Figure 2.4-4 Ten Mile Sector/Zone Diagram
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Figure 2.4-5 Fifty Mile Sector/Zone Diagram o Figure 2.4-6 Map and Description Showing Land Usage o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 2.4-7 Map and Description of the Area Showing Public Utilities Figure 2.4-8 Map and Description of the Area Showing Sewage Systems 2.5 FIGURES Figure No. Title Plant Drawing Rev Figure 2.5-1 Map & Description Showing Location of Sources of Potable
& Industrial Water Supplies & Watershed Areas Figure 2.5-2 Hudson River Drainage Basin 2.6 FIGURES Figure No. Title Plant Drawing Rev Figure 2.6-1 Diurnal Variation of Mean Vector Wind for Virtually Zero Pressure Gradient Conditions Figure 2.6-2 Diurnal Variation of Mean Vector Wind for 24 Hr Periods of Weak Pressure Gradient Conditions Figure 2.6-3 Steadiness of Wind as a Function of Time of Day for Indicated Pressure Gradient Conditions 3.2 FIGURES Figure No. Title Plant Drawing Rev Figure 3.2-1 Typical Power Peaking Factor Versus Axial Offset Figure 3.2-2 Rod Cluster Groups - Cycle 1 [Historical]
Figure 3.2-3 Assembly Average Power & Burnup, Cycle 1 Calculations, BOL, Unrodded Core [Historical]
Figure 3.2-4 Assembly Average Power & Burnup, Cycle 1 Calculations, EOL, Unrodded Core [Historical]
Figure 3.2-5 Assembly Average Power Distribution Cycle 1 Calculations, BOL, Group C4 Inserted [Historical]
o Figure 3.2-6 Assembly Average Power Distribution Cycle 1 Calculations,
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BOL Part-Length Rods In [Historical]
o Figure 3.2-7 Cycle 1 Maximum Fa X Power Versus Axial Height During o Normal Operation [Historical]
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 3.2-7A Deleted - See Unit 2 COLR For Normalized K (Z) - Fq Vs.
Axial Height For Cycle 17 Figure 3.2-8 Burnable Poison & Source Assembly Locations - Cycle 1
[Historical]
Figure 3.2-9 Burnable Poison Rod Locations - Cycle 1 [Historical]
Figure 3.2-10 Moderator Temperature Coefficient Vs Moderator Temperature - EOL, Cycle 1 [Historical]
Figure 3.2-11 Moderator Temperature Coefficient Vs Moderator Temperature - BOL, Cycle 1 Full Power [Historical]
Figure 3.2-12 Moderator Temperature Coefficient Vs Moderator Temperature - BOL, Cycle 1 Zero Power [Historical]
Figure 3.2-13 Doppler Coefficient Vs Effective Fuel Temperature - Cycle 1
[Historical]
Figure 3.2-14 Power Coefficient Vs Percent Power - Cycle 1 [Historical]
Figure 3.2-15 Power Coefficient - Closed Gap Model Figure 3.2-16 Deleted Figure 3.2-17 Deleted Figure 3.2-18 Deleted Figure 3.2-19 Deleted Figure 3.2-20 Deleted Figure 3.2-21 Deleted Figure 3.2-22 Deleted Figure 3.2-23 Deleted Figure 3.2-24 Deleted Figure 3.2-25 Deleted Figure 3.2-26 Deleted Figure 3.2-27 Deleted Figure 3.2-28 Deleted Figure 3.2-29 Deleted Figure 3.2-30 Deleted Figure 3.2-31 Deleted Figure 3.2-32 Deleted o Figure 3.2-33 Deleted
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G) Figure 3.2-34 Deleted o Figure 3.2-35 Deleted o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 3.2-36 Deleted Figure 3.2-37 Deleted Figure 3.2-38 Typical Thermal Conductivity Of U0 2 Figure 3.2-39 High Power Fuel Rod Experimental Program Figure 3.2-40 Typical Comparison Of W-3 Prediction And Uniform Flux Data Figure 3.2-41 Typical W-3 Correlation Probability Distribution Curve Figure 3.2-42 Comparison Of "L" Grid Typical And Thimble Cold Wall Cell Rod Bundle DNB Data For Non-Uniform Axial Heat Flux With Predictions Of W-3 X F'sL Figure 3.2-43 Typical Comparison Of W-3 Correlation With Rod Bundle DNB Data (Simple Grid Without Mixing Vane)
Figure 3.2-44 Typical Comparison Of W-3 Correlation With Rod Bundle DNB Data (Simple Grid With Mixing Vane)
Figure 3.2-44A Typical Measured Versus Predicted Critical Heat Flux-WRB-1 Correlation Figure 3.2-45 Typical Stable Film Boiling Heat Transfer Data And Correlation Figure 3.2-46 Core Cross Section Figure 3.2-47 Reactor Vessel Internals Figure 3.2-48 Core Loading Arrangement - Cycle 1 [Historical]
Figure 3.2-49 Typical Rod Cluster Control Assembly Figure 3.2-50 Rod Control Cluster Assembly Outline Figure 3.2-51 Core Barrel Assembly Figure 3.2-52 Upper Core Support Structure Figure 3.2-53 Guide Tube Assembly Figure 3.2-54 Fuel Assembly And Control Cluster Cross Section - HIPAR, LOPAR, And OFA And VANTAGE+
Figure 3.2-55 HIPAR Fuel Assembly Figure 3.2-56 LOPAR Fuel Assembly Figure 3.2-56A OFA Fuel Assembly Figure 3.2-56B VANTAGE+ Fuel Assembly o Figure 3.2-57 Guide Thimble To Bottom Nozzle Joint
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G) Figure 3.2-58 LOPAR Top Grid To Nozzle Attachment o Figure 3.2-58A OFA And VANTAGE+ Top Grid To Nozzle Attachment o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 3.2-59 Spring Clip Grid Assembly Figure 3.2-60 Mid-Grid Expansion Joint Design Plan View Figure 3.2-61 Elevation View - LOPAR Grid To Thimble Attachment Figure 3.2-61A Elevation View-VANTAGE+ Grid To Thimble Attachment Figure 3.2-61 B Vantage+ Fuel Assembly With Performance+
Enhancements Figure 3.2-61C 15x15 Upgraded Fuel Assembly Figure 3.2-62 Cycle 1 - Neutron Source Locations [Historical]
Figure 3.2-63 HIPAR Burnable Poison Rod Figure 3.2-64 LOPAR Burnable Poison Rod Figure 3.2-65 Control Rod Drive Mechanism Assembl~
Figure 3.2-66 Control Rod Drive Mechanism Schematic Figure 3.2-67 Thimble Location - Fixed Incore Detectors Figure 3.2-68 Cycle 14 Incore Detector, Thermocouple And Flow Mixing Device Locations Figure 3.2-68A GY91@1~g@giQI1AI'l~Fq@IA§§im~IYWQ9§tjql'l§ g~y.
20 Gyql§1§Qqr§Qqmpql1@l'lt§Ad~Fd~§hIB§AWq9§tiQd$
Figure 3.2-68B IIY' Figure 3.2-69 Comparison Of Borosilicate Glass Absorber Rod With WABA Rod Figure 3.2-70 Wet Annular Burnable Absorber Rod 4.2 FIGURES Figure No. Title Plant Drawing Rev Figure 4.2-1 Reactor Coolant System Flow Diagram 9321-2738 116 Figure 4.2-2 Reactor Coolant System Schematic Flow Diagram Figure 4.2-3 Reactor Vessel Figure 4.2-4 Pressurizer Figure 4.2-5 Steam Generator Assembly o Figure 4.2-6 Reactor Coolant Pump
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G) Figure 4.2-7 Reactor Coolant Pump Estimated o Performance Characteristics o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 4.2-8 Flywheel Figure 4.2-9 Reactor Coolant Pump Flywheel Tangential Stress vs. Radius Figure 4.2-10 Pressurizer Relief Tank Figure 4.2-11 Identification & Location of Beltline Region Material for the Indian Point Unit 2 Reactor Vessel Figure 4.2-12 Reactor Vessel Level Instrumentation System Flow Diagram 208798 17 APPENDIX 4C FIGURES Figure No. Title Plant Drawing Rev Figure 4c-1 Primary Nozzle Combustion Engineering Reactor Vessel Figure 4c-2 Primary Nozzle Tampa Steam Generators Figure 4c-3 Spray or Surge Nozzle Tampa Pressurizer 5.1 FIGURES Figure No. Title Plant Drawing Rev Figure 5.1-1 Containment Structure Figure 5.1-2 Containment Building General Arrangement Plans, Sheet 1 9321-2501 25 Figure 5.1-3 Containment Building General Arrangement Plans, Sheet 2 9321-2502 19 Figure 5.1-4 Containment Building General Arrangement Plans, Sheet 3 9321-2503 21 Figure 5.1-5 Containment Building General Arrangement Elevation, 9321-2506 10 Sheet 1 Figure 5.1-6 Containment Building General Arrangement Elevation, 9321-2507 14 Sheet 2 Figure 5.1-7 Containment Building General Arrangement Elevation, 9321-2508 15 Sheet 3 Figure 5.1-8 Deleted Figure 5.1-9 Deleted Figure 5.1-10 Deleted o Figure 5.1-11 Cylinder and Dome-Load Condition (A) - 1.5P
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G) Figure 5.1-12 Cylinder and Dome-Load Condition (B) - 1.25P o Figure 5.1-13 Cylinder and Dome-Load Condition (C) - 1.0P o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 5.1-14 Loading Diagram in Mat-Load Condition (A) - 1.5P Figure 5.1-15 Loading Diagram in Mat-Load Condition (B) - 1.25P Figure 5.1-16 Loading Diagram in Mat-Load Condition (C) - 1.0P Figure 5.1-17 Weld Stud Connection at Panel Low Point Figure 5.1-18 Weld Stud Connection At Panel Low Point Figure 5.1-19 Weld Stud Connection at Panel Center Figure 5.1-20 Wall Section Figure 5.1-21 Cylinder Base Slab Liner Juncture Figure 5.1-22 Typical Base Mat Liner Detail Figure 5.1-23 Base Slab Reinforcing Detail Figure 5.1-24 Reactor Cavity Pit Figure 5.1-25 Equipment Hatch Personnel Lock, Main Steam and Feedwater, Air Purge - Rebar Figure 5.1-26 Torsional Effects Figure 5.1-27 Typical Electrical Penetration Figure 5.1-28 CONAX Penetrations - Outside Containment Weld Figure 5.1-29 CONAX Penetrations -Inside Containment Weld Figure 5.1-30 Typical Piping Penetration Figure 5.1-31 Fuel Transfer Tube Penetration (Conceptual Drawing)
Figure 5.1-32 Containment-Stresses on Penetrations and Liner - Sheet 6 Figure 5.1-33 Containment-Stresses on Penetrations and Liner - Sheet 7 Figure 5.1-34 Assumed Pipe Rupture Accident Break Locations Figure 5.1-35 Steam Generator Support-Section 1-1 Figure 5.1-36 Steam Generator Support-Section 2-2 Figure 5.1-37 Steam Generator Support-Section 3-3 Figure 5.1-38 Steam Generator Support-Section 4-4 Figure 5.1-39 Steam Generator Support-Plan Location Elevation 60 & 63 Figure 5.1-40 Steam Generator Support-Plan Location Elevation 60 & 63 Figure 5.1-41 Pump Support-Section 2-2 and 3-3 Figure 5.1-42 Pump Support-Section 3-3 Figure 5.1-43 Isometric View-Steam Generator Support Figure 5.1-44 Isometric View-Reactor Coolant Pump Support o Figure 5.1-45 Maximum Forces Acting on a Reactor Vessel Support
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G) Figure 5.1-46 Plan View 60 Ft-O In.
o Figure 5.1-47 Typical Layer-Reactor Ring o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 5.1-48 Section 5-5 Figure 5.1-49 Section 18-18 Figure 5.1-50 Plan View at Elevation 19 Ft-7 In.
Figure 5.1-51 Section A-A and Section 8-8 Figure 5.1-52 Deleted Figure 5.1-53 Containment Equipment Hatch Strain Gauge Test Locations Figure 5.1-54 Containment Temporary Opening in NW Quadrant Strain Gauge Test Locations Figure 5.1-55 Containment Strain Gauge Test Locations Figure 5.1-56 Containment Proof Test Gross Deformation Measurements 5.2 FIGURES Figure No. Title Plant Drawing Rev Figure 5.2-1 Containment Isolation System Penetration Schematics Figure 5.2-2 Containment Isolation System Penetration Schematics Figure 5.2-3 Containment Isolation System Penetration Schematics Figure 5.2-4 Containment Isolation System Penetration Schematics Figure 5.2-5 Containment Isolation System Penetration Schematics Figure 5.2-6 Containment Isolation System Penetration Schematics Figure 5.2-7 Containment Isolation System Penetration Schematics Figure 5.2-8 Containment Isolation System Penetration Schematics Figure 5.2-9 Containment Isolation System Penetration Schematics Figure 5.2-10 Containment Isolation System Penetration Schematics Figure 5.2-11 Containment Isolation System Penetration Schematics Figure 5.2-12 Containment Isolation System Penetration Schematics Figure 5.2-13 Containment Isolation System Penetration Schematics Figure 5.2-14 Containment Isolation System Penetration Schematics Figure 5.2-15 Containment Isolation System Penetration Schematics Figure 5.2-16 Containment Isolation System Penetration Schematics Figure 5.2-17 Containment Isolation System Penetration Schematics Figure 5.2-18 Containment Isolation System Penetration Schematics o Figure 5.2-19 Containment Isolation System Penetration Schematics
>> Figure 5.2-20 G) Containment Isolation System Penetration Schematics o Figure 5.2-21 Containment Isolation System Penetration Schematics 235296 &7 o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 5.2-22 Containment Isolation System Penetration Schematics Figure 5.2-23 Containment Isolation System Penetration Schematics Figure 5.2-24 Containment Isolation System Penetration Schematics Figure 5.2-25 Containment Isolation System Penetration Schematics Figure 5.2-26 Containment Isolation System Penetration Schematics Figure 5.2-27 Containment Isolation System Penetration Schematics Figure 5.2-28 Containment Isolation System Penetration Schematics Figure 5.2-29 Containment Isolation System Penetration Schematics 5.3 FIGURES Figure No. Title Plant Drawing Figure 5.3-1 Containment Cooling and Ventilation System 9321-4022 6.2 FIGURES Figure No. Title Plant Drawing Rev Figure 6.2-1 Safety Injection System - Flow Diagram, Sheet 1 9321-2735 136 Sh.1 Figure 6.2-1 Safety Injection System - Flow Diagram, Sheet 2 235296 67 Sh.2 Figure 6.2-2 Primary Auxiliary Building Safety Injection System Piping- g~V.
Schematic Plan 20 Figure 6.2-3 Primary Auxiliary Building Safety Injection System Piping-Schematic Elevations Figure 6.2-4 Containment Building Safety Injection System Piping-Plan Figure 6.2-5 Containment Building Safety Injection System Piping- R$%
Elevation 20 Figure 6.2-6 Safety Injection Pump Performance Figure 6.2-7 Residual Heat Removal Pump Performance Figure 6.2-8 Recirculation Pump Performance Figure 6.2-9 Deleted o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference 6.3 FIGURES Title Plant Drawin Containment Spray Pump Performance Characteristics 6.4 FIGURES Figure No. Title Plant Drawing Rev Figure 6.4-1 Deleted Figure 6.4-2 Deleted Figure 6.4-3 Containment Building Air Recirculation Fan cooler Filter Unit 9321-4026 10
- Plan and Section Figure 6.4-4 Deleted 6.5 FIGURES Figure No. Title Plant Drawing Rev Figure 6.5-1 Isolation Valve Seal- Water System - Flow Diagram 9321-2746 42 Figure 6.5-2 Double Disk Isolation Valve with Seal-Water Injection 6.6 FIGURES Figure No. Title Plant Drawing Rev Figure 6.6-1 Weld Channel and Penetration Pressurization System - 9321-2726 75 Flow Diagram 6.8 FIGURES Figure No. Title Plant Drawing Rev Figure 6.8-1 Passive Hydrogen Recombiners Figure 6.8-2 Containment Hydrogen vs. Time Post-LOCA 9321-2568 14 o Sh.1
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Figure 6.8-2 Containment Hydrogen vs. Time Post-LOCA 9321-2569 15 o Sh.2 o Figure 6.8-3 Postaccident Containment Venting System - Flow Diagram 208879 25 o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 6.8-4 Postaccident Containment Sampling System - Flow 208479 22 Diagram Figure 6.8-5 Deleted Figure 6.8-6 Deleted 6C FIGURES Figure No. Title Plant Drawing Rev Figure 6C-1 Containment Atmosphere Temperature Design Bases Safety Injection Figure 6C-2 Indian Point Unit 2 Postaccident Containment Materials Design Figure 6C-3 Postaccident Core Materials Design Conditions Figure 6C-4 Indian Point Unit 2 Containment Atmosphere Direct Gamma Dose Rate Figure 6C-5 Indian Point Unit 2 Containment Atmosphere Integrated Gamma Dose Rate Figure 6C-6 Titration Curve for TSP in Boric Acid Solution Figure 6C-7 Temperature-Concentration Relation for Caustic Corrosion of Austenitic Stainless Steel Figure 6C-8 Aluminum Corrosion in Design-Basis-Accident Environment Figure 6C-9 Aluminum Corrosion as a Function of pH Figure 6C-10 Solubility of Aluminum Corrosion Products as a Function of pH at 77°F and 150°F Figure 6C-11 Boron Loss from Boron-Concrete Reaction Following a Design-Basis Accident Figure 6C-12 Containment Pressure Transient During Blowdown Phase vs. Time 60 FIGURES Figure No. Title Plant Drawing Rev o Figure 60-1 Temperature - Concentration Relations for Caustic
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Corrosion of Austenitic Stainless Steel o Figure 60-2 Effect of Carbon Dioxide on Corrosion of Iron in NaOH o Solution o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference 7.1 FIGURES Figure No. Title Plant Drawing Rev Figure 7.1-1 Environmental Conditions For Equipment Testing - Pressure Vs Time Figure 7.1-2 Environmental Conditions For Equipment Temperature Vs Time Figure 7.1-3 Instantaneous Gamma Dose Rate Inside The Containment As A Function Of Time After Release - Tid - 14844 Model Figure 7.1-4 Integrated Gamma Dose Level Inside The Containment As A Function Of Time After Release - Tid - 14844 Model Figure 7.1-5 Deleted Figure 7.1-6 Deleted Figure 7.1-7 Deleted Figure 7.1-8 Deleted 7.2 FIGURES Figure No. Title Plant Drawing Rev Figure 7.2-1 Index And Symbols - Logic Diagram 225094 03 Figure 7.2-2 Reactor Trip Signals - Logic Diagram 225095 09 Figure 7.2-3 Turbine Trip Signals - Logic Diagram 225096 @6 Figure 7.2-4 6900 Volt Bus Automatic Transfer - Logic Diagram 225097 05 Figure 7.2-5 Nuclear Instrumentation Trip Signals - Logic Diagram 225098 04 Figure 7.2-6 Nuclear Instrumentation Permissives And Blocks - Logic 225099 OS Diagram Figure 7.2-7 Emergency Generator Starting - Logic Diagram 225100 05 Figure 7.2-8 Safeguard Sequence - Logic Diagram 225101 11 Figure 7.2-9 Pressurizer Trip Signal - Logic Diagram 225102 05 Figure 7.2-10 Steam Generator Trip Signals - Logic Diagram 225103 08 Figure 7.2-11 Primary Coolant System Trip Signals And Manual Trip - 225104 06 o Logic Diagram
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Figure 7.2-12 Safeguard Actuation Signals - Logic Diagram 225105 10 o Figure 7.2-13 Feedwater Isolation - Logic Diagram 225106 07 o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 7.2-14 Rod Stops And Turbine Loads Cutbacks - Logic Diagram 225107 09 Figure 7.2-15 Safeguards Actuation Circuitry And Hardware 243318 03 Channel ization Figure 7.2-16 Simplified Diagram For Overall Logic Relay Test Scheme 243319 02 Figure 7.2-17 Analog And Logic Channel Testing 243320 Q$
Figure 7.2-18 Reactor Protection Systems - Block Diagram 243321 03 Figure 7.2-19 Core Coolant Average Temperature Vs Core Power Figure 7.2-20 Pressurizer Level Control And Protection System 243313 01 Figure 7.2-21 Pressurizer Pressure Control And Protection System 243314 03 Figure 7.2-22 STEAM FLOW ~P Vs POWER 243315 01 Figure 7.2-23 Design Philosophy To Achieve Isolation Between Channels Figure 7.2-24 Cable Tunnel - Typical Section 243317 01 Figure 7.2-25 Typical Analog Channel Testing Arrangement 243322 01 Figure 7.2-26 Typical Simplified Control Schematic 243323 02 Figure 7.2-27 Analog Channels 243324 02 Figure 7.2-28 Analog System Symbols 243311 02 Figure 7.2-29 Deleted Figure 7.2-30 Reactor Trip Breaker Actuation Schematic Figure 7.2-31 Deleted Figure 7.2-32 Steam Generator Level Control And Protection System 243328 01 Figure 7.2-33 Illustrations Of Overpower And Temperature ~T Trips High Sh.1 Temperature Operation Figure 7.2-33 Illustrations Of Overpower And Temperature ~T Trips Low Sh.2 Temperature Operation Figure 7.2-34 Tavi~T CONTROL AND PROTECTION SYSTEM 243330 01 7.3 FIGURES Figure No. Title Plant Drawing Rev Figure 7.3-1 Simplified Block Diagram Of Reactor Control Systems Figure 7.3-2 Deleted o 7.4 FIGURES
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G) o I Figure No. I Title I Plant Drawing I Rev o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 7.4-1 Neutron Detectors And Range Of Operation Figure 7.4-2 Nuclear Instrumentation System Figure 7.4-3 Plan View Indicating Detector Location Relative To Core 7.5 FIGURES Figure No. Title Plant Drawing Rev Figure 7.5-1 Reactor Coolant Wide Range Pressure Instrument System -
Flow Diagram 7.6 FIGURES Figure No. Title Plant Drawing Rev Figure 7.6-1 Typical Arrangement Of Moveable Miniature Neutron Flux Detector System Figure 7.6-2 Arrangement Of Incore Flux Detector Figure 7.6-3 Incore Instrumentation - Details 7.7 FIGURES Figure No. Title Plant Drawing Figure 7.7-1 Deleted 8.2 FIGURES Figure No. Title Plant Drawing Rev Figure 8.2-1 Electrical One-Line Diagram 250907 23 Figure 8.2-2 Electrical Power System Diagram 250907 23 Figure 8.2-3 Main One-Line Diagram 208377 14 Figure 8.2-4 345-KV Installation at Buchanan Figure 8.2-5 6900-V One-Line Diagram 231592 18 o Figure 8.2-6 480-V One-Line Diagram
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G) Figure 8.2-7 Single Line Diagram 480-V Motor Control Centers 21, 22, 9321-3004 84 o 23,25,25A o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 8.2-7a Single Line Diagram - 480-V Motor Control Centers 24 and 249956 15 24A Figure 8.2-8 Single Line Diagram - 480-V Motor Control Centers 27 and 9321-3005 109 27A Figure 8.2-9 Single Line Diagram - 480-V Motor Control Centers 28 and 208507 39 210 Figure 8.2-9a Single Line Diagram - 480-V Motor Control Centers 29 and 249955 21 29A Figure 8.2-10 Single Line Diagram - 480-V Motor Control Centers 28A and 208241 45 211 Figure 8.2-11 Single Line Diagram - 480-V Motor Control Centers 26A and 9321-3006 94 26B Figure 8.2-11 a Single Line Diagram - 480-V Motor Control Center 26C 248513 11 Figure 8.2-12 Single Line Diagram - 480-V Motor Control Centers 26AA 208500 43 and 26BB and 120-V AC Panels No.1 and 2 Figure 8.2-13 Single Line Diagram - 118-VAC Instrument Buses No. 21 208502 61 thru 24 Figure 8.2-14 Single Line Diagram - 118-VAC Instrument Buses No. 21A 208503 34-thru 24A Figure 8.2-15 Single Line Diagram - DC System Distribution Panels No. 208501 38 21, 21A, 21 B, 22, and 22A Figure 8.2-16 Single Line Diagram - DC System Power Panels No. 21 thru 9321-3008 &6 24 Figure 8.2-17 Single Line Diagram of Unit Safeguard Channeling and 208376 11 Control Train Development Figure 8.2-18 Cable Tray Separations, Functions, and Routing 208761 03 8.3 FIGURES Figure No. Title Plant Drawing Figure 8.3-1 Deleted o 9.2 FIGURES
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G) o Figure No. I Title I Plant Drawing o Figure 9.2-1 Sh. I Chemical and Volume Control System - Flow Diagram, I 9321-2736 o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference 1 Sheet 1 Figure 9.2-1 Sh. Chemical and Volume Control System - Flow Diagram, 208168 53 2 Sheet 2 Figure 9.2-1 Sh. Chemical and Volume Control System - Flow Diagram, 9321-2737 53 3 Sheet 3 Figure 9.2-1 Sh. Chemical and Volume Control System - Flow Diagram, 235309 33 4 Sheet 4 Figure 9.2-2 Primary Water Makeup System - Flow Diagram 9321-2724 55 9.3 FIGURES Figure No. Title Plant Drawing Rev Figure 9.3-1 Sh. Auxiliary Coolant System - Flow Diagram, Sheet 1 227781 79 1
Figure 9.3-1 Sh. Auxiliary Coolant System - Flow Diagram, Sheet 2 9321-2720 86 2
Figure 9.3-1 Sh. Auxiliary Coolant System - Flow Diagram, Sheet 3 251783 28 3
9.4 FIGURES Figure No. Title Plant Drawing Rev Figure 9.4-1 Sh. Primary Sampling System - Flow Diagram, Sheet 1 9321-2745 58 1
Figure 9.4-1 Sh. Primary Sampling System - Flow Diagram, Sheet 2 227178 14 2
Figure 9.4-2 Secondary Sampling System - Flow Diagram 9321-7020 39 9.5 FIGURES Figure No. Title Plant Drawing Rev Figure 9.5-1 Fuel Transfer System o Figure 9.5-2 Spent Fuel Storage Rack Layout
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Figure 9.5-3 Spent Fuel Storage Cell Region 1 o Figure 9.5-4 Region I Cell Cross-Section o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference I Figure 9.5-5 I Region II Cross-Section 9.6 FIGURES Figure No. Title Plant Drawing Rev Figure 9.6-1 Sh. Service Water System - Flow Diagram, Sheet 1 9321-2722 115 1
Figure 9.6-1 Sh. Service Water System - Flow Diagram, Sheet 2 209762 66 2
Figures 9.6-2 Deleted Through 9.6-4 Figure 9.6-5 Sh. City Water System - Flow Diagram, Sheet 1 192505 @8 1
Figure 9.6-5 Sh. City Water System - Flow Diagram, Sheet 2 192506 4d 2
Figure 9.6-5 Sh. City Water System - Flow Diagram, Sheet 3 193183 27 3
Figure 9.6-6 Instrument Air - Flow Diagram 9321-2036 95 Figure 9.6-7 Station Air - Flow Diagram 9321-2035 43 9.9 FIGURES Figure No. Title Plant Drawing Rev Figure 9.9-1 Central Control Room HVAC (Heating, Ventilation, and Air 252665 & 138248 16 Conditioning) 09 10.1 FIGURES Figure No. Title Plant Drawing Rev Figure 10.1-1 Load Heat Balance Diagram at 1,078,200 kWe Figure 10.1-2 Deleted Figure 10.1-2a Deleted o Figure 10.1-3 Deleted
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Figure 10.1-4 Deleted o Figure 10.1-5 Deleted o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference ure 10.1-6 Deleted ure 10.1-7 Load Heat Balance Diaaram at 1,034,072 kWe 10.2 FIGURES Figure No. Title Plant Drawing Rev Figure 10.2-1 Main Steam Flow Diagram, Sheet 1 227780 52 Sh.1 Figure 10.2-1 Main Steam Flow Diagram, Sheet 2 9321-2017 83 Sh.2 Figure 10.2-1 Main Steam Flow Diagram, Sheet 3 235308 47 Sh.3 Figure 10.2-2 Turbine Generator Building General Arrangement, 9321-2004 09 Operating Floor Figure 10.2-3 Turbine Generator Building General Arrangement, Cross 9321-2008 07 Section Figure 10.2-4 Condenser Air Removal and Water Box Priming - Flow 9321-2025 55 Diagram Figure 10.2-5 Condensate and Boiler Feed Pump Suction - Flow Diagram, 9321-2018 140 Sh.1 Sheet 1 Figure 10.2-5 Condensate and Boiler Feed Pump Suction Flow Diagram, 235307 31 Sh.2 Sheet 2 Figure 10.2-6 Deleted Sh.1 Figure 10.2-6 Deleted Sh.2 Figure 10.2-7 Boiler Feedwater Flow Diagram 9321-2019 112 Figure 10.2-8 Steam Turbine-Driven Auxiliary Feedwater Pump Estimated Performance Characteristics Figure 10.2-9 Motor-Driven Auxiliary Feedwater Pump Estimated Performance Characteristics o 11.1 FIGURES
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G) o Figure No. I Title Plant Drawing o Figure 11.1-1 I Waste Disposal System Process Flow Diagram, Sheet 1 9321-2719 o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Sh.1 Figure 11.1-1 Waste Disposal System Process Flow Diagram, Sheet 2 9321-2730 71 Sh.2 11.2 FIGURES Figure No. Title Plant Drawing Rev Figure 11.2-1 Deleted Figure 11.2-2 Deleted Figure 11.2-3 Deleted Figure 11.2-4 Deleted Figure 11.2-5 Deleted Figure 11.2-6 Deleted 118 FIGURES Figure No. Title Plant Drawing Rev Figure 11 B-1 lodine-131 Concentration vs Days After Burst Release From Indian Point for 1 Curie Release Figure 11 B-2 lodin-131 Concentration vs Chelsea vs Days After Burst Release From Indian Point for 1 Curie Release Figure 11 B-3 Maximum Concentration vs Distance Upstream for 1 Curie Release Figure 11 B-4 Maximum Concentration at Chelsea vs Half-Life for 1 Curie Release Figure 11 B-5 Time to Reach Peak Concentration at Chelsea vs Half-Life for 1 Curie Release 11D FIGURES Figure No. Title Plant Drawing Rev Figure 110-1 Deleted o Figure 110-2 Deleted
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Indian Point 2 UFSAR & Plant Drawing Cross Reference 11E FIGURES Figure No. Title Plant Drawing Rev Figure 11 E-1 Deleted Figure 11 E-2 Deleted 12.1 FIGURES Figure No. Title Plant Drawing Rev Figure 12.1-1 Deleted Figure 12.1-2 Deleted 14.0 FIGURES Figure No. Plant Drawing Figure 14.0-1 Reactivity Insertion vs. Time for Reactor Trip 14.1 FIGURES Figure No. Title Plant Drawing Rev Figure 14.1-1 Uncontrolled RCCA Withdrawal From A Subcritical Condition Nuclear Power vs. Time Figure 14.1-2 Uncontrolled RCCA Withdrawal From A Subcritical Condition Heat Flux vs. Time, Avg. Channel Figure 14.1-3 Uncontrolled RCCA Withdrawal From A Subcritical Condition Fuel Average Temperature vs. Time At Hot Spot Figure 14.1-4 Uncontrolled RCCA Withdrawal From a Subcritical Condition Clad Inner Temperature vs. Time At Hot Spot Figure 14.1-5 Uncontrolled RCCA Bank Withdrawal From Full Power With Minimum Reactivity Feedback (70 pcm/sec Withdrawal Rate)
Figure 14.1-6 Uncontrolled RCCA Bank Withdrawal From Full Power With o Minimum Reactivity Feedback (70 pcm/sec Withdrawal
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Rate) o Figure 14.1-7 Uncontrolled RCCA Bank Withdrawal From Full Power With o Minimum Reactivity Feedback (70 pcm/sec Withdrawal o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Rate)
Figure 14.1-8 Uncontrolled RCCA Bank Withdrawal From Full Power With Minimum Reactivity Feedback (1 pcm/sec Withdrawal Rate)
Figure 14.1-9 Uncontrolled RCCA Bank Withdrawal From Full Power With Minimum Reactivity Feedback (1 pcm/sec Withdrawal Rate)
Figure 14.1-10 Uncontrolled RCCA Bank Withdrawal From Full Power With Minimum Reactivity Feedback (1 pcm/sec Withdrawal Rate)
Figure 14.1-11 Minimum DNBR Versus Reactivity Insertion Rate, Rod Withdrawal From 100 Percent Power Figure 14.1-12 Minimum DNBR Versus Reactivity Insertion Rate, Rod Withdrawal From 60 Percent Power Figure 14.1-13 Minimum DNBR Versus Reactivity Insertion Rate, Rod Withdrawal From 10 Percent Power Figure 14.1-14 Dropped Rod Incident Manual Rod Control Nuclear Power and Core Heat Flux at BOL (Small Negative MTC) for Dropped RCCA of Worth - 400 PCM Figure 14.1-15 Dropped Rod Incident Manual Rod Control Core Average and Vessel Inlet Temperature at BOL (Small Negative MTC) for Dropped RCCA of Worth - 400 PCM Figure 14.1-16 Dropped Rod Incident Manual Rod Control Pressurizer Pressure at BOL (Small Negative MTC) for Dropped RCCA Worth of 400 PCM Figure 14.1-16a Deleted Figure 14.1-17 Dropped Rod Incident Manual Rod Control Nuclear Power and Core Heat Flux at EOL (Large Negative MTC) for Dropped RCCA of Worth - 400 PCM Figure 14.1-18 Dropped Rod Incident Manual Rod Control Core Average and Vessel Inlet Temperature at EOL (Large Negative MTC) for Dropped RCCA of Worth - 400 PCM Figure 14.1-19 Dropped Rod Incident Manual Rod Control Pressurizer Pressure at EOL (Large Negative MTC)for Dropped RCCA Worth of 400 PCM Figure 14.1-20 Loss of One Pump Out of Four Nuclear Power and Core o
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Heat Flux vs. Time o Figure 14.1-21 Loss of One Pump Out of Four Total Core Flow and Faulted o Loop Flow vs. Time o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 14.1-22 Loss of One Pump Out of Four Pressurizer Pressure and DNBR vs. Time Figure 14.1-23 Four Pump Loss of Flow - Undervoltage Nuclear Power and Core Heat Flux vs. Time Figure 14.1-24 Four Pump Loss of Flow - Undervoltage Total Core Flow and RCS Loop Flow vs. Time Figure 14.1-25 Four Pump Loss of Flow - Undervoltage Pressurizer Pressure and DNBR vs. Time Figure 14.1-26 Four Pump Loss of Flow - Underfrequency Nuclear Power and Heat Flux vs. Time Figure 14.1-27 Four Pump Loss of Flow - Underfrequency Total Core Flow and RCS Loop Flow vs. Time Figure 14.1-28 Four Pump Loss of Flow Underfrequency Pressurizer Pressure and DNBR vs. Time Figure 14.1-29 Locked Rotor Nuclear Power and RCS Pressure vs. Time Figure 14.1-30 Locked Rotor Total Core Flow and Faulted Loop Flow vs.
Time Figure 14.1-30a Locked Rotor Fuel Clad Inner Temperature vs. Time Figure 14.1-31 Loss of Load With Pressurizer Spray and PORV - Nuclear Power and Pressurizer Pressure vs. Time Figure 14.1-32 Loss of Load With Pressurizer Spray and PORV - Average Coolant Temperature and Pressurizer Water Volume vs.
Time Figure 14.1-33 Loss of Load With Pressurizer Spray and PORV - ON BR vs.
Time Figure 14.1-34 Deleted Figure 14.1-35 Deleted Figure 14.1-36 Deleted Figure 14.1-37 Loss of Load Without Pressurizer Spray and Power Operated Relief Valves - Nuclear Power and Pressurizer Pressure vs. Time Figure 14.1-38 Loss of Load Without Pressurizer Spray and Power o Operated Relief Valves - Average Coolant Temperature and
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Pressurizer Water Volume vs. Time o Figure 14.1-39 Loss of Load Without Pressurizer Spray and Power o Operated Relief Valves - Steam Pressure vs. Time o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 14.1-40 Deleted Figure 14.1-41 Deleted Figure 14.1-42 Deleted Figure 14.1-43 Loss of Normal Feedwater, Offsite Power Available, High Sh.1 Tavg Program, Pressurizer Pressure and Pressurizer Water Volume vs. Time Figure 14.1-43 Loss of Normal Feedwater, Offsite Power Available High Sh.2 Tav~ Program, Nuclear Power and Core Heat Flux vs. Time Figure 14.1-43 Loss of Normal Feedwater, Offsite Power Available, High Sh.3 Tavg Program, Loop 21 Temperature and Loop 23 Temperature vs. Time Figure 14.1-43 Loss of Normal Feedwater, Offsite Power Available, High Sh.4 Tavg Program, Steam Generator 21 Pressure and Steam Generator 23 Pressure vs. Time Figure 14.1-43 Loss of Normal Feedwater, Offsite Power Available, High Sh.5 Tavg Program, Total RCS Flow and Pressurizer Relief vs.
Time Figure 14.1-44 Deleted Sh.1 Figure 14.1-44 Deleted Sh.2 Figure 14.1-44 Deleted Sh.3 Figure 14.1-44 Deleted Sh.4 Figure 14.1-44 Deleted Sh.5 Figure 14.1-45 Feedwater System Malfunction Excessive Feedwater Flow -
Sh.1 HFP Conditions Manual Rod Control Nuclear Power, and Core Heat Flux vs. Time Figure 14.1-45 Feedwater System Malfunction Excessive Feedwater Flow -
Sh.2 HFP Conditions Manual Rod Control Pressurizer Pressure and DNBR vs. Time o
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Figure 14.1-45 Feedwater System Malfunction Excessive Feedwater Flow -
o Sh.3 HFP Conditions Manual Rod Control, Loop Delta - T, and o Core Tava vs. Time o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 14.1-46 Deleted
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Sh.1 Figure 14.1-46 Deteted
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Sh.2 Figure 14.1-47 Deleted
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Sh.1 Figure 14.1-47 Deleted
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Sh.2 Figure 14.1-48 Deleted
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Sh.1 Figure 14.1-48 Deleted Sh.2 Figure 14.1-49 Deleted Sh.1 Figure 14.1-49 Deleted
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Sh.2 Figure 14.1-50 Loss of all AC Power, High Tavg Program, Pressurizer Sh.1 Pressure and Water Volume vs. Time Figure 14.1-50 Loss of all AC Power, High Tavg Program, Nuclear Power Sh.2 and Core Heat Flux vs. Time Figure 14.1-50 Loss of all AC Power To The Station Auxilaries, High Tavg Sh.3 Program, Loop 21 Temperature and Loop 23 Temperature Figure 14.1-50 Loss of all AC Power To The Station Auxilaries, High Tavg Sh.4 Program, Steam Generator 21 Pressure and Steam Generator 23 Pressure Figure 14.1-50 Loss of all AC Power To The Station Auxilaries, High Tavg Sh.5 Program, Total RCS Flow and Pressurizer Relief vs. Time Figure 14.1-51 Deleted
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Sh.1 Figure 14.1-51 Deleted
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Sh.2 Figure 14.1-51 Deleted o Sh.3
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Figure 14.1-51 I Deleted o Sh.4 o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Sh.5 Figure 14.1-52 Deleted through 14.1-57 Figure 14.1-58 Deleted Figure 14.1-59 Deleted Sh.1 Figure 14.1-59 Deleted Sh.2 Figure 14.1-60 Deleted Figure 14.1-61 Deleted Figure 14.1-62 Tracking 88-95/96 Stop Valve (SV) Type 1 Failures Stop Valve Disc Fails Figure 14.1-63 Tracking 88-95/96 Stop Valve (SV) Type 2 Failures Stop Valve Spring Fails Figure 14.1-64 Tracking 88-95/96 Stop Valve (SV) Type 3 Failures Stop Valve Sticks Open Figure 14.1-65 Tracking 88-95/96 Control Valve (CV) Type 4 Failures CV Spring 80lt Fails Figure 14.1-66 Tracking 88-95/96 Control Valve (CV) Type 5 Failures Control Valve Sticks Open Figure 14.1-67 Annual Frequency of Destructive Overs peed for Various 88-95/96 Turbine Valve Test Interval 14.2 FIGURES Figure No. Title Plant Drawing Rev Figure 14.2-0 Steam Generator Tube Rupture, 8reak Flow and Safety Injection Flow vs. Reactor Coolant System Pressure Figure 14.2-1 Steam Line Valve Arrangement Schematic Figure 14.2-2 Steam Line Rupture Offsite Power Available, EOL, Core o Sh.1 Heat Flux and Core Reactivity vs. Time
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Figure 14.2-2 Steam Line Rupture Offsite Power Available, EOL, Reactor o Sh.2 Coolant Pressure and RV Inlet Temperature vs. Time o Figure 14.2-2 Steam Line Rupture Offsite Power Available, EOL, Steam o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Sh.3 Flow and Steam Generator Pressure vs. Time Figure 14.2-2 Steam Line Rupture Offsite Power Available, EOL, Core Sh.4 Boron Concentration vs. Time Figure 14.2-3 Deleted Sh.1 Figure 14.2-3 Deleted Sh.2 Figure 14.2-3 Deleted Sh.3 Figure 14.2-4 Deleted Sh.1 Figure 14.2-4 Deleted Sh.2 Figure 14.2-4 Deleted Sh.3 Figure 14.2-5 Deleted Sh.1 Figure 14.2-5 Deleted Sh.2 Figure 14.2-5 Deleted Sh.3 Figure 14.2-6 Deleted Sh.1 Figure 14.2-6 Deleted Sh.2 Figure 14.2-7 Containment Pressure Time History (Double - Ended Main 1%$%
Steam Line Break Main FCV Failure Maximum Containment 20 Safeguards)
Figure 14.2-8 Deleted Through 14.2-10 Figure 14.2-11 Rod Ejection Accident, BOL-HFP, Nuclear Power vs. Time Figure 14.2-12 Rod Ejection Accident, BOL-HFP, Fuel Temperatures vs.
o Time
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Figure 14.2-13 Rod Ejection Accident, BOL-HZP, Nuclear Power vs. Time o Figure 14.2-14 Rod Ejection Accident, BOL-HZP, Fuel Temperatures vs.
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 14.2-15 Rod Ejection Accident, EOL-HZP, Nuclear Power vs. Time Figure 14.2-16 Rod Ejection Accident, EOL-HZP, Fuel Temperatures vs.
Time Figure 14.2-17 Rod Ejection Accident, EOL-HFP, Nuclear Power vs. Time Figure 14.2-18 Rod Ejection Accident, EOL-HFP, Fuel Temperatures vs.
Time Figure 14.2-19 Deleted Thru Figure 14.2-22 14.3 FIGURES o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 14.3-27 I Deleted throuah 14.3-52 Figure 14.3-53a I Deleted through 14.3-58b o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 14.3-64 2.0" Small Break Loca Hot Rod Clad Average Temperature Figure 14.3-65 4.0" Small Break Loca RCS Pressure Figure 14.3-65 4.0" Small Break Loca RCS Pressure Figure 14.3-66 4.0" Small Break Loca Core Mixture Level Figure 14.3-67 4.0" Small Break Loca Hot Rod Clad Average Temperature Figure 14.3-68 Deleted Through 14.3-100 Figure 14.3-101 Reactor Vessel Internals Figure 14.3-102 RPV Shell And Support System Figure 14.3-103 Deleted Figure 14.3- Reactor Vessel Internals Core Barrel Assembly 103a Figure 14.3- Reactor Internals and Fuel 103b Figure 14.3-104 RPV System Model Figure 14.3- Deleted 104a Figure 14.3- Deleted 104b Figure 14.3-104c Deleted Figure 14.3- Deleted 104d Figure 14.3- Deleted 104e Figure 14.3-104f Deleted Figure 14.3- Deleted 104g Figure 14.3- Deleted 104h Figure 14.3-104i Deleted Figure 14.3-104j Deleted o Figure 14.3-104k Deleted
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G) Figure 14.3-105 Double-Ended Pump Suction Break for 3216 Mwt Minimum o Safeguards Integrated Wall Heat Removal o
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 14.3-106 Double-Ended Pump Suction Break for 3216 Mwt Minimum Safeguards Integrated Fan Cooler Heat Removal Figure 14.3-107 Double-Ended Pump Suction Break for 3216 Mwt Minimum Safeguards Integrated Spray Heat Removal Figure 14.3-108 Double-Ended Pump Suction Break for 3216 Mwt Minimum Safeguards Structural Heat Transfer Coefficient Figure 14.3-109 Double-Ended Pump Suction Break for 3216 Mwt Minimum Safeguards Containment Pressure Figure 14.3-110 Double-Ended Pump Suction Break for 3216 Mwt Minimum Safeguards Containment Temperature Figure 14.3-111 Double-Ended Pump Suction Break for 3216 Mwt Maximum Safeguards Containment Pressure Figure 14.3-112 Double-Ended Pump Suction Break for 3216 Mwt Maximum Safeguards Containment Temperature Figure 14.3-113 Double-Ended Hot Leg Break for 3216 Mwt Containment Pressure Figure 14.3-114 Double-Ended Hot Leg Break for 3216 Mwt Containment Temperature Figure 14.3-115 Fan Cooler Heat Removal as a Function of Containment Temperature 95°F Service Water, 1600 GPM SW Flow Figure 14.3-116 Deleted Figure 14.3-117 Deleted Figure 14.3-118 Deleted Figure 14.3-119 Deleted Figure 14.3-120 Deleted Figure 14.3-121 Deleted Figure 14.3-122 Deleted Figure 14.3-123 Deleted Figure 14.3-124 Deleted Figure 14.3-125 Deleted Figure 14.3-126 Deleted Figure 14.3-127 Deleted o Figure 14.3-128 Deleted
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G) Figure 14.3-129 Radiation Levels Surrounding 14-ln. Residual Heat Removal o Pipe [FIGURE RETAINED FOR HISTORICAL PURPOSES]
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Indian Point 2 UFSAR & Plant Drawing Cross Reference 14.4 FIGURES Figure No. Title Plant Drawing Rev Figure 14.4-1 Deleted Figure 14.4-2 Deleted Figure 14.4-3 Deleted Figure 14.4-4 Deleted Figure 14.4-5 Deleted Figure 14.4-6 Deleted Figure 14.4-7 Deleted Figure 14.4-8 Deleted Figure 14.4-9 Deleted Figure 14.4-10 Deleted Figure 14.4-11 Deleted Figure 14.4-12 Deleted Figure 14.4-13 Deleted Figure 14.4-14 Deleted Figure 14.4-15 Deleted Figure 14.4-16 Deleted Figure 14.4-17 Deleted Figure 14.4-18 Deleted Figure 14.4-19 Deleted Figure 14.4-20 Deleted Figure 14.4-21 Deleted Figure 14.4-22 Deleted Figure 14.4-23 Deleted Figure 14.4-24 Deleted Figure 14.4-25 Deleted Figure 14.4-26 Deleted Figure 14.4-27 Deleted Figure 14.4-28 Deleted o Figure 14.4-29 Deleted
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Indian Point 2 UFSAR & Plant Drawing Cross Reference Figure 14.4-32 Deleted Figure 14.4-33 Deleted Figure 14.4-34 Deleted Figure 14.4-35 Deleted Figure 14.4-36 Deleted Figure 14.4-37 Deleted o
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INDIAN POINT ENERGY CENTER (IPECl UNIT 2 UPDATED FINAL SAFETY ANALYSIS REPORT (UFSARl Table of Contents Chapter 1 - Introduction and Summary Chapter 2 - Site and Environment Chapter 3 - Reactor Chapter 4 - Reactor Coolant System Chapter 5 - Containment Spray Systems Chapter 6 - Engineered Safety Features Chapter 7 - Instrumentation and Control Chapter 8 - Electrical Systems Chapter 9 - Auxiliary and Energy Systems Chapter 10 - Steam and Power Conversion System Chapter 11 - Waste and Radiation Protection System Chapter 12 - Conduct of Operations Chapter 13 - Test and Operation Chapter 14 - Safety Analysis OAG10000215_0037
IP2 FSAR UPDATE CHAPTER 1 INTRODUCTION AND
SUMMARY
1.1 INTRODUCTION
The Final Safety Analysis Report (FSAR) was submitted in support of an application by Consolidated Edison Company of New York, Inc. (Con Edison) for a license to operate the Indian Point Unit 2 nuclear power plant (AEC Docket 50-247, Permit No. CPPR-21). It provided pertinent technical information in accordance with Section 50.34 of 10 CFR 50 requirements to obtain a nuclear power plant operating license. Westinghouse Electric Corporation was the primary contractor and had turnkey responsibility for the design, construction, testing, and initial startup of the facility. Westinghouse had contracted with United Engineers and Constructors as architect-engineer to provide engineering assistance in the design of and construction of the structural and civil works.
This document is the Updated FSAR and is submitted in accordance with Section 50.71(e) of 10 CFR Part 50. The revision history is summarized in Section 1.9.2.
The following paragraphs contain a summary of the report's scope:
The unit employs a pressurized water reactor nuclear steam supply system (NSSS) furnished by Westinghouse Electric Corporation.
The reactor was originally licensed at a maximum thermal power of 2758 MW. By letter dated September 30, 1988 Con Edison initiated a request to authorize an increase in the licensed thermal power level to 3071.4 MW. The NRC documented their acceptance of this request in a Safety Evaluation Report dated March 7, 1990. By letter dated December 12, 2002, Entergy Nuclear Operations, Inc. initiated a request for a Measurement Uncertainty Recapture power uprate of 1.4%, for a licensed thermal power level of 3114.4 MW. The NRC documented their acceptance of this request in a Safety Evaluation Report dated May 22, 2003.
Thereact()ris presently licensed to operate until September 28, 2013. The reactor power plus 12,4 MimttQf energy from the reactor coolant pumps give a total rated heat input to the NSSS of 3126.6 MW, which corresponds to a design electric output from the turbine-generator of approximately 1034.1 MW.
The plant heat removal systems have been designed for the equivalent guarantee rating of 3126.6 MWt; some of the portions of the safety analysis dependent on heat removal capacity of plant and safeguards systems have assumed the maximum calculated power of 3216 MWt as have the evaluations of activity release and radiation exposure.
By NRC order dated August 27, 2001 (Reference 1), Con Edison's ownership/operation of Indian Point 1 and 2 was transferred to Entergy Nuclear Indian Point 2 (ENIP2), LLC, as the owner of Indian Point 1 and 2 plants, and Entergy Nuclear Operations (ENO), Inc. as the operator of Indian Point 2 and maintainer of Indian Point 1. Consequently, references to Con Edison (or derivatives thereof) in this document remain only when used in historical context.
Chapter 1, Page 1 of 72 Revision 20, 2006 OAG10000215_0038
IP2 FSAR UPDATE The remainder of Chapter 1 of this report summarizes the principal design features and parameters of the plant, pointing out the similarities and differences with respect to other pressurized water nuclear power plants presently in operation. A general description of the plant is included as well as a statement and summary of all the General Design Criteria.
Chapter 2 contains a description and evaluation of the site and environs, supporting the suitability of the site for a reactor of the size and type described. Chapters 3 and 4 describe and evaluate the reactor and the reactor coolant system; Chapter 5, the containment system; Chapter 6, the engineered safety features; Chapter 7, plant instrumentation and control; Chapter 8, the electrical system; Chapter 9, the auxiliary and emergency system; Chapter 10, the steam and power conversion system; Chapter 11, radioactive waste disposal and radiation protection. Chapter 12 and 13 are conduct of operations and initial test and operations, respectively. They describe plant organization, training programs, and startup administrative procedure. Chapter 14 is a safety evaluation summarizing the analyses, which demonstrate the adequacy of the reactor protection system and the containment and engineered safety features, and which show that the consequences of various postulated accidents are within applicable limits.
REFERENCES FOR SECTION 1.1
- 1. NRC letter to Consolidated Edison, Indian Point Nuclear Generating Unit No.s 1 and 2 - Order Approving Transfer of Licenses from the Consolidated Edison Company of New York, Inc., to Entergy Nuclear Indian Point 2, LLC, and Entergy Nuclear Operations, Inc. and Approving Conforming Amendments (TAC Nos.
MB0743 and MB0744), August 27,2001.
1.2
SUMMARY
PLANT DESCRIPTION 1.2.1 Site Indian Point Unit 2 is adjacent to and north of Unit 1 on a site of approximately 239 acres of land on the east bank of the Hudson River at Indian Point, Village of Buchanan in upper Westchester County, New York. Indian Point Unit 3 (owned and operated by Entergy Nuclear and Entergy Nuclear Operations, Inc.) is adjacent to and south of Unit 1. The site is about 24 miles north of the New York City boundary line. The nearest city is Peekskill, 2.5 miles northeast of Indian Point. An aerial photograph, [Historical] Figure 2.2-1, shows the site and about 58 mile2 of the surrounding area.
1.2.1.1 Meteorology Meteorological conditions in the area of the site were determined during a 2-year test program.
These data were used in evaluating the effects of gaseous discharges from the plant during normal operations and during the postulated loss-of-coolant accident. In addition, data supplied by the U.S. Weather Bureau at the Bear Mountain Station, regarding the meteorological conditions during periods of precipitation, were used to evaluate the rainout of fission gases into surface water reservoirs following the postulated loss-of-coolant accident. The evaluations indicate that the site meteorology provides adequate diffusion and dilution of any released gases.
Chapter 1, Page 2 of 72 Revision 20, 2006 OAG10000215_0039
IP2 FSAR UPDATE 1.2.1.2 Geology and Hydrology Geologically, the site consists of a hard limestone in a jointed condition, which provides a solid bed for the plant foundation. The bedrock is sufficiently sound to support any loads, which could be anticipated up to 50 tons per fe, which is far in excess of any load, which may be imposed by the plant. Although it is hard, the jointed limestone formation is permeable to water.
Thus, if water from the plant should enter the ground (an improbable event since the plant is designed to preclude any leakage into the ground) it would percolate to the river rather than enter any ground water supply. Additional studies by geology consultant, Thomas W. Fluhr, and examination of soil borings confirmed the above conclusions.
In the Hudson River, about 80,000,000 gallons of water flow past the plant each minute during the average tidal flow. This flow provides additional mixing and dilution for liquid discharges from the facility. In fact, however, this aspect is superfluous since the assumption in the plant design is to treat the river water as if it were used for drinking and thus to reduce radioactive discharges, by dilution with ordinary plant effluent, to concentrations that would be tolerable for drinking water. There is minimal danger of flooding at the site as discussed in Section 2.5.
1.2.1.3 Seismology Seismic activity in the Indian Point area is limited to low-level microseismicity. Detailed field investigations (e.g., Ratcliffe, 1976, 1980; Dames and Moore, 1977) have been conducted in the immediate vicinity of Indian Point and along the major faults in the region. To date, no evidence has been found in the rocks exposed at the surface or sediments overlying fault traces or in cores obtained in the vicinity of Indian Point, that might support a conclusion that displacement has occurred along major fault systems within the New York Highlands, the Ramapo or its associated branches during Quaternary time (the last 1.5 million years). In the vicinity of Indian Point, evidence that no displacement has occurred in the last 65 million years (since the Mesozoic) along specific major structures has been observed.
The plant is designed to withstand an earthquake of Modified Mercalli Intensity VII. The validity of the selection of an Intensity VII earthquake was adjudicated before the Atomic Safety and Licensing Appeal Board. The Appeal Board's decision (ALAB-436) verified Intensity VII as the design basis earthquake for the plant.
1.2.1.4 Environmental Radiation Monitoring Environmental radioactivity has been measured at the site and surrounding area for nearly 20 years in association with the operation of the three Indian Point Units. These measurements will be continued and reported. The radiation measurements of fallout, water samples, vegetation, marine life, etc., have shown no significant postoperative increase in activity.
Noticeable increases in fallout have coincided with weapons-testing programs and appear to be related almost entirely to those programs. The New York State Department of Health in an independent 2-year postoperative study found that environmental radioactivity in the vicinity of the site is no higher than anywhere else in the State of New York.
1.2.1.5 Conclusions Consideration of all the items mentioned above, plus the containment design and the engineered safety features included in the plant design lead to the conclusion of appropriate suitability of the site for safe operation of the Indian Point Unit 2 nuclear power plant. Accident Chapter 1, Page 3 of 72 Revision 20, 2006 OAG10000215_0040
IP2 FSAR UPDATE analyses presented in Chapter 14 verify that the maximum expected doses at or beyond the site boundary are within applicable limits.
1.2.2 Plant Description The unit incorporates a closed-cycle pressurized water nuclear steam supply system, a turbine generator and their necessary auxiliaries. A radioactive waste disposal system, fuel handling system and all auxiliaries, structures, and other onsite facilities required for a complete and operable nuclear power plant are provided for the unit.
The general arrangement of the plant is shown on historical Figures 1.2-1 and 1.2-4, and Figure 2.2-2. Other general plant arrangement drawings have been removed due to security reasons following September 11, 2001 and can be viewed as plant drawings 9321-2510, 9321-2511, 9321-2514, 9321-2517, 9321-3052, and 209812.
1.2.2.1 Nuclear Steam Supply System The nuclear steam supply system consists of a pressurized water reactor, reactor coolant system, and associated auxiliary fluid systems. The reactor coolant system is arranged as four closed reactor coolant loops, each containing a reactor coolant pump and a steam generator, connected in parallel to the reactor vessel. An electrically-heated pressurizer is connected to one of the loops.
The reactor core is composed of uranium-dioxide pellets enclosed in zircaloy tubes with welded end plugs. The tubes are supported in assemblies by a spring clip grid structure. The mechanical control rods consist of clusters of stainless steel clad absorber rods and guide tubes located within the fuel assembly. The core was initially loaded in three regions of different enrichments with new fuel being introduced into the outer region at successive refuelings and discharged to spent fuel storage, following burnup.
The steam generators are vertical U-tube units employing Inconel tubes. Integral separating equipment reduces the moisture content of the steam leaving the steam generators to 0.25-percent or less.
The reactor coolant pumps are vertical, single stage, centrifugal pumps equipped with controlled leakage shaft seals.
Auxiliary systems are provided to perform the following functions:
- 1. Charge the reactor coolant system.
- 2. Add makeup water.
- 3. Purify reactor coolant water.
- 4. Provide chemicals for corrosion inhibition and reactor control.
- 5. Cool system components.
- 6. Remove residual heat when the reactor is shut down.
- 7. Cool the spent fuel storage pool.
- 8. Sample reactor coolant water.
- 9. Provide for emergency core cooling.
- 10. Collect reactor coolant drains.
- 11. Provide containment spray.
- 12. Provide containment ventilation and cooling.
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IP2 FSAR UPDATE
- 13. Dispose of liquid, gaseous and solid wastes.
1.2.2.2 Reactor and Plant Control The reactor is controlled by a coordinated combination of chemical shim and mechanical control rods. The control system allows the unit to accept step load changes of +/- 10-percent and ramp load changes of +/- S-percent per min over the load range of 1S-percent to, but not exceeding, 100-percent power under nominal operating conditions subject to xenon limitations.
1.2.2.3 Turbine and Auxiliaries The turbine is a tandem-compound, comprising one high pressure and three low pressure cylinders, 1800 rpm unit having 4S-in. exhaust blading in the low pressure cylinders. There are four moisture preseparators located at the four high-pressure turbine exhaust lines and six combination moisture separator-reheater units that are employed to dry and superheat the steam between the high and low pressure turbine cylinders. The turbine generator is capable of a SO-percent loss of external electrical load without turbine or reactor trip. The turbine auxiliaries include deaerating surface condensers, steam jet air ejector, turbine-driven main feedwater pumps, motor-driven condensate pumps, and six stages of feedwater heating.
The original turbine generator had a guaranteed capability of 1,021,793 kWe at 1.S-in. Hg absolute exhaust pressure with zero percent makeup and six stages of feedwater heating.
1.2.2.4 Electrical System The main generator feeds electrical power through an isolated phase bus to two half-sized main power transformers. Station auxiliaries receive power during normal operation from either the station auxiliary transformer (i.e., offsite power) or the unit auxiliary transformer (i.e., unit main power transformers).
The auxiliary electrical system provides power to those auxiliary components required to operate during normal or emergency conditions of operation. Standby power required during plant startup, shutdown, and after reactor trip is supplied to the station auxiliary transformer from the Con Edison 138-kV system by either of two separate overhead lines from the Buchanan substation approximately O.SO mile from the plant. Alternate feeds from the 13.8-kV system are also available for immediate manual connection to the auxiliary buses. In addition, three gas turbines designed for blackstart (no auxiliary power) capability are available.
Emergency power supply for vital instruments and controls is from four 12S-V station batteries.
The system design provides sufficient independence, isolation capability, and redundancy between the different power sources to avoid complete loss of auxiliary power.
1.2.2.S Control Room The plant is provided with a reactor and turbine-generator control room containing all necessary instrumentation for the operation of the plant under normal and accident conditions.
Adequate shielding and air conditioning facilities permit occupancy during all operating or accident conditions.
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IP2 FSAR UPDATE 1.2.2.6 Diesel Generators Three diesel-generator sets supply emergency power for shutdown or essential safeguards operation in the event of a loss of all other alternating current auxiliary power.
1.2.2.7 Waste Disposal System The waste disposal system collects and processes liquids, gaseous, and solid waste from plant operation for removal from plant site. All removals are made in accordance with government guidelines for the process.
1.2.2.8 Fuel Handling System The fuel handling system provides the ability to fuel and refuel the reactor core. Carefully established administrative procedures plus the design of the system minimizes the probability of potential fission product release during the refueling operation.
The system also includes the following features:
- 1. Safe accessibility for operating personnel.
- 2. Provisions to prevent fuel storage criticality.
- 3. Visual monitoring of the refueling procedures at all times.
1.2.2.9 Engineered Safety Features The engineered safety features for this plant have sufficient redundancy of component and power sources such that under the conditions of a hypothetical loss-of-coolant, the system can, even when operating with partial effectiveness, maintain the integrity of the containment and keep the exposure of the public within applicable limits.
The major engineered safeguards systems are as follows:
- 1. The containment system, which incorporates continuously pressurized and monitored penetrations and liner weld channels and a seal water injection system, which provides a highly reliable, essentially leaktight barrier against the escape of radioactivity, which might be released to the containment atmosphere.
- 2. The safety injection system (which constitutes the emergency core cooling system) provides borated water to cool the core in the event of a loss-of-coolant accident.
- 3. The containment air recirculation cooling system provides a heat sink to cool the containment atmosphere.
- 4. The containment spray system provides a spray of cool, borated water to the containment atmosphere that is a heat sink and also provides iodine removal capability.
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IP2 FSAR UPDATE 1.2.2.10 Structures The major structures are the reactor containment building, the primary auxiliary building, the control building, the fuel storage building, the turbine building, and the maintenance and operations building. General layouts of the reactor containment interior components arrangement are shown on Plant Drawings 9321-2501, 9321-2502, 9321-2503, 9321-2506, 9321-2507, 9321-2508 [formerly UFSAR Figures 5.1-2 through 5.2-7]. General layouts and interior components arrangement of the primary auxiliary building, control building, fuel storage building, and holdup tank building were removed from the UFSAR due to security reasons following September 11, 2001 and can be viewed on plant drawings.
1.2.2.11 Containment The reactor containment is a steel-lined reinforced concrete cylinder with a hemispherical dome and a flat base. The containment is designed to withstand the internal pressure accompanying a loss-of-coolant accident, is virtually leaktight, and provides adequate radiation shielding for both normal operation and accident conditions.
When required, the containment isolation valve seal water system permits automatic rapid sealing of pipes, which penetrate the containment so that in the event of any loss-of-coolant accident, leakage from containment to the environment is minimal.
Ground accelerations for the operational basis earthquake used for containment design purposes and all seismic Class I structures (Section 1.11) are 0.10g applied horizontally and 0.05g applied vertically. In addition, ground accelerations for the design basis earthquake of 0.15g horizontal and 0.1 Og vertical are used to analyze the no loss-of-function concept.
1.2 FIGURES Figure No. Title Figure 1.2-1 Indian Point Nuclear Generating Units 1 & 2 [Historical]
Figure 1.2-2 Deleted Figure 1.2-3 Deleted Figure 1.2-4 Cross Section of Plant [Historical]
Figure 1.2-5 Deleted Figure 1.2-6 Deleted Figure 1.2-7 Deleted Figure 1.2-7 Deleted Figure 1.2-8 Deleted Figure 1.2-9 Deleted 1.3 GENERAL DESIGN CRITERIA (GDC)
The General Design Criteria define or describe safety objectives and approaches incorporated in the design of this plant. These General Design Criteria, tabulated explicitly in the pertinent systems sections in this report, comprised the proposed Atomic Industrial Forum versions of the criteria issued for comment by the AEC on July 11, 1967. Also included in this section, are brief descriptions of related plant features, which are provided to meet the design objectives reflected Chapter 1, Page 7 of 72 Revision 20, 2006 OAG10000215_0044
IP2 FSAR UPDATE in the criteria at the time of the initial license application. The descriptions are more fully developed in those succeeding sections of the report indicated by the references.
More recently, Con Edison completed a study of compliance with 10 CFR Parts 20 and 50 in accordance with the Commission's Confirmatory Order of February 11, 1980. The detailed results of the evaluation of Indian Point Unit 2 compliance with the then current General Design Criteria established by the Nuclear Regulatory Commission (NRC) in 10 CFR 50 Appendix A, were submitted to the NRC by Con Edison on August 11, 1980 (Reference 1). Commission concurrence was received on January 19, 1982.
The parenthetical numbers following the section headings indicate the numbers of their related proposed Atomic Industrial Forum versions of the General Design Criteria as described in the first paragraph of this section.
1.3.1 Overall Plant REQUIREMENTS (GDC 1 - GDC 5)
All systems and components of the facility are classified according to their importance. Those items vital to safe shutdown and isolation of the reactor or whose failure might cause or increase the severity of a loss-of-coolant accident or result in an uncontrolled release of excessive amounts of radioactivity are designated Class I. Those items important to reactor operation but not essential to safe shutdown and isolation of the reactor or control of the release of substantial amounts of radioactivity are designated Class II. Those items not related to reactor operation or safety are designated Class III.
Class I systems and components are essential to the protection of the health and safety of the public. Consequently, they are designed, fabricated, inspected, erected, and use materials selected to the applicable provisions of recognized codes, good nuclear practice and to quality standards that reflect their importance.
All systems and components designated Class I are designed so that there is no loss of function in the event of the maximum potential ground acceleration acting in the horizontal and vertical directions simultaneously. The working stresses of both Class I and Class II items are kept within code allowable values for the operational basis earthquake. Similarly, measures are taken in the plant design to protect against high winds, sudden barometric pressure changes, flooding, and other natural phenomena.
Reference sections:
Section Title Section Meteorology 2.6 Geology and Seismology 2.7 Reactor Coolant System; Design Bases 4.1 Containment System Structures; Design Bases 5.1.1 Electrical Systems; Design Bases 8.1 Design Criteria for Structures and Equipment 1.11 Fire prevention in all areas of the nuclear electric plant is provided by structure and component design, which maximizes the use of fire-resistant materials, optimizes the containment of combustible materials and maintains exposed combustible materials below their ignition Chapter 1, Page 8 of 72 Revision 20, 2006 OAG10000215_0045
IP2 FSAR UPDATE temperature in the design atmosphere. Fixed and portable fire fighting equipment is provided with capacities proportional to the energy that might credibly be released by fire.
Reference sections:
Section Title Section Operating Control Stations 7.7 Facility Service System 9.6 The only structures, systems, or components important to safety that are shared between Units 2 and 3 are:
- 1. The cooling water discharge channel, which carries the service water discharge to the river.
- 2. The Emergency Fuel Supply to the Emergency Diesel Generators.
Since the channel is designed to handle the discharge flow from both operating units, sharing of this structure will not impair the ability of Unit 2 safety systems to perform their safety functions.
Units 2 and 3 are required by Technical Specification to maintain a designated amount of fuel reserve onsite or at the Buchanan substation, which is dedicated for emergency diesel generator use at that unit. Therefore sharing the Emergency Fuel Supply to the Emergency Diesel Generators will not impair the ability of Unit 2 safety systems to perform their safety functions.
Reference sections:
Section Title Section Emergency Fuel Supply 8.2.3.2 Service Water System 9.6.1 A complete set of facility plant and system diagrams including arrangements, plans, and structural plans and records of initial tests and operation are maintained throughout the life of the reactor. A set of all the quality assurance data generated during fabrication and erection of the essential components of the plant, as defined by the quality assurance program, is retained.
Reference sections:
Section Title Section Records 12.4 Initial Tests and Operation 13 Quality Assurance Program 1.10 1.3.2 Protection By Multiple Fission Product Barriers (GDC 6 - GDC 10)
The reactor core, with its related control and protection system, is designed to function throughout its design lifetime without exceeding acceptable fuel damage limits. The core design, together with reliable process and decay heat removal systems, provides for this capability under all expected conditions of normal operation with appropriate margins for uncertainties and anticipated transient situations.
Chapter 1, Page 9 of 72 Revision 20, 2006 OAG10000215_0046
IP2 FSAR UPDATE The reactor control and protection instrumentation is designed to actuate a reactor trip for any anticipated combination of plant conditions when necessary to ensure DNBR remains at or above the applicable safety analysis DNBR limit and fuel center temperature below the melting point of uranium-dioxide.
Reference sections:
Section Title Section Reactor; Design Bases, 3.1 Reactor Design 3.2 Instrumentation and Control; Protective Systems 7.2 Safety Analysis 14 The design of the reactor core and related protection systems ensures that power oscillations, which could cause fuel damage in excess of acceptable limits are not possible or can be readily suppressed.
Low frequency spatial xenon oscillations may occur in the axial dimension. However, the core is expected to be (and has proven to be) stable to xenon oscillations in the X-Y dimension. Ex-core instrumentation is provided to monitor any xenon induced oscillations. Incore instrumentation is used periodically to calibrate and verify the information provided by the ex-core instrumentation.
The moderator temperature and overall power coefficients in the power operating range were maintained negative by inclusion of burnable poison shims in the first core loading. The overall power coefficient in the power operating range is always negative (as discussed in Section 14.1.11.2).
Reference sections:
Section Title Section Reactor, Design Bases, 3.1 Reactor Design 3.2 The reactor coolant system in conjunction with its control and protective provisions is designed to accommodate the system pressures and temperatures attained under all expected modes of plant operation or anticipated system interactions, and maintain the stresses within applicable code stress limits. The materials of construction of the pressure retaining boundary of the reactor coolant system are protected by control of coolant chemistry from corrosion phenomena, which might otherwise reduce the system structural integrity during its service lifetime.
System conditions resulting from anticipated transients or malfunctions are monitored and appropriate action is initiated automatically to maintain the required cooling capability and to limit system conditions so that continued safe operation is achieved.
The system is protected from overpressure by means of pressure relieving devices, as required by Section III of the ASME Boiler and Pressure Vessel Code and by an overpressure protection system intended to ensure compliance with 10 CFR 50, Appendix G.
Chapter 1, Page 10 of 72 Revision 20, 2006 OAG10000215_0047
IP2 FSAR UPDATE Isolable sections of the system are provided with overpressure relieving devices discharging to closed systems such that the system code allowable relief pressure within the protected section is not exceeded.
Reference section:
Section Title Section Reactor Coolant System; Design Bases 4.1 The design pressure of the containment exceeds the peak pressure occurring as the result of the complete blowdown of the reactor coolant through any pipe rupture of the reactor coolant system up to and including the hypothetical severance of a reactor coolant system pipe.
All piping systems, which penetrate the vapor barrier are anchored so that the penetration is structurally adequate to resist the piping loads and the vapor barrier will not be breached due to a hypothesized pipe rupture. The lines (with the exception of sample tubing) connected to the reactor coolant system that penetrate the vapor barrier are restrained near the secondary shield walls and are each provided with at least one valve between the shield wall and the reactor coolant system. These restraints are designed to withstand the thrust moment and torque resulting from a hypothesized rupture of the attached pipe.
All isolation valves are supported to withstand, without impairment of valve operability, the combined loading of the design-basis accident and design seismic conditions.
Reference section:
Section Title Section Containment System Structures; Design Bases 5.1.1 1.3.3 Nuclear And Radiation Controls (GDC 11 - GDC 18)
The plant is equipped with a control room, which contains all controls and instrumentation and facilities necessary for continuous operation of the reactor and turbine generator under normal and accident conditions.
Sufficient shielding, ventilation control and filtration, and containment integrity are provided to ensure that control room personnel will not be subjected to doses under postulated accident conditions during occupancy of the control room, which in the aggregate, would exceed the applicable limits.
Instrumentation and controls essential to avoid undue risk to the health and safety of the public are provided to monitor and maintain neutron flux, reactor coolant pressure, flow rate, temperature, and control rod positions within prescribed operating ranges.
The non-nuclear regulating, process and containment instrumentation measures temperatures, pressures, flows, and levels in the reactor coolant system, steam systems, containment, and other auxiliary systems.
Chapter 1, Page 11 of 72 Revision 20, 2006 OAG10000215_0048
IP2 FSAR UPDATE The quantity and types of process instrumentation provided ensures safe and orderly operation of all systems and processes over the full operating range of the plant.
The operational status of the reactor is monitored from the control room. When the reactor is subcritical, the neutron source multiplication is continuously monitored and indicated by proportional counters located in instrument wells in the primary shield adjacent to the reactor vessel. The source detector channels are checked prior to operations in which criticality may be approached by the use of an incore source. Any appreciable increase in the neutron source multiplication, including that caused by the maximum physical boron dilution rate, is slow enough to give ample time to start corrective action (boron dilution stop and/or emergency boron injection) to prevent the core from becoming critical (as discussed in Sections 14.1.5.2.3 and 14.1.5.3).
When the reactor is critical, means for showing the relative reactivity status of the reactor is provided by display of all rod bank positions in the control room. Periodic samples of the coolant boron concentration are taken, the variation of which provides a further check on the reactivity status of the reactor including core depletion during life.
Instrumentation and controls provided for the protective systems are designed to trip the reactor when necessary to prevent or limit fission product release from the core, to limit energy release, to signal closure of containment isolation valves, and to control the operation of engineered safety features equipment.
During reactor operation in the startup and power modes, redundant safety limit signals will automatically actuate two reactor trip breakers, which are in series with the rod drive mechanism coils. This action would interrupt power and initiate reactor trip.
Reference sections:
Section Title Section Instrumentation and Control; General Design Criteria, 7.1 Protective Systems, 7.2 Nuclear Instrumentation, 7.4 Operating Control Stations 7.7 The reactor protection system receives signals from plant instrumentation, which are indicative of an approach to an unsafe operating condition, actuates alarms, prevents control rod motion, initiates load cutback, and/or opens the reactor trip breakers, depending on the severity of the condition.
The basic reactor tripping philosophy is to define a region of power and coolant temperature conditions allowed by the primary tripping functions (e.g., overpower i1T trip, overtemperature i1T trip, and nuclear flux trips). The allowable operating region within these trip settings is shown to prevent any combination of power, temperature, and pressure, which would result in DNB with all reactor coolant pumps in operation. Additional tripping functions such as low and high pressurizer pressure trips, high pressurizer level trip, loss of flow trip, steam and feedwater flow mismatch trip, steam generator low-low level trip, turbine trip, safety injection trip, nuclear flux trips (source, intermediate, and high range), and manual trip are provided as backup to the primary tripping functions for specific accident conditions and mechanical failures.
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IP2 FSAR UPDATE Intermediate Range and Power Range rod stops, Overtemperature boT and Overpower boT rod stops, are provided to prevent abnormal power conditions which could result from excessive control rod withdrawal initiated by operator violation of administrative procedures.
Reference sections:
Section Title Section Safety Injection System 6.2 Protective Systems 7.2 Positive indications in the control room of leakage of coolant from the reactor coolant system to the containment are provided by equipment, which permits continuous monitoring of containment air activity and humidity, and of runoff from the condensate collecting pans under the cooling coils of the containment air recirculation units. The basic design criterion is the detection of deviations from normal containment environmental conditions including air particulate activity, radiogas activity, humidity, condensate runoff and in addition, in the case of gross leakage, the liquid inventory in the process systems and containment sump.
The containment atmosphere, the plant vent, the containment fan-coolers service water discharge, the waste disposal system liquid effluent, and the component cooling loop are monitored for radioactivity concentration during all normal operations, anticipated transients, and accident conditions.
For the case of leakage from the reactor containment under accident conditions, the plant area radiation monitoring system supplemented by portable survey equipment to be kept in the control room provide adequate monitoring of accident releases.
Monitoring and alarm instrumentation is provided for fuel and waste storage and handling areas to detect excessive radiation levels. The permanent record of activity releases is provided by radiochemical analysis of known quantities of waste.
A controlled ventilation system removes gaseous radioactivity from various areas of the plant and discharges it to the atmosphere via the plant vent. Radiation monitors are in continuous service in these areas to actuate high activity alarms on the control board annunciator and initiate containment isolation.
Reference sections:
Section Title Section Engineered Safety Features; Safety Injection System, 6.2 Isolation Valve Seal Water System, 6.S Leakage Detection and Provisions for the Primary and Auxiliary Coolant Loops 6.7 Auxiliary Coolant System 9.3 Radiation Protection 11.2 Chapter 1, Page 13 of 72 Revision 20, 2006 OAG1000021S_00S0
IP2 FSAR UPDATE 1.3.4 Reliability And Testability Of Protection Systems (GDC 19 - GDC 26)
Upon a loss of power to the magnetic-type control rod drive mechanisms, the rod cluster control (RCC) assembly is released and falls by gravity into the core. The reactor internals, fuel assemblies, RCC assemblies and drive system components are designed as Class I equipment.
The RCC assemblies are fully guided through the fuel assembly and for the maximum travel of the control rod into the guide tube. Furthermore, the RCC assemblies are never fully withdrawn from their guide thimbles in the fuel assembly. Due to this and the flexibility designed into the RCC assemblies, abnormal loadings and misalignments can be sustained without impairing operation of the RCC assemblies.
All reactor protection channels are supplied with sufficient redundancy to provide the capability for channel calibration and test at power. The analog channel trip bistables have the capability to be bypassed for surveillance testing. This "test in bypass" feature enables analog logic relays in a protective channel to be bypassed while testing actuation of the associated bistable.
Bistable testing does not preclude the protective action provided by concurrent channels.
Two reactor trip breakers are provided to interrupt power to the rod drive mechanisms. The breaker main contacts are connected in series with the power supply to the mechanism coils.
Opening either breaker interrupts power to all full length rod mechanisms. Each breaker is opened through an undervoltage trip coil. Each protection channel actuates two separate trip logic trains, one for each reactor trip breaker undervoltage trip coil. The protection system is thus inherently safe in the event of a loss of rod control power.
Channel independence is carried throughout the system extending from the sensor to the relay providing the logic. In most cases, the safety and control functions when combined are combined only at the sensor. A failure in the control circuitry does not affect the safety channel.
This approach is used for pressurizer pressure and water level channels, steam generator water level, Tavg and i1T channels, steam flow, and nuclear power range channels. The power supplied to the channels is fed from four vital instrument buses. All four of the buses are supplied by static inverters.
The initiation of the engineered safety features provided for loss-of-coolant accidents (e.g., high head safety injection, residual heat removal pumps, and containment spray systems) is accomplished from redundant signals derived from reactor coolant system and containment instrumentation. The initiation signal for containment spray comes from coincidence of two sets of two-out-of-three high-high containment pressure signals. On loss of voltage to the safety features equipment buses, the emergency diesel generators will be automatically started and connected to their respective buses.
Trip signals for the containment isolation valves are derived from either a high-high containment pressure signal (phase B), and/or a safety injection signal (phase A), and/or a containment ventilation isolation signal.
The components of the protection system are designed and laid out so that the mechanical and thermal environment accompanying any emergency situation in which the components are required to function does not interfere with that function.
Each protection channel in service at power is capable of being calibrated and tripped independently by simulated signals for test purposes to verify its operation.
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IP2 FSAR UPDATE Redundancy is provided in that there are three emergency diesel-generator sets capable of supplying separate 480-V buses. The minimum complement of safety features equipment is supplied from any two of the three emergency diesel generators.
The ability of the emergency diesel-generator sets to start within the prescribed time and to carry load is periodically checked.
An open circuit or loss of reactor trip channel power causes the system to go into its trip mode.
In a two-out-of-three circuit, the three channels are equipped with separate primary sensors and each channel is energized from independent electrical buses.
The signal for containment isolation is developed from a two-out-of-three circuit in which each channel is separate and independent and which signals for containment isolation upon loss of power. The failure of any channel to de-energize when required does not interfere with the proper functioning of the isolation circuit.
Diesel engine cranking is accomplished by a stored energy system supplied solely for the associated emergency diesel generator. The undervoltage relay scheme is designed so that loss of 480-V power does not prevent the relay scheme from functioning properly.
Reference sections:
Section Title Section Instrumentation and Control; Protection Systems 7.2 Electrical Systems 8 1.3.5 Reactivity Control (GDC 27 - GDC 32)
Reactivity control is achieved by two independent systems, the rod cluster control assemblies and the chemical and volume control system, which regulates the concentration of boric acid solution neutron absorber in the reactor coolant system. The system is designed to prevent, under anticipated system malfunction, uncontrolled or inadvertent reactivity changes, which might stress the system beyond design limits.
The reactivity control systems provided are capable of making and holding the core subcritical from any hot standby or hot operating condition, including those resulting from power changes.
The maximum excess reactivity expected for the core occurs for the cold, clean condition at the beginning-of-life (as discussed in Section 14.1.11.2).
The full length rod cluster control assemblies are divided into two categories comprising control banks and shutdown banks. The control bank of the rod cluster control assemblies is used to compensate for short-term reactivity changes at power produced due to variations in reactor power or in coolant temperature. The chemical shim control is used to compensate for the more slowly occurring changes in reactivity throughout core life such as those due to fuel depletion and fission product buildup and decay.
The shutdown banks are provided to supplement the control banks of the rod cluster control assemblies to make the reactor at least 1-percent subcritical (keff = 0.99) following trip from any credible operating condition to the hot, zero power condition assuming the most reactive rod cluster control assembly remains in the fully withdrawn position.
Chapter 1, Page 15 of 72 Revision 20, 2006 OAG10000215_0052
IP2 FSAR UPDATE Boron injection from the safety injection system supplements rod insertion and prevents exceeding core safety limits in the event of the maximum credible steam break, namely opening of a safety valve. This is accomplished with maximum worth rod fully withdrawn.
Any time that the plant is at power, the quantity of boric acid retained in the boric acid storage tanks and ready for injection always exceeds that quantity required for the normal cold shutdown.
The boric acid solution is transferred from the boric acid storage tanks into the reactor coolant by the boric acid transfer pumps and charging pumps, which can be operated from emergency diesel-generator power on loss of primary power. Boric acid can be injected by either boric acid transfer pumps and one of the three charging pumps to shut down the reactor from full power with no rods inserted. In addition, boric acid can be injected to compensate for xenon decay (xenon decay below the equilibrium operating level will not actually begin until approximately 20 hr after shutdown). Additional boric acid is employed if it is desired to bring the reactor to cold shutdown conditions.
The reactor protection systems will limit reactivity transients such that DNBR remains at or above the applicable safety analysis DNBR limit due to any single malfunction in the reactor coolant deboration controls.
Limits, which include considerable margin, are placed on the maximum reactivity worth of control rods by limiting position of insertion as a function of power and on rates at which reactivity can be increased to ensure that the potential effects of a sudden or large change of reactivity cannot (a) rupture the reactor coolant pressure boundary or (b) disrupt the core, its support structures, or other vessel internals so as to lose capability to cool the core.
The rod cluster drive mechanisms are wired into preselected groups and are prevented from being withdrawn in other than their respective groups. The rod drive mechanism is of the magnetic latch type and the coil actuation is sequenced to provide variable speed rod travel.
The maximum reactivity insertion rate is analyzed in the detailed plant analysis assuming two of the highest worth banks to be accidentally withdrawn at maximum speed, yielding reactivity insertion rates of the order of 8 x 10-4 i1k1sec, which is well within the capability of the overpower-overtemperature protection circuits to prevent core damage (as discussed in Section 14.1.1).
Reference sections:
Section Title Section Reactor; Design Bases 3.1 Instrumentation and Control; Protective Systems, 7.2 Regulating Systems 7.3 Chemical and Volume Control System 9.2 1.3.6 Reactor Coolant Pressure Boundary (GDC 33 - GDC 36)
The reactor coolant pressure boundary is shown to be capable of accommodating without further rupture, the static and dynamic loads imposed as a result of a sudden reactivity insertion such as rod ejection (as discussed in Section 14.2.6.10).
Chapter 1, Page 16 of 72 Revision 20, 2006 OAG10000215_0053
IP2 FSAR UPDATE The operation of the reactor is such that the severity of an ejection accident is inherently limited.
Since the rod cluster control assemblies are used to control load variations only and core depletion is followed with boron dilution, only the rod cluster control assemblies in the controlling groups are inserted in the core at power, and at full power these rods are only partially inserted.
A rod insertion limit monitor is provided as an administrative aid to the operator to ensure that this condition is met.
By using the flexibility in the selection of control rod groupings, radial locations and positions as a function of load, the design limits the maximum fuel temperature for the highest worth ejected rod to a value, which precludes any resultant damage to the primary system pressure boundary.
The failure of a rod mechanism housing causing a rod cluster control assembly to be rapidly ejected from the core is evaluated as a theoretical, though not a credible, accident. The analysis is discussed in Section 14.2.6.
In the core region of the reactor vessel, the V-notch toughness of the material will change during operation as a result of fast neutron exposure, which results in a shift in the nil ductility transition temperature (NOTT). This is factored into the operating procedures in such a manner that full operating pressure is not obtained until the affected vessel material is above the increased design transition temperature (OTT) and in the ductile material region. The pressure during startup and shutdown at the temperature below NOTT is maintained below the threshold of concern for safe operation.
The OTT is a minimum NOTT plus 60°F and dictates the procedures to be followed in the hydrostatic test and in station operations to avoid excessive cold stress. The value of the OTT is increased during the life of the plant as required by the expected shift in the NOTT, and as confirmed by the experimental data obtained from irradiated specimens of reactor vessel materials during the plant lifetime.
The design of the reactor vessel and its arrangement in the system provide the capability for accessibility during service life to the entire internal surfaces of the vessel and the external zones of the vessel including the nozzle to reactor coolant piping welds and the top and bottom heads. The reactor arrangement within the containment provides sufficient space for inspection of the external surfaces of the reactor coolant piping, except for the area of piping within the primary shielding concrete.
Determination of the NOTT of the core region plate forgings, weldments, and associated heat treated zones is performed in accordance with ASTM E-185, Recommended Practice for Surveillance Tests on Structural Materials in Nuclear Reactors. Samples of reactor vessel plate material have been retained and catalogued in case future engineering development shows the need for further testing.
The material properties surveillance program includes not only the conventional tensile and impact tests, but also fracture mechanics specimens. The observed shifts in NOTT of the core region materials with irradiation will be used to confirm the calculated limits to startup and shutdown transients.
To define permissible operating conditions below OTT, a pressure range is established, which is bounded by a lower limit for pump operation and an upper limit, which satisfies reactor vessel stress criteria. Since the normal operating temperature of the reactor vessel is well above the Chapter 1, Page 17 of 72 Revision 20, 2006 OAG10000215_0054
IP2 FSAR UPDATE maximum expected OTT, brittle fracture during normal operation is not considered to be a credible mode of failure.
Reference sections:
Section Title Section Reactor Coolant System; System Design and Operation, 4.2 Safety Limits and Conditions, 4.4 Inspections and Tests, 4.5 Determination of Reactor Pressure Vessel NDTT Appendix4A 1.3.7 Engineered Safety Features (GDC 37 - GDC 65)
The design, fabrication, testing and inspection of the core, reactor coolant pressure boundary, and their protection systems give assurance of safe and reliable operation under all anticipated normal, transient, and accident conditions. However, engineered safety features are provided in the facility to back up the safety provided by these components.
These engineered safety features have been designed to cope with any size reactor coolant pipe break up to and including the circumferential rupture of any pipe in that boundary assuming unobstructed discharge from both ends as discussed in Section 14.3.3.3. They are also designed to cope with any steam or feedwater line break up to and including the main steam or feedwater headers as discussed in Section 14.2.5. The total loss of all offsite power is assumed concurrent with these accidents.
The release of fission products from the reactor fuel is limited by the safety injection system, which by cooling the core and limiting the fuel clad temperature, keeps the fuel in place and substantially intact with its essential heat transfer geometry preserved and limits the metal-water reaction to an insignificant amount.
The basic design criteria for ensuring that the core geometry remains in place and substantially intact so that effective cooling of the core is not impaired following a loss-of-coolant accident:
- 1. The cladding temperature is to be less than:
- a. The melting temperature of Zircaloy-4
- b. The temperature at which gross core geometry distortion, including fragmentation, may be expected.
- 2. The total core metal-water reaction will be limited to less than 1-percent.
The safety injection system (which constitutes the emergency core cooling system) consists of high and low head centrifugal pumps driven by electric motors, and passive accumulator tanks, which are self-energized and which act independently of any actuation signal or power source.
The release of fission products from the containment is limited in three ways:
- 1. Blocking the potential leakage paths from the containment. This is accomplished by:
Chapter 1, Page 18 of 72 Revision 20, 2006 OAG10000215_0055
IP2 FSAR UPDATE
- a. A steel-lined concrete reactor containment with continuously pressurized double-sealed penetrations and liner weld channels, which form a virtually leaktight barrier to the escape of fission products should a loss-of-coolant accident occur.
- b. Isolation of process lines by the containment isolation system, which imposes double barriers in each line penetrating the containment.
- 2. Reducing the fission product concentration in the containment atmosphere. This is accomplished by containment spray, which removes elemental iodine vapor and particulates from the containment atmosphere by washing action.
- 3. Reducing the containment pressure and thereby limiting the driving potential for fission product leakage. This is accomplished by cooling the containment atmosphere by the following independent systems, each with adequate heat removal capacity:
- a. Containment spray system
- b. Containment air recirculation cooling system.
A comprehensive program of plant testing is performed for all equipment systems and system control vital to the functioning of engineered safety features. The program consists of performance tests of individual pieces of equipment in the manufacturer's shop and integrated tests of the system as a whole, and periodic tests of the actuation circuitry and mechanical components to assure reliable performance, upon demand, throughout the plant lifetime. In the event that one of the components should require maintenance as a result of failure to perform during the test according to prescribed limits, the necessary corrections or minor maintenance will be made and the unit retested immediately.
The plant is supplied with emergency power sources as follows:
- 1. Three independent emergency diesel generators, located in the Diesel Generator Building adjacent to the Primary Auxiliary Building, supply emergency power to the engineered safety features buses in the event of a loss of AC auxiliary power. There are no automatic bus ties associated with these buses. Each diesel generator is started automatically on a safety injection signal or upon the occurrence of an undervoltage condition on any vital 480-V switchgear bus. The system is sufficiently redundant such that any two diesels have adequate capacity to supply the engineered safety features for the design basis accident concurrent with a loss of offsite power. One diesel is adequate to provide power for a safe and orderly plant shutdown in the event of a loss-of-offsite electrical power.
- 2. Emergency power for vital instrumentation and control and for emergency lighting is supplied from the 125 VDC system via four independent DC channels. The station batteries supply emergency power to the instrumentation and control systems when their associated battery chargers are not available.
Chapter 1, Page 19 of 72 Revision 20, 2006 OAG10000215_0056
IP2 FSAR UPDATE For such engineered safety features as are required to ensure safety in the event of an accident, protection from dynamic effects or missiles is considered in the layout of plant equipment and missile barriers.
Layout and structural design specifically protect injection paths leading to unbroken reactor coolant loops against damage as a result of the maximum reactor coolant pipe rupture.
Injection lines penetrate the main missile barrier, and the injection headers are located in the missile-protected area between the missile barrier and the containment outside wall. Individual injection lines, separated to the maximum extent practicable, are connected to the injection header, pass through the barrier and then connect to the loops. Movement of the injection line, associated with rupture of a reactor coolant loop, is accommodated by line flexibility and by the design of the pipe supports such that no damage outside the missile barrier is possible.
In 1989, the NRC approved changes to the design basis with respect to dynamic effects of postulated primary loop ruptures, as discussed in Section 4.1.2.4.
Each engineered safety feature provides sufficient performance capability to accommodate any single failure of an active component and still function in a manner to avoid undue risk to the health and safety of the public.
Under the hypothetical accident conditions, the containment air recirculation cooling system, and the containment spray system are designed and sized so that either system, operating alone at its rated capacity, is able to supply the necessary postaccident cooling capacity to reduce rapidly the containment pressure following blowdown and cooling of the core by safety injection. Together these two systems provide the single failure protection for the containment cooling function as analyzed in Chapter 14.
All active components of the safety injection system (exception: injection line isolation valves) and the containment spray system are located outside the containment and not subject to containment accident conditions.
Instrumentation, motors, cables, and penetrations located inside the containment are selected to meet the most adverse accident conditions to which they may be subjected. These items are either protected from containment accident conditions or are designed to withstand, without failure, exposure to the worst combination of temperature, pressure, radiation, and humidity expected during the required operational period.
The reactor is maintained subcritical following a primary system pipe rupture accident.
Introduction of borated cooling water into the core results in a net negative reactivity addition (as discussed in Section 14.3.2). The control rods insert and remain inserted, except for the large break loss of coolant accident analysis where it is conservatively assumed that the control rods do not insert as discussed in Section 14.3.3.2. The delivery of cold safety injection water to the reactor vessel following accidental expulsion of reactor coolant was evaluated to ensure that this does not cause further loss of integrity of the reactor coolant system boundary (as discussed in Section 14.3.4.3.3).
Design provisions have been made to the extent practical to facilitate access to the critical parts of the reactor vessel internals, injection nozzles, pipes, valves, and safety injection pumps for visual or boroscopic inspection for erosion, corrosion, and vibration wear evidence, and for non-destructive test inspection where such techniques are desirable and appropriate.
Chapter 1, Page 20 of 72 Revision 20, 2006 OAG10000215_0057
IP2 FSAR UPDATE Design provisions are made so that active components of the safety injection system can be tested periodically for operability and functional performance. The safety injection pumps can be tested periodically during plant operation using the minimum flow recirculation lines provided.
The residual heat removal pumps are used every time the residual heat removal loop is put into operation.
An integrated safety injection system test is performed at refueling outage intervals. This test does not introduce flow into the reactor coolant system but demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry upon initiation of safety injection.
The accumulator tank pressure and level are continuously monitored during plant operation and flow from the tanks can be checked at any time using test lines.
The accumulators and the safety injection piping up to the final isolation valve are maintained full of borated water at refueling water concentration while the plant is in operation. Flow in each of the high head injection headers and in the main flow line for the residual heat removal pumps is monitored by flow and pressure instrumentation.
The design provided for capability to test initially, to the extent practical, the full operational sequence up to the design conditions for the safety injection system to demonstrate the state of readiness and capability of the system. These functional tests provided information to confirm valve operating times, pump motor starting times, the proper automatic sequencing of load addition to emergency diesel generators, and delivery rates of injection water to the reactor coolant system.
The following general criteria were followed to ensure conservatism in computing the required containment structural load capacity:
- 1. In calculating the containment pressure, rupture sizes up to and including a double-ended severance of reactor coolant pipe were considered.
- 2. In considering postaccident pressure effects, various malfunctions of the emergency systems were evaluated. Contingent mechanical or electrical failures were assumed to disable one of the emergency diesel electric generators, two of the fan coolers and one of the containment spray pumps (as discussed in Section 14.3.5.3.7).
- 3. The pressure and temperature loading obtained by analyzing various loss-of-coolant accidents, when combined with operating loads and maximum wind or seismic forces, does not exceed the load-carrying capacity of the structure, its access opening or penetrations.
Discharge of reactor coolant through a double-ended rupture of the main loop piping, followed by operation of only those engineered safety features, which can run simultaneously with power from two of the three emergency onsite diesel generators, results in a sufficiently low radioactive materials leakage from the containment structure that there is no undue risk to the health and safety of the public.
The concrete containment is not susceptible to a low temperature brittle fracture. The containment liner is enclosed within the containment and thus is not exposed to the temperature extremes of the environs. The containment ambient temperature during operation is between Chapter 1, Page 21 of 72 Revision 20, 2006 OAG10000215_0058
IP2 FSAR UPDATE 90°F and 130°F, which is well above the NDTT + 30°F for the liner material. Containment penetrations, which can be exposed to the environment are also designed to the NDTT + 30°F criterion.
The reactor coolant pressure boundary does not extend outside of the containment. Isolation valves for all fluid system lines penetrating the containment provide at least two barriers against leakage of radioactive fluids to the environment in the event of a loss-of-coolant accident.
These barriers, in the form of isolation valves or closed systems, are defined on an individual line basis. In addition to satisfying containment isolation criteria, the valving is designed to facilitate normal operation and maintenance of the systems and to ensure reliable operation of other engineered safety features.
After completion of the containment structure and installation of all penetration and weld channels, an initial integrated leakage rate test was conducted at the peak calculated accident pressure, maintained for a minimum of 24 hr, to verify that the leakage rate was not greater than 0.1-percent by weight of the containment volume per day. This leakage rate test was performed using the reference vessel method.
A leak rate test at the peak calculated accident pressure using the same method as the initial leak rate test can be performed at any time during the operational life of the plant, provided the plant is not in operation and precautions are taken to protect instruments and equipment from damage.
Penetrations are designed with double seals so as to permit pressurization of the interior of the penetration whenever a leak test is required. The system utilizes a supply of clean, dry, compressed air which places all the penetrations under an internal pressure above the peak calculated accident pressure (Peak calculated accident pressure is discussed in Section 14.3.5.1.1). Leakage from the system is checked by measurement of the integrated makeup air flow. In the event excessive leakage is discovered, each penetration can then be checked separately.
Capability is provided to the extent practical for testing the functional operability of valves and associated apparatus during periods of reactor shutdown.
Initiation of containment isolation employs coincidence circuits, which allow checking of the operability and calibration of one channel at a time.
The main steam and feedwater barriers and isolation valves in systems, which connect to the reactor coolant system are hydrostatically tested to measure leakage.
Design provisions are made to the extent practical to facilitate access for periodic visual inspection of all important components of the containment air recirculation cooling and containment spray systems.
The containment pressure-reducing systems are designed to the extent practical so that the spray pumps, spray injection valves, spray nozzles can be tested periodically and after any component maintenance for operability and functional performance.
Permanent test lines for the containment spray loop are located so that all components up to the isolation valve at the spray nozzle may be tested. These isolation valves are checked separately.
Chapter 1, Page 22 of 72 Revision 20, 2006 OAG10000215_0059
IP2 FSAR UPDATE The air test lines for checking that spray nozzles are not obstructed connect upstream of the isolation valve. Air flow through the nozzles is verified by periodic testing in accordance with the Technical Specifications.
Capability is provided to test initially to the extent practical the operational startup sequence beginning with transfer to alternate power sources and ending with near design conditions for the containment air recirculation cooling and containment spray systems.
Reference sections:
Section Title Section Containment System 5 Engineered Safety Features 6 Electrical Systems; Design Bases, 8.1 Electrical Systems Design 8.2 1.3.8 Fuel And Waste Storage Systems (GDC 66 - GDC 69)
The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations. The spent fuel storage pit is filled with borated water, normally at a similar concentration to that used in the reactor cavity and refueling canal during refueling operations. The fuel is stored vertically in an array with sufficient neutron absorbers and distance between assemblies to assure keff <1.0 even if unborated water were used to fill the pit and ::;0.95 when filled with water borated;:::: 2000 ppm boron.
During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concentration is maintained at not less than that required to maintain the reactor subcritical by 5-percent i1k1k with all the rods inserted. The refueling water boron concentration is periodically checked to ensure the proper shutdown margin.
The design of the fuel handling equipment incorporating built-in interlocks and safety features, the use of detailed refueling instructions and observance of minimum operating conditions provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety. The refueling system interlocks are verified to be functioning each refueling shutdown prior to refueling operations.
The refueling water provides a reliable and adequate cooling medium for spent fuel transfer.
Heat removal from the refueling water is provided by an auxiliary cooling system.
Adequate shielding for radiation protection is provided during reactor refueling by conducting all spent fuel transfer and storage operations underwater. This permits visual control of the operation at all times while maintaining low radiation levels for periodic occupancy of the area by operating personnel. Pit water level is alarmed and water to be removed from the pit must be pumped out as there are no gravity drains when the pit is isolated from the refueling canal.
Shielding is provided for waste handling and storage facilities to permit operation within requirements of 10 CFR 20.
Gamma radiation is continuously monitored in the fuel storage building. A high level signal is alarmed locally and is annunciated in the control room.
Chapter 1, Page 23 of 72 Revision 20, 2006 OAG10000215_0060
IP2 FSAR UPDATE Auxiliary shielding for the waste disposal system and its storage components was also designed to limit the dose rate.
All fuel and waste storage facilities are contained and equipment designed so that accidental releases of radioactivity directly to the atmosphere are monitored and will not exceed the applicable limits.
The refueling canal and spent fuel storage pit are reinforced concrete structures with a seam-welded stainless steel plate liner. These structures are designed to withstand any anticipated earthquake loadings as seismic Class I structures so that the liner should prevent leakage even in the event the reinforced concrete develops cracks.
Reference sections:
Section Title Section Sampling System 9.4 Waste Disposal System 11.1 Radiation Protection Systems 11.2 1.3.9 Plant Effluents (GDC 70)
Liquid, gaseous, and solid waste disposal facilities are designed so that discharge of effluents and offsite shipments are in accordance with applicable governmental regulations.
Radioactive fluids entering the waste disposal system are collected in sumps and tanks until determination of subsequent treatment can be made. They are sampled and analyzed to determine the quantity of radioactivity, with an isotopic breakdown if necessary. Before any attempt is made to discharge, they are processed as required and then released under controlled conditions. The system design and operation are characteristically directed toward minimizing releases to unrestricted areas. Discharge streams are appropriately monitored and safety features are incorporated to preclude excessive releases.
Radioactive gases are pumped by compressors through a manifold to one of the gas decay tanks where they are held a suitable period of time for decay. Cover gases in the nitrogen blanketing system are reused to minimize gaseous wastes. During normal operation, gases are discharged intermittently at a controlled rate from these tanks through the monitored plant vent.
The system is provided with discharge controls so that environmental conditions do not restrict the release of radioactive effluents to the atmosphere. Liquid wastes are processed to remove most of the radioactive materials. The spent resins from the demineralizers and the filter cartridges are packaged and stored onsite until shipment offsite for disposal.
Reference section:
Section Title Section Waste Disposal System
11.1 REFERENCES
FOR SECTION 1.3 Chapter 1, Page 24 of 72 Revision 20, 2006 OAG10000215_0061
IP2 FSAR UPDATE
- 1. Letter from P. Zarakas, Con Edison, to H. Denton, NRC,
Subject:
Actions Taken to Comply with NRC Confirmatory Order of February 11, 1980, dated August 11, 1980.
- 2. Deleted 1.4 DESIGN PARAMETERS AND PLANT COMPARISON 1.4.1 Design Highlights The original design of the plant is based upon proven concepts, which have been developed and successfully applied in the construction of pressurized water reactor systems. In subsequent paragraphs, the original design features of the plant are discussed.
1.4.1.1 Power Level The initial license application power level for Indian Point Unit 2 was 2758 MWt. The increase in this power rating over 1473 MWt for Connecticut Yankee was achieved by a 44-percent increase in heat transfer surface area and a 31-percent increase in average heat flux. The increased heat transfer surface area is due to 22-percent more fuel rods, each 20-percent longer.
The increase in maximum heat flux and the 13.4 kW/ft linear heat generation rate (LHGR) resulting are justified by the results of incore experiments by Westinghouse and others.
1.4.1.2 Reactor Coolant Loops The reactor coolant system for the Indian Point Unit 2 consists of four loops as compared with three loops for San Onofre and four loops for Connecticut Yankee. The use of four loops in the Indian Point Unit 2 for the production of 2758 MWt requires an attendant increase in the size and capacity of the reactor coolant system components such as the reactor vessel, reactor coolant pumps, piping, and steam generators. These increases represent reasonable engineering extrapolations of existing and proven designs.
1.4.1.3 Peak Specific Power Based on values for hot channel factors, reactivity coefficients, and other design parameters, which were established in the PSAR and are supported by the previous experience with other plants of the same type, this reactor can be operated safely at power levels at least as high as the license application rating.
1.4.1.4 Fuel Cladding The fuel rod design for the plant employs zircaloy as a cladding material. This clad has proven successful in numerous operating facilities. The fuel rod dimensions are identical to those in Ginna, Salem, and Zion Station Units 1 and 2.
Chapter 1, Page 25 of 72 Revision 20, 2006 OAG10000215_0062
IP2 FSAR UPDATE 1.4.1.5 Fuel Assembly Design The fuel assembly incorporates the rod cluster control concept in a canless 15 x 15 fuel rod assembly using a spring clip grid to provide support for the fuel rods. Extensive out-of-pile and in-pile tests have been performed on this concept; operating experience is available from numerous facilities.
1.4.1.6 Moderator Temperature Coefficient of Reactivity The reactor has a negative moderator temperature coefficient of reactivity at operating temperature at all times throughout core life (as discussed in Section 14.1.11.2).
1.4.2 IP2 - IP3 Design Differences An NRC Confirmatory Order (Reference 1) required the Licensees (Consolidated Edison and the Power Authority) to jointly review and identify the significant differences between Indian Point Units 2 and 3 and evaluate these differences in light of the regulatory standards and requirements in existence at the time. Consolidated Edison determined, evaluated, and provided justification for each design difference in submittals to the NRC for acceptance (References 2, 3). These design differences were found acceptable in an NRC Safety Evaluation Report (Reference 4).
REFERENCES FOR SECTION 1.4
- 1. Letter from Harold R. Denton, NRC, to Consolidated Edison, "Confirmatory Order", dated February 11, 1980.
- 2. Letter from William J. Cahill, Consolidated Edison, to Harold R. Denton, NRC, "Confirmatory Order", dated May 9, 1980.
- 3. Letter from John D. O'Toole, Consolidated Edison, to Steven A. Varga, NRC, "Confirmatory Order", dated May 27, 1982.
- 4. Letter from Steven A. Varga, NRC, to John D. O'Toole, Consolidated Edison, "Confirmatory Order", dated December 1, 1982.
1.5 RESEARCH AND DEVELOPMENT REQUIREMENTS Research and development were conducted relating to finalization of core design details and parameters, air recirculation system halogen filters, failure of core cooling systems and means to ameliorate consequences, emergency core cooling system, control rod ejection analysis, and reactor coolant pump controlled leakage seals.
The detailed final core design and thermal-hydraulics and physics parameters have been completed. The nuclear design including fuel configuration and enrichments, control rod pattern and worths, reactivity coefficients and boron requirements are presented in Section 3.2.1 and the final thermal-hydraulics design parameters are in Section 3.2.2. Section 3.2.3 presents the final fuel, fuel rod, fuel assembly, and control rod mechanical design. The core design incorporates fixed burnable poison rods 1 in the initial loading to ensure a negative moderator Chapter 1, Page 26 of 72 Revision 20, 2006 OAG10000215_0063
IP2 FSAR UPDATE reactivity temperature coefficient at operating temperature. This improves reactor stability and lessens the consequences of a rod ejection or loss-of-coolant accident.
Core stability has been analyzed 2-4 and design provisions for detection and control of potential xenon oscillations have been finalized 3 . The original core design incorporated part-length control rods for controlling these xenon oscillations and shaping the axial power distribution.
These have since been found unnecessary and removed from the reactor. X-Y control is not required and therefore not provided. Tests in operating reactors demonstrate the ability of the out-of-core instrumentation to give accurate indication of power redistribution and provide the operator information necessary to monitor redistributions and control axial oscillations by moving the rods in a prescribed pattern 3 . This capability was verified during startup tests in the Indian Point Unit 2 Plant.
Full-size filter tests were conducted for the Connecticut Yankee Atomic Power Company to demonstrate the efficiency for iodine removal under the most extreme conditions anticipated in the postaccident containment environment. The results of these tests 5 filed with the former U.
S. Atomic Energy Commission (now the U. S. Nuclear Regulatory Commission) under Docket No. 50-213, are directly applicable to the charcoal filter system originally employed in this plant.
The charcoal filters are no longer required by the Technical Specifications and the radiological consequences analysis presented in Section 14.3.6 does not credit the filters.
A program for development of a crucible system design, which would contain the reactor core assuming failure of the core cooling system to prevent a core meltdown, was undertaken. A scheme for containing the molten core in a water submerged high melting point refractory lined steel crucible resulted. Refractory materials and crucible physical design were investigated along with analysis of the temperature distribution expected with the molten core and crucible refractory, and steam and water recirculation paths. As a result of uncertainties in material properties at the high application temperature, the lack of experimental proof that the boiling core mass would dissipate its heat upward through the water cover, and the possibility of violent liquid metal-water reactions, it became apparent that the proper emphasis for research and development on the loss-of-coolant accident should be increased emphasis placed on research and development for emergency core cooling system improvement in order to eliminate need for a crucible. This is supported by the conclusions of the Report of Advisory Task Force on Power Reactor Emergency Cooling, "Emergency Core Cooling," USAEC.
This additional development effort on emergency core cooling system design resulted in the modification of the system to include pressurized accumulator tanks for rapid core reflooding.
This increased flooding capability limits the clad temperature after a loss-of-coolant accident to well below the melting temperature of Zircaloy-4, minimizes metal-water reaction and ensures that the core remains in place and intact thereby ensuring preservation of essential heat transfer geometry. The system design incorporates redundancy of components such that the minimum required water addition rates can be met assuming any active component to fail concurrent with the loss-of-coolant accident or, over the long term period of postaccident core decay heat removal a passive or active component failure in either the safety injection or service water systems, or an active failure in the component cooling water system. Details of system design and operation are given in Chapters 6 and 9, and analysis of the loss-of-coolant accident is presented in Section 14.3. Because of the incorporation of this revised emergency core cooling system, the reactor pit crucible was deleted from the plant design. Although it is not required to provide cooling for molten fuel in the bottom of the reactor vessel with the upgraded emergency core cooling system performance, the clearance between the insulation and the instrumentation penetrations with the pressure relief holes in the insulation at the top of the vessel provide Chapter 1, Page 27 of 72 Revision 20, 2006 OAG10000215_0064
IP2 FSAR UPDATE assurance that water in the flooded reactor vessel cavity will be in contact with the vessel. No other provisions for direct vessel cooling are provided or required.
A control rod ejection analysis was performed for the final core design, rod worths, rod position limits, and moderator reactivity temperature coefficient. As mentioned above, the addition of the burnable poison rods eliminates power operation with a positive moderator temperature coefficient and reduces the severity of the ejected rod accident, hence, lessening the need for research and development on this subject. The analysis is presented in Section 14.2.6.
The reactor coolant pump controlled leakage seal design for this plant has been fully developed.
A full scale mock-up of this seal was operated for over 100 hr to confirm that seal deflection and leak rate under load were acceptable. The full scale mock-up has been used during the development of the controlled leakage seal to provide information related to long-term performance. One of the two seals used in this plant was operated about 300 hr and the other about 100 hr, each in its pump motor unit. During hot functional testing in the plant, before the core was loaded, additional operation brought the total operating time for each seal to well over 500 hr. Successful operation of similar seals had previously been demonstrated with over 5000 hr total running time in San Onofre and over 3000 hr in Haddam Neck.
REFERENCES FOR SECTION 1.5
- 1. P. M. Wood, E. A. Bassler, et aI., Use of Burnable Poison Rods in Westinghouse Pressurized Water Reactors, WCAP-7113, October 1967.
- 2. P. M. Wood, J. M. Gallagher, R. M. Metz, R. A. Dean, Use of Part-Length Absorber Rods in Westinghouse Pressurized Water Reactors, WCAP-7072, June 1967.
- 3. Westinghouse Electric Corporation, Power Distribution Control of Westinghouse Pressurized Water Reactors, WCAP-7208, October 1968.
- 4. Westinghouse Electric Corporation, Power Maldistribution Investigations, WCAP-7407 -L (proprietary), January 1970.
- 5. Connecticut Yankee Charcoal Filter Tests, CYAP-101, December 1966.
1.6 IDENTIFICATION OF CONTRACTORS [Historical Information Only1 The Indian Point Unit 2 was designed and built by the Westinghouse Electric Corporation as prime contractor for Con Edison. Westinghouse undertook to provide a complete, safe, and operable nuclear power plant ready for commercial service. The project was directed by Westinghouse from the offices of its Atomic Power Division in Pittsburgh, Pennsylvania, and by Westinghouse representatives at the plant site during construction and plant startup.
Westinghouse engaged United Engineers and Constructors of Philadelphia, Pennsylvania, to provide the design of certain portions of the plant.
The plant construction was under the general direction of Westinghouse through United Engineers and Constructors, which was responsible for the management of all site construction activities and either performed or subcontracted the work of construction and equipment Chapter 1, Page 28 of 72 Revision 20, 2006 OAG10000215_0065
IP2 FSAR UPDATE erection. Preoperational testing of equipment and systems at the site and initial plant operation was performed by Con Edison personnel under the technical direction of Westinghouse.
1.7 PROJECT REORGANIZATION - DECEMBER 1969 [Historical Information Only]
This section describes a reorganization in project management, which was implemented by Westinghouse in December 1969.
Westinghouse formed a wholly-owned subsidiary corporation, called WEDCO Corporation, to perform certain functions at the Indian Point site of Con Edison. Westinghouse remained the prime contractor and continued to exercise overall control and to have full responsibility for the Indian Point 2 project. WEDCO performed, under Westinghouse, project management, engineering, quality assurance, construction, and procurement functions for Indian Point Unit 2.
These functions were previously carried out by Westinghouse or United Engineers and Constructors (UE&C).
The entire Westinghouse senior management organization, which prior to the advent of WEDCO, was responsible for the Westinghouse effort at Indian Point Unit 2, remained responsible. All other personnel within Westinghouse senior management who, prior to WEDCO, carried any responsibility in any area for Indian Point Unit 2 continued to carry those responsibilities, regardless of the formation of WEDCO or changes in title or designation.
Furthermore, WEDCO had behind it the full organization and strength of the Westinghouse Electric Corporation: Westinghouse engineering, legal, and other personnel continued to be available for the project.
The functional relationships among Con Edison, Westinghouse, WEDCO, and UE&C are shown in Historical Figure 1.7-1.
Westinghouse WEDCO-UE&C Relationship.
Westinghouse retained UE&C as its architect-engineer-constructor to perform certain work and services in connection with the plant. Initially, UE&C performed services within its scope in the following areas:
- 1. Design and Engineering
- 2. Procurement
- 3. Construction Management and Construction
- 4. Quality Assurance (including Home Office Quality Control Engineering, Vendor Surveillance and Onsite Quality Control)
Westinghouse removed items (2) and (3) from the scope of work to be performed by UE&C and assigned these functions to WEDCO. In these areas, however, UE&C provided qualified personnel to assist in effectuating the transition of work to Westinghouse and WEDCO.
UE&C continued to have responsibility for all of the design and engineering functions and all of the quality assurance functions, including home office quality control engineering, vendor surveillance and onsite quality control, for which it had responsibility prior to the advent of Chapter 1, Page 29 of 72 Revision 20, 2006 OAG10000215_0066
IP2 FSAR UPDATE WEDGO. UE&G continued to have direct corporate responsibility to Westinghouse for all of the work within its scope.
In its organizational structure, WEDCO exercised a high level quality and engineering reliability function. This function included the activities previously performed by the Nuclear Power Service Staff Resident Quality Assurance Engineer, and in addition included the centralization and overall management for quality assurance activities previously performed by various organizations. This function was carried out by a Reliability Manager who was located at the site. The Reliability Manager was responsible for surveillance visits to selected shops or suppliers. This function was previously delegated to the Westinghouse Nuclear Power Services Group. In addition, the Reliability Manager continually audited the quality assurance efforts of UE&G. In effect, a new reliability management function over and above those previously set forth was established while all existing organizational functions and responsibilities for quality assurance were maintained.
The quality control functions previously performed at various Westinghouse organizational levels continued to be performed. At the Westinghouse headquarters level, the staff quality assurance audit team reviewed periodically the quality control program for Indian Point Unit 2 as it had done in the past. At the Westinghouse PWR Systems Division level, the quality control functions performed by that division for the nuclear steam supply system continued as before.
Con Edison.
The project reorganization described did not in any way alter the ultimate responsibility of Con Edison for the quality assurance program. There was no basic change in the Con Edison program. However, the following minor procedural changes were made in view of the existence of WEDCO:
- 1. Con Edison's monitoring function included monitoring the activities of WEDGO.
- 2. Con Edison forwarded the United States Testing Company quality assurance reports to Westinghouse and/or WEDCO.
- 3. Con Edison contacted Westinghouse and/or WEDCO for necessary corrective action.
1.7 FIGURES Figure No. Title Figure 1.7-1 Functional Relationships [Historical]
1.8 PROJECT REORGANIZATION - MARCH 1970 [Historical Information Only1 This section describes a change, which was implemented by Westinghouse in the spring of 1970 in the project organization.
The changes made in December 1969 (see Section 1.7) involved the creation of WEDCO and the delegation to WEDCO by Westinghouse of certain functions at the Indian Point site Chapter 1, Page 30 of 72 Revision 20, 2006 OAG10000215_0067
IP2 FSAR UPDATE previously carried out by Westinghouse or United Engineers and Constructors (UE&C).
Following the December 1969 reorganization, UE&C retained the following functions within its scope of work as architect-engineer-constructor:
- 1. Design and Engineering.
- 2. Quality Assurance (including Home Office Quality Control Engineering, Vendor Surveillance, and Onsite Quality Control).
The change consisted of the removal of the vendor surveillance and onsite quality control portions of item (2) from the scope of work to be performed by UE&C, and the assigning of these functions to WEDCO.
There was little change of personnel involved in the transfer of the onsite quality control function. WEDCO employed a Manager of Vendor Surveillance and other personnel for this work. The transition in this respect was gradual. New personnel were phased in and the UE&C personnel were used during the transition period to ensure continuity of the surveillance program. The transfer was made on a purchase-order-by-purchase-order basis, with UE&C personnel working with new personnel in performing the surveillance during the transition.
To assure that a level of quality assurance review was not lost, the organization of the WEDCO reliability group was structured to provide for an independent, internal audit of the two quality assurance functions transferred to WEDCO. The Vendor Surveillance Group and Onsite Quality Control Group each reported directly to the Reliability Manager.
The activities of both the Vendor Surveillance Group and the Onsite Quality Control Group were audited by a Systems Reliability Group. The Systems Reliability Group reported directly to the Reliability Manager to ensure its functional independence. Historical Figure 1.8-1 shows these organizational relationships in chart form.
1.8 FIGURES Figure No. Title Figure 1.8-1 Organization Chart WEDCO Reliability Group {Historical]
1.9 SUPPLEMENTS AND REVISIONS TO ORIGINAL FSAR 1.9.1 Supplements Supplement 1 to the Indian Point Unit 2 Final Safety Analysis Report consisted of responses to questions from the Atomic Energy Commission as contained in two letters. The first letter from Peter A. Morris, Director of the Division of Reactor Licensing, on March 5, 1969, to Mr. Donham Crawford of Con Edison, requested additional information on the medical plans and facilities at Indian Point. The questions and responses are found following Tab I of Volume 5 of the original FSAR. These responses were incorporated into Section 11.2.5 of the original FSAR as page changes. The responses to the questions in Volume 5 indicate where the specific answer may be found in the page change.
The second letter to Arthur N. Anderson of Con Edison from Peter A. Morris, dated August 4, 1969, requested additional information on Chapters 1, 2, 3, 4, 5, 6, 7, 8, 11, 12, and 14 of the Chapter 1, Page 31 of 72 Revision 20, 2006 OAG10000215_0068
IP2 FSAR UPDATE original FSAR. Supplement 1 responded to several of the questions in the second letter found behind Tab II of Volume 5 of the original FSAR. The responses consisted of questions and answers given in Volume 5 of the original FSAR and also of page changes to the original text of the FSAR in some instances.
Supplement 2 to the Indian Point Unit 2 Final Safety Analysis Report consisted of responses to questions from the Atomic Energy Commission and page changes to the report. The questions were contained in a letter to Arthur N. Anderson of Con Edison from Peter A. Morris, Director of the Division of Reactor Licensing, dated August 4, 1969. The responses consisted of questions and answers added to Volume 5 of the original FSAR in the proper order behind Tab II. Page changes for the FSAR were included with Supplement No.2.
Supplement 3 to the Indian Point Unit 2 Final Safety Analysis Report consisted of responses to questions from the Atomic Energy Commission and page changes to the report. The questions were contained in a letter to Arthur N. Anderson of Con Edison from Peter A. Morris, Director of the Division of Reactor Licensing, dated August 4, 1969. The responses consisted of questions and answers added to Volume 5 of the original FSAR in the proper order behind Tab II. This supplement responded to several questions concerning Chapters 1, 4, 5, 7, 8, and 11 of the report.
Supplement 4 to the Indian Point Unit 2 Final Safety Analysis Report consisted of responses to questions from the Atomic Energy Commission and page changes to the report. The questions were contained in a letter to Arthur N. Anderson of Con Edison from Peter A. Morris, Director of the Division of Reactor Licensing, dated August 4, 1969. The responses consisted of questions and answers added to Volume 5 of the original FSAR in the proper order behind Tab II. Also included with this supplement was a description of the project reorganization within Westinghouse. This supplement also responded to several questions concerning Chapters 4, 5, 7, 11, and 14 of the report.
Supplement 5 to the Indian Point Unit 2 Final Safety Analysis Report consisted of responses to questions from the Atomic Energy Commission. The questions were contained in a letter to Arthur N. Anderson of Con Edison from Peter A. Morris, Director of the Division of Reactor Licensing, dated August 4, 1969, and a letter to William J. Cahill, Jr., of Con Edison from Peter A. Morris dated November 13, 1969. The responses consisted of questions and answers added to Volume 5 of the original FSAR in the proper order behind Tab II. The supplement responded to several questions concerning Chapters 1,4,6, 11, 12, and 14 of the report.
Supplement 6 to the Indian Point Unit 2 Final Safety Analysis Report consisted of responses to questions from the Atomic Energy Commission. The questions were contained in a letter to Arthur N. Anderson of Con Edison from Peter A. Morris, Director of the Division of Reactor Licensing, dated August 4, 1969, and a letter to William J. Cahill, Jr., of Con Edison from Peter A. Morris dated November 13, 1969. The responses consisted of questions and answers added to Volume 5 of the original FSAR in the proper order behind Tab II. The supplement responded to several questions concerning Chapters 1, 3, 4, 6, 9, and 14 of the report. Also included with this supplement was the Indian Point Unit 2 Containment Design Report.
Supplement 7 to the Indian Point Unit 2 Final Safety Analysis Report consisted of responses to questions from the Atomic Energy Commission and page changes to the report. The questions were contained in a letter to Arthur N. Anderson of Con Edison, from Peter A. Morris, Director of the Division of Reactor Licensing, dated August 4, 1969, and a letter to William J. Cahill, Jr., of Chapter 1, Page 32 of 72 Revision 20, 2006 OAG10000215_0069
IP2 FSAR UPDATE Con Edison, from Peter A. Morris, dated November 13, 1969. This supplement responded to several questions concerning Chapters 4,5,6,9, 13, and 14 of the report.
Supplement 8 to the Indian Point Unit 2 Final Safety Analysis Report consisted of responses to questions from the Atomic Energy Commission and page changes to the report. The questions were contained in a letter to Arthur N. Anderson of Con Edison, from Peter A. Morris, Director of the Division of Reactor Licensing, dated August 4, 1969, and a letter to William J. Cahill, Jr., of Con Edison, from Peter A. Morris, dated November 13, 1969. The responses consisted of questions and answers added to Volume 5 of the original FSAR. The supplement responded to questions concerning Chapters 4,6,7, and 13 of the report.
$yppl~m~dt$,1Q,1@,14,§d~@Q to the Indian Point Unit 2 Final Safety Analysis Report consisted of corrections and additional information for the original FSAR in the form of page changes.
Supplement No. 11 to the Indian Point Unit 2 Final Safety Analysis Report provided the proposed Technical Specifications for operation of the facility in accordance with the rules of practice, 10 CFR 50.36.
Supplement 13 to the Indian Point Unit 2 Final Safety Analysis Report consisted of responses to questions from the Atomic Energy Commission contained in a letter from Peter A. Morris, Director of the Division of Reactor Licensing, on July 24, 1970, to William J. Cahill, Jr., of Con Edison. The letter requested additional information on Chapters 1, 4, 7, 8, 12, and 14 of the original FSAR. The responses consisted of questions and answers given in Volume 5 of the FSAR and also of page changes to the original text of the FSAR in some instances.
Supplement 15 to the original Final Safety Analysis Report consisted of correction pages that updated certain areas where final design parameters were available and where design modifications had resulted from AEC review. In addition, a cross-reference index was submitted for each chapter of the FSAR where required. The index referenced the responses to questions in Volumes 5 and 6 where additional information could be found concerning specific sections.
The proposed Technical Specifications were reissued in their entirety with this supplement.
This issue superseded the specifications submitted in Supplement 11.
Supplement 18 to the original Final Safety Analysis Report consisted of the relocation of information from the site Custom Technical Specifications into the UFSAR for items and topics that were no longer found in the Improved Technical Specifications. It also updated references to the new Technical Specification sections, to information relocated from the Technical Specifications into the Off Site Dose Calculation Manual (ODCM) and added cross references to the new Technical Requirements Manual (TRM).
Supplement 19 to the original Final Safety Analysis Report consisted of corrections and additional information for the original FSAR in the form of changes to reflect several plant modifications, changes to reflect 10 CFR 100.11, the new fuel design and new core design for Cycle 17 and Cycle 16 Core Reload Design, the permanent increase in Tave to 562°F, and the approved alternate source term fuel handling accidents (FHB & VC) which take no credit for charcoal filtration. Changes were also included from NRC approved projects, including Appendix "K" Power Uprate [1.4% Power Uprate] with the re-analysis of some of the Chapter 14 accidents to account for the 1.4% power uprate, re-analysis of the Loss of Electrical Load transients and LONE/LOOP transients, and the re-analysis of the Feedwater System Chapter 1, Page 33 of 72 Revision 20, 2006 OAG10000215_0070
IP2 FSAR UPDATE Malfunction with a step increase of 120% of nominal feedwater flow to one steam generator, and to reflect the approved Stretch Power Uprate to 3216 MWt.
1.9.2 Revisions Pursuant to 10 CFR 50.71(e), Con Edison submitted an updated Final Safety Analysis Report for Indian Point Unit 2 on July 22, 1982, reflecting changes made up to a maximum of 6 months prior to the submittal date. In addition, the following revisions to the updated Final Safety Analysis Report have been submitted to date:
Revision 1, July 1983 Revision 2, July 1984 Revision 3, July 1985 Revision 4 July 1986 Revision 5, June 1987 Revision 6, June 1988 Revision 7, June 1989 Revision 8, June 1990 Revision 9, June 1991 Revision 10, June 1992 Revision 11, June 1993 Revision 12, June 1994 Revision 13, December 1995 Revision 14, December 1997 Revision 15, December 1999 Revision 16, July 2001 Revision 1 May 2003 Revision 18, October 2003 Revision 19,I\IIC3y~995 R~NI'i§iql'l@Q,NQV~mb~1"400P Revision 2, in addition to reflecting required changes, incorporated a major editorial effort to standardize the format of the updated FSAR and to correct typographical, grammatical, and syntax errors, so that the material is presented in a more uniform and clear manner. The changes of technical content and some major editorial changes were marked in the margins with the numeral 2. The majority of editorial changes were minor and were not marked individually. A changed page, however, was indicated by the label Revision 2 at the lower right hand corner.
1.10 QUALITY ASSURANCE PROGRAM 1.10.1 General Entergy's Quality Assurance Program (QAP) for Indian Point Unit 2 is in accordance with the quality assurance requirements of 10 CFR 50 Appendix B. The QAP is described in a Quality Assurance Program Manual, which satisfies the criteria of Appendix B. Changes to the program description are submitted to the NRC in accordance with the provisions of 10 CFR 50.54(a)(3).
Chapter 1, Page 34 of 72 Revision 20, 2006 OAG10000215_0071
IP2 FSAR UPDATE 1.10.2 Scope The Quality Assurance Program provides control for activities affecting the quality of structures, systems, and components of the plant and their operation to the extent consistent with their importance to safety. Those structures, systems, and components of the plant that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public are designated "Class A" as described in the Quality Assurance Program. All items and activities affecting safety addressed in Regulatory Guide 1.29 "Seismic Design Classification" revision 3, September 1978, are also governed by the Quality Assurance Program. A list of Class A items is maintained. Elements of the Quality Assurance Program are also applicable to activities and items affecting safety as defined in Licensing commitments. 2 It is recognized that not every portion of each of the listed systems and components affect the safety related function. Therefore, allowance is made for subcomponents of systems to be declassified. When such is the case, the agreement is appropriately documented identifying the parts or subcomponents concerned and showing appropriate concurrences.
1.10.3 Organization And Responsibilities The major organizations or groups participating in the Quality Assurance Program are: Nuclear Quality Assurance and Oversight, Nuclear Power, Nuclear Power Engineering, and the Safety Review Committee. The duties and responsibilities of the individuals participating in the Quality Assurance Program are described in procedures, or manuals.
REFERENCES FOR SECTION 1.10
- 1. Deleted
- 2. Letter from John D. O'Toole, Con Edison, to Director of Nuclear Reactor Regulation, NRC,
Subject:
Response to NRC letter of September 23, 1980 to Mr.
Zarakas requesting information on the Quality Assurance Program for Indian Point Unit 2 dated March 11, 1981.
TABLE 10.1-1 DELETED 1.11 DESIGN CRITERIA FOR STRUCTURES AND COMPONENTS 1.11.1 Definition Of Seismic Design Classifications All structures and components are classified as seismic Class I, Class II, or Class III as recommended in:
- 1. TID-7024, "Nuclear Reactors and Earthquakes," August 1963 and,
- 2. G. W. Housner, "Design of Nuclear Power Reactors Against Earth-quakes,"
Proceedings of the Second World Conference on Earthquake Engineering, Volume I, Japan, 1960, Pages 133, 134 and 137.
Chapter 1, Page 35 of 72 Revision 20, 2006 OAG10000215_0072
IP2 FSAR UPDATE Class I Seismic Class I is defined as those structures and components including instruments and controls whose failure might cause or increase the severity of a loss-of-coolant accident or result in an uncontrolled release of radioactivity causing more than 10 rem to the thyroid or 10 rem whole body to the average adult beyond the nearest site boundary. Also included are those structures and components vital to safe shutdown and isolation of the reactor.
Class II Class II is defined as those structures and components, which are important to reactor operation but not essential to safe shutdown and isolation of the reactor and whose failure could result in the release of radioactivity causing more than 1.0 rem to the thyroid or 0.5 rem whole body dose to the average adult beyond the nearest site boundary.
Class III Class III is defined as those structures and components, which are not directly related to reactor operation or containment. In Indian Point Unit 2, the only portions of the plant, which are not seismic Class I and which might carry substantial radioactivity because of required safeguards operation or requirements for safe shutdown and isolation of the reactor are portions of the chemical and volume control system and waste disposal system.
The specific components in the chemical and volume control system are the volume control tank, holdup tank, and the concentrates holding tank with associated piping, valves and supports. These components are all seismic Class I. In addition, the design of the system tanks and their location were based upon the commitment that a vessel rupture would not cause doses in excess of 10 CFR 20 limits at the exclusion radius.
The specific components in the waste disposal system are the gas decay tanks with their associated piping, valves and supports. These components are all seismic Class I. In addition, the gas decay tanks of the waste disposal system have been designed such that the failure of any tank will not exceed 10 CFR 20 doses at the exclusion radius.
The analysis showing that the rupture of the volume control tank or a gas decay tank does not exceed the special dose limits selected for Indian Point Unit 2 is found in Section 14.2.3 of the FSAR.
Those components of the chemical and volume control system that are not seismic Class I are as follows: batching tank, monitor tanks, monitor tank pumps, chemical mixing tank, and resin fill tank. In addition, the boric acid evaporator and the condensate demineralizer are not seismic Class I.
Those components of the waste disposal system, which are not seismic Class I include: waste condensate tank and pumps.
Failure of these components will not result in offsite doses in excess of 10 CFR 20 limits at the site exclusion radius.
All components, systems, and structures classified as seismic Class I are designed in accordance with the following criteria:
Chapter 1, Page 36 of 72 Revision 20, 2006 OAG10000215_0073
IP2 FSAR UPDATE
- 1. Primary steady state stresses, when combined with the seismic stress resulting from the response to a ground acceleration of 0.05g acting in the vertical and 0.10g acting in the horizontal planes simultaneously, are maintained within the allowable stress limits accepted as good practice and, where applicable set forth in the appropriate design standards, e.g., ASME Boiler and Pressure Vessel Code, USAS B31.1 Code for Pressure Piping, ACI 318 Building Code Requirements for Reinforced Concrete, and AISC Specifications for the Design and Erection of Structural Steel for Buildings.
- 2. Primary steady state stresses when combined with the seismic stress resulting from the response to a ground acceleration of 0.10g acting in the vertical and 0.15g acting in the horizontal planes simultaneously, are limited so that the function of the component, system or structure shall not be impaired as to prevent a safe and orderly shutdown of the plant.
All Class II structures and components are designed on the basis of a static analysis for a ground acceleration of 0.05g acting in the vertical and 0.10g acting in the horizontal directions simultaneously.
The structural design of all Class III structures meets the requirements of the applicable building code, which is the "State Building Construction Code" State of New York, 1961. This code does not reference the Uniform Building Code.
The Original Steam Generator Storage Facility (OSGSF) has been constructed for the storage of the original steam generators. The OSGSF is a seismic Class III structure, designed in accordance with the requirements of the State of New York Official Compilation of Codes, Rules and Regulations, Title 9, Subtitle S, 1995 edition, copyright 1999, and the American Concrete Institute (ACI) 318, Building Code Requirements for Structural Concrete, 1999.
Table 1.11-1 gives the damping factors used in the design of components and structures.
The design of seismic Class I structures and components utilizes the "response spectrum" approach in the analysis of the dynamic loads imparted by the earthquake. The analysis is based upon the response spectra shown on Figures 1.11-1 and 1.11-2.
The following method of analysis is applied to seismic Class I structures and components, including instrumentation:
- 1. The natural period of vibration of the structure or component is determined.
- 2. The response acceleration of the component to the seismic motion is taken from the response spectrum curve at the appropriate period.
- 3. Stresses and deflections resulting from the combined influence of normal loads and the seismic load due to the design earthquake (0.05g acting in the vertical and 0.10g acting in the horizontal planes simultaneously) are calculated and checked against the limits imposed by the design standard.
- 4. Stresses and deflections resulting from the combined influence of normal loads and the seismic loads due to the maximum potential earthquake (0.10g acting in Chapter 1, Page 37 of 72 Revision 20, 2006 OAG10000215_0074
IP2 FSAR UPDATE the vertical and 0.15g acting in the horizontal planes simultaneously) are calculated and checked to verify that deflections do not cause loss of function and that stresses do not produce rupture.
Where the vibrator system is of a highly complex geometric shape, such as piping systems, the maximum response from the response curve with the appropriate damping factor is selected. By using this conservative value and demonstrating that the stresses are satisfactory, it becomes unnecessary to perform any further analysis to determine the natural periods of the system.
For a further discussion of the models and methods used for the seismic Class I design of structures, equipment, piping, instrumentation and controls, see Section 1.11.4.
1.11.2 Classification Of Particular Structures And Equipment Examples of particular structure and equipment classifications are given below. These classifications are not intended to be all-inclusive.
Item Buildings and Structures Containment (including all penetrations and airlocks, the concrete shield, the liner, and the interior structures)
Spent fuel pit Control Building Diesel Generator Building Intake structure (to the extent that water is always available to the service water pumps)
Service water screenwell Primary Auxiliary Building Turbine Building III Buildings containing conventional facilities Such as the Maintenance and Outage Building III Original Steam Generator Storage Facility III Equipment. Piping, and Supports
[Note - Class I components (equipment, piping, instrumentation, etc.) located in or supported on a Class /I structure are protected from earthquake damage or are backed up by other Class I components located in or supported by a Class I structure.]
Reactor control and protection system Chapter 1, Page 38 of 72 Revision 20, 2006 OAG10000215_0075
IP2 FSAR UPDATE Radiation monitoring system Process instrumentation and controls Reactor Vessel and its supports Vessel internals Fuel assemblies Rod cluster control assemblies and drive mechanisms Supporting and positioning members Incore instrumentation structure Reactor coolant system Piping and valves (including safety and relief valves)
Steam generators Pressurizer Reactor coolant pumps Supporting and positioning members Main Steam system, up to and including the isolation valve Engineered safety features Safety injection system (including safety injection and residual heat removal pumps, refueling water storage tank, accumulator tanks, residual heat removal heat exchangers and connecting piping and valving)
Containment spray system (including spray pumps, spray headers, and connecting piping and valving)
Containment air recirculation cooling system (including fans, coolers, ducts, valves, and demisters)
Auxiliary building ventilation system Condensate storage tanks Pressurizer relief tank II Residual heat removal loop Containment penetration and weld channel pressurization system Component cooling loop Instrument air system (essential sections)
Isolation valve seal water system Chapter 1, Page 39 of 72 Revision 20, 2006 OAG10000215_0076
IP2 FSAR UPDATE Sampling system II Spent fuel pit cooling loop II Fuel transfer tube Emergency power supply system Diesel generators and fuel oil storage tank DC power supply system Power distribution lines to equipment required for transformers and switchgear supplying the engineered safety features Control panel boards Motor control centers Control Equipment, facilities and lines necessary for the above seismic Class I items Waste disposal system Chemical drain tank Waste holdup tanks Gas decay tanks Reactor coolant drain tank Compressors Waste holdup tank pumps Interconnecting waste gas piping Waste disposal system II or III All elements not listed as seismic Class I Containment crane Manipulator and other cranes III Conventional equipment, tanks and piping, other than Classes I and II III Auxiliary boiler feed and service water pumps and piping The chemical and volume control system is considered seismic Class I except for the components listed below:
Batching tank II Monitor tanks III Chapter 1, Page 40 of 72 Revision 20, 2006 OAG10000215_0077
IP2 FSAR UPDATE Monitor tank pumps III Chemical mixing tank II Resin fill tank III 1.11.3 Design Criteria For Seismic Class I Structures And Equipment The criteria for functional adequacy of structures, equipment, piping, instrumentation, and controls follow.
No loss of function implies that rotating equipment will not freeze, pressure vessels will not rupture, supports will not collapse under the load, systems required to be leaktight will remain leaktight, and components required to respond actively (such as valves and relays) will respond actively.
The criteria for functional adequacy of the structures state stresses will not exceed yield when subjected to a 0.15g ground acceleration. The manner in which these criteria have been met is by limiting stresses in seismic Class I structures to meet the above criteria.
For all seismic Class I piping and their supports, the criteria for functional adequacy and the manner in which the criteria are met are the following:
with a ground acceleration of 0.15g horizontal, the spectral acceleration corresponding to the maximum point on the 0.5-percent critical damping response curve was used to calculate an equivalent static force imparted to the pipe at its support points. This resulted in a seismic design load approximately equal to 0.6W horizontally and O.4W vertically taken simultaneously, where W is the weight of the pipe including static forces. The sum of the resulting additional stress plus the normal stresses was limited to 1.2 times the 831.1 code allowable. The stresses in the pipe supports and hangers were likewise limited to 1.2 times the 831.1 code allowable.
Since all the buildings containing seismic Class I piping are essentially rigid structures, no amplification is expected.
For seismic Class I equipment and tanks the same method was used to arrive at an equivalent static force. In each case, the total of seismic and normal stresses was limited to the applicable code allowable. The refueling water storage tank and condensate storage tank were designed in accordance with the stress limitations of American Water Works Association Standard D100.AII components of the reactor coolant system and associated systems are designed to the standards of the applicable ASME code or USAS code. The loading combinations, which are employed in the design of seismic Class I components of these systems, i.e., vessels, piping, supports, vessel internals and other applicable components, are given in Table 1.11-2.
Table 1.11-2 also indicates the stress limits, which are used in the design of the listed equipment for the various loading combinations. The original design criteria given above and in Table 1.11-2 have been modified in certain instances in accordance with NRC guidance given in References 3 and 4. Generic Letter 87-11 allows for the elimination of pipe whip restraints and jet impingement shields, which were installed to mitigate the effects of arbitrary intermediate pipe ruptures, provided certain criteria are met.
Chapter 1, Page 41 of 72 Revision 20, 2006 OAG10000215_0078
IP2 FSAR UPDATE These design criteria have also been modified in certain instances by the application of "leak before break" technology, as discussed in Section 4.1.2.4.
To be able to perform their function, i.e., allow core shutdown and cooling the reactor vessel, internals must satisfy deformation limits, which are more restrictive than the stress limits shown in Table 1.11-2. For this reason the reactor vessel internals are treated separately.
1.11.3.1 Piping, Vessels, and Supports The reasoning for selection of the load combinations and stress limits given in Table 1.11-2 is as follows. For the design earthquake, the nuclear steam supply system is designed to be capable of continued safe operation, i.e., for the combination of normal loads and design earthquake loading. Critical equipment and supports needed for this purpose are required to operate within normal design limits as shown in line 2 of Table 1.11-2.
In the case of the maximum potential earthquake, it is only necessary to ensure that critical components do not lose their capability to perform their safety function, i.e., shut the plant down and maintain it in a safe condition. This capability is ensured by maintaining the stress limits as shown in line 3 of Table 1.11-2. No rupture of a seismic Class I pipe can be caused by the occurrence of the maximum potential earthquake. With respect to the seismic design of the piping supports, relative displacement between anchor points has been considered in the seismic analysis of the main steam lines, where largest relative displacements are expected.
Analysis indicates that the stress at the highest seismically stressed point is affected by less than 10-percent when relative anchor displacements are considered. The seismic supports installed have been verified to agree with the design location and therefore the locations used in the analyses.
Careful design and thorough quality control during manufacture and construction and periodic inspection during plant life, ensures that the independent occurrence of a reactor coolant pipe rupture is extremely remote. If it is assumed that a reactor coolant pipe ruptures, the stresses in the unbroken leg will be as noted in line 4 of Table 1.11-2.
1.11.3.2 Reactor Vessel Internals 1.11.3.2.1 Design Criteria for Normal Operation The internals and core are designed for normal operating conditions and subjected to loads of mechanical, hydraulic, and thermal origin. The response of the structure under the design earthquake is included in this category.
The stress criteria established in Section III of the ASME Boiler and Pressure Vessel Code, Article 4, have been adopted as a guide for the design of the internals and core with exception of those fabrication techniques and materials, which are not covered by the code, such as the fuel rod cladding. Seismic stresses are combined in the most conservative way and are considered primary stresses.
The members are designed under the basic principles of: (1) maintaining distortions within acceptable limits, (2) keeping the stress levels within acceptable limits, and (3) prevention of fatigue failures.
Chapter 1, Page 42 of 72 Revision 20, 2006 OAG10000215_0079
IP2 FSAR UPDATE 1.11.3.2.2 Design Criteria for Abnormal Operation The abnormal design condition assumes blowdown effects due to a reactor coolant pipe double-ended break.
For this condition the criteria for acceptability are that the reactor be capable of safe shutdown and that the engineered safety features are able to operate as designed. Consequently, the limitations established on the internals for these types of loads are concerned principally with the maximum allowable deflections. The deflection limits for critical internals structures under abnormal operation are presented in Table 14.3-14.
1.11.3.3 Reactor Vessel The criteria for movement of the reactor vessel, under the worst combination of loads, i.e.,
normal plus the maximum potential earthquake or normal plus reactor coolant pipe rupture loads, ensure that the radial movement of the reactor vessel will not exceed the clearance between the reactor coolant piping and the surrounding concrete.
The relative motions between reactor coolant system components are controlled by the structures, which are used to support the reactor vessel, the steam generators, the pressurizer and the reactor coolant pumps.
The maximum movement of the reactor vessel under the worst combination of loads, i.e.,
normal plus maximum potential earthquake or normal plus reactor coolant pipe rupture loads comprises an end deflection on the safety injection piping, which is small even in comparison with that resulting from thermal growth during plant heatup, and is well within the flexibility of the design of the piping system.
The supports are designed to limit the stresses in the pipe to the stress limits given in Table 1.11-2.
1.11.4 Models And Methods For Seismic Class I Design The variety of design problems associated with the seismic analysis of all Class I structures, systems and equipment were approached by various methods. For the design of the reactor, recirculating pumps, and Class I piping an amplification factor of 4.0 was used with respect to ground motion of 0.15g. This amplification factor was based on the maximum for a one-half percent damping of the ground response spectrum. The fundamental frequency of the reactor building internal structure is approximately 17 cycles/sec. As can be seen from Figure 1.11-2 for this frequency level, no significant building amplification of the ground response is encountered.
With the exception of the containment, primary auxiliary building, and electrical cable tunnel, no dynamic analyses were performed on Indian Point Unit 2 structures, hence no mathematical models were developed. The following methods were used in the seismic design of Class I structures.
1.11.4.1 Containment Building See Sections 2.0, 3.0, and 4.0 of the Containment Design Report for Indian Point Unit 2 containment building structures and components.
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IP2 FSAR UPDATE 1.11.4.1.1 Steel In the design of the steel, 100-percent of the dead load and SO-percent of the live load were considered. The peak of the response curve for 0.1Sg ground acceleration and 1.0-percent critical damping was used to obtain the seismic forces, which were distributed by the method described in the Containment Design Report and resisted by the bracing. The 1.0-percent critical damping is conservative since the structure is shop welded and field bolted to the columns. The actual critical damping value would be between 1.0-percent (welded) and 2.S-percent (bolted). A one-third increase over working stress was allowed in the design of the bracing.
1.11.4.1.2 Concrete In the design of the concrete, 100-percent of the dead load and SO-percent of the live load were considered. The Modified Rayleigh Method was used to calculate the natural period and the base shear was distributed by the same method described in the Containment Design Report.
The forces determined from the response curve for a 0.1Sg ground acceleration with S-percent critical damping were applied at the node points where the masses were lumped for the Rayleigh approach. These loads were resisted by the vertical walls, which acted as shear walls, and horizontal reinforcing, which resisted the moment. The Ultimate Strength Design method of ACI 318-63 was used for the design and construction of the containment building.
1.11.4.2 Control Building The dead load and equipment loads were considered. The period was determined from the formula T = 0.1 n, where n = number of stories (Design of Multistory Reinforced Concrete Building for Earthquake Motions by N. M. Newmark, et. al.). The response curve for 0.1Sg ground acceleration with 2.S-percent critical damping was used to determine the base shear.
This base shear was distributed at the floor levels by the same method described in the Containment Design Report and resisted by a rigid frame structure with a one-third increase on allowable working stresses. The design was controlled by a deflection limitation due to the adjacent Unit 1 control building.
1.11.4.3 Diesel Generator Building Due to the light weight of the structure, the wind load controlled the design.
1.11.4.4 Fan House One hundred percent of the dead load and SO-percent of the live load were considered. The peak of the response curve for 0.1Sg ground acceleration with S-percent critical damping was used for the concrete structure and the corresponding 2.S-percent was used for the steel superstructure. A one-third increase in allowable working stresses was allowed.
1.11.4.S Boric Acid Evaporator Building One hundred percent of the dead load was considered. For method of design, see fan house.
Without allowing a one-third stress increase for seismic design, the controlling factor for reinforcing design was the minimum temperature steel requirements of the ACI-318 Building Code.
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IP2 FSAR UPDATE 1.11.4.6 Intake Structure One hundred percent of the live and dead load were considered. The peak of the response curve for 0.1g (OBE) ground acceleration with S-percent critical damping was used to obtain the seismic loads. The effect of water sloshing was considered in the earthquake analysis (per TI 0-7024 "Nuclear Reactors and Earthquakes," Section 6.S). Although DBE was not explicitly considered in the calculation (the seismic forces used in the design shows that DBE is not governing), the controlling factor in the design of the intake structure was the service load with the worst combination being one chamber empty and the adjacent chamber filled with water.
1.11.4.7 Waste Holdup Tank Pit One hundred percent of the dead load and SO-percent of the live load were considered (including the tank dead weight on the roof). The peak of the response curve for 0.1Sg ground acceleration with S-percent damping was used to determine the base shear. Using working stress limits for the seismic design, service loads controlled the design of the top slab. The bottom slab and wall of the pit were designed for earthquake loads with stresses limited to yield multiplied by the factors recommended in Section IV-B of the ACI-318-63 "Building Code."
Consideration was given to the tanks in the pit when designing the base slab.
1.11.4.8 Spent Fuel Pit The seismic loads, as determined in TID-7024 "Nuclear Reactors and Earthquakes," Section 6.S, were resisted by the reinforced concrete walls and base slab. Working stresses were used except for the moment at the base of the walls where ultimate strength design was considered with stresses limited to 0fy. The effects of water in the pool are accounted for in this design approach. Ground acceleration of 0.1Sg was used. In 1990, new high density spent fuel storage racks were installed. Prior to their installation, the spent fuel pit was reanalyzed (Ref.
6). The new racks were also analyzed (Ref. Sand 6).
1.11.4.9 Electrical Penetration Tunnel The peak of the response curve for 0.1Sg ground acceleration with S-percent critical damping was considered using working stress design limits. The load was considered to act at 2/3 L, where L = the height of the tunnel. Temperature of steel considerations controlled the design of the concrete while service loads controlled the structural steel.
1.11.4.10 Pipe Penetration Tunnel One hundred percent of the dead load, plus SO-percent of the live load were considered. The peak of the response curve for 0.1Sg ground acceleration with S-percent damping was used to find the shear, which was considered as a concentrated load applied at the top slab of the tunnel. A one-third increase on working stress allowables was used in the design.
1.11.4.11 Electrical Cable Tunnel One hundred percent of the dead load, SO-percent of the surcharge, and SO-percent of live load in the tunnel were considered. The Modified Rayleigh Method was used to determine the natural period and the loads were distributed as described in the Containment Design Report.
The response curve for 0.1Sg ground acceleration with S-percent critical damping was used. A Chapter 1, Page 45 of 72 Revision 20, 2006 OAG1000021S_0082
IP2 FSAR UPDATE one-third stress increase was permitted on working stress allowables when considering the effect of seismic loads.
1.11.4.12 Shield Wall The peak of the response curve for 0.15g ground acceleration with 5-percent critical damping was used. The pipe break loads controlled the design.
1.11.4.13 Retaining Wall At Equipment Entrance The wall was designed for soil pressure including a 1000 psf surcharge applied during reactor loading. The load combination that includes a seismic factor governs the design. It has been shown that the wall design was adequate.
1.11.4.14 Primary Water Storage Tank and Refueling Water Storage Tank Foundation The seismic loads on the circular wall and center pier were those supplied by the tank manufacturer. The shear force from the earthquake on the water in the tank was applied at 3/4 L above the top slab. The shear force from the earthquake on the tank was applied at Ll2 above the top slab, where L =the height of the tank. The horizontal shear force from the earthquake effect on the dead weight of the foundation was determined by using the peak of the response curve for 0.15g ground acceleration with 5-percent critical damping. A triangular distribution was used. The earthquake effect of the backfill was also considered. The load was applied to the walls as the resultant of a triangular pressure distribution. The stresses were limited to working stress design limits. The temperature steel considerations controlled the design of the walls and center pier.
1.11.4.15 Condensate Water Storage Tank Foundation The seismic loads on the spread footing foundation were those supplied by the tank manufacturer. The shear forces from the earthquake on the water in the tank were applied at 3/4 L above the footing, where L = the height of the tank. The shear force from the earthquake on the tank was applied at Ll2 above the top of the footing. The stresses were limited to working stress design limits.
A multi degree-of-freedom modal analysis was performed on all Class I structures for Indian Point Unit 3. The results indicated that all structures except the containment structure were rigid. The only significant differences between the structural design of Units 2 and 3 seismic Class I buildings are the control building and the steel structural portion of the primary auxiliary building for Indian Point Unit 2, which are flexible steel structures. On Unit 3 they are rigid concrete structures. All seismic Class I structures on Indian Point Unit 2 except control building and containment shell are rigid and move with zero period ground acceleration. However, the design of all seismic Class I structures on Unit 2 were standardized and based on the peak acceleration of the ground response spectrum, which is extremely conservative for rigid structures.
In the preceding designs, limits have been placed on stresses to ensure that all structures will respond elastically to the earthquake. If for some reason inelastic response were to occur, the period of the structure would be expected to increase. Since the majority of the structures were designed for the peak of the response curve, the effect of any change in period would be to decrease the coefficient of spectral acceleration and thus lower all shears and moments.
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IP2 FSAR UPDATE Mathematical models were not used for seismic design of instrumentation. Ability to withstand the seismic condition is determined by actual vibration type testing of typical instrumentation equipment under simulated seismic accelerations to demonstrate its ability to perform its functions. The seismic testing is reported in Westinghouse report WCAP-7397-L (Reference 1).
The locations of protection and safeguards control and electrical equipment in Indian Point Unit 2 have been identified. The most adverse location, seismically, is the control building floor at elevation 53-ft, which supports the nuclear instrumentation, radiation monitoring, process instrumentation, and safeguards logic racks. Dynamic analyses of this building for the plant design basis earthquake of 0.15g show that the significant horizontal and vertical accelerations of this floor are within the specified low seismic test envelopes given in WCAP-7397-L (Reference 1).
Seismic analysis of Class I equipment including heat exchangers, pumps, tanks, valves, motors, and electrical equipment components was performed using one of the following four methods:
- 1. Equipment, which is rigid and rigidly attached to its support structure, was analyzed for a "g" loading equal to the peak acceleration of the supporting structure at the appropriate elevation.
- 2. Equipment, which is not rigid and therefore a potential for response to the support motion exists, was analyzed for the peak of the floor response curve for appropriate damping values.
- 3. In some instances nonrigid equipment was analyzed using a multi degree-of-freedom modal analysis. All contributing modes were considered. In addition, a sufficient number of masses was included in the mathematical models to ensure that coupling effects of members within the component were properly considered.
The results of these analyses indicated that the models contained more masses than necessary, and that future analyses of comparable equipment could be considerably simplified by considering fewer masses. The method of dynamic analysis used a proprietary computer code called WESTDYN. This code used as input the inertia values, member sectional properties, elastic characteristics, support and restraint data characteristics, and the appropriate seismic response spectrum. Both horizontal and vertical components of the seismic response spectrum were applied simultaneously. The modal participation factors were combined with the mode shapes and the appropriate seismic response spectra to obtain the structural response for each mode. The internal forces and moments were computed for each mode from which the modal stresses are determined.
The stresses were then summed using the root mean square method.
- 4. Type testing of selected electrical equipment has been conducted to demonstrate seismic design adequacy as described in WCAP-7397-L (Reference 1).
For the analysis of equipment to resist the vertical seismic component, two-thirds of the horizontal response spectrum curves were used to determine the acceleration appropriate to the vertical frequency.
Engineered safeguards tanks, e.g., boric acid, accumulator and surge tanks, were analyzed using method 3 above for combined horizontal and vertical seismic excitation occurring Chapter 1, Page 47 of 72 Revision 20, 2006 OAG10000215_0084
IP2 FSAR UPDATE simultaneously and in conjunction with normal loads without exceeding allowable stresses.
Hydrodynamic analyses of these tanks have been performed using the methods described in Chapter 6 of the "U.S. Atomic Energy Commission - TID 7024." The stresses for these components due to the above-mentioned load combinations were found to be within allowable limits. Heat exchangers associated with the engineered safeguard systems, e.g., component cooling and residual heat removal, were analyzed using method 3 above, and the results show that stresses and deflections are within allowable limits.
Selected critical engineered safeguards valves were analyzed using method 3 above and the results indicated that their fundamental natural frequency was sufficiently separated from the building frequency. The results further indicated that the total stress, considering all modes, was far below the allowable stress limit.
Appendages, such as motors attached to motor-operated valves, were included in the mathematical models.
1.11.4.16 Class I Piping Systems Class I piping systems were designed and analyzed as described in the succeeding paragraphs. However, in an attempt to correlate the simplified method of analysis suggested by the AEC for the H. B. Robinson Nuclear Generating Station, the following discussion is presented:
If no dynamic analysis is performed on Class I piping systems, these systems for H. B.
Robinson plant were to be checked to determine whether the results conform to the following formula:
1.3* K Ss + Sn :::; 1.8 Sa
[Note - *The 1.3 factor was recommended by the AEC to represent the contributions of higher modes above the fundamental mode. Detailed dynamic analyses performed on Indian Point Unit 2, and described later, indicate that where significant stresses exist in piping systems, a more realistic modal contribution factor would be 1. 1. However, for the present discussion we will adhere to the 1.3 factor for additional conservatism.]
where:
Ss - represents seismic stress including effects of valve motors, from design calculations Sn - represents normal primary and bending stresses for loadings other than seismic, from design calculations 1.8 Sa - equals 1.8 times the allowable stress or yield stress, whichever is higher for code listed materials.
K- ratio of peak acceleration of floor response spectra to acceleration used in the piping design The piping design criteria limited the deadweight and seismic stresses to 0.2 Sa. The longitudinal pressure stress is 0.5 Sa.
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IP2 FSAR UPDATE 1.3 K (0.2 Sa) + 0.5 Sa :s; 1.8 Sa Solving, the K-factor becomes:
K=5 This factor combined with the 1.3 modal contribution factor gives a combined factor of 6.5, which is more than double the original suggested multiplier of 3.
Indian Point Unit 2 conservatively meets the criteria suggested for application on the H. B.
Robinson Plant for seismic Class I piping.
However, a different and more detailed method of analysis was actually undertaken to illustrate the conservatism of design approach used for Indian Point Unit 2. This approach is described in detail below:
It is obviously necessary to use simplifying assumptions when performing initial design of piping systems, including restraints, rather than a dynamic analysis involving a trial and error procedure. Simplified design procedures are not uncommon and often suggested in codes, i.e.,
USAS B31.1 - Power Piping Code.
A complete flexibility analysis involving detailed modeling of Class I piping systems is unnecessary if the conservatism of the simplifying assumptions used in the initial design can be demonstrated. A "third party" review was conducted to establish the adequacy and conservatism of the original design criteria for Class I piping systems as performed by the architect/engineer (United Engineers and Constructors, Inc.) and the seismic restraint supplier (Bergen-Paterson Pipe Support Corp.). The review involved the following steps:
- 1. Representatives from Westinghouse and United Engineers and Constructors, Inc., visited the Indian Point Unit 2 site and inspected the Class I piping systems.
- 2. Based upon their best engineering judgment, representative worst-case lines were selected for detailed dynamic analyses.
- 3. In exercising their engineering judgment, these representatives looked for the following characteristics, which would indicate possible sources of problems.
- a. Amplification due to the location and elevation in building.
- b. Large concentrated masses such as overhung motor-operated valves, particularly in what appear to be flexible sections of the pipe.
- c. Complexity of configuration of the piping system itself such that application of the original design criteria would be difficult.
- d. Manual excitation of the pipe by pushing or kicking indicated excessive flexibility either in the pipe excited or the piping attached to it.
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- 4. The results of the dynamic analyses were compared with original design values to determine whether the design approach was conservative. Besides analysis of the reactor coolant loop, portions of the following systems were analyzed:
- a. Safety injection.
- b. Main steam.
- d. Service water.
- e. Accumulator discharge.
- g. Component cooling.
1.11.4.16.1 Design Approach The design and placement of seismic restraints were predicated on the principle of containing the seismic stresses without restricting the free thermal expansion of the piping system. The systems were designed to have sufficient flexibility to prevent the movements from causing failure of piping or anchors from overstress.
Two fundamental principles underlie the design approach, namely:
- 1. The system be designed such that its fundamental natural frequency does not coincide with the exciting frequency.
- 2. The maximum seismic stresses in piping be less than the USAS B31.1 code allowable value. The seismic stresses were limited to 0.2 S allowable (3000 psi).
This is extremely conservative since the longitudinal pressure stress accounts for approximately 0.5 S allowable leaving a margin of safety of 0.5 S allowable, which is unused. (Note-this is based on a maximum allowable of 1.2 Sa)
These fundamental principles should ensure that stresses will be within code allowable stress limits, and that the piping will not go into resonance with the exciting frequency. Tables of recommended maximum spacing of supports, for straight runs of pipe, were developed. The recommended spacing of supports was modified near bends and concentrated masses (i.e.
valves) to account for additional weight and flexibility.
1.11.4.16.2 Analysis Approach In order to determine whether the design procedure resulted in an acceptable system, selected worst case Class I piping systems were modeled and a dynamic flexibility analysis performed.
A detailed description of the method of analysis is given below.
The analysis was performed using a proprietary computer code called WESTDYN. The code uses as input system geometry, inertia values, member sectional properties, elastic characteristics, support and restraint data characteristics, and the appropriate Indian Point seismic floor response spectrum for O.S-percent critical damping. Both horizontal and vertical components of the seismic response spectrum are applied simultaneously.
With this input data, the overall stiffness matrix of the three-dimensional piping system is generated (including translational and rotational stiffness's). The modal participation factors are Chapter 1, Page 50 of 72 Revision 20, 2006 OAGI000021 5_0087
IP2 FSAR UPDATE computed and combined with the mode shapes and the appropriate seismic response spectra to give the structural response for each mode.
Each piping run is modeled as a three-dimensional system, which consists of straight segments, curved segments, and restraints. Straight segments are distinguished from curved segments during data output.
The computer code requires that the piping be represented by a discrete mass model. Each mass includes the contribution of both the steel encasement and conveyed fluid. Where valves or other concentrated masses exist in the piping system, they were included in the model.
Restraints were included in the model at their proper location. The directionality of the restraints was also considered. The detailed dynamic analyses of selected worst case Class I piping indicated that the method used to design the seismic restraints was conservative. Based on this critical review of the selected worst case systems and the consistent application of the same design procedure to all completely engineered seismic Class I systems, the seismic design of other Class I systems, not analyzed, was deemed adequate.
The maximum stresses imposed by the normal loads plus loads associated with the design-basis earthquake (DBE) are below 1.2S, where S is the allowable stress limit obtained from the Power Piping Code - USAS B31.1.0 - 1955.
Some of the items of conservatism employed in the seismic design of Class I piping systems for Indian Point Unit 2 were:
- 1. The maximum longitudinal stress due to seismic excitation was limited to 0.2S rather than the usual 0.7S.
- 2. The maximum allowable stress was limited to 1.2S. If the combination of normal and DBE loads were considered as a faulted condition, the allowable membrane and bending stresses could be chosen as those corresponding to 20-percent to 40-percent of the material uniform strain at temperature, respectively. This would give more than a factor of 2 margin between the allowable and the maximum actual stresses.
- 3. A low value of the fraction of critical damping was adopted (0.5-percent). Dr. N.
M. Newmark recommended a value of 2-percent for vital piping at or just below the yield point. This would reduce the maximum amplification of the ground acceleration.
- 4. The maximum longitudinal stresses due to pressure, deadweight, and seismic loads were presumed to occur at the same cross-section and some point in the cross-section.
Some averaging of the response spectra was performed to smooth out the erratic response of the earthquake's random behavior. At the high frequency end of the spectra, the acceleration levels of the smoothed spectra converge to the values of the unsmoothed spectra.
It is therefore concluded that the design procedure used to design seismic Class I restraints for Indian Point Unit 2 is conservative.
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IP2 FSAR UPDATE NRC IE Bulletin (lEB) No. 79-07 was concerned with inadequacies identified in the seismic analysis of certain piping systems at several power reactors. The inadequate treatment of piping loads from earthquakes was attributed to the fact that some piping analysis codes used an algebraic summation of the loads predicted separately by computer code for both the horizontal components and the vertical component of seismic events. In accordance with the IEB, such co-directional loads should not be algebraically added unless certain more complex time-history analyses are performed. The IEB emphasized that to properly account for the effects of earthquakes on systems important to safety, such loads should be combined absolutely or by using techniques such as the sum of the squares.
In response to IE Bulletin No. 79-07, eight (8) Indian Point Unit No.2 lines were reanalyzed using the UE&C-ADLPIPE-2 dynamic seismic computer code. This code utilizes the worst-case two-dimensional evaluation technique and uses the square root of the sum of the squares option for combining both intramodal and intermodal responses.
The difference between the newly calculated total pipe stress and the originally calculated total pipe stress is not significant. Even after applying a 1.3 "adjustment" factor to the calculated seismic stress component, the total pipe stress remains below the allowable stress limit.
Furthermore, the loads on the pipe supports and equipment nozzles were re-evaluated on the basis of the confirmatory reanalysis and found to be acceptable, as documented in Reference 9.
1.11.4.17 Reactor Coolant System Analysis for Combination Loading of Design-Basis Earthquake and Design-Basis Accident [Historical Information Only]
The Indian Point Unit 2 reactor coolant system was not committed to be designed for the combination of the seismic and blowdown loads. However, an analysis for this combination of loadings was performed for the original configuration of the Indian Point Unit 3 reactor coolant system, which was identical to the original Unit 2 configuration.
The analysis was performed as outlined below:
- 1. A lumped mass dynamic mathematical model of the primary coolant loop and support system was developed.
- 2. This dynamic model was subjected to multiple simultaneous time history hydraulic forcing functions for the blowdown analysis. The double-ended ruptures were located at places of large change in flexibility. Time history response of the total structure to these conditions was computed and reduced to time history stresses.
- 3. The dynamic model was then subjected to a floor response spectra earthquake analysis.
- 4. The loads as determined above were used for an evaluation of the stresses along the piping system.
- 5. The stresses as determined from the basis described above were lower than the allowable stresses calculated by using the approach described in WCAP-5890, Revision 1 (Reference 2) and the following parameters:
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IP2 FSAR UPDATE
- a. 20-percent of the uniform strain on the allowable membrane and average strain.
- b. 23,100 psi as the at-temperature yield in the axial direction. This value was based on the minimum value of the at-temperature yield in the loop direction as measured with samples from the Unit 2 piping, and increased by 10-percent for the increase in strength in going from the loop to the axial direction. The tensile tests on the Unit 2 piping material at-temperature yielded at a minimum value of 20,900 psi, a maximum of 29,700 psi, and an average of eleven samples of 23,300 psi.
Based on the above analysis, it was conduded that the Unit 2 reactor coolant system can stand the combination of blowdown and seismic loads within acceptable stress limits.
In 1989, the NRC approved changes to the design bases with respect to dynamic effects of postulated primary loop pipe ruptures, as discussed in Section 4.1.2.4. In 2000, an analysis of the Unit 2 reactor coolant loop and its component supports, which incorporates the NRC approved changes, was performed with the replacement steam generators and sixteen of the original twenty-four steam generator support frame hydraulic snubbers removed (Reference 11).
In line with the older analysis for Unit 3 described above, the Unit 2 analysis of 2000 induded the effects of the now controlling feedwater line break at the steam generator nozzle in a similar fashion as described for the Unit 3 analysis. Based on this revised analysis, it was concluded that the Unit 2 reactor coolant system can withstand the combination of blowdown and seismic loads within acceptable stress limits.
This 2000 analysis has since been updated to include a power uprate to a core power level of 3216 MWt. Combination of blowdown and seismic loads were not considered in this latest evaluation.
1.11.4.18 Service Water Lines The service water lines consist of two 24-in. diameter carbon steel pipes. They run in a common trench, which is backfilled. Assuming that the ends of a pipe are free to displace vertically but not rotate and that the maximum permissible stress is restricted to 30,000 psi, a parametric study showed that the following maximum allowable relative displacements may occur during a seismic disturbance without overstressing the pipe:
Length, ft 1 10 25 50 75 100 Displacement, in. 0.002 0.20 1.25 5.01 11.27 20.04 It is therefore concluded that the service water lines could withstand, without being overstressed, relative bedrock displacements associated with the earthquakes defined for the Indian Point site.
1.11.4.19 Seismic Evaluation of the Fan Cooler and Passive Hydrogen Recombiner Systems The seismic analysis of the fan cooler system was completed in two parts.
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IP2 FSAR UPDATE
- 1. Analysis of the structural steel enclosure of the fan cooler units to include the effect of supported equipment.
The structural analysis considering simultaneous incident pressures and earthquake forces was conducted on particular members, plates, and connections, which are, considered critical to the structural performance of the reactor containment fan coolers. This analysis included consideration of the mass of all components supported partially or wholly by the enclosure. The fan, fan motor, and fan motor heat exchanger, although entirely within the enclosure, are independently supported from the concrete floor that makes up the base of each unit.
Earthquake loadings were treated using the response spectra techniques. A horizontal force of 0.6W and a vertical force of OAW, where W is the weight of the member including static forces, were concentrated at the center of gravity of each member. A negative differential pressure of 1.5 psig was applied to the portion of the unit from the inlet up through the fan compartment; a negative differential pressure of 6.3 psig was applied to the charcoal filter compartment.
Both of these values are consistent with unit geometry and containment environment following a loss of coolant accident or main steam line break. The charcoal filter compartment pressure is limited by a pressure equalization device installed during the 1997/1998 Maintenance Outage. All loadings were assumed to act simultaneously and comparison was made with allowable stresses consistent with specifications for installed materials. An increase in allowables was considered for loads associated with accident conditions. Where applicable, allowable concrete stresses were taken from ACI Specifications.
Results of the analysis on the enclosure demonstrated that the design is adequate.
- 2. Evaluation of the fan motor system and its foundation.
The fan motor and its supporting structural system was evaluated using acceleration values for a maximum hypothetical earthquake. These values are 0.6g for the horizontal direction and O.4g for the vertical direction. These accelerations were assumed to act simultaneously.
The failure modes considered for the motor unit were excess deflection of the rotor shaft, which results in rubbing against the housing or by bearing failure.
The failure modes for the fan are failure of the fan shaft support bearings or deflection of the fan housing and fan wheel causing binding. In addition, the potential for shear and overturning failure of the motor fan assembly at the foundation anchorage was evaluated.
Based on analyses made on similar fan motor systems, it was concluded the fan cooler units in the containment are adequately designed to resist the seismic loading defined for the site and supporting building structure.
The two hydrogen recombiners are located in the containment at the 95-ft elevation. The hydrogen recombiners are as shown on Figure 6.8-1. The Passive Hydrogen Recombiners Chapter 1, Page 54 of 72 Revision 20, 2006 OAG10000215_0091
IP2 FSAR UPDATE (PHRs) are seismic class I 344-1987.
1.11.4.20 Masonry Walls In response to IE Bulletin 80-11, safety related masonry walls were evaluated to demonstrate the ability to withstand the specified design load conditions without impairment of wall integrity or the performance of required safety functions. NRC acceptance of this evaluation is documented in reference 10. As a result of this evaluation, certain walls in the control building, the Unit No. 1 Superheater building, the boric acid evaporator building, the fan house, and the fuel storage building have been reinforced.
1.11.S Wind Effects The IP2 licensing basis does not include tornado protection for the design of the buildings, structures and components. Tornado protection is not a design criterion for IP2. However, the following structures were evaluated for tornado loads: containment building, primary auxiliary building, control building, fuel storage building (including the spent fuel pit), and the intake structure.
Detailed information on the containment structure is found in Appendix B of the Containment Design Report. The containment structure will not be penetrated by a 4-in. x 12-in. x 12-ft wood plank traveling at 300 mph, or by a 4000-pound auto traveling at SO mph less than 2S-ft above the ground.
With respect to the primary auxiliary building, control building, and fuel storage building, information from the siding manufacturer indicates that siding panels will blowout at 170 psf, which is equivalent to a 1.18 psi negative pressure. Panels fail at 60 psf external pressure, which is equivalent to a 162 mph external wind load (60 psf controls the external loading condition). The grits will fail at 90 psf, which is equivalent to a 0.62 psi negative pressure. The 3.2S-in. thick siding panels are not capable of resisting any tornado-generated missiles.
Spent fuel pit tornado protection is discussed in proprietary WCAP-7313-L. The intake structure is capable of resisting any wind or missile loads generated by a tornado. This is true for the structure itself, but does not necessarily include associated equipment.
1.11.6 Structural Effects The potential for damage to Class I structures due to failure of nearby Class II or Class III structures, or due to failure of Class III cranes, has been considered.
The only Class I structures and components that could be endangered by failure of Class III structures are the control building, main steam piping, and feedwater piping, which could be endangered by failure of the Class III turbine building. No special provisions were provided in the original plant design except in the case of the main steam and feedwater lines up to the isolation valves, which are protected by the shield wall and the structural frame at the north end of the shield wall. Evaluations were performed and bracing was added to the turbine building during construction, as described in section 1.11.6.S, to preclude such catastrophic failures.
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IP2 FSAR UPDATE The only Class III crane whose failure could endanger any Class I function is the fuel storage building crane. Failure of this crane will not impair a safe and orderly shutdown. The wheels of the bridge and the trolley are shaped such that sliding perpendicular to the rail would not be possible. The lateral load from an earthquake on the trolley crane rail is about 50-percent greater than the lateral loads from impact specified by the AISC Code for design within working stress limits. The stresses on the crane rail are low due to the earthquake load. For this reason no failure of the crane rail is anticipated.
The Class III manipulator crane in the containment building is restrained from overturning and will not endanger Class I structures.
The turbine building and the fuel handling building are functionally Class III structures.
However, these structures have been analyzed using a multidegree of freedom modal dynamic analysis method to ensure that there is no potential for gross structural collapse of these structures as a result of the maximum hypothetical earthquake. The results of the analyses are given below. A value of 7-percent structural damping was assumed in the analysis. Total response of the structure was determined on the basis of the square root sum of the squares basis of each mode contribution. A similar dynamic analysis was also performed to ensure that no potential gross failure of the Indian Point Unit 1 stack or superheater building could occur for the maximum hypothetical earthquake, or for the design-basis tornado for Indian Point Unit 2.
The resultant dead, live, and seismic design stresses in the basic building structure is limited to 0.9 yield of the steel.
The results of specific analyses are discussed in the following sections.
1.11.6.1 Seismic Analysis of the Indian Point Unit 2 Turbine Building A spectrum response analysis was performed for the turbine building considering the design-basis earthquake (DBE), which has a peak horizontal ground acceleration of 0.15g. The associated earthquake response spectrum is shown in Figure 1.11-2.
The foundation was considered rigid since the footings for the structural frames of the building are underlaid by either rock or a lean concrete, which bears on rock. Also, in the analysis, interaction between the turbine and the structural frame for the building was neglected. The analysis, as performed, represents a linear elastic system.
The analysis of the turbine building was performed under the assumption that the north-south motions, east-west motions and vertical motions will be uncoupled. The dynamic analysis effort was limited only to horizontal motions in the east-west and north-south directions. However, vertical components of the earthquake were considered by adding a 0.13g component to dead loads. Each of the models was simulated for the computer program called STARDYNE. A description of the modeling capabilities of STARDYNE are contained in "STARDYNE Structural Analyses Systems Users' Manual" prepared by Mechanics Research, Inc., for Control Data Corporation.
The STARDYNE program was used in three ways. First, the portal frames were analyzed for a static unit force at each portal to determine their resistance to horizontal motions resulting from the turbine bay crane. This information was incorporated into the model for the analysis of the crane girder to determine the distribution of horizontal turbine bay crane loads to the various east-west portal frames. Secondly, the program was used to determine the forces induced in the frames as a result of gravity forces, and, thirdly, the STARDYNE program was used to Chapter 1, Page 56 of 72 Revision 20, 2006 OAG10000215_0093
IP2 FSAR UPDATE determine the fundamental frequencies of each of the models and the characteristic shapes. In addition, the STARDYNE program is also capable of determining the modal member forces for each of the fundamental frequencies. This information for each model and mode was stored on tape along with the gravity forces for each model and later used in an earthquake analysis program to determine the maximum probable deflection, acceleration, member forces, member stresses, and the combined gravity plus earthquake member stress responses. Dynamic characteristics of the turbine building are shown in Table 1.11-4.
Results of the analysis indicated that the 0.9 Fy combined load allowable stress was not violated except locally in the flange of columns where cross bracing framed in eccentric to other joint members. Reduction of stresses to allowable values is accomplished by the addition of flange cover plates.
While allowable stresses in the cross bracing did not exceed the 0.9 yield stress allowable, it was determined that most of the "x" cross bracing would buckle at very low compressive stress due to high £ Ir ratios. In order to assure the lateral stiffness of the bents and load carrying capacity as determined in the analysis, cover plates were attached to the bracing equal to the original area of the "x" crossing bracing. This assures design adequacy with only "x" cross bracing in tension assumed to be active in carrying lateral load.
1.11.6.2 Seismic Evaluation of the Fuel Storage Building Structure Above the Spent Fuel Pit The fuel storage building for Indian Point Unit 2 consists of the spent fuel pit constructed of reinforced concrete and founded on rock. The fundamental frequency of the pit is approximately 22 cps and therefore can be considered rigid. The steel superstructure above the pit encloses the pit and supports the fuel cask handling crane. This superstructure was designed as a Class III structure. The seismic loads used in the analysis of the steel superstructure were as follows:
- 1. Zero period ground acceleration: 0.15g horizontal, 0.1 Og vertical.
- 2. 7-percent damping.
- 3. Response spectrum curve as defined in Figure 1.11-2.
- 4. Inertial forces for each mass point are determined on the basis of the square root of the sum of the squares.
A dynamic multidegree of freedom, modal analysis of the structure was constructed as shown in Historical Figures 1.11-3 and 1.11-4. The stiffness properties of the elements were determined by the combined stiffness of the frame bents in the north-south and east-west directions taken separately. The stiffness of each bent was determined by the computer program STRUDL. The total inertial forces determined by the dynamic analysis were distributed to each individual bent and resultant member stresses were determined. The crane was assumed fully loaded.
Evaluation of these seismic stresses show maximum stresses occurring in diagonal bracing.
The maximum stress thus determined in the cross bracing was 18.5 ksi. The maximum combined dead and seismic column load stress determined by the analysis was 12.8 psi compression.
On the basis of these results it was determined that the fuel storage building superstructure was adequately designed to carry the seismic load defined for the site.
Chapter 1, Page 57 of 72 Revision 20, 2006 OAG10000215_0094
IP2 FSAR UPDATE In addition to the analysis of the building structure, the fuel crane bridge was evaluated to determine the potential for the crane bridge to lift off its track support in the event of a seismic disturbance. The vertical mode fundamental frequency of the fuel storage building is approximately 9 cps.
The crane bridge has also been analyzed dynamically both loaded and unloaded and for various positions of the trolley. It was determined that the crane with the trolley at the end of the span and unloaded would have a fundamental frequency of approximately 9 cps. Considering potential resonance with the fundamental vertical mode of the building at 9 cps the resulting g-loading was 1.05g. The only potential for crane lift-off will be in the unloaded condition with the trolley parked near the support. Since the unloaded crane will not be parked over the pool no potential hazard exists and vertical restraints are not required.
1.11.6.3 Seismic and Wind Analysis of the Superheater Stack of Indian Point Unit 1 The Indian Point Unit 1 superheater stack has been analyzed for seismic, tornado, and vortex-shedding wind load effects. The results of this analysis are summarized below. As a result of this analysis on the existing stack it is concluded:
- 1. The stack can withstand a tornado wind load of approximately 300m ph prior to buckling failure of the stack steel shell.
- 2. The maximum stress in the stack at the critical vortex-shedding frequency wind velocity is 7660 psi, which provided a 3.64 factor of safety against stack failure by this mode.
- 3. The maximum combined dead and seismic stress for the earthquake parameters defined for the site is 19,140 psi, which provides a 1.46 factor of safety against stack failure by this mode.
1.11.6.3.1 Load Case 1 - Tornado I. Load Criteria Wind = 300 mph L= D+W where:
L =Total load D = Dead load W= Tornado load II. Method of Load Analysis As prescribed in ASCE Paper 3269 for uniform wind velocity with height; no gust factor.
III. Allowable Stress Criteria Chapter 1, Page 58 of 72 Revision 20, 2006 OAG10000215_0095
IP2 FSAR UPDATE 0.72Et Cia = 27,900 psi
]t(1 - v 2 )r where:
Cia = allowable stress (psi)
E = modulus of elasticity (psi) t =shell thickness (in.)
v = Poisson's ratio r = radius of stack (in.)
IV. Stress Determination D W'yr Ci = + = 1.54 + 25.75 = 27.29 ksi A I where:
y =centroidal height of stack (in.)
=moment of inertia of stack (in.4)
A =cross sectional area of stack (in. 2
)
0' 27.9 Factor of Safety = _a = = 1.02 0' 27.29 1.11.6.3.2 Load Case 2 - Seismic I. Load Criteria a) Zero period ground acceleration: 0.15 g horizontal; 0.10 g vertical.
b) Damping 7-percent.
c) Ground response curve - Figure 1.11-2.
L = D + E' h' = E'v Chapter 1, Page 59 of 72 Revision 20, 2006 OAG10000215_0096
IP2 FSAR UPDATE where:
E' h = load resulting from horizontal earthquake component E'v =Ioad resulting from vertical earthquake component II. Method of Load Analysis Multidegree of freedom modal analysis of the superheater building and stack as shown in Figure 1.11-5. The square root of the sum of the squares of seismic inertia forces at mass points is used to determine resultant shears and moments in the stack.
III. Allowable Stress Criteria See Load Case 1, item III.
IV. Stress Determination D E'v EhXr cr = +- +- -
A A I a= 1.54 + 0.20 + 17.4 = 19.14 (j 27.9 Factor of Safety = _a (j 19.14
= 1.46 where:
x = lever arm of node inertia force 1.11.6.3.3 Load Case 3 - Vortex-Shedding I. Expression for maximum uniformly distributed force due to vortex-shedding.
P= (MF) 1/2pv 2 x CL x D x L~
8 CL = Lift coefficient for a stationary circular cylinder MF = A multiplying factor applied to the lift coefficient to account for a vibrating cylinder D = Average stack diameter (ft)
Chapter 1, Page 60 of 72 Revision 20, 2006 OAG10000215_0097
IP2 FSAR UPDATE L = Length of stack (ft) o= Logarithmic decrement p =Air density (0.0023385 Ib - sec2fft4) v = F1 x Vc Vc =Critical vortex-shedding velocity (fps)
F1 = A correction factor, which accounts for the fact that stack oscillations have occurred as high as 30-percent above shedding velocity fx D Vc =
5 S =Stronhal number f = Fundamental frequency (cps)
II. Pertinent parameters CL =0.1 MF =4.0 D =20-ft L = 334.5-ft 0= 0.04rc (2-percent critical damping)
Vc =42.7 fps F1 = 1.2 S =0.27 f =0.576 cps III. Stress criteria (j = -D + -Phr = 1.54 + 6.12 = 7.66ksl.
A 21 Factor of Safety = (J"a = 27.9 =3.64 (J" 7.66 Chapter 1, Page 61 of 72 Revision 20, 2006 OAG10000215_0098
IP2 FSAR UPDATE In addition to the analysis performed for the existing stack it was determined that the stack with 80-ft removed from the top would have the capacity to resist a 360 mph wind for the criteria as defined in Load Case I; the seismic as defined in Load Case II; and the vortex-shedding as defined in Load Case III.
1.11.6.4 Seismic and Tornado Evaluation of the Superheater Building at Indian Point Unit 1 A spectrum response analysis was performed for the superheater building considering the design basis earthquake, which has a maximum horizontal ground motion of 0.15g. A dampening coefficient equal to seven percent was assumed for all modes. The earthquake response spectra used is shown in Figure 1.11-2 normalized to 0.15g zero period ground acceleration. In the analysis no interaction with the foundation was considered since the footings for the structural frame for the building are underlaid by rock. Also, in the analysis, the stiffness interaction between the turbine building and the structural frame for the superheater building was neglected, but the mass of the turbine building was included in the dynamic analysis. The analysis, as performed, represents a linear elastic system.
The analysis of the superheater building was performed under the assumption that the north-south motions, east-west motions, and vertical motions were uncoupled. The analysis effort was limited only to horizontal motions in the east-west and north-south directions, and no attempt was made to model vertical motions or to combine vertical and horizontal motions.
However, vertical seismic motions have been considered in the results by increasing the dead load stress in building members by a factor equal to two thirds of the combined mode horizontal inertial g-Ioad as determined in either the east-west or north-south direction.
In each direction, north-south and east-west, the column lines were modeled in detail. These structural models were developed for elastic-static analyses obtained from the computer program STRUDL. They were used for two purposes: to develop the master stiffness matrices associated with the two directions, east-west and north-south, used in the dynamic analyses; and to determine resultant member stresses using the equivalent static seismic forces determined from the dynamic analyses.
The dynamic characteristics, frequencies, and mode shapes of the superheater building were determined using the Westinghouse computer program SAND. The equivalent static forces resulting from the dynamic response were developed using a response spectrum seismic analysis performed by the Westinghouse computer program SPECTA.
The equivalent static force associated with a particular mass resulting from a dynamic response is defined as the square root of the sum of the squares of the equivalent static forces associated with that mass for each mode. The equivalent static force associated with a mode and a mass point is defined as the value of the mass times the maximum acceleration associated with the mass point for that particular mode. The maximum acceleration associated with a mode and mass point is defined as follows:
Chapter 1, Page 62 of 72 Revision 20, 2006 OAG10000215_0099
IP2 FSAR UPDATE Where:
n = Refers to mode n r = Refers to mass r 0'rn= Component of 0rn in the direction of the earthquake 0rn = Component of mode shape n for mass r Mr = Mass lumped at point r (An)Max= Maximum modal acceleration for mode n Sa n= Spectral acceleration for mode n from response curve for 7-percent damping (fj rn)Max = Maximum acceleration in mode n for mass point r rn = Modal participation factor for mode n Sectional views in the north-south and east-west directions are shown in Figures 1.11-5 and 1.11-6. A typical column line modeled for STRUDL to determine overall column line stiffness and permit determination of resultant seismic stresses is shown in Figure 1.11-7. In Figure 1.11-8 is presented the dynamic model used to determine inertial forces.
Results of the analysis showed several column lines contained diagonal bracing with stresses, which exceeded the allowable stress value of 0.9 fy . In addition several of the cross bracings showed compressive stress levels, which exceeded the expected buckling stress as determined by the f! Ir ratio for the member. Overstressed members can be strengthened by attaching cover plates to the angle bracing. In a few instances columns were found to be locally overstressed due to eccentric positioning of cross bracing. These areas can be reinforced by flange cover plates. Approximately 30 tons of additional plate will strengthen the structure.
With respect to tornado resistance of the structure, total lateral load in the north-south direction is approximately 10-percent, and in the east-west direction 20-percent, less than the seismic-induced lateral load on the structure.
Tornado loads were based on a 360-mph wind using the shape factors for a rectangular building as defined in ASCE Paper 3269. It was assumed that 20-percent of the wall area of the building was still intact as a reaction surface for the wind in addition to the total surface area of major equipment and the stack at its existing height. On the basis of this analysis, the building has Chapter 1, Page 63 of 72 Revision 20, 2006 OAG10000215_0100
IP2 FSAR UPDATE approximately the same resistance capacity to a 360-mph tornado wind as it does for the 0.15g earthquake.
1.11.6.5 Evaluation of Structural Modifications In the analysis of the superheater and turbine buildings under lateral loads, the following connections were examined:
- 1. Gusset plates.
- 2. Check of connections between beams and columns to determine their adequacy to transfer horizontal shear load.
- 3. Check of connections at column bases in the foundation to determine their ability to transfer the given horizontal shear load. For those column base connections subjected to a net uplift load, an analysis has been performed to ensure that they are adequate for these loads.
If it was found that a connection was inadequate to support the given load, it was redesigned.
It is not necessary to reanalyze the turbine building after the redesign because the building stiffness characteristics are essentially the same as those assumed in the initial analysis. This is because the significant fixes involved the cross bracing system, which is made up of pairs of cross bracing members. In the initial analysis, both sets of cross bracing were assumed active.
However, the bracing system was such that cross members would buckle under a very small compressive load. Therefore, lateral building load must be carried in tension by the bracing system.
The fix used in the redesign was to double the area of cross bracing. The bracing in compression, due to buckling, is not active in resisting lateral building load. Therefore, only half of the cross bracing assumed in the initial analysis, which is in tension, resists this load.
However, since the area of cross bracing has been doubled, the resultant effective lateral resistance is the same as that assumed in the original analysis.
An initial analysis was made of the superheater building using the existing design parameters.
After completion of the analysis, the overstressed members were strengthened and a dynamic reanalysis made.
Tables 1.11-5, 1.11-6, and 1.11-7 give the relative comparisons in stiffness, horizontal inertial load, and frequency between the initial analysis and the reanalysis.
Subsequently, retired Unit 1 superheater-associated equipment has been removed from certain areas of the superheater building and the areas refurbished to provide permanent administrative facilities. These areas do not contain any safety-related equipment. The total loading on the superheater building has been reduced from the original design loading due to the removal of superheater-associated equipment. Therefore, the administrative facilities will not adversely affect the response of the superheater building during a safe-shutdown earthquake.
Chapter 1, Page 64 of 72 Revision 20, 2006 OAG10000215_0101
IP2 FSAR UPDATE 1.11.7 Seismic Qualification For Safe Shutdown In response to NRC Generic Letter (GL) 87-02 and Supplement No.1 to GL 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue A-46," Con Edison committed to implement Generic Implementation Procedure (GIP-2) including the clarifications, interpretations, and exceptions in the NRC's Supplemental Safety Evaluation Report (SSER-2). The NRC accepted Consolidated Edison's response and commitments regarding this issue as documented in their Safety Evaluation Report (SER) dated November 19, 1992. Consolidated Edison has verified the seismic capabilities of equipment required for safe shutdown, as documented in Reference 7. The verification utilized the Generic Implementation Procedure developed by the Seismic Qualification Utility Group and approved by the NRC. A Summary Report, including the Seismic Evaluation and Relay Evaluation Reports, was submitted to the NRC on December 31, 19967 .
Indian Point 2 site specific SER, dated November 8, 2000, Reference 15, provides NRC conclusions that the licensee may revise its licensing basis by incorporating the SQUG methodology.
Revision 3 of the Generic Implementation Procedure (GIP-3), Reference 12, as modified and supplemented by the NRC's Supplemental Safety Evaluation Report No.2 (SSER-2), Reference 13, and No.3 (SSER-3), Reference 14, may be used as an alternative to existing methods for the seismic design and verification of modified, new, and replacement equipment. Only those portions of GIP-3 as described in Section 5 of "Implementation Guidelines for Seismic Qualification of New and Replacement Equipment/Parts (NARE) using the Generic Implementation Procedure", Reference 16, shall apply to the seismic design and verification of mechanical and electrical equipment, electrical relays, tanks and heat exchangers, and cable and conduit raceway systems.
1.11.8 Protection from Flooding of Equipment Important to Safety In response to NRC Guidelines for Protection from Flooding of Equipment Important to Safety, Consolidated Edison identified the potential sources of flooding outside containment that could affect safety-related equipment. The areas containing safety-related equipment that could be subject to flooding from postulated failure of water systems that are not seismic Class I were evaluated. The plant is designed so as to minimize or eliminate the vulnerability of safety-related equipment to this flooding. Modifications were made to install water level alarm switches in the adjoining Unit 1 condenser pit area, and add flap panels in doors from the primary auxiliary building and the auxiliary feed pump room. These modifications along with the implementation of an alternate safe shutdown capability, as discussed in Section 8.3, serve to mitigate the consequences of the postulated flooding. Additionally, operator action would be taken in the event of flooding from the circulating water system to prevent damage to the 480 volt switchgear in the control building.
In their Safety Evaluation Report (SER) dated December 18, 1980, the NRC determined that design features and operating procedures provide assurance that the plant can be safely shut down in the event of flooding outside containment from a non-seismic component or pipe and that their guidelines (contained in Appendix A to the SER) have been satisfied. 8 REFERENCES FOR SECTION 1.11 Chapter 1, Page 65 of 72 Revision 20, 2006 OAG10000215_0102
IP2 FSAR UPDATE
- 1. E. L. Vogeding, Topical Report - Seismic Testing of Electrical and Control Equipment, WCAP-7397-L, Westinghouse Electric Corporation, January 1970.
- 2. Westinghouse Electric Corporation, Ultimate Strength Criteria to Ensure No Loss of Function of Piping and Vessels Under Earthquake Loading, WCAP-5890, Revision 1.
- 3. NRC Generic Letter, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements, G.L. 87-11, dated June 19,1987.
- 4. NRC Branch Technical Position MEB 3-1, Postulated Rupture Locations in Fluid System Piping Inside and Outside Containment.
- 5. Letter (Attachment B) from S. Bram, Con Edison, to NRC,
Subject:
Request for License Amendment to Technical Specification Modifying Spent Fuel Storage Requirements, dated June 20, 1989.
- 6. Letter (Attachment I) from S. Bram, Con Edison, to NRC,
Subject:
Indian Point Unit No.2 Spent Fuel Storage Capacity Increase, dated January 19, 1990.
- 7. Letter from Quinn, Con Edison, to NRC,
Subject:
Summary Report for Resolution of USI-A-46, Seismic Qualification, dated December 31, 1996.
- 8. Letter from S. A. Varga, NRC, to J. D. O'Toole, Con Edison,
Subject:
Safety Evaluation Report Susceptibility of Safety-Related Systems to Flooding from Failure of Non-Category I Systems, dated December 18, 1980.
- 9. Letter from Cahill, Con Edison, to A. Schwencer, Director of Nuclear Reactor Regulation NRC,
Subject:
Supplemental Response to IE Bulletins 79-02 and 79-07, dated November 27, 1979.
- 10. Letter from Steven A Varga, NRC to John D. O'Toole Con Edison,
Subject:
Completion of IE Bulletin 80-11, "Masonry Wall Design" for Indian Point Nuclear Generating Unit No.2 (lP2), (Safety Evaluation Report included) dated October 19,1983.
- 11. Altran Corporation, Technical Report No. 00222-TR-001, Rev. 1, Reactor Coolant Loop Analysis for Replacement Steam Generators and Snubber Reduction
- 12. Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Plant Equipment, Revision 3, Updated 05/16/97 (GIP-3). Prepared by SQUG and sent to the NRC by letter dated May 16,1997.
- 13. Supplement No.1 to Generic Letter (GL) 87-02 that transmits Supplemental Safety Evaluation Report No.2 (SSER-2) on SQUG Generic Implementation Procedure Revision 2, as Corrected on February 14, 1992 (GIP-2), May 22, 1992.
Chapter 1, Page 66 of 72 Revision 20, 2006 OAG10000215_0103
IP2 FSAR UPDATE
- 14. NRC letter to SQUG dated December 4, 1997, Supplemental Safety Evaluation Report No.3 (SSER-3), on the Review of Revision 3 to the Generic Implementation Procedure for Seismic Verification of Nuclear Power Plant Equipment, Updated 05/16/97 (GIP-3)
- 15. Indian Point Nuclear Generating Station No.2 - "Plant Specific Safety Evaluation Report for Unresolved Safety Issue A-46 Program Implementation", November 8, 2000.
- 16. "Implementation Guidelines for Seismic Qualification of New and Replacement Equipment/Parts (NARE) Using the Generic Implementation Procedure (GIP)",
Revision 4, by MPR Associates Chapter 1, Page 67 of 72 Revision 20, 2006 OAG10000215_0104
IP2 FSAR UPDATE TABLE 1.11-1 Damping Factors PERCENT OF COMPONENT CRITICAL DAMPING Containment structure 2.0 Concrete support structure of reactor 2.0 vessel Steel assemblies:
2.5 Bolted or riveted 1.0 welded Vital piping systems 0.5 Concrete structures above ground 5.0 Shear Wall 5.0 Rigid Frame Chapter 1, Page 68 of 72 Revision 20, 2006 OAG10000215_0105
IP2 FSAR UPDATE TABLE 1.11-2 Loading Combinations and Stress Limits Loading Combinations Vessels1 Piping Supports
- 1. Normal Pm~Sm Pm~S Working stresses loads PL + PB ~ 1.5 S m PL + PB ~ S or applicable factored load design values
- 2. Normal + Pm~Sm Pm~1.2S 1-1/3 working design PL + PB ~ 1.5 Sm PL + PB ~ 1.2S stresses or earthquake applicable loads factored load design values
- 3. Normal + Pm ~ l.2Sm Pm ~ 1.2S Deflections and maximum PL + PB ~ 1.2 (1.5 Sm) PL + PB ~ 1.2 (1.5 S) stresses of potential supports limited earthquake to maintain loads supported equipment within their stress limits
- 4. Normal + Pm ~ l.2Sm Pm ~ 1.2S Deflections and pipe PL + PB ~ 1.2 (1.5 Sm) PL + PB ~ 1.2 (1.5 S) stresses of rupture supports limited loads to maintain supported equipment within their stress limits Where: Pm = primary general membrane stress; or stress intensity PL = primary local membrane stress; or stress intensity PB primary bending stress; or stress intensity Sm = stress intensity value from ASME Band PV Code,Section III S = allowable stress from USAS B31.1 Code for Pressure Piping notes:
- 1. Limited to vessels designed to ASM E,Section III, Class A (or Class 1) rules. Otherwise use piping for stress limits.
Chapter 1, Page 69 of 72 Revision 20, 2006 OAG10000215_0106
IP2 FSAR UPDATE TABLE 1.11-3 DELETED TABLE 1.11-4 Dynamic Characteristics of the Turbine Building MODE Frequency Values No. (cps) 1 0.5042 0.08 2 1.6141 0.12 3 2.2849 0.19 4 4.3292 0.2 5 5.2813 0.2 6 8.2814 0.18 7 12.1704 0.15 9 15.1274 0.15 10 20.754 0.15 11 22.4809 0.15 12 23.8001 0.15 13 27.3040 0.15 14 33.9678 0.15 TABLE 1.11-5 Relative Stiffness Percentages Percentage Increase In Stiffness Between First And Second Analysis (Percent)
RELATIVE LOCATION IN EAST-WEST NORTH-SOUTH SUPERHEATER DIRECTION DIRECTION BUILDING BOTTOM 8 56.7 MIDDLE 18.3 41.4 TOP 19.9 10.4 Chapter 1, Page 70 of 72 Revision 20, 2006 OAG10000215_0107
IP2 FSAR UPDATE TABLE 1.11-6 Inertial Loads Relative Inertial Loads for First and Second Analysis Location in (Units: Kips)
Superheater East-West North-South Building Direction Direction Original Reanalysis Original Reanalysis Bottom 908 908 1091 1102 Middle 1888 1914 1687 1803 Top 1242 1271 1082 1181 TABLE 1.11-7 Frequencies Frequencies For First And Second Analysis (Units: cps)
EAST-WEST NORTH-SOUTH DIRECTION DIRECTION MODE ORIGINAL REANALYSIS ORIGINAL REANALYSIS 1 0.94 1.0 0.72 0.88 2 2.07 2.15 1.58 2.13 3 4.08 4.19 3.47 4.12 1.11 FIGURES Figure No. Title Figure 1.11-1 Ten Percent of Gravity Response Spectra Figure 1.11-2 Fifteen Percent of Gravity Response Spectra Figure 1.11-3 Fuel Storage Building North-South Model [Historical1 Figure 1.11-4 Fuel Storage Building East-West Model [Historical1 Figure 1.11-5 Indian Point Unit 1 Superheater Building North-South Section Figure 1.11-6 Indian Point Unit 1 Superheater Building East-West Section Figure 1.11-7 Column Line "G" Figure 1.11-8 Representation of Lumped Mass Model of Superheater Building Used in Dynamic Analysis Chapter 1, Page 71 of 72 Revision 20, 2006 OAG10000215_0108
IP2 FSAR UPDATE 1.12 INSERVICE INSPECTION AND TESTING PROGRAMS 1.12.1 General The lSI Program complies with the requirements of 10 CFR SO.SSa and is based upon the requirements set forth in ASME Boiler and Pressure Vessel Code,Section XI and by the applicable Code year. This program is also responsive to pertinent provisions of applicable Regulatory Guides.
The Indian Point Unit 2 Inservice Inspection (lSI) and Testing (1ST) Programs for the ten year interval are controlled by Program Plans and plant procedures.
1.12.2 Application The lSI program applies to Quality Groups A, B, and C systems, components (including supports),
and pumps and valves as classified in accordance with Regulatory Guide 1.26, Revision 3.
1.12.3 Program Summary The lSI and 1ST programs identify the specific systems, components, or parts thereof to be examined and the specific pumps and valves to be tested.
1.13 CONTROL OF HEAVY LOADS In response to a December 22, 1980 Generic Letter and to NRC Staff guidelines provided in NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," Con Edison performed evaluations of provisions for the handling and control of heavy loads in the vicinity of irradiated fuel or safe shutdown equipment. Control of heavy loads in the Fuel Storage Building is addressed in section 9.S.6. The NRC documented their acceptance of Con Edison's assessments in a Safety Evaluation Report dated February 19, 1985.
Chapter 1, Page 72 of 72 Revision 20, 2006 OAG1000021S_0109
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NOTE: STIFFNESS MATRICES USED TO DEFINE THE STIFFNESS RELATIONSHIP BETWEEN MASS POINTS IN THE E-W AND N-S DIRECTION INDIAN POINT UNIT No. 2 UFSAR FIGURE 1.11-8 REPRESENTATION OF LUMPED MASS MODEL OF SUPERHEATER BUILDING USED IN DYNAMIC ANALYSIS MIC. No. 1999MC3569 IREV. No. 17A OAGI0000215 0121
IP2 FSAR UPDATE CHAPTER 2 SITE AND ENVIRONMENT 2.1
SUMMARY
AND CONCLUSIONS This section of the FSAR sets forth the site and environmental data, which together formed the basis for the criteria for designing the facility and for evaluating the routine and accidental release of radioactive liquids and gases to the environment. These data support the conclusion that there will be no undue risk to public health and safety with the plant as designed and the environmental characteristics as described. This conclusion rests not only upon the data, but upon the scientific documentation of several independent consultants in their particular area of expertise-health physics, demography, geology, seismology, hydrology, and meteorology.
Environmental characteristics of the area have been documented by field measurements and studies conducted since 1958. These studies quantified the effects on the environment of the operation of nuclear power plants.
Conservative projections have been made of the probable growth of population in the area, and these projections have been taken into account in plant design both as to control of accidents and as to assumptions about operation.
[Historical Information] According to 1980 population estimates, about 50 people reside within a 1100-m radius of Unit 2 (most of them to the east-southeast), and approximately 2600 live within 1-mile. Approximately 75,000 people reside within a 5-mile radius of the facility. The largest concentration of population is in the City of Peekskill, the center of which is about 2.5-miles northeast of the site. The most densely populated 15-degree sector, within 5-miles, is toward Peekskill to the northeast.
The 1960 population within a 1S-mile radius of the site was approximately 352,000, whereas in the year 2000 the estimated population is 1,107,195. The projections do not indicate, and there is no reason to conclude otherwise, that the land usage within this radius will shift appreciably during the intervening period. (The land is now zoned principally for residential and state park use, although there is some industrial activity and minor or isolated agricultural and grazing activity.)
The outer boundary of the low-population zone has been set at 1100 m from Unit 2.
Geologically, the site consists of a hard limestone in a jointed condition that provides a solid bed for the plant foundation. The bedrock is sufficiently sound to support any loads that could be expected up to 50 tons/ft2 , which is far in excess of any load that may be imposed by the plant.
Although it is hard, the jointed limestone formation is permeable to water. Thus, if water from the plant should enter the ground (an improbable event since the plant is designed to preclude any leakage into the ground), it would percolate to the river rather than enter any ground-water supply.
About 80 million gallons of Hudson River water flow past the plant each minute during the peak tidal flow. This flow will provide additional mixing and dilution for liquid discharges from the facility.
The assumption in the plant design is to treat the river water as if it were used for drinking and thus to reduce radioactive discharge, by dilution with ordinary plant effluent, to concentrations that would be tolerable for drinking water. There is a very low probability of flooding at the site.
[Historical Information] Seismic activity in the Indian Point area is limited to low-level microseismicity. Detailed field investigations1-3 have been conducted in the immediate vicinity of Chapter 2, Page 1 of 119 Revision 20, 2006 OAG10000215_0122
IP2 FSAR UPDATE Indian Point and along the major faults in the region. To date, no evidence has been found in the rocks exposed at the surface or sediment overlying fault traces or in cores obtained in the vicinity of Indian Point that might support a conclusion that displacement has occurred along major fault systems within the New York Highlands, the Ramapo, or its associated branches during Quaternary time (the last 1.5 million years). In the vicinity of Indian Point, evidence that no displacement has occurred in the last 65 million years (since the Mesozoic) along specific major structures has been observed.
The plant is designed to withstand an earthquake of Modified Mercalli Intensity VII as required by Appendix A to 10 CFR 100 "Seismic and Geologic Siting Criteria for Nuclear Power Plants." The validity of the selection of an Intensity VII earthquake was adjudicated before the Atomic Safety Licensing Appeal Board. The Appeal Board's decision (ALAB-436) verified Intensity VII as the plant's design-basis earthquake.
Meteorological conditions in the area of the site were determined during a 2-year program (1955 to 1957). The validity of these conclusions has been verified by several programs, including that performed by the Atmospheric Services Department of York Services Corporation in completing a meteorological update for Consolidated Edison Company in 1981 (see Appendix 2A).
These data have been used in evaluating the effects of gaseous discharges from the plant during normal operations and during the postulated loss-of-coolant accident. The evaluations indicate that the site meteorology provides adequate diffusion and dilution of any released gases.
Environmental radioactivity has been measured at the site and surrounding area since 1958 in association with the operation of Indian Point Unit 1 and the construction and operation of Indian Point Units 2 and 3. Unit 3 is owned by Entergy Nuclear Indian Point 3, LLC. These measurements will be continued and reported. The radiation measurements of fallout, water samples, vegetation, marine life, etc., have shown no perceptible post-operative increase in activity. Noticeable increases in fallout have coincided with weapons-testing programs and appear to be related almost entirely to those programs. The New York State Department of Health in an independent 2-year postoperative survel,5 found that environmental radioactivity in the vicinity of the site is no higher than anywhere else in the State of New York.
[Historical Information1 ConSUltants who have participated in the preparation of the various reports, measurements, and conclusions appearing in this chapter include Dr. Merril Eisenbud, director of Environmental Radiation Laboratory, Institute of Industrial Medicine, New York University; Dr. Benjamin Davidson (deceased), meteorologist and director, Geophysical Science Laboratory, New York University College of Engineering; Dr. James Halitsky, senior research scientist, Department of Meteorology and Oceanography, New York University, College of Engineering; Dr. Edgar M. Hoover, Regional Economic Development Institute, Inc.; Metcalf and Eddy Engineers, hydrology specialists; Quirk, Lawler, and Matusky Engineers, Environmental Science and Engineering Consultants; Mr. Karl R. Kennison, consulting civil and hydraulic engineer; and Woodward-Clyde Consultants, consulting engineers, geologists and environmental scientists.
REFERENCES FOR SECTION 2.1
- 1. Ratcliffe 1976.
- 2. Ratcliffe 1980.
- 3. Dames & Moore 1977.
Chapter 2, Page 2 of 119 Revision 20, 2006 OAG10000215_0123
IP2 FSAR UPDATE
- 4. Hollis S. Ingraham, Consolidated Edison Indian Point Reactor Environmental and Post Operational Survey - August, 1965, Division of Environmental Health Services, New York State Department of Health.
- 5. Hollis S. Ingraham, Consolidated Edison Indian Point Reactor Environmental and Post Operational Survey - July, 1966, Division of Environmental Health Services, New York State Department of Health.
2.2 LOCATION 2.2.1 General
[Historical Information] Indian Point is a multiunit site consisting of approximately 239 acres of land on the east bank of the Hudson River at Indian Point, village of Buchanan, in upper Westchester County, New York. Indian Point Units 2 and 3 (see Section 2.2.3) are located north and south, respectively, of Unit 1, which has been retired. The site is about 24-miles north of the New York City boundary line. The nearest city is Peekskill, located 2.5-miles northeast of Indian Point, with a population of about 20,000. An aerial photograph, Historical Figure 2.2-1, shows the site and about 58-mile2 of the surrounding area.
2.2.2 Access The site is accessible by several roads in the village of Buchanan. A paved road links the eastern boundary of the site to the existing plant. The existing wharf is used to receive heavy equipment as needed. The site is not served by rail.
2.2.3 Site Ownership And Control ENIP2 owns Units 1 and 2 while Entergy Nuclear Indian Point 3, LLC (ENIP3) owns Unit 3. The Algonquin Gas Transmission Company has a right-of-way running east to west through the property, 2840-ft long and 65-ft wide. Unit 2 is 1450-ft north of the 26-in. Algonquin gas main. The Georgia-Pacific Corporation has an easement, 1610-ft long and 30-ft wide, through the southerly part of the Indian Point site. The Georgia-Pacific easement is used for overhead electrical power and telephone lines and underground gas, water, and sewer lines.
Units 1, 2, and 3 have a security fence surrounding the "protected" areas. Access to the protected areas is controlled via security buildings that are manned on a 24-hr basis. In addition, spaces within the protected area designated as "vital areas" are provided with additional access control.
All roads within the site are continuously patrolled by security personnel. A site plot plan is shown in Historical Figure 2.2-2.
2.2.4 Activities On The Site The principal activities on the site are the generation, transmission, and distribution of electrical energy; associated service activities; activities relating to the controlled conversion of the nuclear energy of fuel to heat energy by the process of nuclear fission; and the storage, use, and production of special nuclear source and byproduct materials.
Chapter 2, Page 3 of 119 Revision 20, 2006 OAG10000215_0124
IP2 FSAR UPDATE 2.2 FIGURES Figure No. Title Figure 2.2-1 Aerial Photo of Indian Point Site and Surrounding Area
[Historical1 Figure 2.2-2 Indian Point Building Identification [Historical]
2.3 TOPOGRAPHY
[Historical Information] The Indian Point Generating Station is on the east bank of the Hudson River. The river runs northeast to southwest at this point but turns sharply northwest approximately 2-miles northeast of the plant. The west bank of the Hudson is flanked by the steep, heavily-wooded slopes of the Dunderberg and West Mountains to the northwest (elevations 1086 and 1257-ft, respectively, above mean sea level) and Buckberg Mountain to the west-southwest (elevation 793-ft). These peaks extend to the west and gradually rise to slightly higher peaks.
The general orientation of this high ground is northeast to southwest. One mile northwest of the site, Dunderberg bulges to the east. North of Dunderberg and the site, high grounds reaching 800-ft form the east bank of the Hudson River. At this location the Hudson River makes a sharp turn to the northwest. To the east of the site, peaks are generally lower than those to the north and west. Spitzenberg and Blue Mountains average about 600-ft in height, and there is a weak, poorly-defined series of ridges that run in a north-northeast direction. To the west of the site there are the Timp Mountains at an elevation of 846-ft. To the south of the site, elevations of 100-ft or less gradually slope towards Verplanck. The river south of the site makes another sharp bend to the southeast and then widens as it flows past Croton and Haverstraw.
Historical Figure 2.3-1 shows topographic features of the site and the surrounding areas.
2.3 FIGURES Figure No. Title Figure 2.3-1 Topographical Map of Indian Point and Surrounding Area [Historical]
2.4 POPULATION AND LAND USE 2.4.1 Overview The population within a 50-mile radius of the Indian Point site has been estimated for 1990.
These population estimates were taken from statistics recently released by the U.S. Census Bureau. The population within the 50-mile radius of Indian Point has increased from the 1980 estimates by approximately 68,000 people, less than half of one percent.
2.4.2 Population And Land Use According to 1990 estimates, approximately 15.465 million people live within a 50-mile radius of the Indian Point site. A major part of this number live in New York City, an area 25 to 50-miles Chapter 2, Page 4 of 119 Revision 20, 2006 OAG10000215_0125
IP2 FSAR UPDATE south of the plant. Approximately 16S0 persons, concentrated in sectors south to southeast of the station, live within 1-mile of the plant. Approximately 74,000 persons live within S-miles of the plant.
The area surrounding the Indian Point site is generally residential with some large parks and military reservations. Some increased commercial development has occurred within a mile of the station since 1980. Most of the area to the east of the Hudson River within 1S-miles of the site is zoned for residential uses. West of the Hudson within a 1S-mile radius, the Palisades Interstate Park and residential areas are the dominant land uses. The only agricultural areas within 1S-miles are south or northwest of the plant on the west side of the River.
Several maps and tables are included to illustrate the population distribution and land use of the area. Figure 2.4-1 and Figure 2.4-2 show the sector/zone approach to the population data and the area within a SO-mile radius of the Indian Point site. Figure 2.4-3 through Figure 2.4-S illustrate the population distribution radially by sectors out to SO-miles from the plant site. Figure 2.4-6 through Figure 2.4-8 show, respectively, the land uses based on official zoning maps, areas served by public utilities, and areas served by sewage systems, all as of 1970. Table 2.4-1 explains the sector/zone designations for the population maps and tables that follow. Table 2.4-2 through Table 2.4-18 give the 1990 estimated populations for all sector/zones within a SO-mile radius of the Indian Point site.
The New York State Department of Commerce projects no substantial increases in population from 1986 to the year 2013 in any of the four counties in the vicinity of Indian Point.
Table 2.4-19 and Table 2.4-20 show the estimated and projected land uses by County for 1960 and 1980, respectively. These estimates were developed by the Regional Economic Development Institute, Inc., from Regional Planning Association data.
2.4.3 Low-Population Zone About SO people reside within a 1100-m radius of Unit 2, most of them to the east-southeast.
This distance was used as the outer boundary of the low population zone in the analysis of a postulated fission product release. The water boundary (Peekskill Bay) of the more densely populated area of Peekskill was used as the population center distance, which exceeds 1-1/3 times the distance from the reactor to the outer boundary of the low-population zone. A low-population zone outer boundary radius of 1100-m satisfies both 10 CFR 100.11 (a)(3) and 10 CFR SO.67. The low-population zone population in the year 2010 is projected to be approximately 88.
2.4.4 Exclusion Area The exclusion area for Indian Point Unit 2 includes plant property within a S20-m radius of the reactor containment. An exclusion radius of S20-m satisfies both 10 CFR 100.3(a) and 10 CFR SO.67.
2.4.S Population Data Sources The population data used in this section were developed from the following sources:
Chapter 2, Page 5 of 119 Revision 20, 2006 OAG1000021S_0126
IP2 FSAR UPDATE
- 1. 1978 Official Population Projections for New York State Counties, prepared by the Economic Development Board, New York State Department of Commerce.
- 2. Population by Municipality 1970-2000, prepared by the Westchester County Department of Planning, October 1979.
- 3. Population of Rockland County, Capacity and Forecast, 1970-2000, prepared by the Rockland Planning Board, April 1978.
- 4. Population Estimate and Projections, Orange County, New York, prepared by the Orange County Planning Department, March 1980.
- 5. Putnam County Population Projections, prepared by the Putnam County Planning Board, 1977.
- 6. New Jersey Revised Total and Interim Age and Sex Population Projections, 1980-2000, prepared by the New Jersey Department of Labor and Industry, Division of Planning and Research, Office of Demographic and Economic Analysis, April 1979.
- 7. State of Connecticut Population Projections for Connecticut Municipalities and Regions to the Year 2000, prepared by the Office of Policy and Management, Comprehensive Planning Division, February 1980.
- 8. Pennsylvania Projection Series, Summary Report, Employment by Labor Market Area, and Population and Labor Force by County for 1980, 1985, 1990, 1995 and 2000, Report No. 78, PPS-1, prepared by the Office of State Planning and Development, State Economic and Social Research Data Center, June 1978.
Chapter 2, Page 6 of 119 Revision 20, 2006 OAG10000215_0127
IP2 FSAR UPDATE TABLE 2.4-1 Sector and Zone Designators for Population Distribution Map1 Sector Nomenclature Zone Nomenclature Centerline of Sector in Degrees True 22.5° Sector2 Miles From Facility Zone North From Facility o and 360 A 0-1 1 22.5 B 1-2 2 45 C 2-3 3 67.5 D 3-4 4 90 E 4-5 5 112.5 F 5-6 6 135 G 6-7 7 157.5 H 7-8 8 180 J 8-9 9 202.5 K 9-10 10 225 L 10-15 15 247.5 M 15-20 20 270 N 20-25 25 292.5 P 25-30 30 315 Q 30-35 35 337.5 R 35-40 40 Notes:
- 1. An area is identified by a sector and zone alphanumeric designator (refer to Figure 2.4-1). Thus, area A 1 is that area, which lies between 348.75- and 11.25-degrees true north from the facility out to a radius of 1-mile. Area G4 would be that area between 123.75-to 146.25-degrees and the 3- and 4-mile arcs from the facility.
- 2. The letters I and 0 have been omitted from sector designators so as to eliminate possible confusion between letters and numbers.
Chapter 2, Page 7 of 119 Revision 20, 2006 OAG10000215_0128
IP2 FSAR UPDATE TABLE 2.4-2 Population Estimates, 1990, For All Sectors Zone Population 1 1,644 2 15,130 3 18,428 4 14,225 5 24,508 6 25,922 7 28,096 8 25,967 9 36,930 10 46,488 15 342,852 20 488,652 25 920,850 30 2,171,399 35 2,276,172 40 3,451,123 45 3,416,140 50 2,199,601 Chapter 2, Page 8 of 119 Revision 20, 2006 OAG10000215_0129
IP2 FSAR UPDATE TABLE 2.4-3 Population Estimates, 1990, for Sector A (North)
Sector, Zone Population A1 0 A2 70 A3 0 A4 0 A5 400 A6 390 A7 5,301 A8 5,898 A9 2,474 A10 874 A15 4,132 A20 36,987 A25 31,000 A30 57,873 A35 39,998 A40 20,100 A45 17,689 A50 40,853 Chapter 2, Page 9 of 119 Revision 20, 2006 OAG10000215_0130
IP2 FSAR UPDATE TABLE 2.4-4 Population Estimates, 1990, for Sector B (North-Northeast)
Sector, Zone Population B1 0 B2 54 B3 139 B4 143 B5 1,721 B6 1,553 B7 867 B8 246 B9 2,123 B10 1,187 B15 4,343 B20 7,982 B25 20,310 B30 16,651 B35 4,800 B40 6,991 B45 8,457 B50 5,761 Chapter 2, Page 10 of 119 Revision 20, 2006 OAG10000215_0131
IP2 FSAR UPDATE TABLE 2.4-5 Population Estimates, 1990, for Sector C (Northeast)
Sector, Zone Population C1 0 C2 4,879 C3 9,102 C4 4,159 C5 5,534 C6 3,895 C7 2,382 C8 1,594 C9 630 C10 1,034 C15 10,371 C20 9,685 C25 8,200 C30 12,479 C35 13,687 C40 13,067 C45 7,901 C50 6,621 Chapter 2, Page 11 of 119 Revision 20, 2006 OAG10000215_0132
IP2 FSAR UPDATE TABLE 2.4-6 Population Estimates, 1990, for Sector 0 (East-Northeast)
Sector, Zone Population 01 49 02 2,379 03 2,691 04 1,899 05 2,324 06 2,272 07 4,667 08 4,713 09 5,982 010 3,900 015 32,854 020 14,721 025 8,961 030 82,240 035 21,876 040 18,762 045 12,991 050 60,032 Chapter 2, Page 12 of 119 Revision 20, 2006 OAG10000215_0133
IP2 FSAR UPDATE TABLE 2.4-7 Population Estimates, 1990, for Sector E (East)
Sector, Zone Population E1 59 E2 560 E3 0 E4 289 E5 279 E6 345 E7 1,769 E8 1,138 E9 3,287 E10 3,762 E15 17,702 E20 5,099 E25 22,465 E30 20,987 E35 15,730 E40 159,720 E45 162,993 E50 101,121 Cha pter 2, Page 13 of 119 Revision 20, 2006 OAG10000215_0134
IP2 FSAR UPDATE TABLE 2.4-8 Population Estimates, 1990, for Sector F (East-Southeast)
Sector, Zone Population F1 147 F2 305 F3 336 F4 689 F5 260 F6 987 F7 475 F8 860 F9 758 F10 1,999 F15 19,121 F20 11,728 F25 49,821 F30 120,701 F35 58,734 F40 33,691 F45 0 F50 29,199 Chapter 2, Page 14 of 119 Revision 20, 2006 OAG10000215_0135
IP2 FSAR UPDATE TABLE 2.4-9 Population Estimates, 1990, for Sector G (Southeast)
Sector, Zone Population G1 575 G2 2,298 G3 1,295 G4 769 G5 420 G6 3,702 G7 3,892 G8 2,672 G9 2,159 G10 6,890 G15 27,939 G20 23,849 G25 86,999 G30 44,001 G35 17,093 G40 79,903 G45 240,102 G50 328,012 Chapter 2, Page 15 of 119 Revision 20, 2006 OAG10000215_0136
IP2 FSAR UPDATE TABLE 2.4-10 Population Estimates, 1990, for Sector H (South-Southeast)
Sector, Zone Population H1 109 H2 1,782 H3 1,363 H4 741 H5 93 H6 0 H7 0 H8 78 H9 5,039 H10 5,752 H15 22,162 H2O 103,969 H25 226,002 H30 252,482 H35 209,921 H40 535,969 H45 723,004 H50 469,960 Chapter 2, Page 16 of 119 Revision 20, 2006 OAG10000215_0137
IP2 FSAR UPDATE TABLE 2.4-11 Population Estimates, 1990, for Sector J (South)
Sector, Zone Population J1 531 J2 650 J3 20 J4 129 J5 1,351 J6 4,012 J7 3,133 J8 4,308 J9 5,189 J10 4,321 J15 40,993 J20 55,102 J25 220,032 J30 954,691 J35 1,472,384 J40 1,907,927 J45 1,601,010 J50 702,739 Chapter 2, Page 17 of 119 Revision 20, 2006 OAG10000215_0138
IP2 FSAR UPDATE TABLE 2.4-12 Population Estimates, 1990, for Sector K (South-Southwest)
Sector, Zone Population K1 174 K2 1,245 K3 1,282 K4 2,049 K5 8,093 K6 4,124 K7 2,526 K8 2,531 K9 6,291 K10 9,371 K15 86,297 K20 72,902 K25 146,895 K30 427,391 K35 321,209 K40 534,296 K45 444,572 K50 353,770 Chapter 2, Page 18 of 119 Revision 20, 2006 OAG10000215_0139
IP2 FSAR UPDATE TABLE 2.4-13 Population Estimates, 1990, for Sector L (Southwest)
Sector, Zone Population L1 0 L2 63 L3 1,621 L4 2,694 L5 2,184 L6 4,059 L7 2,876 L8 902 L9 2,087 L10 4,021 L15 26,019 L20 28,753 L25 41,514 L30 94,167 L35 31,725 L40 89,824 L45 124,188 L50 54,722 Chapter 2, Page 19 of 119 Revision 20, 2006 OAG10000215_0140
IP2 FSAR UPDATE TABLE 2.4-14 Population Estimates, 1990, for Sector M (West-Southwest)
Sector, Zone Population M1 0 M2 359 M3 188 M4 399 M5 169 M6 274 M7 170 M8 15 M9 96 M10 271 M15 5,139 M20 4,976 M25 14,343 M30 8,817 M35 21,625 M40 18,889 M45 47,849 M50 30,319 Cha pter 2, Page 20 of 119 Revision 20, 2006 OAG10000215_0141
IP2 FSAR UPDATE TABLE 2.4-15 Population Estimates, 1990, for Sector N (West)
Sector, Zone Population N1 0 N2 292 N3 214 N4 0 N5 0 N6 0 N7 0 N8 23 N9 438 N10 63 N15 3,321 N20 8,827 N25 10,234 N30 7,794 N35 14,233 N40 9,028 N45 9,007 N50 2,109 Chapter 2, Page 21 of 119 Revision 20, 2006 OAG10000215_0142
IP2 FSAR UPDATE TABLE 2.4-16 Population Estimates, 1990, for Sector P (West-Northwest)
Sector, Zone Population P1 0 P2 85 P3 52 P4 0 P5 32 P6 58 P7 9 P8 626 P9 357 P10 2,004 P15 17,997 P20 9,983 P25 12,394 P30 47,277 P35 5,927 P40 9,121 P45 3,960 P50 3,917 Cha pter 2, Page 22 of 119 Revision 20, 2006 OAG10000215_0143
IP2 FSAR UPDATE TABLE 2.4-17 Population Estimates, 1990, for Sector Q (Northwest)
Sector, Zone Population Q1 0 Q2 0 Q3 125 Q4 189 Q5 55 Q6 0 Q7 29 Q8 321 Q9 0 Q10 1,039 Q15 7,023 Q20 9,872 Q25 10,745 Q30 12,244 Q35 10,160 Q40 7,942 Q45 5,653 Q50 6,962 Cha pter 2, Page 23 of 119 Revision 20, 2006 OAG10000215_0144
IP2 FSAR UPDATE TABLE 2.4-18 Population Estimates, 1990, for Sector R (North-Northwest)
Sector, Zone Population R1 0 R2 109 R3 0 R4 76 R5 1,593 R6 251 R7 0 R8 42 R9 20 R10 0 R15 17,439 R20 44,219 R25 10,935 R30 12,144 R35 17,070 R40 5,893 R45 6,764 R50 3,504 Cha pter 2, Page 24 of 119 Revision 20, 2006 OAG10000215_0145
IP2 FSAR UPDATE TABLE 2.4-19 Estimated Land Use in 1960 and Projected Land Use in 19801 Within a 55-Mile Radius Intensive 1960 and 1980 Nonintensive 1960 Nonintensive 1980 1 2 3 4 5 6 7 8 9 10 11 12 Public Community Public Industrial/ Institutional Rights- Facilities Parks Rights- Grand Residential Commercial Total and Park of-Way Total Institutions Recreation of-Way Total Open Totals 1960 Square miles 1032 216 1248 696 418 1114 4062 6424 Percentage of total developed land 43 9 52 29 19 48 High 58 12 45 22 Low 32 2 15 15 1980 Square miles 2040 368 2408 876 784 682 2342 1674 6424 Percentage of total developed land 43 8 51 19 16 14 49 1960-1980 Square miles of land to be developed 1400 220 1620 1228 Percentage of total land to be developed 58 42 Notes:
- 1. The averages were derived from the data in "Table 3. The Use of Developed Land in Selected Areas of the Regions." RPA Bulletin Number 100, Page 21, September 1962. The data for square miles excludes Monmouth County from the original Regional Plan Association (RPA) totals.
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Chapter 2, Page 25 of 119 1U'l o Revision 20, 2006
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IP2 FSAR UPDATE TABLE 2.4-20 (Sheet 1 of 2)
Land Use Projection by County for 1980 in Sguare Miles Within a 55-Mile Radius Counties in Con Ed Study Area Intensive Low Intensive Community Outside Industrial! Facilities Parks Public Rights-State In RPA Region RPA Region Residential Commercial Community Recreation Of Way Open Institutions Conn. Fairfield 183 33 92 83 71 171 Litchfield [30h [6] [3] [3] [2] [5]
New Haven [88] [19] [73] [65] [55] [134]
N.J. Bergen 118 22 20 19 16 38 Essex 83 16 6 6 5 12 Hudson 26 5 3 3 2 6 Middlesex 126 (58h 22 (10) 18 16 14 34 Morris 130 23 69 63 54 129 Passaic 75 14 23 21 18 43 Somerset 71 (24) 13 (4) 16 15 12 30 Sussex [34] [8] [107] [97] [83] [199]
Union 63 12 6 6 5 11 Warren [3] [1] [9] [9] [7] [18]
N.Y. Dutchess 106 19 152 138 117 283 Nassau 230 41 5 4 4 9 Orange 110 20 154 140 119 286 Putnam 37 6 42 38 32 79 Rockland 56 10 25 23 19 46 0
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IP2 FSAR UPDATE TABLE 2.4-20 (Sheet 2 of 2)
Land Use Projection by County for 1980 in Sguare Miles Within a 55-Mile Radius Counties in Con Ed Intensive Low Intensive Study Area Community Outside Industrial! Facilities Parks Public Open State In RPA RPA Region Residential Commercial Community Recreation Rights-Region Of Way N.Y. Suffolk 279 (199h 50 (35) 92 84 72 172 Sullivan [8h [4] [117] [106] [90] [217]
Ulster [53] [12] [207] [188] [160] [386]
Westchester 162 31 53 48 42 99 Bronx 25 4 3 3 2 5 Kings 42 7 4 4 4 8 New York 14 2 1 1 1 3 Queens 65 11 7 7 6 13 Richmond 39 7 3 2 2 5 P.A. Pike [7] [1] [76] [69] [59] 142 Total RPA 2040 368 79 4 3* 724 3* 6 173* 14823*
Region3 Total Consolidated 2078 383 1385 1261 1073 2583 Edison Area Notes:
- 1. Figures in brackets are for those counties outside RPA's Region. They are added to the total for Con Ed's area.
- 2. Figures in parentheses are those portions of the RPA Region contained in the Con Ed area.
o 3. Total RPA Region figures followed by
- indicate that only the portion of the counties in Con Ed's area are included.
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IP2 FSAR UPDATE 2.4 FIGURES Figure No. Title Figure 2.4-1 Schematic Sector/Zone Diagram Figure 2.4-2 Indian Point Station Ten and Fifty Mile Radius Map Figure 2.4-3 Five Mile Sector/Zone Diagram Figure 2.4-4 Ten Mile Sector/Zone Diagram Figure 2.4-5 Fifty Mile Sector/Zone Diagram Figure 2.4-6 Map and Description Showing Land Usage Figure 2.4-7 Map and Description of the Area Showing Public Utilities Figure 2.4-8 Map and Description of the Area Showing Sewage Systems Cha pter 2, Page 28 of 119 Revision 20, 2006 OAG10000215_0149
IP2 FSAR UPDATE 2.5 HYDROLOGY The hydrologic features of the Indian Point site are relevant to the analysis of radioactive liquid discharges from the plant. These features are the Hudson River, ground water and wells, and surface-water reservoirs. During normal plant operation liquid wastes are discharged to the Hudson River through the circulating water discharge canal. Ground-water contamination from accidental ground seepage or leakage from the plant is not expected because such leakage should flow to the river because of the higher elevation of the plant relative to the river.
The hydrology in the environs of the Indian Point site has been extensively studied for Con Edison by numerous consultants, augmenting the data base established through the investigations of various governmental agencies. The initial Con Edison study was conducted in 1955 by Kennison,1 who analyzed the flow characteristics of the river at the site. Metcalf and Eddl further examined the river flow, and also investigated local groundwater hydrology and surface-water reservoirs. The salient aspects of these and other studies 3 ,4 are reported below.
The Hudson River below the dam at Troy (immediately below the confluence of the Hudson and Mohawk Rivers) is a tide-influenced, estuarine waterway. (see Figure 2.5-1.) Fresh water from the combined Hudson and Mohawk Rivers, as well as from numerous tributaries, discharges directly into the tidal portion of the river. Seawater enters the extreme lower reaches of the river through the Narrows and the Harlem/East River. The distribution of saltwater is influenced by fresh water flow, tides, physical characteristics of the river channel, and weather.
Flow in the Hudson River is controlled more by the tides than by the runoff from the tributary watershed. River width opposite the plant ranges from 4500 to 5000-ft. Water depths within 1000-ft of the shore near the site are variable with an average depth of 65-ft; at some points the depth exceeds 85-ft. River cross-sectional areas in the vicinity of the site range from 165,000 to 170,000-ff. Tidal flow past the plant is about 80 million gpm about 80-percent of the time, and it has been estimated that this frequency flow is at least 9 million gpm in a section 500-600-ft wide immediately in front of the facility. Mean tidal flow in the vicinity of the site is over 70 million gpm.
The average downstream flow (for a 17-year period preceding 1930) is in excess oe
- 11,700,000 gpm 20-percent of the time.
- 6,800,000 gpm 40-percent of the time.
- 4,710,000 gpm 60-percent of the time.
- 3,100,000 gpm 80-percent of the time.
- 1,800,000 gpm 98-percent of the time.
The plant is designed to use the dilution characteristics of the large tidal flow and will be operated such that discharges into the river would not contravene regulatory limitations.
Historical flow patterns were further examined by Quirk, Lawler, and Matuskl,4 who reported both long-term (monthly) river discharges and potential drought flows. Quirk, Lawler, and Matusky also analyzed and reported on the hydraulic conveyance properties of the estuary and the effects of tide and salinity on movement in the estuary.
Review of historical records indicates that flooding at the site is non-existent. Flood stages are primarily the effect of tidal influence, with the secondary influence of runoff. The highest recorded water elevation in the vicinity of the site was 7.4-ft above mean sea level (MSL), which occurred Cha pter 2, Page 29 of 119 Revision 20, 2006 OAG10000215_0150
IP2 FSAR UPDATE during an exceptionally severe hurricane in November 1950. Since the river water elevation would have to reach 15-ft 3-in. above MSL before it would seep into any of the Indian Point buildings, the potential for any flooding damage at the site appears to be extremely remote.
Seven different flooding conditions governing the maximum water elevation at the site were investigated, including the following:
- 1. Flooding resulting from runoff generated by a probable maximum precipitation over the entire Hudson River drainage basin upstream of the site.
- 2. Flooding caused by the occurrence of any upstream dam failure concurrent with heavy runoff generated by a standard project flood.
- 3. Flooding due to the occurrence of a probable maximum hurricane concurrent with a spring high tide in the Hudson River.
The severest flooding condition revealed by the study results from the simultaneous occurrence of a standard project flood, a failure of the Ashokan Dam and a storm surge in New York Harbor at the mouth of the Hudson River resulting from a standard project hurricane. The water level under these conditions would reach 14-ft above MSL. Local wave action due to wind effects has been determined to add 1-ft to the river elevation producing a maximum water elevation of 15-ft above MSL at the Indian Point Site. Since this maximum water elevation is 3-in. lower than the critical elevation of 15-ft 3-in. noted earlier, it is reasonable to conclude that flooding in the Hudson River will not present a hazard to the safe operation of Indian Point.
The three most severe hurricanes to hit New York Harbor (September 21, 1821; November 25, 1950 (mentioned previously); and September 12, 1960) produced tidal surges at the Battery of 11-ft, 8.2-ft and 6.3-ft, respectively. Accordingly, these surges would appear as 7.5-ft, 5.5-ft, and 4.3-ft surges at Indian Point. The 5.5-ft surge predicted for the November 25, 1950, hurricane agrees well with the actual surge that produced the 7.4-ft-high watermark recorded for Indian Point on that date.
The Quirk, Lawler and Matusky report indicated that the combination of the maximum probable runoff, upstream dam failures and maximum ebb tide in the Hudson River is a less severe condition than the one postulated above. This latter scenario would cause the water level at Indian Point to be 11.7-ft above MSL, also below the critical control elevation.
The report also indicates that the combination of probable maximum hurricane, spring high tide, and wave run-up will cause the water level at Indian Point to reach 14.5-ft above MSL. This is also below the critical control elevation of 15-ft-3-in. Table 2.5-1 summarizes the Indian Point water surface elevations resulting from the various combinations.
In view of the recorded hydrologic history of the Hudson River and New York Harbor and the predicated maximum hurricane surge at Indian Point, flooding at the site is a highly unlikely possibility.
Within a 5-mile radius of the plant only one municipal water supply uses ground water. Other wells in this area are used for industrial and commercial purposes. The rock formations in the area and elevations of wells relative to the plant are such that accidental ground leakage or Cha pter 2, Page 30 of 119 Revision 20, 2006 OAG10000215_0151
IP2 FSAR UPDATE seepage percolating into the ground at Indian Point will not reach these sources of ground water, but will flow to the river.
Only two reservoirs within a 5-mile radius are used for municipal water supplies. The first, Camp Field Reservoir, is the raw-water receiving basin for the system, which serves the city of Peekskill. This system uses the Catskill Aqueduct and Montrose Water District as alternative sources of water supply. The second reservoir, the impounding reservoir for the Stony Point water system, serves the towns of Stony Point and Haverstraw, and the villages of Haverstraw and West Haverstraw. The Stony Point system is connected to the Spring Valley Water Company to provide an alternative source of supply. A third reservoir within 5-miles of the plant, Queensboro Lake, supplies water to a state park area only. The location of these reservoirs, and others within a 15-mile radius of the site, are shown on Figure 2.5-2. The city of New York's Chelsea Pumping Station is located about 1-mile north of Chelsea, New York, on the east bank of the Hudson River, about 22-miles upstream of the site. Water will be pumped from intakes in the river at the rate of up to 100 million gal per day into the city reservoir system as required to supplement the primary supply from watersheds during severe drought conditions. This source, however, was not used during the recent 1981 drought.
The discharge of any contaminant into a tidal estuary will result in its distribution throughout the estuary. Factors affecting this distribution include tide amplitude and current, river geometry, salinity distribution, and freshwater discharge. Quirk, Lawler, and Matusky investigated for Con Edison the influence of these factors and determined the effect of radioactive discharges on overall river concentrations, and specifically conditions at Chelsea Pumping station, as discussed in Section 11.1. During normal operations, the plant discharge will not exceed its maximum permissible concentration. Compliance with regulatory release limits is further discussed in Section 11.1.
REFERENCES FOR SECTION 2.5
- 1. Letter report of Karl L. Kennison, Civil and Hydraulic Engineer, to G. R. Milne, Con Edison, November 28, 1955.
- 2. Metcalf and Eddy Engineers, Hydrology of Indian Point Site and Surrounding Area, October 1965.
- 3. Quirk, Lawler, and Matusky Engineers, Transport of Contaminants in the Hudson River above Indian Point Station, May 1966.
- 4. Quirk, Lawler, and Matusky Engineers, Evaluation of Flooding Conditions at Indian Point Nuclear Generating Unit No.3, April 1970.
Chapter 2, Page 31 of 119 Revision 20, 2006 OAG10000215_0152
IP2 FSAR UPDATE TABLE 2.5-1 Water Surface Elevation at Indian Point Resulting From Stated Flow and Elevation Conditions Elevation at Indian Elevation at Elevation at Point Including the Battery - Flow at Indian Point - Local Oscillatory Component Flow Datum MSL Indian Point Datum-MSL Wave Height Datum at Indian Point (ft) (million cfs) (ft) MSL (ft)
- 1. Probable MSL 0.00 1.100 12.7 13.7 maximum flood
- 2. Probable High water 1.014 12.4 13.4 maximum flood +/-2.20 and tidal flow
- 3. Probable Low water 1.165 13.0 14.0 maximum flood -2.20 and tidal flow
- 4. Standard project MSL 0.00 0.705 7.2 8.2 flood and Ashokan Dam failure
- 5. Standard project Standard project 0.550 13.0 14.0 flood hurricane
+11.00
- 6. Standard project Standard project 0.705 14.0 15.0 flood and Ashokan hurricane Dam failure (+11.00)
- 7. Probable Probable maximum 13.5 14.5 maximum hurricane hurricane and +17.5 0 spring high tide
>>
G) 0 0
0 0
N
--"
(J1 Cha pter 2, Page 32 of 119 I Revision 20, 2006 0
--"
(J1 0)
IP2 FSAR UPDATE 2.5 FIGURES Figure No. Title Figure 2.5-1 Map & Description Showing Location of Sources of Potable &
Industrial Water Supplies & Watershed Areas Figure 2.5-2 Hudson River Drainage Basin 2.6 METEOROLOGY 2.6.1 General
[Historical Information] Meteorological parameters related to atmospheric transport and diffusion have been extensively investigated in the Indian Point area since 1955. Studies of the wind flow characteristics, induced by the topography surrounding the site, illustrated the unique valley wind system and the channeling of low level winds.
Meteorological studies1-3 were conducted from 1955 to 1957 by the Research Division of New York University, under the direction of Prof. Ben Davidson, in support of Unit 1 licensing activities. Data from these studies illustrated the channeling of the air flow by the terrain into downvalley (north-northeast) and upvalley (south-southwest) regimes. Historical data collected by the U.S. Weather Bureau in 1932 also illustrated the valley wind system.
Subsequent meteorological investigations were conducted from 1968 through 1972 by New York University School of Engineering and Science, Department of Meteorology and Oceanography, under the direction of Dr. James Halitsky and Mr. Edward J. Kaplin. These studies supported the earlier findings of the valley wind system by Prof. Davidson and are documented in Appendix 2A of this FSAR and in the FSAR for Indian Point Unit 3.
The most recent meteorological programs and data analyses conducted in the Indian Point environs since 1972 were documented in a York Services Corporation report (Meteorological Update, September 1981). This report is included in Appendix 2A. The 10-m elevation on the 100-ft meteorological tower used for the Unit 2 siting studies is the backup tower to the 400-ft (122-m) primary tower. The 10-m tower installed at the Buchanan Service Center is also available as an additional contingency.
The York Services Report summarizes the meteorological activities conducted for Indian Point from 1955 to 1981. Included are topographic effects, wind correlations, data collection, diumal wind distribution, trajectory analyses, atmospheric stability, and wind distributions. The report substantiates previous studies conducted on the existence of the valley wind system in the environs of Indian Point.
2.6.2 Application of Site Meteorology to Safety Analysis of Loss-Of-Coolant Accident The atmospheric dispersion factors required for the safety analysis of Chapter 14 have been computed for the worst possible meteorological conditions that could prevail at the Indian Point site.
A search of the records indicates that the most protracted consecutive period during which the wind direction was substantially from the same direction was 5 days. The winds in this case Cha pter 2, Page 33 of 119 Revision 20, 2006 OAG10000215_0154
IP2 FSAR UPDATE were from the northwest and speeds ranged from 15 to 30 mph. Therefore, this case does not represent the most conservative meteorology associated with the loss-of-coolant accident.
The most frequent wind flow at low heights under inversion conditions is down the axis of the valley. This direction, roughly 10- to 30-degrees, is also the direction of maximum wind frequency. Because of the relatively high frequency of inversion conditions associated with this wind direction, the safety analysis assumes that the distribution of wind speed and thermal stability during the hypothetical accident is exactly that measured at the 100-ft tower level for the 5- to 20-degree wind direction.
The valley wind is diurnal in nature, that is, upvalley during unstable hours and downvalley during stable hours. In general, these local winds are most frequent under clear sky and relatively light prevailing wind conditions. The diurnal variation of the vector mean wind as measured 70-ft above river during September-October 1955 is shown in Figure 2.6-1 for conditions in which the large-scale flow was virtually zero (12 days) and in Figure 2.6-2 for conditions in which the large-scale flow (geostropic wind) was less than 16 mph (35 days). It may be seen that for these virtually stagnant prevailing wind conditions, there is a regular diurnal shift in wind direction and that the mean vector wind associated with the downvalley flow is approximately 6 mph.
A measure of the magnitude of the diurnal shift in wind direction is shown in Figure 2.6-3, where the steadiness of the wind (vector) mean speed over the mean scalar speed is shown as a function of time and the strength of the prevailing flow. Where the steadiness is close to one, the persistence of a given wind direction is very high. These data indicate that a consecutive 24-hr downvalley flow with light wind speeds and inversion conditions is extremely improbable due to the diurnal variation of the steadiness.
The safety analysis of the loss-of-coolant accident assumes that the accident occurs during downvalley inversion flow conditions and that this condition persists for 24 hr with average wind speeds slightly less than 2 m/sec. Figures 2.6-1 and 2.6-2 indicate that the duration of the downvalley flow is about 12 hr rather than 24 hr and that the vector mean wind speeds are approximately 2.5 m/sec.
In view of the discussion above, it must be concluded that the safety analysis for the first 24 hr is conservative to within a factor of about 2.
The remainder of the safety analysis assumes that for the next 30 days, 35-percent of the winds are in the 20-degree sector corresponding to the nocturnal downvalley flow and that wind speed and thermal stability are as observed over the period of 1 year as measured at the 100-ft tower location. If the observations were distributed uniformly throughout the year, slightly over 100 hr per month of 5- to 20-degree winds could be expected to occur. The analysis assumes that 276 hr of 5- to 20-degree winds occur in the first 31 days after the accident, and that about 130 of these hours are characterized by inversion conditions. Approximately 35 weak-pressure gradient days were observed in September-October 1955 or about 430 hr per month. From Figure 2.6-3, the hour during which the downvalley flow is quite persistent under weak-pressure gradient conditions are from 0 to 8 hr. Assuming that the steadiness is 1.0 during these hours (it is in fact about 0.9 or less), the number of downvalley inversion winds per month during September and October is on the order of 140 hr per month. This indicates that the meteorology assumed in the safety analysis beyond the first 24 hr is reasonable for the worst Cha pter 2, Page 34 of 119 Revision 20, 2006 OAG10000215_0155
IP2 FSAR UPDATE months (September and October) and is undoubtedly conservative with varying degrees of conservatism for the remainder of the year.
The inversion frequency assumed for the 30-day accident case is conservative because the evaluation is made from concurrent assumptions concerning the postulated meteorological conditions, namely:
- 1. Inversion conditions prevail for 42.4-percent of the time.
- 2. The wind direction is within a narrow 20-degree sector for 35-percent of the time.
This is equivalent to assuming that in the model 20-degree sector, the inversion frequency is 14.8-percent for the 30-day period. The observed annual maximum inversion frequency for a 20-degree sector is 6.2-percent (p. 29, Table 3-3, Section 1.6 of Reference 3). If we assume that the inversion frequency is spread uniformly throughout the year, almost 3 months worth of inversions in the model 20-degree sector are considered to occur in the first 31-day month after the accident. The assumption of uniform spread of inversion frequency over the year is examined above where an attempt is made to isolate those local meteorological conditions at Indian Point, which might yield the highest 30-day dose. It is concluded that the "worst" meteorological conditions are associated with the nocturnal downvalley flow, which is most frequent during September and October.
REFERENCES FOR SECTION 2.6
- 1. New York University, Research Division, A Micrometeorological Survey of the Buchanan, NY., Area, NYU Technical Report 372.1, November 1955, which was Exhibit L-1, Docket 50-3, given in its entirety. The topography of the area surrounding the site is described and the effects of the topography on meteorological conditions are discussed. The types of data collected, the methods and frequencies of collection, the description and location of the equipment, and the general scope of the meteorological program are indicated in this report. Seasonal wind characteristics, including speeds, directions, and frequencies are tabulated.
- 2. New York University, Research Division, Evaluation of Potential Radiation Hazard Resulting From Assumed Release of Radioactive Wastes to the Atmosphere From Proposed Buchanan Nuclear Power Plant, Sections 1, 2, and 3 of NYU Technical Report 372.3, April 1957. This report was submitted to the NRC in its entirety as Exhibit L-5, Docket 50-3. These sections discuss the diffusive conditions and the climatological data of the site. The basis for evaluating the diffusion parameters selected for the safety analysis is given on pages 19 to 21. Section 3 contains tables of frequency distribution of diffusion classes and wind directions, and also wind roses.
- 3. New York University, Research Division, Summary of Climatological Data at Buchanan, N.Y., 1956-1957, NYU Technical Report 372.4, March 1958, was Exhibit L-6, Docket 50-3. This report summarizes the final meteorological testing at Indian Point.
Chapter 2, Page 35 of 119 Revision 20, 2006 OAG10000215_0156
IP2 FSAR UPDATE 2.6 FIGURES Figure No. Title Figure 2.6-1 Diurnal Variation of Mean Vector Wind for Virtually Zero Pressure Gradient Conditions Figure 2.6-2 Diurnal Variation of Mean Vector Wind for 24 Hr Periods of Weak Pressure Gradient Conditions Figure 2.6-3 Steadiness of Wind as a Function of Time of Day for Indicated Pressure Gradient Conditions 2.7 GEOLOGY AND SEISMOLOGY
[Historical Information1 The Indian Point site and surrounding area were studied in 1955 by Sidney Paige, consulting geologist, before the construction of Unit 1 In 1965, Thomas W. Fluhr, P.E., an engineering geologist, reviewed the geology of the site and made additional borings at the location of Unit 2.
In 1982, a report by Woodward-Clyde Consultants was done to update Section 2.7 of the FSAR.
The previous studies are listed in the reference list of the report. The report is included in Appendix 2B.
A seismic monitoring network exists in the vicinity of the site and data from this network is periodically evaluated.
2.8 ENVIRONMENTAL RADIOACTIVITY Monitoring for environmental radioactivity in the vicinity of the Indian Point Station began in 1958, approximately 4 years before the operation of Unit 1. Measurements since that time have indicated that the present operation of Units 2 and 3 and the past operation of Unit 1 have had no significant effect on the environment. The monitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and the modeling of the environmental exposure pathways. Measurements of radioactivity in the environment are summarized in the Annual Radiological Environmental Operating Report, which is submitted annually as required by the plant's Technical Specifications.
Determinations of radioactivity in the environment are made regularly. Samples include drinking water from two nearby reservoirs and the New York City Aqueduct, river water, sediments, fish, shellfish, lake and river aquatic vegetation, land vegetation, soil from locations in the vicinity of the site, shoreline soil, air, and milk along with direct gamma radiation measurements in various locations.
The overall objectives of the environmental monitoring program are as follows:
- 1. To establish a sampling schedule for Indian Point Units 1 and 2 that will recognize changes in radioactivity in the environs of the plant.
Cha pter 2, Page 36 of 119 Revision 20, 2006 OAG10000215_0157
IP2 FSAR UPDATE
- 2. To ensure that the effluent releases are kept as low as is reasonably achievable (ALARA) and within allowable limits in accordance with 10 CFR 20.
- 3. To verify projected and anticipated radioactivity concentrations in the environment and related exposure from releases of radioactive material from the Indian Point site.
Results of environmental surveys conducted by Con Edison have been verified by the Bureau of Radiological Health Service of the New York State Health Department in previous years and presently, by the New York State Bureau of Environmental Radiation. 1,2 Environmental surveys have also been confirmed by Dr. Merril Eisenbud, Director of Environmental Radiation Laboratory, Institute of Environmental Medicine, New York University Medical Center, who has found that the levels of environmental radioactivity are associated with natural background and fallout of nuclear weapons testing. 2 In a study of the radioactivity in the Hudson River, Mr. Sherwood Davis, Director, Bureau of Radiological Health Service, New York State Department of Health, et aI., have concluded that the discharges from Indian Point Unit 1 "are a minute fraction of the federallimits."4 The above results were obtained in preoperational and operational periods of Units 1 and 2 in the late 1950s and in the 1960s. In the more recent years of the late 1970s, radiological impact evaluations have shown similar results. These evaluations of actual plant releases have been performed for inclusion in the effluent release reports and have shown that operation of the Unit 2 plant has had an insignificant impact on the environs.
REFERENCES FOR SECTION 2.8
- 1. New York State Department of Health, Division of Environmental Health Services, Consolidated Edison Indian Point Reactor, Post Operational Survey, August 1965.
- 2. New York State Department of Health, Division of Environmental Health Services, Consolidated Edison Indian Point Reactor, Post Operational Survey, July 1966.
- 3. New York University Medical Center Institute of Environmental Medicine, Ecological Survey of the Hudson River: Progress Report No.2, submitted to Division of Radiological Health, USPHS, Contract PHS 86-95, Neg. 141, December 1966.
- 4. F. Cosolito, et aI., Radioactivity in the Hudson River, Symposium on Hudson River Ecology, Hudson River Valley Commission of New York, October 4-5, 1966.
Chapter 2, Page 37 of 119 Revision 20, 2006 OAG10000215_0158
IP2 FSAR UPDATE NOTE: This information is classified as Historical Information APPENDIX2A FACILITY SAFETY ANALYSIS REPORT (FSAR)
CONSOLIDATED EDISON COMPANY OF NEW YORK, INCORPORATED INDIAN POINT NUCLEAR GENERATING UNIT NO.2 METEOROLOGICAL UPDATE SEPTEMBER, 1981 YSC PROJECT + 01-4122 prepared by:
EDWARD J. KAPLAN BRUCE R. WUEBBER Senior Scientist Staff Meteorologist ATMOSPHERIC SERVICES DEPARTM ENT Cha pter 2, Page 38 of 119 Revision 20, 2006 OAG10000215_0159
IP2 FSAR UPDATE TABLE OF CONTENTS LIST OF FIGURES LIST OF TABLES 1.0 GENERAL 1.1 Historical 1.1.1 Introduction 1.1.2 Tower Siting and Instrumentation 1.1.2.1 Hudson River 1.1.2.2 General Topography 1.1.2.3 Site Configuration 2.0 METEOROLOGICAL PROGRAM AT INDIAN POINT 2.1 122m Meteorological Tower 2.1.1 Siting 2.1.2 Instrumentation 2.1.2.1 Sensor Configuration 2.1.2.2 Instrumentation Specifications 2.1.2.2.1 Climatronics F460 Wind Speed Transmitter 2.1.2.2.2 Climatronics F460 Wind Direction Transmitter 2.1.2.2.3 Climatronics TS-10 and TS-10WA Motor Aspirated Shields 2.1.2.2.4 Climatronics 100087, 100087 Temperature-Delta Temperature 2.1.2.2.5 Climatronics DP-1 0 Dew Probe 2.1.2.2.6 Climatronics 1000971 - Heated Rain -
Snow Gauge 2.1.2.2.7 Data Collection Systems 2.1.3 Meteorological Support System 2.2 Data Log 2.2.1 Indian Point Tower IP3 2.2.2 122 Meter Meteorological Tower Cha pter 2, Page 39 of 119 Revision 20, 2006 OAG10000215_0160
IP2 FSAR UPDATE 3.0 ANALYSES DATA 3.1 Indian Point Tower IP3 3.2 122M Meteorological Tower (I P4) 3.2.1 October 1, 1973 to August 31,1974 3.2.2 August 1,1978 - July 31, 1979 3.2.2.1 Surface Air Trajectories Analyses -
Summary 3.2.3 March 1980 - December 1980 3.2.3.1 General 3.2.3.2 Wind Frequency Distributions 3.2.3.3 Diurnal Wind Distributions 3.2.3.4 Resultant and Concurrent Hourly Winds 3.2.3.5 Summary - Trajectory II Study 3.2.4 January 1, 1979 through December 31, 1980 3.2.4.1 Data Analyses 3.2.4.2 Wind Frequency Distributions 3.2.4.3 Diurnal Wind Direction Distributions 3.2.4.4 Wind Speed Distributions 3.2.4.5 Wind Velocities and Atmospheric Stability 3.2.4.5.1 Joint Frequency Distribution of Wind Direction and Stability 3.2.4.5.2 Frequency of Occurrence of Stability Categories 3.2.4.5.3 Average Wind Speed and Diurnal Variation as a Function of Stability Categories 4.0
SUMMARY
4.1 Meteorology 4.1.1 General 4.1.2 122 Meter Meteorological Tower System 4.1.3 Local Meteorological Characteristics 4.1.4 Conclusion
5.0 REFERENCES
Cha pter 2, Page 40 of 119 Revision 20, 2006 OAG10000215_0161
IP2 FSAR UPDATE LIST OF FIGURES Figure No. Title Figure 1 Ground Contours at Elevation 200 Feet Figure 2 Ground Contours at Elevation 400 Feet Figure 3 Elevations in the Indian Point Region Figure 4 Water Courses in the Indian Point Region Figure 5 Existing and Historical Meteorological Towers at Indian Point Figure 6 Indian Point Meteorological Site Figure 7 Tower Configuration Figure 8 Station Configuration Figure 9 Indian Point - Meteorological Support Systems Figure 10A Two Station Wind Correlation Data Period - October 1973 Figure 10B Two Station Wind Correlation Data Period - December 1973 Figure 11 Position of One Mile Grid in Relation to Topographic Features Figure 12 Position of Wind Files on Grid Figure 13 Average March, 1980 East and West Bank Diurnal Wind Distributions Figure 14 Average June, 1980 East and West Bank Diurnal Wind Distributions Figure 15 Average December, 1980 East and West Bank Diurnal Wind Distributions Figure 16 Locations of Monitoring Sites in Relation to One Mile Grid Figure 17 Comparison of 10M Level Diurnal Wind Distributions Figure 18 Comparison of 122M Level Diurnal And Wind Distribution Figure 19 Diurnal Distribution of Wind Speeds Figure 20 Percent Probability Distribution of Wind Speeds Chapter 2, Page 41 of 119 Revision 20, 2006 OAG10000215_0162
IP2 FSAR UPDATE LIST OF TABLES Table 1 Tower and Instrumentation Record Table 2 Valid Data Log Table 3 Comparison of Annual Percent Occurrence of Stability Categories Table 4 Summary of Trajectory End-Points Table 5 Summation of Trajectory End Points - August, 1978 Table 6 Summation of Trajectory End Points - January, 1979 Table 7 Summation Trajectory Occurrences South of Indian Point Table 8 Locations of Stations Relative to Indian Point Table 9 Valid Data for Trajectory Wind Sites Table 10 Frequency Distribution of 24 Hour Resultant Wind Directions Table 11 Summary of Two-Station Wind Correlations Piermont (Site 1)
Referenced to Selected Monitoring Locations (Site 2)
Table 12 Concurrence of Two-Station Wind Directions Table 13 Diurnal Distribution of Occurrences of Eight-Hour Trajectories With On Grid Reversals Table 14 Summary of Trajectory End-Point Counts Table 15 Summary of Trajectory End-Points (Percent)
Table 16A Historical Comparisons of Wind Frequency Distributions - March Table 16B Historical Comparisons of Wind Frequency Distributions - July Table 16C Historical Comparisons of Wind Frequency Distributions - December Table 17 Comparison of Percent Wind Frequency Distributions - Summer Table 18 Comparison of Percent Wind Frequency Distributions - Winter Table 19 Comparison of Diurnal Resultant Wind Directions Table 20 Indian Point (10M) Wind Speed (MPH) - Summer Season Table 21 Indian Point (10M) Wind Speed (MPH) - Winter Season Table 22 Indian Point (122M) Wind Speed (MPH) - Summer Season Table 23 Indian Point (122M) Wind Speed (MPH) - Winter Season Table 24 Maximum Diurnal Wind Speed (MPH)
Table 25 Annual Summary of Wind Direction Percent Frequency Distribution As A Function of Stability - 10M Level Table 26 Summary of Wind Direction Percent Frequency Distribution As A Function of Stability - Summer Season Table 27 Summary of Wind Direction Percent Frequency Distribution as A Function of Stability - Winter Season Table 28 Historical Comparisons of Percent Occurrence of Stability Table 29 Comparison of Percent Occurrence of Stability on 122 Meter Tower Table 30 Diurnal Variation of Stability Class and Wind Speed (10M)
Table 31 Diurnal Variation of Stability Class and Wind Speed (122M)
Table 32 Diurnal Variation of Stability Class and Wind Speed (Delta-T 400'-200')
Table 33 Comparisons of Average Wind Speeds (MPH) As A Function of Stability Cha pter 2, Page 42 of 119 Revision 20, 2006 OAG10000215_0163
IP2 FSAR UPDATE FACILITY SAFETY ANALYSIS REPORT (FSAR)
CONSOLIDATED EDISON COMPANY OF NEW YORK, INC.
I NOlAN POINT NUCLEAR GENERATING UNIT 2 METEOROLOGICAL UPDATE 1.0 GENERAL 1.1 HISTORICAL BACKGROUND 1.1.1 Introduction Meteorological data were initially collected and evaluated with respect to Indian Point, Buchanan, New York and its environment during the period from 1955 through 1957. This work was accomplished under the direction of Professor Benjamin Davidson of New York University under contract with the Consolidated Edison Company of New York, Inc. (Con Edison). The data and studies during this period were the bases for the Environmental Reports relevant to Indian Point Nuclear Generating Units 1, 2 and 3.
With respect to the Facility Safety Analysis Report (FSAR) for Unit 2, the Environmental Report Supplement, Appendices Volume 1 as Appendices C, 0 and E contains:
- NYU Technical Report 372.1 (November, 1955), B. Davidson
- NYU Technical Report 372.3, Section 2 and 3, (April, 1957), B. Davidson and J.
Halitsky
- NYU Technical Report 372.4 (March, 1958), B. Davidson In 1968 under the direction and supervision of Dr. James Halitsky of New York University, Con Edison contracted to establish experimental meteorological monitoring, data collection and evaluation activities at the Indian Point site and at specified sites in its environment (Halitsky, Laznowand Leahy, February 1970). The original purpose of the above investigation, as noted in the reference, was modified after the studies had begun in order to provide the AEC Construction Hearings for Indian Point Unit 3 with clarification of aspects of the 1956-1957 meteorological data base for the Units 1, 2 and 3 diffusion models. This phase of data collection began in December, 1969.
A report dealing with the results of this latter phase (Halitsky, Kaplin and Laznow, NYU GSL Technical Report No. TR 7103, May 1971) appears as Appendix G in the FSAR Unit 2 Environmental Report Supplement Appendices Volume 1 and as Supplement 1 in the FSAR for Unit 3. The focus of the above report was to validate present site meteorology as representing no significant change in relation to site meteorology from the 1955-1957 period.
Data collection and evaluations continued under this program and a report was submitted by Kaplin and Laznow (1972) representing the data collection period from 1 January, 1970 through 31 December 1971. A copy of this report appears in FSAR for Unit 3 as Supplement 10 (January, 1973).
With respect to Indian Point on-site meteorological measurements, there were for the purpose of the reports that have been cited, three different meteorological towers at three different Cha pter 2, Page 43 of 119 Revision 20, 2006 OAG10000215_0164
IP2 FSAR UPDATE locations. These are specifically delineated in Figure 1 of Halitsky, Kaplin and Laznow {1971}.
The meteorological data collected during the 1955-1957 were from the 300 Foot Tower designated as IP1. Meteorological data collected and reported by Halitsky, Kaplin and Laznow (1971) and Kaplin and Laznow {1972} were from the 100 Foot Tower designated as IP3. The base of IP3 Tower was about 200 feet from the base location of the originallP1 Tower.
Using input data meteorological data from Indian Point Tower (IP3), Bowline Point and the Cape Charles along with sequences of upper air pilot balloon observations Kaplin, Laznow and Wurmbrand (1972) provided Con Edison with input information for the location of the 90-percent probability air monitoring sites, overlay patterns for the prediction of the distribution of gaseous releases and evaluation of the requirements of the AEC Safety Guide 23 {1972}.
All data on the IP3 Tower were collected in accordance with U.S. AEC Safety Guide No. 23, On-Site Meteorological Programs as delineated 2/17/71. The IP3 Tower was maintained from March 1972 through December 1973 by York Research Corporation, Stamford, Connecticut under contract with Con Edison. The last formal report for this tower was prepared and submitted to Con Edison by Kaplin and Kitson (1974) for the 1973 data collection period.
In April of 1973, York Research Corporation under contract with Con Edison began work on the erection of an on-site 400 Foot Meteorological Tower approximately 1725 feet-S and 1750 feet-W of the IP3 Tower. The function of this tower was to develop micro-climatological data suitable for the design of cooling towers and the evaluation of their potential environmental impact on the Indian Point site and its environs. Concurrent studies were conducted to develop three dimensional aspects of the local valley flow using pibal balloons, constant level tetroons and balloon-sondes. In addition, a concurrent study was conducted to develop background levels of ambient air salt concentrations. The results of these studies were submitted in two reports:
Kaplin, Kozenko and Kirshner (1974) and Kaplin, Kitson and Kozenko (1974). This latter report compared meteorological data from the IP3 Tower. At the conclusion of 1973, the primary source of reduced on-site meteorological data were from the IP4, 400 Foot Meteorological Tower. IP3 Tower systems maintenance was continued in accordance with Safety Guide 23 through October 1, 1976 and meteorological data were recorded on analog charts. Data collection was transferred to the IP4 Tower on July 1, 1976. The 400 Foot Tower (IP4) servicing, maintenance and data collection and data selective processing is on-going. Its systems have been updated to meet present requirements of NUREG-0654 Appendix 2 {1980}
and proposed Revision 1 to NRC Safety Guide 1.23 (1980).
The continued maintenance services, etc., of the 400 Foot Meteorological Tower by York Research CorporationlYork Services Corporation was continuous under contract with Con Edison until September 30, 1978, and from October 1, 1978 through the present time under contract with the Power Authority of the State of New York (PASNY).
In September, 1979, York Services Corporation under contract with Con Edison began a study of north to south surface air trajectories analyses and evaluations based on "real-time" wind data available from the Indian Point vicinity. This study incorporated local wind velocity data from the 400 Foot Meteorological Tower at Indian Point, the Orange and Rockland Utilities, 350 Foot Meteorological Tower in Haverstraw, New York, as well as from selected U.S. Department of Commerce, NOAA, Weather Stations. For this study, meteorological data were analyzed and evaluated for the period from August 1, 1978 through July 31, 1979. The final results of this study were presented in reports: Kaplin, E.J. and B. Wuebber, (1979) and Kaplin, E.J. (1979).
As an outgrowth of these studies an expended network of surface (10M) wind velocity Cha pter 2, Page 44 of 119 Revision 20, 2006 OAG10000215_0165
IP2 FSAR UPDATE monitoring stations were sited at key locations along the Hudson River Valley north and south of Indian Point and inland to the east. This network consisted of new anemometer stations in addition to the Indian Point 400 Foot Meteorological Tower, the Bowline 350 Foot Meteorological Tower and the U.S. Department of Commerce, NWS Station at Westchester County. These wind data were digitalized and evaluated by New York Services Corporation under contract to Con Edison for the purposes of defining surface air flow patterns within a 10-15-mile range of Indian Point with emphasis on generating refined estimated of southward movements. As completed, ten consecutive months (March 1, 1980 - December 31, 1980) were evaluated and a total of 7,264 eight-hour parcel trajectories were created objectively using appropriate local one-hour wind velocity averages on a real time basis. The results of this study were submitted to Con Edison: Kaplin, Edward J. and B. Wuebber (1981).
For the purpose of this FSAR, second meteorological update, meteorological data have been analyzed and evaluated from the 400 foot (122 Meter) Indian Point Meteorological Tower for the two year period from January 1, 1979 through December 31, 1980.
1.1.2 Tower Siting and Instrumentation 1.1.2.1 Hudson River There are a number of pertinent facts about the Hudson River itself that are relevant to its ability to induce and/or influence mesoscale flow phenomena that are dominant factors in the Indian Point environs. The most important factor is that it is not a river but, rather a tidal estuary. From New York City, 154-miles north to Troy, there is no drop in the surface elevation of the river.
Except for spring runoff from the Andirondacks, which can smother the tide down to Albany, there is almost imperceptible downstream current.
Since there is no slope to the river surface, it will not support its own gravity flow. Any air movement within its canyons during minimal atmospheric pressure gradient periods can be strictly local cells, which may actually block continuous horizontal air movement over the water surface.
Thermally induced air movement of the Atlantic sea breeze follows the natural path of the river.
It has been noted, however, that lona Island 45-miles north of the tip of Manhattan is considered the point of maximum inland intrusion. The northward movement of sea breeze does not proceed up the Hudson River Valley and Hackensack River Valley at the same speed. The inland movement along the Hackensack Valley lags the Hudson Valley movement. The Hackensack River is on the west side of the Hudson and is specifically delineated because its headwaters are just south of the South Mountains and isolated from the Hudson River by the Hook Mountains and the Palisades. The South Mountains are the east-west extension of the Hook Mountains. The South Mountains abut the Ramapo Range and form a sheer wall from the Southern boundary of the west bank community of Haverstraw.
1.1.2.2 General Topography Each of the reports cited in the Section 1.1.1 describe in some detail general topography in the Indian Point environs. The most recent was provided by Kaplin and Wuebber, 1981. Indian Point is located in the lower Hudson River Valley 27-miles due north of northern boundary of New York City (Manhattan Island).
Cha pter 2, Page 45 of 119 Revision 20, 2006 OAG10000215_0166
IP2 FSAR UPDATE The Indian Point area has been described by Halitsky, e1. aL, 1970, as being located roughly on the axis of a north-south valley enclosed by the Dunderberg and Buckberg Mountains to the west and Blue Mountain and Prickly Pear Hill on the east. The shape of the valley at the 200 foot and 400 foot elevation levels are given in Figures 1 and 2. At the 200 foot contour level the valley width is two miles at Dunderberg Mountain and opens southward to a width of five miles at Prickly Pear Hill.
Cha pter 2, Page 46 of 119 Revision 20, 2006 OAG10000215_0167
IP2 FSAR UPDATE The Hudson River, flowing southward through the valley, resembles a gourd with its curved %-
mile thick northern neck nestling at the base of Dunderberg Mountain while the bulbous three mile thick body fills the southern part of the valley between South Mountain and Prickly Pear Hill. The Indian Point peninsula lies in the hollow of the curved neck.
Beyond its northern end, the valley is split into two branches by Manitou Mountain. The Hudson River passes through the steep, narrow northwest branch between Manitou and Dunderberg Mountains. The northeast branch, between Manitou and Blue Mountains, is about 1.5-miles wide at Manitou Mountain but degenerates with distance into three tributary valleys containing Annsville Creek, Sprout Brook and Peekskill Hollow Brook with sources in the mountainous region north of Peekskill.
South of Haverstraw Bay, the valley opens up rapidly to the southeast while the west bank of the Hudson River follows the blocking of the east-west orientated South Mountains to assume a southward course along the Hook Mountains to the Palisades Mountains.
At elevations higher than 200 feet the solidity of the eastern wall of the valley breaks, first between Blue Mountain and Prickly Pear Hill to form at 300 feet the irregular drainage system, which supplies Furnace Brook, and then at 400 feet into an irregular array of mountain tops.
The western wall is till fairly solid at 300 feet but breaks at 400 feet into two well-defined valleys containing Cedar Pond and Minisceongo Creeks. However, just to the west are Ramapo Mountains whose elevations exceed 1000 feet Figures 3 and 4 (Kaplin and Wuebber, 1981) show the elevations of significant mountain peaks and water courses in the region of Indian Point 1.1.2.3 Site Configuration The location of the specific meteorological towers associated with earlier studies in the Indian Point environs are available in Figure 1 of Kaplin and Laznow (1972), FSAR 3, Supplement 10, 1973 and in Figures 1a and 1d of Kaplin, Kitson and Kozenko (1974).
Table 1 lists the operational periods and the instrumentation associated with each tower. Not included in this listing are those wind monitoring sites that were used for the most recent study (Kaplin and Wuebber 1981, Sec. 2.4 & 2.5). In this study, an 11 site monitoring network was established equipped with Climatronics Mark III, wind speed and direction systems. This network was operable for all or part of the period from March, 1980 through December, 1980.
Chapter 2, Page 47 of 119 Revision 20, 2006 OAG10000215_0168
IP2 FSAR UPDATE TABLE 1 TOWER AND INSTRUMENTATION RECORD (INCLUDES PARAMETERS NOT REQUIRED BY PROPOSED SAFETY GUIDE 1.23)
Meteoro log ica I Base Elevation 0l2erational Period Exposure Station Ft. MSL From To Parameter Instrument M. Above Grade Indian Pt (IP1) 130 1956 1957 Wind Aerovane 91 & 30 Temp. Diff. Honeywell 91-2 & 30-2 USS Jones (J) 0 1956 1957 Wind Aerovane 21 Indian Pt (IP2) 60 1968 1969 Wind Climet 30 Temp. Diff. Bristol 29 - 1.5 Montrose (MP) 60 1968 6/71 Wind Climet 30 Bowline Pt CBP) 5 9/68 11/69 Wind Climet 30 11/69 8/72 Wind Aerovane 30 9/68 11/69 Temp. Diff. Honeywell 30 - 3 11/69 2/72 Temp. Diff. Bristol 30 - 3 2/72 8/72 Temp. Diff. Climet- 30 - 3 (Rosemont)
Bowline Tower 10 Note 2 Present Wind Climatronics 100,50 & 10 Note 2 Present Temp. Diff. Climatronics 100-10 & 50-10 Trap Rock (TR) 90 1969 7/72 Wind Climet 30 USS Cape Charles 0 3/70 9/70 Wind Aerovane 30 (CC)
Indian Pt (IP3) 120 11/69 9/76 Wind Aerovane 30 11/69 9/76 Wind Climet 30 6/73 9/76 Wind Climet 10 Backup Met 12/81 Present Wind Climatronics 10 System (IP3) 11/69 10/71 Temp. Diff. Honeywell 29 - 2 8/72 9/76 Temp. Diff. Climet- 30-3 & 9-3 (Rosemont) 8/72 9/76 Amb. Temp. Climet- 9 (Rosemont) 8/72 9/76 Dew Point Climet- 30,9 & 3 (Foxboro) 8/72 9/74 Net Radiation Teledyne 9 Geotech 5/70 12/70 Turbulence Bivane* 30 Note 2 Bowline Tower is located at approximately Latitude 41 0 13'N and Longitude 730 58' W. This location is about 3000 feet NW of the earlier Bowline Point Tower. It began operation in the 1972/73 time period.
- Intermittent usage.
Cha pter 2, Page 48 of 119 Revision 20, 2006 OAG10000215_0169
IP2 FSAR UPDATE TABLE 1 (Cont'd)
Meteorological Base Elevation OQerational Period Exposure Station Ft. MSL From To Parameter Instrument M. Above Grade Indian Pt (IP4) 117 9/73 Present Wind Climatronics 122,38 & 10 122 M. Tower 9/73 6/80 Wind Climatronics 85 - Note 1*
6/80 Present Wind Climatronics 60 9/73 9/79 Amb. Temp. EG&G 10 9/79 Present Amb. Temp. C1imatronics 10 9/73 9/79 Dew Pt. EG&G 122,61 & 10 9/79 Present Dew Pt. C1imatronics 10 9/73 9/79 Temp. Diff. EG&G 122-10 & 61-10 9/79 Present Temp. Diff. C1imatronics 122-10 & 60-10 1/74 Present Net Rad. Teledyne 10 Geotech 7/80 Present Precipitation Climatronics 1 Indian Pt (IP4) ** 117 9/73 7/77 Visual Range EG &G FSM 10 10M Tower Emergency 135 7/24/80 11/81 Wind C1imatronics 11.8 Control Center #
Note 1* - 85 meter wind speed and wind direction moved from the 85 meter level to the 60 meter level as required by proposed Revision 1 to NRC Safety Guide 1.23.
- Tower and System Removed 07/22/80.
- Tower and System Removed 07/22/80.
Cha pter 2, Page 49 of 119 Revision 20, 2006 OAG10000215_0170
IP2 FSAR UPDATE 2.0 122M METEOROLOGICAL TOWER 2.1.1 Siting The relative locations of the existing meteorological towers in historic perspective are shown in Figure 5. Specific details of site location are shown in Figure 6.
2.1.2 Instrumentation 2.1.2.1 Sensor Configuration The sensor configuration and exposure on the existing operational 122M Meteorological Tower are shown in Figures 7 and 8.
2.1.2.2 Instrumentation Specifications The following specifications apply to specific operational sensors that are a part of the total meteorological support systems at Indian Point.
2.1.2.2.1 Climatronics F460 Wind Speed Transmitter Accuracy: 0.07 MIS or 1%
Range: 0-56 MIS Threshold: 0.22 MIS Distance Constant: 1.5 M 2.1.2.2.2 Climatronics F460 Wind Direction Transmitter Accuracy: +/-2° Range: 0-540° Threshold: 0.22 MIS Distance Constant: 1.5 M Damping Ratio: 0.4 at 10° initial angle of attack 2.1.2.2.3 Climatronics TS-l0 and TS-l0WA Motor Aspirated Shields Shield Effectiveness: Under radiation intensities of 110 W/m 3 (1.6 cal/cm 2/min) radiation error not exceeding O.l°C Aspiration Rate: 3 MIS at sensor location 2.1.2.2.4 Climatronics 100087, 100087 Temperature-Delta Temperature (matched thermistor)
Temperature:
Range: -34 to +500 C Accuracy: +/- 0.2°C Time Constant: 10 sec. To 63% (in TS-l0 Shield)
Linearity: +/-0.2%
Cha pter 2, Page 50 of 119 Revision 20, 2006 OAG10000215_0171
IP2 FSAR UPDATE Delta Temperature:
Range: +/- 10°F Sensitivity: 0.02°F Accuracy: O.l°F or +/- 5% of delta-T not to exceed O.3°F Response Time: 10 sec. To 63% in TS-I0 Shield 2.1.2.2.5 Climatronics DP-I0 Dew Probe (YSI Lithium Chloride)
Range: -40° to 42°C Accuracy: +/- 0.5°C Response Time: 1°C/min.
2.1.2.2.6 Climatronics 1000971 - Heated Rain - Snow Gauge (Tipping Bucket)
Accuracy: +/- 1% up to 3"/hr Resolution: 0.01" Size: 8" diameter x 24" height Conversion Accuracy: +/-0.2%
2.1.2.2.7 Data Collection Systems Analog:
Wind Systems: Esterline-Angus Model Ell02R - Rectigraph Recorders Temperature, Dew POint, Delta Temperature: Tracer Westronics Model M l1E, Multipoint Precipitation: Esterline-Angus Model MS 401C Digital:
Climatronics Data Processor: lMP/801 Tape Collection Interface: Tandeberg TDl 10-50 2.1.3 Meteorological Support System The meteorological systems at Indian Point are equipped, maintained and operated in compliance with the specification of NUREG-0654, Appendix 2 (1980); Proposed Revision 1 to NRC Regulatory Guide 1.23 (1980); and applicable regulatory requirements. The total system as presently operated is outlined in Figure 9.
2.2 DATA LOG 2.2.1 Indian Point Tower IP3 Meteorological data from the IP3 Tower were reduced and evaluated (Kaplin and Kitson, 1974) through December 1973. from the period 1974 through September 1976 the tower system was maintained, as previously noted. Analog records were provided to Con Edison for storage. Tower removed from service in September, 1976. (Reactivated as site for Backup Wind System: 12/01/81.)
2.2.2 122 Meter Meteorological Tower (IP4)
Chapter 2, Page 51 of 119 Revision 20, 2006 OAG10000215_0172
IP2 FSAR UPDATE The data log for the 122 Meter Meteorological Tower for the period from October 1973 through August 1974 can be found in Kaplin, et. aL, 1974. Subsequent to the completion of the above report and the data contained therein, the meteorological analog charts and the data collection magnetic tape were documented and transmitted to Con Edison for storage.
Commencing in August 1977, wind velocity data at the 10 and 122 meter levels and the delta temperatures: 50-10M and 122-10M were reduced to hourly averages and transmitted to Con Edison in addition to the analog charts. The summary of valid data for these parameters for the period from August 1977 - July 1981 is shown in Table 2 on concurrent and total hours basis. The concurrent basis assumes that if any parameter is missing. The total basis relates an individual missing data hour to the total number of possible data hours in a month.
On the concurrent basis, the average valid data collection was 92.4 +/- 10.8-percent. On a total hour basis, the average valid data collection was 98.2 +/- 2.5.
Cha pter 2, Page 52 of 119 Revision 20, 2005 OAG10000215_0173
IP2 FSAR UPDATE TABLE 2 VALID DATA LOG*
1977 1978 1979 1980 1981 Month Concurrent Total Concurrent Total Concurrent Total Concurrent Total Concurrent Total January N/A N/A 89.8 98.2 94.8 99.0 98.4 99.7 99.9 99.9 February N/A N/A 92.1 97.4 98.1 99.7 90.5 98.4 67.7 89.2 March N/A N/A 95.1 98.9 97.4 99.6 97.3 99.1 98.0 99.3 April N/A N/A 98.1 99.6 96.7 99.4 100.0 100.0 100.0 100.0 May N/A N/A 95.3 98.6 90.3 98.0 88.4 98.0 100.0 100.0 June N/A N/A 86.8 95.6 95.0 99.1 96.7 99.0 100.0 100.0 July N/A N/A 94.1 96.8 92.6 98.8 71.1 94.1 99.4 99.8 August 94.2 98.7 95.3 99.2 52.3 92.0 77.7 92.7 September 84.7 94.4 99.7 99.9 57.5 92.3 98.3 99.7 October 95.0 98.9 98.1 99.4 94.0 98.2 96.1 99.4 November 98.9 99.5 98.8 99.6 98.3 99.7 99.6 99.9 December 75.3 95.5 96.4 99.4 100.0 100.0 98.9 99.8 Concurrent Average: 92.4 +/- 10.8%
Total Parameter Hours: 98.2 +/- 2.5%
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- Based on six (6) parameters: wind data at 10 and 122M and delta-temperatures: 60-lOM and 122*10M o NA - Values not available.
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Cha pter 2, Page 53 of 119 1U'l o Revision 20, 2006
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IP2 FSAR UPDATE 3.0 ANALYSES DATA 3.1 INDIAN POINT TOWER IP3 A FSAR documented study with respect to the Indian Point IP3 Meteorological Tower was prepared by Kaplin and Laznow (1972) [Referenced FSAR 3 Supplement 10, 1973]. This report covered the data collection period through 1971. Additional data were subsequently provided through 1972 to provide composite joint wind velocity frequency distribution for Pasquill Stability Categories (FSAR 3 Supplement 13 and 16, 1973). Kaplin and Kitson (1974) provided an analyses of IP3 for the period March 1973 through December, 1973.
This report confirmed the earlier study that wind data in the Indian Point environs based on monthly diurnal wind distributions, wind frequency distributions and joint wind stability categories are comprised of two "seasons" with little apparent transition.
The "winter season" reflects little or no average diurnal variation in the hourly resultant winds, dominant winds from the west to north. The "summer season" is characterized by dominant north-northeast winds during the evening and early morning hours with a sharp transition to south to southwest winds during the day and another transition in late afternoon to the evening pattern.
The wind frequency distributions and joint frequencies as a function of Pasquill Stability Categories were comparable in 1973 with data collected in 1970 and 1971.
It is noted that the temperature gradients on the IP3 Tower were derived from delta-temperatures: 99-7 feet and wind measured at 105 feet above a grade elevation of 120 feet MSL.
Kaplin, et. aL, (1974) compared three months (October - December, 1973) of IP3 wind data as measured at 105 feet above grade with concurrent three months of data from the 125 foot level on the 122 Meter Meteorological Tower (IP4) (grade elevation: approximately 117 feet MSL) using a two station wind correlation program (Appendix B, Kaplin, et. ai, 1974).
Figures lOA and lOB show the relationships obtained for October and December, 1973. The November results were similar with the directional relationships falling between that obtained for October and December. The maximum variations between the two sites occurred with winds from E-ESE and SW to WSW. These corresponded to sectors of minimum average wind speeds. Deviations between the two sites can be attributed to local factors including terrain elevation, land use and ground cover.
o The wind direction displacement effects found in the two station correlations were confirmed in the monthly diurnal analyses.
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Cha pter 2, Page 54 of 119 1U'l o....Jo. Revision 20, 2006
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IP2 FSAR UPDATE 3.2.1 October 1, 1973 to August 31. 1974 The purpose of the 122M Meteorological Tower at Indian Point (IP4) was to develop a three dimensional micro-climatological data file to be used to assess the impact of proposed cooling towers and to provide the basis for design criteria as required.
The results of one year of operation of this tower were presented in a final report (with Appendices) by Kaplin, et. ai., 1974.
Except as noted in the previous section, the meteorological data collected and evaluated were not compared at the time of this study with historical meteorological data associated with FSAR 2.
ThiS study determined that the two distinctive seasonal patterns existed at each of the four levels of wind velocity measurement: 10M, 38M, 85M and 122M. Wind directions tended to back with elevation assuming an orientation parallel to general terrain contours.
In times of weak synoptic pressure gradient patterns, there were abrupt transitions in the diurnal flow patterns consistent with valley flow winds particularly during the summer season. These transitions began at the surface and progressed up to the 122M level.
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Chapter 2, Page 55 of 119 1U'l o Revision 20, 2006
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IP2 FSAR UPDATE TABLE 3 COMPARISON OF ANNUAL PERCENT OCCURRENCE OF STABILITY CATEGORIES Stabilit~ Cate90r~
Year Tower Gradient eM) 8- ~ k D g E G 1970 IP3 29 - 3 21.68 2.20 3.39 33.35 24.75 9.01 5.62 1971 IP3 29 - 3 19.17 2.75 2.97 22.74 30.87 11.69 9.75 1970-72 IP3 29 - 3 16.25 1.82 2.95 29.71 26.61 1327 9A5 1970-72* IP3 29 - 3 6.76 2.67 2.13 32.65 40.57 11.78 3.31 1973 IP3 29 - 3 23.14 3.16 3.70 20.87 25.02 13.89 10.23 1973-74 122M-IP4 60 - 10 10.35 3.21 2.94 25.38 44.86 11.35 1.91 1979 122M-IP4 60 - 10 12.27 3.25 3.86 29.30 40.39 8.83 1.31 1980 122M-IP4 60 - 10 13.32 4.06 4.60 29.81 33.97 11.34 2.07 1979-80 122M-IP4 60 - 10 12.80 3.66 4.23 29.56 37.17 10.08 1.69 o
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- Temperature difference corrected by a factor of 0.605; (FSAR 3, Supplement 16, April 1973) o o
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Cha pter 2, Page 56 of 119 1U'l o Revision 20, 2006
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IP2 FSAR UPDATE The morning transition during the summer season was sharply defined. At 0800 EST all levels had approximately the same resultant wind direction. The evening transition began just after 1800 EST and was sharply defined a the 10M and 38M levels. The 10M level reached its nocturnal northeast drainage wind by 2100 EST along with the somewhat more erradic 38M level. The 85M and 122M levels rotated systematically and did not reach their nocturnal directions (NNE-N) until 0200 EST. The systematic rotations are referenced to the average diurnal distributions on a real time basis, the upper wind levels could be "disconnected" from the lower wind levels with an intermediate shear zone generated by winds up to 1800 out of phase.
The resultant winds of the 122M Tower (IP4) associated with the diurnal variation curves for the summer season, veered after the morning transition until about noontime, then steadied out at SW-WSW before backing into the nocturnal pattern after the evening transition. The resultant summer season winds for the IP3 Tower (Kaplin and Laznow, 1972) in 1970 and 1971 veered throughout the entire day. In Kaplin and Kitson (1974), the summer IP3 diurnal resultant winds exhibited the veering - backing trait. Kaplin and Laznow (1972) indicated that the question of backing or veering was related, on any given day, not only to strength of the valley drainage flow wind but also to relative strength of local land-sea circulations.
On an annual basis there were no significant differences between the percent frequency distribution of occurrences of stability categories (Pasquill) between the adjusted composite year 1970-1972 for IP3 (FSAR 3, Supplement 16, April 1973) and lower temperature gradient level on the 122M Tower (IP4). These comparisons are shown in Table 3. It is presumed that if all individual years of IP3 data were Similarly adjusted prior to classification, they would also be reasonably comparable to results based on the temperature gradient 60-10M on the 122M Tower.
3.2.2 August 1. 1978 - July 31. 1979 Wind veloCity data (10M level) from the 122M Meterological Tower at Indian Point and the Orange and Rockland Utilities, Inc. 100M Meteorological Tower were used to evaluate the path of air parcels in the Indian Point environs without considering stability (Kaplin and Wuebber, 1979 and Kaplin, 1979).
Each hour a parcel movement was initialized from Indian Point. Each parcel was projected forward for eight consecutive hours in hourly increments. The average wind velocity at 10M level of the 122M Tower was used to determine the speed and direction of the parcel for its initial hour increment. Subsequent movement of each parcel was determined by the location of the parcel after the initial hour on a zone of influence file that assigned a wind vector to that location: Indian Point or Bowline.
Prior to usage the wind velocity for selected 1978-1979 data were assessed by comparison with historical data files (1973-1974) at Indian Point. There were no variations that could not be accounted for by climatological variations of at least synoptic scale when assessed with reference to U.S. Department of Commerce, NOAA, EDS, LOCAL CLIMATOLOGICAL DATA for LA GUARDIA AIRPORT, NEW YORK and SIKORSKY AIRPORT, Bridgeport, CON NECTICUT.
In considering persistent southward movement of an air parcel from Indian Point assuming that Bowline would be representative of air movement south of Grassy Point, an examination of resultant winds for August 1978 and January, 1979 (typical "summer" and Cha pter 2, Page 57 of 119 Revision 20, 2006 OAG10000215_0178
IP2 FSAR UPDATE "winter" seasons) indicated that such movements did not occur. While concurrent hourly average north winds were found 14 times in August and 17 times in January, these occurrences represent only 13.3-percent and 16.S-percent of all north winds relative to all north winds at Bowline. These results were anticipated, particularly during periods of light winds and weak synoptic pressure fields, from the opposing patterns of the diurnal variation curves for the two monitoring sites.
3.2.2.1 Surface Air Trajectories Analyses - Summary Trajectory end points were derived on an objective basis using surface wind data from monitoring stations. The use of observed wind data appropriate to the moving air pa rcel's location at a given time is important since these data inherently account for local wind pattern aberrations that may be of topographic and/or unique micro-meteorological origin.
No individual atmospheriC stability category was explicitly considered.
The ability of the derived trajectories to generate realistic movement patterns is contingent on having sufficient wind monitoring sites to define the actual wind flow field in and around the area of interest on a concurrent real time basis.
The area of interest was limited to ten miles south of Indian Point. For practical purposes the study area was 21 x 21 square miles subdivided in a one mile grid as shown in Figure
- 11. Indian Point was located near the top center of the area at grid point 10, 16. This allowed for lS-miles of due south movement. The South, Hi Tor and Hook Mountains are emphasized because of the barrier that they form for air movement due south of Indian Point.
Trajectories were generated for each of the 12 months in the data file based on Indian Point and Bowline wind data. These were the only available data applicable to the study area.
Trajectories were created for up to eight consecutive hours of movement.
For the first hour of the trajectory, Indian Point was used as the origin of an air parcel, which would travel a distance and in a direction determined by the hourly average wind velocity at the 33' (10M) level of the Meteorological Tower. Subsequent movement depended on the location of the trajectory end point after the first hour's movement. Zones for which the Bowline and Indian Point wind velocity measurements were considered as representative had been previously assigned. Trajectories for each hour of the month were computed. The end points were accumulated as summations of occurrences in their appropriate grid squares. In the process of generation, all end pOints were moved and accumulated whether or not they were in the 21 x 21-mile square. Only those end pOints within the study area boundary appear on tabular printouts. In any given period, an end point could pass out of the grid and move back in at a subsequent time interval.
For August, 1978, and January, 1979, two different patterns of weather station representative areas were used as shown in Figure 12. Pattern 1, which was used for all 12 months of data, had Bowline winds dominating after passage of a line three miles south of Indian Point (through Grassy Paint). Pattern 2, used for August and January only, moved this line one mile further north (through Stony Paint). The influence patterns are the same for the first hour's movement.
A summary of the August, 1978 and January, 1979, results in terms of percent of total possible trajectory end pOints remaining in the 21 x 21-mile area for selected trajectory time periods is shown in Table 4.
Cha pter 2, Page 58 of 119 Revision 20, 2006 OAG10000215_0179
IP2 FSAR UPDATE TABLE 4
SUMMARY
OF TRAJECTORY END-POINTS August! 1978 January, 1979 Pattern I No. of Occur. % Total No. of Occur. % Total Hour 1 722 (729) 99.0 707 (721) 98.1 Hour 2 617 (728) 84.8 486 (720) 67.5 Hour 4 420 (726) 57.9 225 (718) 31.3 Hour 6 312 (724) 43.1 157 (716) 21.9 Hour 8 206 (722) 28.5 113 (714) 15.8 Pattern II Hour 1 722 (729) 99.0 707 (721) 98.1 Hour 2 595 (728) 81.7 486 (720) 67.5 Hour 4 414 (726) 57.0 226 (718) 31.5 Hour 6 308 (724) 42.5 157 (716) 21.9 Hour 8 207 (722) 28.7 115 (714) 16.1
( ) =Tota I number of trajectories generated.
The actual number of pOints within the grid network does not differentiate between those pOints that have never left the network and those that have recirculated. This feature takes on added importance if total distance of travel is a consideration.
Summaries of occurrences within designated grid sectors are shown in Table 5 for August, 1978 and Table 6 for January, 1979. In terms of totals in the grid area, there is no significant effect of influence pattern assignment. This effect does show up in Tables 5 and 6 when the occurrences south of Indian Point are totaled. For this purpose, a SW sector is defined encompassing the area below Indian Point from the grid edge to ordinate Line 9.
The S sector extends one mile south of Indian Point along ordinate Cha pter 2, Page 59 of 119 Revision 20, 2006 OAG10000215_0180
IP2 FSAR UPDATE TABLE 5 SUMMATION OF TRAJECTORY END POINTS AUGUST, 1978 SECTOR KEY 17-21 NW N NE 16 W LP. E 1-15 SW S SE 1-9 10 11-21
% TOTAL 35 164 133 46.0 HR 1 26 8 15 6.8 232 84 25 47.2
% TOTAL 40.6 35.4 24.0 99.0 PATTERN 1 PATTERN 2
% TOTAL % TOTAL 52 93 109 41.2 52 93 109 42.7 HR2 20 10 7 6.0 18 10 7 5.9
-ill -21 ~ 52.8 -ill -.2Q 108 51.4
% TOTAL: 42.8 25.0 32.2 84.8 % TOTAL 36.6 25.7 37.6 81.7
% TOTAL % TOTAL 43 38 40 28.8 44 38 36 28.5 HR4 7 2 2 2.6 7 1 2 24.4
--1£§. -.1Z --.ldJ. 68.6 -1ll ~ 146 69.1
% TOTAL: 42.3 16.0 41.7 57.9 % TOTAL 39.1 16.4 44.4 57.0
% TOTAL % TOTAL 33 19 18 22.4 30 19 19 22.1 HR6 7 0 1 2.6 6 0 1 2.3
.lQ.Q ---.lQ ~ 75.0 ----.Jlli ----.l.Z 130 75.6
% TOTAL: 44.9 9.3 45.8 43.1 % TOTAL 39.6 11.7 48.7 42.5
% TOTAL % TOTAL 28 6 4 18.4 28 6 4 18.4 HR8 3 0 1 1.9 3 1 1 2.4
~ _7 ----.2Q 79.6 ~ ----.1.2 ----.2§. 79.2
% TOTAL: 47.6 6.3 46.1 28.5 % TOTAL 40.6 10.6 48.8 28.7 Cha pter 2, Page 60 of 119 Revision 20, 2006 OAG10000215_0181
IP2 FSAR UPDATE TABLE 6 SUMMATION OF TRAJECTORY END POINTS JANUARY, 1979 SECTOR KEY 17-21 NW N NE 16 W LP. E 1-15 SW S SE 1-9 10 11-21
% TOTAL 25 52 86 23.1 HR 1 10 10 27 6.7 154 72 271 70.3
% TOTAL 26.7 19.0 54.3 98.1 PATTERN 1 PATTERN 2
% TOTAL % TOTAL 22 21 58 20.8 24 21 56 20.8 HR2 14 5 16 7.2 12 5 16 6.8
-ill ~ --.J2§. 72.0 -1QQ ~ 207 72.4
% TOTAL: 34.8 13.8 51.4 67.5 % TOTAL 28.0 14.6 57.4 67.5
% TOTAL % TOTAL 17 4 19 17.7 14 4 21 17.3 HR4 7 0 3 4.4 7 0 3 4.4
---2Z -1§. ~ 77.8 -2Z -.l1 106 78.3
% TOTAL: 40.4 8.9 50.7 31.3 % TOTAL 34.5 8.0 57.4 31.5
% TOTAL % TOTAL 16 2 7 15.9 15 2 8 15.9 HR6 2 1 0 3.2 2 1 0 1.9
~ _8 ----.n. 80.9 ~ __ 8 ~ 82.1
% TOTAL: 40.8 7.0 51.0 21.9 % TOTAL 35.7 7.0 57.3 21.9
% TOTAL % TOTAL 9 1 4 12.4 6 0 4 8.7 HR8 3 1 0 3.5 3 1 0 3.5
~ ---..lQ ----.2.1 84.1 ~ __ 9 ~ 87.8
% TOTAL: 40.7 10.6 48.7 15.8 % TOTAL 33.0 8.7 58.3 16.1 Chapter 2, Page 61 of 119 Revision 20, 2006 OAG10000215_0182
IP2 FSAR UPDATE Line 10 to the grid base. The SE sector comprises the remaining area to the east of the S line and below Indian Point. These results are summarized below in terms of number of occurrences and percentage of total possible observations:
TABLE 7 SUMMATION TRAJECTORY OCCURRENCES SOUTH OF IN DIAN POINT August, 1978 Pattern 1 Pattern 2 Southwest South Southeast Southwest South Southeast Occur % Occur 0/0 Occur %
Occur %
Occur 0/0 Occur a/a Hour 1 232 31.8 84 11.5 25 3.4 232 31.8 84 11.5 25 3.4 Hour 2 192 26.4 51 7.0 83 11.4 148 20.3 50 6.9 108 14.8 Hour 4 128 17.6 27 3.7 133 18.3 111 15.3 29 4.0 146 20.1 Hour 6 100 13.8 10 1.4 124 17.1 86 11.9 17 2.3 130 18.0 Hour 8 67 9.3 7 1.0 90 12.5 53 7.3 15 2.1 96 13.3 JANUARY, 1979 Hour 1 154 21.4 72 10.0 271 37.6 154 21.4 72 10.0 271 37.6 Hour 2 133 18.5 41 5.7 176 24.4 100 13.9 45 6.3 207 28.8 Hour 4 67 9.3 16 2.2 92 12.8 57 7.9 14 1.9 106 14.8 Hour 6 46 6.4 8 1.1 73 10.2 39 5.4 8 1.1 82 11.5 Hour 8 34 4.8 10 1.4 51 7.1 29 4.1 9 1.3 63 8.8 The effect of the pattern change is not so much as to alter the total; rather, it is to shift the number of occurrences from the SW sector to the Sand SE sectors. There are anomalies found that may be associated with recirculation.
After five miles of southward movement from Indian Point, the results seem to indicate the anomaly of surface wind impaction against the South Mountain and High Tor Ridges. This anomaly occurred since there were no local wind measurements available to induce deflections.
Historical studies have shown such deflections do exist. The present results cannot account for terrain unless the trajectory paths are deflected by observed surface winds. This requires a larger monitoring network, strategically placed, than was available. This need for further definition of local wind field is confirmed by the differences that appear in the results generated by Patterns 1 and 2.
At the present time, based on historical studies, Pattern 2 is probably the better representation of local trajectories for the available data.
Assuming a continuous 1 MIS wind speed (2.2 MPH), the number of occurrences in the south sector represent those parcels that have traveled with the effective speed (neglecting recirculation). Of the totals given, only four have traveled greater than ten miles for August (1); five for August (2); six for January (2); six of January (1); and three for January (2).
These pOints would have to had passed through or over the South, High Tor and Hook Mountain Ridge lines.
Cha pter 2, Page 62 of 119 Revision 20, 2006 OAG10000215_0183
IP2 FSAR UPDATE 3.2.3 March 1980 - December 1980 3.2.3.1 General The results of the Trajectory II Study conducted by York Services Corporation for Con Edison have been recently submitted (Kaplin and Wuebber, 1981).
It was concluded from the initial trajectory study (Kaplin and Wuebber, 1980; Kaplin, 1980) that a lack of directional persistence of low speed surface winds (10M) at Indian Point and Bowline make recirculation of local air probable. There were indications of both convergence and divergence of local air streams. Objectivity created surface air parcel trajectories generated anomalies by passing over or through abrupt terra in features. The two local monitoring sites available were unable to resolve these anomalies.
In the Trajectory II Study, a supplemental network of ten surface wind monitoring stations were established for the express purpose of objectively assessing the southward movement of air parcels from Indian Point (see Figures 3 and 4). Sites were selected, specifically, in an attempt to resolve anomalous flow patterns with respect to terrain and tributary river drainage basins. A listing of sites used is shown in Table 8. A listing of valid data collected for the period is shown in Table 9.
3.2.3.2 Wind Frequency Distributions An historical evaluation of the representativeness of the data collected in 1980 was made for Indian Point and Bowline. Variations in wind frequency distributions were found to be associated with climatological variations on the synoptic-cyclonic scale.
These variations can be naturally expected between any given year or set of years. For example: Over a 20 year period (1960-1979) prior to 1980 Bridgeport, WBAS, for the month of July had an average wind directional frequency for a north wind of 6.5 +/- 2.7-percent with an absolute maximum of 12.9-percent in 1974 and an absolute minimum of 3.0-percent in 1979. In 1980[ the frequency set a new low of 2.8-percent. A further extreme example was found at La Guardia WBAS. Based on an eight year average (1972-1979) the northwest wind has a frequency of 12.2 +/- 4.2-percent with minimum of 6.1-percent (1973) and a maximum of 16.5 (1974). In 1980, a new maximum of 18.2-percent was observed while in 1979, the frequency was 7.9-percent, which was the second lowest value in the period.
It was noted that the climatic variations of wind frequencies at Indian Point were generally minimal and less pronounced. This was attributed to topographic confinement. The wind frequency data for Indian Point and Bowline for the data collection period were adjudged to be representative (Figures 6.1-6.4, Kaplin and Wuebber, 1981). It was assumed that all concurrently collected wind velocities were representative of respective monitoring sites and relationships between sites could be evaluated.
It was found that wind frequency distribution patterns in themselves were deceptive representations of the continuity of air movement in the lower Hudson River valley unless there was an understanding of the patterns of wind velocity variations on a temporal basis.
Cha pter 2, Page 63 of 119 Revision 20, 2006 OAG10000215_0184
IP2 FSAR UPDATE TABLE 8 LOCATIONS OF STATIONS RELATIVE TO INDIAN POINT Distance (miles) Direction Station from Indian Point (degrees)
Iona Island 2.50 334 Annsville 2.20 020 Watch Hill Road 3.15 132 Jurka 6.65 122 Croton Point 6.40 155 Ossining 8.80 151 Grassy Point 3.20 191 Bowline Point 4.15 190 South Nyack 13.60 174 Piermont 15.95 173 Kingsland 13.00 163 Eastview 11.90 145 Westchester County Airport 20.00 135 Cha pter 2, Page 64 of 119 Revision 20, 2006 OAG10000215_0185
IP2 FSAR UPDATE TABLE 9 YORK SERVICES CORPORATION ONE RESEARCH DRIVE, STAMFORD, CT CLIENT: CONSOLIDATED EDISON OF NEW YORK VALID DATA FOR TRAJECTORY WIND SITES PERIOD OF RECORD: 1980 SITE PERCENT VALID DATA MARCH APRIL MAY JUNE JULY AUGUST SEPT OCT NOV DEC 01-Piermont 79.23 78.75 63.71 66.67 100.00 100.00 71.25 37.10 0.00 0.00 02-0ssining 98.59 100.00 100.00 100.00 100.00 99.80 99.93 96.77 100.00 100.00 03-lona Island 96.77 53.75 85.28 69.72 53.76 95.63 100.00 99.46 100.00 82.80 04-Jurka/Grassy 78.76 55.97 0.00 0.00 0.00 0.00 0.00 81.85 100.00 84.95 OS-Kingsland 87.57 94.65 17.47 59.79 98.66 100.00 100.00 100.00 74.03 93.55 06-Watch Hill 74.46 0.00 0.00 0.00 48.79 70.56 100.00 85.48 100.00 100.00 07-South Nyack 79.17 100.00 83.87 100.00 100.00 100.00 100.00 100.00 100.00 100.00 08-Annsville 100.00 100.00 90.86 100.00 77.28 100.00 71.81 98.92 100.00 100.00 09-Eastview 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 100.00 10-Croton Point 99.73 100.00 89.85 100.00 98.12 96.77 81.39 47.04 100.00 36.96 U-West Cty Apt 99.87 100.00 100.00 99.86 100.00 100.00 100.00 100.00 100.00 100.00 12-Indian Point 99.66 100.00 94.82 99.24 95.41 94.69 99.17 99.93 100.00 99.46 13-Bowline Point 72.45 99.72 98.91 85.97 92.41 98.52 96.53 96.98 96.46 86.16 0
>>
G) 0 0
0 0
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(J1 Cha pter 2, Page 65 of 119 I Revision 20, 2006 0
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IP2 FSAR UPDATE 3.2.3.3 Diurnal Wind Distributions The diurnal va riation curves for Indian Point and Bowline for the data collection period (March through December 1980) were found to be historically representative. For selected winter and summer months, they demonstrated all the attributes of the two "season" characteristics (Figure 6.21-6.27, Kaplin and Wuebber, 1981).
With some variation at selected monitoring sites, it was found that the diurnal wind distributions were not only seasonally characteristic but characteristic of the monitoring site locations. They could, almost without exception, be uniquely categorized as Hudson River "west bank"; Hudson River "east bank"; or "inland" (Figures 6.28-6.33, Kaplin and Wuebber, 1981).
The characteristics of this uniqueness were examined by combining all appropriate sites to generate "average" east and west bank diurnal wind distributions (Figures 6-38-6.42, Kaplin and Wuebber, 1981). The average diurnal curves for March, June and December, 1980 are shown in Figures 13, 14, and 15. A computer check revealed that individual days at any given site could be found that had observed 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> diurnal wind variation patterns that matched their own monthly average distributions andlor the appropriate east or west bank average diurnal distribution based on the criteria 16 or more hours of fit +/- 45° (not necessarily consecutive).
While there are unique common characteristics to the diurnal wind distribution patterns in the Indian Point environs, variations in local meso-scale factors dictate that the ultimate path of an air parcel whose movement is determined by surface (10M level) wind velocities is governed by time of departure as well as point of departure. Between wind velOCity monitoring sites in the region, perSistent wind direction and wind speeds are not supported.
This is most obvious during the "summer" season or at any time that the area is under the influence of a weak synoptic-cyclonic pressure gradient pattern. Between individual monitoring sites there is apparent divergence and convergence of surface air.
3.2.3.4 Resultant and Concurrent Hourly Winds A first approach at the evaluation of southward movement of air for prolonged periods of time was made for the data collection period March 1, 1980 through December 31, 1980, by examining the frequency distribution of the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> resultant winds (Kaplin and Wuebber, 1981). These results are shown in Table 10 as a function of persistence category (the ratio of the resultant to arithmetic average wind speeds). At perSistence levels greater than 0.9, a north wind was found in only four out of 273 possible valid cases (1.4%). The average wind speed 2.75 MIS. The high wind speeds associated with all northerly winds implies strong synoptic-cyclonic scale pressure gradient systems are the generating mechanism.
Simple liner relationships between high persistent, 24-hour resultant winds between Indian Point and Piermont (bearing 173° about 16-miles from Indian POint) showed that an average 24-hour resultant wind direction of 012 +/- 22° at Indian Point was related to an average 24-hour resultant wind at Piermont of 359 +/- 24°. At the same time, for corresponding cases, the average resultant wind speed at Indian Point was about 2.5 +/- 0.9 MIS and the concurrent average resultant wind speed at Piermont was 5.6 +/- 1.2 MIS. The angular offset implies terrain tracking and high average resultant wind speeds indicate the necessity for a strong pressure gradient field.
Cha pter 2, Page 66 of 119 Revision 20, 2006 OAG10000215_0187
IP2 FSAR UPDATE Such replacements were also implied when concurrent hourly average wind data from the selected monitoring sites were correlated to Piermont. These results a re shown in Table 11 for the available concurrent data collected during the period from March 1, 1980 through October 31, 1980. Out of 4,394 valid data hours, there were only 56 (1.3%) in which Indian Point a nd Piermont had concurrent winds from the north (350-011°). Almost half of these cases (26:0.59%) occurred in May, 1980. There was only one such hour out of 742 valid data hours in July, 1980. For July, in fact, for 3,355 concurrent data hours from five southern sites there was only one additional hour in which a site, South Nyack, had a north wind direction concurrent with Piermont.
Chapter 2, Page 67 of 119 Revision 20, 2006 OAG10000215_0188