ML14339A653
| ML14339A653 | |
| Person / Time | |
|---|---|
| Site: | Kewaunee  |
| Issue date: | 11/24/2014 |
| From: | Dominion Energy Kewaunee |
| To: | Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14339A626 | List: |
| References | |
| 14-572 | |
| Download: ML14339A653 (68) | |
Text
Revision 25-11/26/14KPS USARB-iAppendix BSpecial Design Procedures Revision 25-11/26/14KPS USARB-iiIntentionally Blank Revision 25-11/26/14KPS USARB-iiiB.1Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1B.2Classification of Structures and Components. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1B.2.1Definition of Nuclear Safety Design Classifications (NSDC) . . . . . . . . . . . . B-2B.3Design Codes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-10B.4Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-11B.4.1Environmental Loads. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12B.4.2Tornado Loads. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12B.4.3Live Loads. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12 B.4.4Dead Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12 B.4.5Seismic Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-13 B.4.6Design Basis Accident (DBA) Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-13B.4.7Other Loads. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-13 B.4.8Seismic Design and Verification of Modified, New and Replacement Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-13B.5Protection of Class I Items. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-14B.6Design Criteria for Structures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-15B.6.1Load Combinations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-15 B.6.2Stress Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-17B.6.3Structural Design Basis. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-17B.7Design Criteria for Components. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-33B.7.1Load Combinations. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-33 B.7.2Design Criteria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-34B.8Protection Against Crane Toppling and Control of Heavy Loads. . . . . . . . . . . . . . . B-50B.8.1Protection Against Crane Toppling. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-50B.8.2Control of Heavy Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-50B.8.3Design Criteria for Upgraded Auxiliary Building Crane . . . . . . . . . . . . . . . B-51B.9Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53B.9.1Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53 B.9.2Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53 B.9.3Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53 B.9.4Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53Appendix B: Special Design ProcedureTable of ContentsSectionTitlePage Appendix B: Special Design ProcedureTable of Contents (continued)SectionTitlePageRevision 25-11/26/14KPS USARB-ivB.10Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55B.10.1Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55B.10.2Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.10.3Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.10.4Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55B.11Internal Flooding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55B.11.1Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.11.2Flooding Design Criteria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.11.3Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.11.4Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.11.5Conclusion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55B.12Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59B.12.1Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59 B.12.2Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59 B.12.3Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59 Revision 25-11/26/14KPS USARB-vAppendix B: Special Design ProceduresList of TablesTableTitlePageB.2-1Classification of Structures, Systems and Components . . . . . . . . . . . . . . . . . .B-4B.6-1Load Combinations for Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-25B.6-2Applicable Code Stresses Class I Structures: Reinforced Concrete - Structural Steel. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-27B.6-3Applicable Code Stresses: Class I Structures. . . . . . . . . . . . . . . . . . . . . . . . . .B-28B.6-4Allowable Stresses: Class I*, II, III*, III and IV Structures. . . . . . . . . . . . . . .B-29B.6-5Damping Factors. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-30B.6-6Tornado-Generated Missiles. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-31B.6-7Internally-Generated Missiles Inside Of Containment. . . . . . . . . . . . . . . . . . .B-32B.7-1Load Combinations For Components Class Of Components. . . . . . . . . . . . . .B-41B.7-2Loading Conditions and Stress Limits: Pressure Vessels. . . . . . . . . . . . . . . . .B-42 B.7-3Loading Conditions and Stress Limits: Pressure Piping in Accordance with USAS B31.1. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-43B.7-4Loading Conditions and Stress Limits: Equipment Supports. . . . . . . . . . . . . .B-45 B.7-5Load Combination and Stress Limits for Class I Components. . . . . . . . . . . . .B-45B.7-6Alternative Design Loading Combinations and Stress Limits: Pressure Class 1, 2, and 3 Piping In Accordance With ASME Section III . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-46B.9-1Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .B-54 Revision 25-11/26/14KPS USARB-viAppendix B: Special Design ProcedureList of FiguresFigureTitlePageB.7-1Typical Stress Strain Curve . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-47B.7-2Comparison Between Design and Collapse Conditions Hoop Stress: 0.90 Sy B-48B.7-3Comparison Between Design and Collapse Conditions Hoop Stress: 0.00 Sy B-49B.11-1Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-57 B.11-2Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-58 Revision 25-11/26/14KPS USARB-1Appendix BSpecial Design ProceduresB.1SCOPEThe special design procedures contained in this appendix apply to all structures, systems(including instruments and controls), and all components.B.2CLASSIFICATION OF STRUCTURES AND COMPONENTSAll structures, systems (including instruments and controls), and components are classified as Class I, I*, II, III, III* or IV according to their function and importance in relation to the safe decommissioning of the facility, with emphasis on the degree of integrity required to protect the public. These are listed in Table B.2-1.The Turbine Building, Administration Building, Auxiliary Building and Shield Building structures are constructed as a contiguous complex. In general, these structures are identified as either Class I or Class III by placing emphasis on the predominant use of the structure in its relation to the safe decommissioning of the station.In some instances there may be more than one classification applicable within a building or structure. This situation is treated as a mixed classification.Individual components or portions of a system may be determined to have a different classification than the system as a whole. This determination would be accomplished by considering design and functionality requirements of both the system and the components/sub-components, consistent with the 10 CFR 50, Appendix B program for the Kewaunee Power (KPS).
Revision 25-11/26/14KPS USARB-2B.2.1 Definition of Nuclear Safety Design Classifications (NSDC)The definition of the nuclear safety design classifications is given in the following paragraphs1:1.Class IThose structures and components including instruments and controls whose failure might cause or increase the severity of a loss-of-coolant accident (LOCA) or result in an uncontrolled release of substantial2 amounts of radioactivity, and those structures and components vital to safe shutdown and isolation of the reactor. Some items in Table B.2-1 are designated as Class I* indicating that these items have been designed to Class I Design Basis Earthquake (DBE) loading (dynamic) only, and that these items are treated as Class III items in all other respects.2.Class IIThose structures and components which are important to reactor operation3 but not essential to safe shutdown and isolation of the reactor and whose failure would not result in the release of substantial amounts of radioactivity.3.Class IIIThose structures and components which are not directly related to reactor operation or containment. Some items in Table B.2-1 are designated as Class III* indicating that the items are Class III by definition, however, these have been designed to Class II seismic loading.4.Mixed ClassificationThis classification includes structures that are combinations of various Class I, II or III structures. Mixed classifications apply only to structures and not to any systems and/or components. The design criteria for mixed classification are detailed in Section B.6 of this Appendix.5.Class I - Part Class IIIThe spent fuel pool is classified as a Class I structure. The Auxiliary Building structure above the spent fuel pool is classified as a Class III* structure. The Technical Support Center (TSC) basement (586 ft - 0 in) is classified as a Class I structure. The upper floors of the TSC (606 ft-0 in and 626 ft-0 in) are classified as Class III* structures.1.For clarity and continuity, the NSDC definitions have not been revised to reflect the permanent shutdown of the station.2.A substantial amount of radioactivity is defined as that amount of radioactive material, which would produce radiation levels at the site boundary in excess of 1.0 percent of 10 CFR 100 guidelines.3.Reactor operation is defined as the condition where the reactor is producing only that power required to maintain the Reactor Coolant System (RCS) at normal operating pressure and temperature.
Revision 25-11/26/14KPS USARB-36.Class III - Part Class IThe Turbine Building is classified as a Class III* structure. Those areas in the Turbine Building that are classified as Class I house the following equipment:a.Basement Floor*Safety significant 480V Switchgear*Air Compressorsb.Mezzanine Floor*BatteriesThe Class I designation applies to the walls, floors, ceilings, structural support and foundations of structures that isolate, support, or are associated with the protection of Class I equipment.In response to Bulletin 80-11, which identified NRC concerns regarding the structural integrity of safety-related masonry walls a detailed study was performed to provide a technical evaluation of the plant's masonry walls, that at the time were classified as safety-related. The basic documents for guidance in this review were the criteria developed by the Structural and Geo-Technical Engineering Branch of the NRC. The review concluded that the safety-related masonry walls in the plant could withstand the loads and load combinations, as specified in the USAR, without exceeding allowable stress limits (see NRC Safety Evaluation Report in Reference 35).Door 142 and Door 143 were modified per DCR 3594 to allow venting of Room 302 and Corridor 304 in the Auxiliary Building to accommodate atmospheric pressure changes due to a tornado. Room 302 and Corridor 304 are partially enclosed with Class I masonry block walls that are not designed to withstand differential pressure loads from a tornado. Venting of Room 302 and Corridor 304 will minimize the pressure loads on the affected masonry block walls and help ensure the structural integrity of the walls during a tornado.Door 49 was modified with breakaway pins per DCR 3597 to provide relay room block wall protection from Atmospheric Pressure Change (APC) which could result from design tornado loads as specified in USAR Appendix B, Section B.4.2, Item 1.
Revision 25-11/26/14KPS USARB-4Table B.2-1 CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTSNote:This table documents system level and major subsystem design requirements. Systems may contain components or subsystems having a nuclear safety design classification different than the system/subsystem level classification cited within this table. Note:Some abandoned systems, structures, or components (SSC) will continue to be listed in this table.ItemClassClassification of Buildings and StructuresReactor containment vessel (including all penetrations, air locks, isolation valves, vacuum relief devices, and internal containment structures performing Class I function)I*Shield Building (including vent and all penetrations)I*Spent Fuel Pool Structure (including fuel transfer tubes and valves)IControl RoomIScreenhouse (including Access Tunnel and areas housing Service Water, Turbine Building Ventilation and Screenhouse Ventilation System Components)IConcrete Encased Electrical (Class 1) Screenhouse Conduit Structure (598'0")ICirculating Water Intake and Discharge StructuresI Auxiliary Building (areas housing Auxiliary Building Special Ventilation System, radwaste storage, and Engineered Safety Features)IAuxiliary Building Support System for craneaI*Auxiliary Building (except Class I or I*)bIII*Turbine Building (areas housing safeguard batteries, safety significant 480V switchgear, air compressor)ITurbine Building Support System for Turbine Building craneI*Turbine Building (Except Class I or I*)III*
Administrative Building basement (586 feet 0 inch), includes diesel generator roomIEmergency Diesel Generation Room Air Inlet and Outlet StructuresIAdministrative Building (first and second floors, 606 feet 0 inch and 626 feet 0 inch)IIITSC basement (586 feet 0 inch)ITSC upper floors (606 feet 0 inch and 626 feet 0 inch)III*Miscellaneous structuresIIISecurity BuildingIIIOffice/Warehouse AnnexIVAdministration & Training FacilityIVHP Loading DockIVAugmented Water System BuildingIV Revision 25-11/26/14KPS USARB-5Table B.2-1 (continued)CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTSNote:This table documents system level and major subsystem design requirements. Systems may contain components or subsystems having a nuclear safety design classification different than the system/subsystem level classification cited within this table. Note:Some abandoned SSCs will continue to be listed in this table.ItemClassClassification of Systems and Components Reactor Plant EquipmentReactor-Reactor pressure vessel and its supportsI*-Vessel internalsI*-Fuel assembliesI*
-Rod Cluster Control Assemblies (RCCAs) and drive mechanismsI*-In-core instrumentation structuresI*Reactor Coolant System-Piping and valves containing full system pressure (including safety and relief valves)I*-Steam generatorsI*-Pressurizer (excluding pressurizer relief tank, piping downstream of pressurizer relief and safety valves)I*-Reactor coolant pumpsI*-Supporting and positioning membersI*-Primary Sampling System (up to second isolation valve)I*-Pressurizer relief tank and piping (downstream of pressurizer relief valves)IIEmergency Core Cooling System-Safety Injection System (including Accumulator Tanks, Safety Injection Pumps, Residual Heat Removal Pumps (RHR), Refueling Water Storage Tank, RHR Heat Exchangers (RHR), and Primary Connecting Piping and Valving)I*Residual Heat Removal SystemI*Internal Containment Spray System (including spray pumps, spray ring headers, and primary connecting piping and valving)I*Primary Sampling System (beyond second isolation valve)IIIComponent Cooling SystemI*Reactor Control and Protection SystemI*Radiation Monitoring System (to the extent that it must function in support of Class I equipment)I*
Revision 25-11/26/14KPS USARB-6Emergency Power Supply System-Diesel GeneratorsI*-Fuel Oil Storage TankI*-Diesel Generator Cooling SystemI*-Safety features busesI*-Emergency Load Distribution SystemI*-DC power supply, batteries, cableI*-Diesel Generator Fuel Oil Vent LinesI*-Fuel Oil Supply Lines to Day TanksI*Instrumentation-Instrumentation and Control (on all Class I systems)I*-Plant Process Computer System (PPCS)III-Turbine Plant System Instrumentation (except portions of Reactor Control and Protection System, which is Class I)IIINuclear Fuel Handling and Storage-New Fuel Storage RacksI*-Spent Fuel StorageI-Spent Fuel Pool LinerI*-Fuel Transfer System (Including Fuel Transfer Carriage, Containment Upender and Auxiliary Building Upender)III-Spent Fuel Pool Cooling System (Piping and valving whose failure could result in significant release of pool water)I*-Spent Fuel Pool Cooling System (portions not Class I)IIIVentilation SystemsShield Building Ventilation SystemI*Auxiliary Building Special Ventilation System (includes Zone SV isolation dampers and boundary ductwork)I*Auxiliary Building Air Conditioning SystemIIIAuxiliary Building Ventilation SystemIIITable B.2-1 (continued)CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTSNote:This table documents system level and major subsystem design requirements. Systems may contain components or subsystems having a nuclear safety design classification different than the system/subsystem level classification cited within this table. Note:Some abandoned SSCs will continue to be listed in this table.ItemClassClassification of Systems and Components (continued)
Revision 25-11/26/14KPS USARB-7-Safeguards Fan Coil Units I*Reactor Building Ventilation System-Containment Purge and Vent System (Containment Isolation Valves are Class I)III-Containment Dome Fans I*-Post-LOCA Hydrogen Control System (Containment Isolation Valves are Class I)III-Containment Vacuum Relief System I*-Containment Fan Coil Units (includes fans, coils, and housings) I*-CRDM Shroud Cooling System II-Reactor Gap and Neutron Detector Cooling System (excluding Class I piping segment in the reactor cavity) II-Reactor Support Cooling System IIControl Room Air Conditioning System with Service Water System cooling water supply. (Includes Relay Room and Mechanical Equipment Room) I*-Control Room Chillers for normal operationIIITurbine Building Ventilation System (General Area)III-Class 1E Battery Rooms Ventilation System I*-Screenhouse Ventilation System I*-Emergency Diesel Generator Rooms Ventilation System I*-Class I Aisle Safeguards Fan Coil Units I*Technical Support Center Ventilation System IIChemical and Volume Control System-All items except those listed below.I*-Boric acid transfer pumpsII-Boric acid filterII-Boric acid heat tracingII-Batch tankIII*-Evaporator condensate demineralizersIII*-Condensate filterIII*-Monitor tanksIII*Table B.2-1 (continued)CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTSNote:This table documents system level and major subsystem design requirements. Systems may contain components or subsystems having a nuclear safety design classification different than the system/subsystem level classification cited within this table. Note:Some abandoned SSCs will continue to be listed in this table.ItemClassClassification of Systems and Components (continued)
Revision 25-11/26/14KPS USARB-8-Monitor tank pumpsIII*-Deborating demineralizersIII*-Concentrates holding tankIII*-Concentrates holding tank transfer pumpsIII-Chemical mixing tankIII-Resin fill tankIIIWaste Disposal System-Waste Hold-Up TankIII*-Sump TankIII*-Gas Decay TanksI*-Reactor Coolant Drain Tank and PumpsII-Waste Gas Compressor PackageI*-Waste Evaporator Feed PumpIII*-Sump Tank PumpsIII*-Interconnecting Piping and Valves Between Class I EquipmentI*-Waste EvaporatorcIII*-Waste Evaporator Condensate TanksIII*-Laundry and Hot Shower TanksIIIAutomatic Gas Analyzer H2 and O2III*Nitrogen Supply ManifoldIIIHydrogen Supply ManifoldIII Miscellaneous Reactor Plant Equipment-Steam Generator Blowdown System upstream of Isolation Valves BT3A and BT3B outside of containmentI*-Steam Generator Blowdown System downstream of Isolation Valves BT3A and BT3BIII-Polar CraneI*-Manipulator CraneIII-Fuel Pool Bridge CraneI*
-Auxiliary Building CraneI*Table B.2-1 (continued)CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTSNote:This table documents system level and major subsystem design requirements. Systems may contain components or subsystems having a nuclear safety design classification different than the system/subsystem level classification cited within this table. Note:Some abandoned SSCs will continue to be listed in this table.ItemClassClassification of Systems and Components (continued)
Revision 25-11/26/14KPS USARB-9-Turbine Building CraneI*-All Other CranesIII-Conventional Equipment, Tanks, Piping (other than Class I and II)IIITurbine Plant-Turbine, Generator, Foundations, Exciter, Oil Purification, Turbine Gland Seal System, Reheaters and Moisture Separators, Hydrogen and CO2 SystemsIIIService Water System-Serving Class I equipmentI*-All that is not Class IIIIMake-Up Water SystemsIII-Reactor Make-Up Water Storage TankIIICirculating Water System-Circulating water pumpsIII
-Intake piping to ScreenhouseI*
-Circulating water pump discharge pipingIII-Condenser discharge pipingIIICondensate and Feedwater Systems-Main CondenserIII
-Condensate SystemIII-Main Feedwater System (excluding Class I piping and isolation valves)IIIAir Removal SystemIIIAuxiliary Feedwater SystemI*Main Steam System-Main Steam System (portions not Class I)III*-Main steam, safety, relief, and isolation valvesI*-Main steam up to isolation valves including steam piping to Turbine Driven AFW PumpI*Steam Dump SystemIII*Table B.2-1 (continued)CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTSNote:This table documents system level and major subsystem design requirements. Systems may contain components or subsystems having a nuclear safety design classification different than the system/subsystem level classification cited within this table. Note:Some abandoned SSCs will continue to be listed in this table.ItemClassClassification of Systems and Components (continued)
Revision 25-11/26/14KPS USARB-10B.3DESIGN CODESThe design and construction of this plant has been in accordance with the following codes, as applicable:Heating Steam System (those portions in diesel generator rooms, battery rooms, screenhouse, and auxiliary building steam exclusion zones)I*Heater and Moisture Separator Drain SystemIIIBleed Steam SystemIIISecondary Sample SystemIII Miscellaneous Power Systems and Plant Equipment-Station and Instrument Air SystemIII
-Instrument Air System - Portions required for safe shutdownI*-Instrument Air System (except portions required for safe shutdown) II-Fire Protection (serving Class I equipment)I*-Fire Protection System including detection and alarm (other than Class I)III-Potable Water SystemIIITransformers-Main Auxiliary TransformerII-Reserve Auxiliary TransformerII-Tertiary Auxiliary TransformerII-4.16-0.480 kV safety features transformersI*-4.16-0.480 kV auxiliary transformer (other than Class I)III-Transformer serving pressurizer heaters from safety features busIIa.For definition of Class I*, refer to Section B.2.1.b.For definition of Class III*, refer to Section B.2.1 .c.No longer in service.Table B.2-1 (continued)CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTSNote:This table documents system level and major subsystem design requirements. Systems may contain components or subsystems having a nuclear safety design classification different than the system/subsystem level classification cited within this table. Note:Some abandoned SSCs will continue to be listed in this table.ItemClassClassification of Systems and Components (continued)
Revision 25-11/26/14KPS USARB-111.American Concrete Institute Code ACI 318-632.American Institute of Steel Construction "Specification for the Design, Fabrication and Erection of Structural Steel Buildings," 1963 Edition13.American Welding Society Code D 1.0 "Standards for Arc and Gas Welding in Building Construction"4.International Conference of Building Officials "Uniform Building Code," 1967 Edition5.Atomic Energy Commission publication TID 7024 "Nuclear Reactors and Earthquakes"6.American Society of Mechanical Engineers "Boiler and Pressure Vessel Code" 7.Piping Code, USAS B31.1.0-1967 with applicable N-code cases to ASA B31.1-19552, 38.Welding Research Council Bulletin No. 107, 1965 Edition 9.Wisconsin Administrative Code: "Rules of Department of Industry, Labor & Human Relations"10.Crane Manufacturers Association of America Specification 70, "Specifications for Top Running Bridge and Gantry Type Multiple Girder Electric Overhead Traveling Cranes," 2004 Edition.11.ASME NOG-1, "Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder)," 2004 Edition.12.Electrical Overhead Crane Institute (EOCI) Standard 61.
13.NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," dated July 1980.14.NUREG-0554, "Single Failure Proof Cranes for Nuclear Power Plants," dated May 1979.B.4LOADSAll Structures and Components in this plant are designed to withstand various kinds and combinations of loads.The different kinds of loads treated in the design are described in the subsequent paragraphs.1.A later edition may be used for plant physical changes provided appropriate reconciliation is documented. 2.An alternative Design Code to USAS B31.1 is ASME Section III (Post 1980 Editions Approved by NRC, reference Table B.7-6). 3.During RFO 28 tubing for penetrations 1, 3, 21, 27E, 27EN, 27N, 27NE, and 36, located between containment and the shield building, was analyzed to ASME Section III, reference Table B.7-6. Analyses were performed to reconcile thermal stresses that may occur during sampling and differences in displacement of the containment and shield buildings due to annual temperature variations and periodic ILRT testing.
Revision 25-11/26/14KPS USARB-12The load combinations are given in Section B.6 for Structures and in Section B.7 for Components.B.4.1 Environmental LoadsThese consist of wind and snow loads.*Snow LoadA snow load of 40 lb per sq. ft of horizontal projected area is used in the design of Structures and Components exposed to snow.*Wind LoadThe design wind speed is 100 mph. Wind pressure, shape factors, gust factors, and variation of winds with height have all been determined in accordance with the procedures given in the American Society of Civil Engineers' paper ASCE 3269 "Wind Forces on Structures."B.4.2 Tornado Loads Tornado loadings used in design consist of the following:1.A differential pressure equal to 3 psi. This pressure is assumed to build up from normal atmospheric pressure in 3 seconds.2.A lateral force caused by a funnel of wind having a peripheral tangential velocity of 300 mph and a forward progression of 60 mph.3.The design tornado-driven missile was assumed equivalent to an airborne 4 in x 12 in x 12 ft -0 in plank travelling end-on at 300 mph, or a 4000 lb. automobile flying through the air at 50 mph and at not more than 25 feet above ground level.The plant site was examined for possible sources of other missiles including building and equipment parts which were evaluated to determine the potentially most damaging missile. The result of this study is reported in Section B.6.3 of this appendix. The conclusion of this study is that no other missiles are as damaging as the design missiles given above.B.4.3 Live LoadsEquipment loads are specified from manufacturer's drawings and floor loads are based upon the intended use of the floor.B.4.4 Dead LoadsDead loads consist of the weight of structural steel, concrete, dead weight of the component, etc., as computed for each case.
Revision 25-11/26/14KPS USARB-13B.4.5 Seismic LoadsSeveral different seismic loads were used in the design of this plant.1.Operational Basis Earthquake (OBE)The OBE was based upon a maximum vertical ground acceleration of 0.04g, a maximum horizontal ground acceleration of 0.06g and the response spectra are given on Plate 8 in Appendix A.2.Design Basis Earthquake (DBE)The DBE was based upon a maximum horizontal ground acceleration of 0.12g and the response spectra are given on Plate 9 in Appendix A.3.Uniform Building Code Earthquake LoadsThe seismic loads for this category are in accordance with the requirements of the Uniform Building Code. This code specifies the location of the plant site to be in a "Zero" earthquake area. However, for conservatism, earthquake loads applicable to Zone 1 areas were used in the design under this category.B.4.6 Design Basis Accident (DBA) LoadsThe DBA for this plant was the instantaneous double-ended rupture of the cold leg of the RCS. This accident transmits loads to structures and equipment, which were designated as DBA loads.In the permanently shutdown and defueled condition, DBA loads, as defined above, are reduced to zero.B.4.7 Other LoadsIn addition to all the above loads listed, other loads were used in the design wherever applicable. Among these were ice loads, jet forces, other pipe rupture loads, etc.B.4.8 Seismic Design and Verification of Modified, New and Replacement EquipmentOn February 19, 1987 the NRC issued Generic Letter (GL) 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46."The Seismic Qualification Utility Group (SQUG) developed Generic Implementation Procedure (GIP) For Seismic Evaluation of Nuclear Plant Equipment (Reference 19 and Reference 25), as modified and supplemented by the U.S. Nuclear Regulatory Commission Supplemental Safety Evaluation Report Nos. 2 and 3 (Reference 14 and Reference 26), which may be used as an alternative to existing methodologies for the seismic design and verification of modified, new or Revision 25-11/26/14KPS USARB-14replacement equipment classified as Seismic Class I. Subsequent revisions to the GIP may be used provided they have received USNRC review and approval.The SQUG methodology may only be applied to certain classes of active mechanical and electrical equipment as specified in the SQUG GIP (Reference 19 and Reference 25), electrical relays, new and replacement cable and conduit raceway systems, tanks and heat exchangers. The GIP criteria and procedures may be applied to modification or repair of existing anchorage (anchor bolts or welds) including one-for-one component replacements and new anchorage designs. However, for new installations and new anchorage designs, the factor of safety currently recommended for new nuclear power plants shall be met when determining allowable anchorage loads.As specified in the GIP, and accepted by the NRC in Reference 14 and Reference 26, the use of 5 percent damped response spectra may be used when performing seismic evaluations in accordance with the GIP. However, as stated in Reference 18, if it is determined that the equipment natural frequency is within +/- 25 percent of the frequency associated with the peak acceleration, the peak acceleration will be used as the input motion for that piece of equipment.Subsequently, in 1998 the NRC accepted Kewaunee's USI A-46 program implementation as described in the NRC Safety Evaluation Report in Reference 24.B.5PROTECTION OF CLASS I ITEMSThe Class I items are protected against damage from:a.Rupture of a pipe or tank resulting in serious flooding to the extent that the Class I function is impaired.b.Deletedc.Earthquake, by having the ability to sustain seismic accelerations adopted for purposes of plant design without loss of function. Protection from interaction with the surrounding buildings is accomplished by providing a separating joint of sufficient size for earthquake displacements. Unless the building is designed to Class I seismic design, an analysis is made to demonstrate that it will not collapse; otherwise, the systems are protected locally.d.Tornado wind loads.e.Other natural hazards. Examples of these hazards are seiche and ice.f.Fire, in such a way that fire and operation of fire-fighting equipment does not cause damage to redundant parts of the system.g.Missiles from different sources. These sources comprise:(i.)Tornado created missiles.
Revision 25-11/26/14KPS USARB-15(ii.)Missiles from components containing moving parts, which could be subjected to overspeed. (Potential sources for such missiles are diesel engines and gas turbines.)(iii.)DeletedNo protection is required if the factors described under item a (non-HELB), item f, and item g cannot affect any Class I systems, or if redundant systems are provided and the physical separation of these systems is sufficient to prevent these factors from damaging both systems. Under item c and item d, redundancy and physical separation may decrease the requirements for protection. If redundancy and physical separation are not used, and if the surrounding building is not designed as a missile barrier, missile protection by shielding is necessary, either by shielding the source itself or by shielding the system.B.6DESIGN CRITERIA FOR STRUCTURES This section describes the general Design Criteria for Structures used in the plant design. Special situations like protection against crane toppling, etc., are given towards the end of this appendix in separate paragraphs.B.6.1 Load CombinationsThe load combinations applicable to Class I, Class I*, Class II, Class III*, Class III, and Class IV Structures are given in detail in subsequent paragraphs and are also listed in Table B.6-1.Class I StructuresClass I Structures are analyzed for each of the following conditions of loading:1.Normal Operating ConditionsThe load combinations consist of Dead and Live loads together with the environmental loads (wind and snow) as specified in Section B.4.1.2.Operational Basis Earthquake ConditionsThe combination consists of Dead, Live, DBA, and snow loads together with the greater of the OBE or Wind loads.3.Design Basis Earthquake ConditionsThe load combination consists of Dead, Live, Snow, and DBA loads together with the DBE loads.4.Tornado Condition Revision 25-11/26/14KPS USARB-16The load combination consists of Dead and applicable Live loads together with the 300-mph design tornado and tornado missile loads, if any. These loads are assumed non-coincident with DBA or Seismic loads.In addition to the above four conditions, windrowed ice loading was considered for the following Class I structures:1.The screenhouse water intake structure is located inland from the shore of the Lake Michigan and is not therefore subjected to ice loading.2.The circulating water intake structure is designed for ice loading.Class I* StructuresThose structures designated as Class I* were analyzed for each of the following conditions of loading:1.Normal Operating ConditionsThe load combination consists of Dead and Live loads together with environmental loads (wind and snow).2.Design Basis Earthquake ConditionsThe load combination consists of Dead, Live, Snow and the DBE loads.Class II StructuresClass II Structures were designed for the greater of the following two combinations of loads:1.Dead, Live, and Environmental loads (wind and snow), or2.Dead, Live, and Uniform Building Code earthquake loads specified in Section B.4.5 of this Appendix.Class III* Structures Class III* Structures were designed for the greater of the two combinations of loads given above for Class II Structures.Class III StructuresClass III Structures were designed for Dead and Live loads together with environmental loads (wind and snow).As a minimum, Class III Structures were designed in accordance with the applicable codes as listed in Section B.3. In accordance with the Wisconsin Public Service Corporation's normal policy for the design of steam-electric generating stations, certain items of power plant structures in the Class III category were designed according to the requirements of a higher classification.
Revision 25-11/26/14KPS USARB-17Class IV StructuresClass IV structures were designed for Dead and Live loads together with environmental loads (wind and snow) in accordance with the State of Wisconsin Administrative Code.B.6.2 Stress Design CriteriaNormal OperationThe allowable stress design criteria that were applied for normal operating conditions were in accordance with the applicable code(s) listed in Section B.3. These code allowable stresses are summarized in Table B.6-2, Table B.6-3, and Table B.6-4.Design Basis Accident and Operational Basis Earthquake The allowable stress design criteria applied for the DBA condition in combination with the OBE were that stresses remain within the allowable limits specified by the applicable code(s) listed herein, except that allowable stresses were not increased for the earthquake condition as is permitted by some codes. These code allowable stresses are summarized in Table B.6-2, Table B.6-3, and Table B.6-4.Safe ShutdownThe design criteria for tornado missiles, the 300-mph design tornado condition, and also for the DBA in combination with the DBE were that the reactor can be safely shut down and that there be no uncontrolled release of radioactivity.To meet these criteria, structures or components were examined for their function in the total system to assure a safe and orderly shut down.These criteria, as applied to tornado winds, and to the DBA condition in combination with DBE loads, will permit some permanent deformation but will not permit loss of structural function. In this sense, structural function is defined to mean that structures will remain intact and continue to support their normal operating loads after an earthquake and/or DBA, but may require repair or replacement for future continued use.
Tornado missiles may result in large local deformations, but the criteria will not permit the missile to breach the barrier so that essential safety features functions are jeopardized.B.6.3 Structural Design BasisClass I StructuresThe designs of Class I structures for seismic, tornado winds, tornado missiles, etc., are given in subsequent paragraphs.
Revision 25-11/26/14KPS USARB-18Seismic DesignFor dynamic analysis, an equivalent multi-mass mathematical model was constructed to approximate the structural system. The effect of the foundation soils was included in the model by means of equivalent springs. The spectral method was then used to determine the maximum response of each mass point for each node, using as input the OBE (Plate 8 in Appendix A) and the damping factors given in Table B.6-5. The total response for each point was determined by the root-mean-square method. From this, a set of curves was developed which show the variation with height of the maximum translational accelerations, displacements, shears and moments in the structure. All of the above work was performed by John A. Blume and Associates and is reported in detail in a separately submitted Topical Report JAB-PS-01(s) (hereinafter referred to as "The Blume Report"). Vertical acceleration equal to two-thirds the horizontal ground acceleration was applied to the structure.Operational Basis EarthquakeUsing the data presented in the Blume Report, stresses were computed for the various parts of the plant structures. The stresses resulting from both horizontal and vertical acceleration were combined to obtain the total earthquake stresses. Earthquake stresses were then added linearly and directly to stresses caused by DBA, snow, dead loads, and the appropriate operating loads to obtain the total stresses. The total stresses were reviewed to ensure that they were within the maximum stress limits as established in Table B.6-2 and Table B.6-3. Direct superposition of stresses has been used for all loads except missile impact and contact points of pipe rupture restraints. For these loads the material is stressed beyond the elastic range. Design procedures for missile impacts are given in the section entitled Tornado Missiles of this appendix.Design Basis Earthquake The forces for the DBE were taken to be two times the forces as determined by the spectral analyses for the OBE. Stresses were combined as before and it was established that they were within limits as indicated in Table B.6-2 and Table B.6-3.Tornado Winds Structures were analyzed for stresses due to tornado and missile loads. Stresses due to tornado loading were combined with stresses due to dead loads and the appropriate operating loads to obtain a total stress. Maximum stresses were limited to those specified in Table B.6-2.Tornado Missiles Spectrum of Missiles ConsideredMany missiles were considered, but only the most damaging missiles were used for design. Missiles were assumed to be generated by explosive injection due to pressure differential, by building component failure, and by aerodynamic lifting, each resulting in an airborne or Revision 25-11/26/14KPS USARB-19free-falling missile. Only objects with large surface-to-weight ratios would remain airborne long enough to attain high horizontal velocities. Table B.6-6 lists the missiles considered and the maximum velocities that would be attained by each.Design for MissilesThe design basis missile protection criteria (Reference 36) states "systems required to shut the plant down and to keep the plant in a safe shutdown condition shall not be prevented from performing their function by external missiles." It also states: "protection of the equipment relied upon to provide reasonable assurance of safe plant operation can be achieved by either housing or making them part of redundant systems with such physical separation that sufficient back-up is provided to assure no loss-of-function of them."
Systems, structures, and components (SSC) were designed to meet the design basis missile protection criteria, considering the spectrum of design basis missiles and the limiting, most damaging design basis missile for that specific SSC. Systems relied upon for safe shutdown and safe plant operation in a tornado event were either placed (housed) in a class 1 structure for protection or designed as a redundant system with sufficient physical separation to ensure that the single limiting, most damaging DB missile would not cause a loss of function of that system.Systems with peripheral, unprotected SSC were evaluated for tornado effects to address the potential vulnerabilities in these systems (References 37, 38, 39). Based on these evaluations it was concluded that: 1) the system design is in compliance with the original design basis tornado missile protection criteria 2) there is adequate physical separation and redundancy in the design to ensure that the system is capable of performing its design function required in a tornado event for safe shutdown 3) the plant will be able to withstand the consequences of the tornado, will retain the capability to achieve and maintain the reactor in a safe shutdown condition, and there will be no uncontrolled release of radioactivity as a result of the tornado event 4) the unprotected peripheral SSC in these systems can reasonably be exempt from the requirement for specific missile barriers without jeopardizing the health and safety of the public. The concept in analysis and design considered impact to be a plastic collision between the missile and the structure.
Tornado missiles generally are of an intermediate energy level. Their total kinetic energy is dissipated by energy absorption of the affected structure as a whole. This results from the elastic and plastic response of the structure to the impact force, energy absorption by the missile itself due to plastic deformation of the missile, and by the building structure missile barrier member due to local plastic deformation.
A missile barrier of reinforced concrete will react to missile impact as a combination of non-ductile concrete and ductile reinforcing steel. The mode of concrete failure will be brittle fracture such as might result from punching shear. Shear cracks will occur at the impact area Revision 25-11/26/14KPS USARB-20perimeter and progress outward as concentric perimetric rings of fracture. A reinforced concrete member will respond elastically and plastically as a moment-resisting reinforced concrete element up to the point of brittle fracture of the concrete, and then the reinforcement will respond as tensile strands in membrane actions, elongating plastically to absorb the kinetic energy.The problem of establishing a missile barrier can be subdivided according to the behavioral response of the characteristic structural element, i.e., slab, wall, beam, and column.Slab and walls can respond by perforating or shear failure, plastic bending, and finally forming a tensile membrane as described above.A comparison was made of various penetration formulas such as the Army Corps of Engineers, Ballistic Research Laboratory, and Modified Petry before selection of the Modified Petry formula as the most commonly used and best fit to the controlling conditions. None of the available formulas developed from empirical ballistic information were particularly suited to tornado missile problem solutions.
In using the Petry formula, the usual rule is to make a slab or wall of a thickness at least twice the penetration determined by the second Modified Petry formula for concrete of finite thickness. This was done assuming all deformation to occur in the concrete (indestructible missile). A correction factor was applied to steel missiles of non-circular or open cross-section, such as steel girts and steel pipe, so that the area used in the Petry formula to determine the theoretical penetration of an indestructible missile was three times the net cross-sectional area of the steel.Assumption of an indestructible missile leads to very high peak loads and shear stresses when making an analysis for impulse loading, therefore, experiments of limited scope were performed which verified that almost all of the local plastic deformation would occur in the wood (for a wood missile) impacting on concrete, and that steel missiles would enter a plastic range while penetrating concrete.To provide a workable solution for applying the Petry formula to a wood missile, a "K" value predicted on plastic deformation (or destruction) of the wood was used to determine the "penetration" or deceleration path, and from this a peak load was obtained. In the case of the steel missile, the peak load is limited by the short-duration yield strength of the steel.Table B.6-6 is a tabulation of tornado-generated missiles, which shows the weight to cross sectional area ratio and gives the impact velocity of the worst case for these missiles. All tornado-generated missiles were assumed to impact end on at 90 degrees to the surface being impacted, and all areas of Class I structures exposed to either falling or horizontally flying tornado missiles are investigated. The tornado missiles were assumed to come from stored material, destruction of lower class structures, off-site construction, etc. The peak loads associated with the various missiles are as follows:
Revision 25-11/26/14KPS USARB-21*Horizontal flying wood plank400 kips*Vertical falling wood plank288 kips
- Steel girts197 and 257 kips*Steel pipe180 kips*Automobile182 kipsUsing the peak load, slabs and walls were analyzed for their response to shear (approximately at the ultimate strength of the concrete, in shear) and ability to develop plastic hinges and a tensile membrane of reinforcing steel. After the shear failure of the slab or wall, the plastic deformation of the longitudinal reinforcing is calculated not to exceed a strain of 5 percent.Reinforced concrete beams in a horizontal plane were analyzed for impulse loading. A rectangular force-time curve was assumed so that the methods contained in Reference 1 could be used. The dynamic system was established, including boundary conditions, size of member, member characteristics, reinforcing, loading, span, etc., to determine the natural frequency and plastic strength of the member. From the peak load previously found and the plastic resistance, the ductility factor was determined and this was conservatively limited to 6. If this is exceeded, the beam is redesigned to limit the ductility factor to 6. The dynamic reactions were calculated for the elastic or plastic strain range, as required, and combined with other loads (Dead loads, etc.). A minimum value of missile impact reaction of 300 kips was used in order to provide a minimum shear strength capability for missiles impacting near a support.The allowable shear stresses used were: 4 f 'cd for reinforced beam webs, 6 f 'cd for d/2 stirrup spacing, 10 f 'cd for d/4 stirrup spacingwhere: = 0.85, and d = 1.25,f 'c = ultimate compressive strength of concrete,and using a minimum web reinforcement of 0.15 percent bs (beam width, b x bar spacing, s). Stirrup stress was limited to 0.85 times 1.25 fy (yield strength of reinforcing bars) and bond stress was limited to 0.15f 'c with 0.85 of the summation of the perimeter of bars.Beams designed by this procedure will have very minimal plastic deflection under tornado missile impact. Beams, which were too small to comply with the above requirements were investigated for the capability to hang from adjacent slabs as a thickened portion of the slab.
Revision 25-11/26/14KPS USARB-22Columns were designed for a 300-kip missile impact load, centered on top, combined with all other applicable loading. The 300-kip load was chosen to establish a minimum strength in columns subject to missile impact and exceeds the dynamic reaction from the beams.The stress level in columns under the above loading was limited to 1.5 times the ACI code allowable stress to provide a higher factor of safety in the columns than that used in beam and slab design.The listed procedures were conservative and provide for missile barriers that can absorb sufficient missile energy to reduce the missile velocity to zero without physical breach of the barrier, and keep cracking and plastic deformation within acceptable levels.Class I* StructuresThe design of Class I* Structures is similar to the design of Class I Structures for seismic loads only, as detailed previously in this section.In all other respects the design requirements of Class I* Structures are identical to the design requirements of Class III Structures, as detailed below.Class II StructuresStructures in this class were designed for the conditions of loading specified in Section B.6.1 and Table B.6-1 and in accordance with the design methods and allowable stresses specified in the codes listed in Section B.3. Stresses were combined as before and reviewed to assure that they were within the limits set forth in Table B.6-4.Class III* StructuresThe design of Class III* Structures is similar to the design of Class II Structures for the condition of loading specified in Section B.6.1 and Table B.6-1.In all other respects the design of Class III* Structures is identical to the design of Class III Structures as detailed below.Class III StructuresStructures in this class were designed for the conditions of loading specified in Section B.6.1 and Table B.6-1, and in accordance with the design methods and allowable stresses specified in the codes listed in Section B.3. Stresses were combined as before and reviewed to assure that they were within the limits set forth in Table B.6-4.
Revision 25-11/26/14KPS USARB-23Class IV StructuresStructures in this class were designed for the conditions of loading specified in Section B.6.1 and Table B.6-1 and comply with the requirements of the State of Wisconsin Administrative Code. These facilities provide staff working space, employee facilities, and material and record storage space. These structures are designed to be independent of other plant structures except for minor loads imposed at the interconnection to existing facilities. These structures connect to Class III or Class III* structures only.Mixed Classification StructuresA Class I area located in a lower class structure was treated as a Class I structural system within the lower class structure.Components of the Class I structural system which were required to meet the total structural function of this system may extend into the lower class area and were analyzed for their Class I function. These components include related foundations, supporting structures and overhead structures.The design provisions made where a structure of a lower seismic design classification is adjacent to a structure of a higher classification to prevent damage to the higher classification structure under conditions associated with design basis seismic or tornado events were as follows:*The mathematical model of the Reactor, Auxiliary and Turbine Buildings including the steel framed structures were all considered as one interconnected structure for the dynamic earthquake analysis. The resultant acceleration displacements, shears and torques have all been included in the design of the interconnected Class II structures, thus making the structural elements higher classification. At joints where seismic separations between adjacent structures were required a gap equal to twice the sum of their respective displacements was provided.*Smaller lower class structures appending the main structures were analyzed under Class II seismic requirements in accordance with this appendix. These structures were reviewed to assure that the effects of a DBE would not damage the higher-grade structures sufficiently to affect the safe and orderly shutdown of the reactor.The Class I concrete structures were analyzed and designed to withstand the effects of a tornado in accordance with the parameters as established in this appendix.Steel-framed structures are enclosed with metal siding and roof decking. The siding and a portion of the roof decking have been attached with pressure relief fasteners to vent the building from tornado pressures and forces. This will prevent the stresses in the main structural frames from exceeding the allowable limits established in this appendix, and thus prevent their collapse onto Class I structures.
Revision 25-11/26/14KPS USARB-24The design assures that a failure of adjacent lower class structures due to earthquake, tornado winds or missiles will not cause a loss of function to the Class I structure by direct or indirect failure of structural components.
Revision 25-11/26/14KPS USARB-25Table B.6-1 LOAD COMBINATIONS FOR STRUCTURESClass of StructuresConditions of LoadingClass IClass I*Class IIClass III*Class IIIClass IVNormal OperatingDead + Live + Wind + SnowDead + Live + Wind + SnowDead + Live + Wind + SnowDead + Live + Wind + SnowDead + Live + Wind + SnowDead + Live + Wind + SnowOperational Basis Earthquake (OBE)Dead + Live +
DBA + Snow +
Greater of the OBE or WindNANANANANADesign Basis Earthquake (DBE)Dead + Live +
Snow + DBA + DBEDead + Live + Snow + DBENANANANATornadoDead + Live + 300 mph Design Tornado + Tornado Missile, if anyNANANANANA Revision 25-11/26/14KPS USARB-26OtherIn addition to above, jet forces ice loads, pipe rupture loads, etc., whichever is applicableNADead + Live + Uniform Building Code Zone 1 earthquake loads (see Section B.4.5)Dead + Live + Uniform Building Code Zone 1 earthquake loads (see Section B.4.5)NAWind + Snow Loads are specified in the State of Wisconsin Administrative Code as 80 mph and 30 lb/ft2Note: N/A = Not ApplicableTable B.6-1 (continued)LOAD COMBINATIONS FOR STRUCTURESClass of StructuresConditions of LoadingClass IClass I*Class IIClass III*Class IIIClass IV Revision 25-11/26/14KPS USARB-27Table B.6-2APPLICABLE CODE STRESSES CLASS I STRUCTURES: REINFORCED CONCRETE - STRUCTURAL STEELLoading ConditionReinforced ConcreteStructural Steel1.Normal Operating Condition: Dead and Live Loads + Environmental Loads (Wind + Snow)ACI 318-63 allowable valuesAISC allowable values2.Operational Basis Earthquake Condition: Dead + Live + DBA + Snow + Greater of the OBE OR WindACI 318-63 allowable valuesAISC allowable values3.Design Basis Earthquake Condition: Dead Loads + Live + Snow Loads + DBA +
DBE Load1 1/2 times ACI 318-63 allowable values1 1/2 times AISC allowable values4.Tornado Condition: Dead Loads + Live Loads + 300 mph Design Tornado (Does not include Tornado Missile)fc = 0.75 f 'cfs = 0.90 Y.S.fs = 0.90 Y.S.Where: f 'c= Minimum 28-day compressive strength of concrete fc = Compressive stress in concrete fs = Tensile Stress in steel Y.S.=Specified minimum yield strength or yield point of steel Revision 25-11/26/14KPS USARB-28Table B.6-3APPLICABLE CODE STRESSES: CLASS I STRUCTURESLoading ConditionCriteriaConcrete Stresses fcReinforced Steel Stresses615 Grade 40A 615 Grade 60Allowable Working Stress psiPercent of Min. Spec. Yield 1Allowable Working Stress psiPercent ofMin. Spec. Yield 1Normal OperatingConditionACI 318-63 Allowable values0.45 f 'c20,0005024,00040Operational Basis EarthquakeACI 318-63 Allowable Values0.45 f 'c20,0005024,00040Design Basis Earthquake11/2 times ACI 318-63 allowable values0.675 f 'c30,0007536,000601.Minimum specified yield points of steel reinforcements are as follows: A615 Grade 4040,000 psi A615 Grade 6060,000 psi Revision 25-11/26/14KPS USARB-29Table B.6-4ALLOWABLE STRESSES: CLASS I*, II, III*, III AND IV STRUCTURESClassLoading ConditionCriteriaI*Item (3), Table B.6-2 and Table B.6-3Item (3), Table B.6-2 and Table B.6-3IIDead load plus live loads, plus greater of wind plus snow or Zone I earthquakeACI 318-63 and AISC allowable stresses with no increase in stresses for earthquake conditionIII*Same as for Class II aboveACI 318-63 and AISC allowable stresses with no increase in stresses for earthquake conditionIIIItem (1), Table B.6-2ACI 318-63 and AISC allowable stressesIVItem (1), Table B.6-2State of Wisconsin Administrative Code Revision 25-11/26/14KPS USARB-30Table B.6-5DAMPING FACTORSItemPercent ofCriticalDamping*Reactor Containment vessel1.0Shield Building2.0Reactor containment vessel internal concrete5.0Steel frame structures2.0Reinforced concrete construction2.0Piping systems0.5Electrical and mechanical equipment evaluated in accordance with the Blume Report (Reference 9)1.0Foundation soils5.0Electrical and mechanical equipment evaluated in accordance with the SQUG GIP**5.0*The maximum percent of critical damping factors given is applied to both the OBE and the DBE.** See Section B.4.8.Note: At and below the mezzanine floor level, the Shield Building, Auxiliary Building, and Containment System are interconnected so as to comprise a monolithic structure. The many shear walls below this level in the Auxiliary Building, the grout under the Reactor Containment Vessel, and the shear walls in the Containment System all combine to form a very stiff connection between the Basement level and the Mezzanine level. For this reason, the mathematical model used for the dynamic analysis of these buildings considers that they are rigid between these two levels. Above mezzanine floor level these concrete buildings are not interconnected and the individual damping values are used (i.e., 5 percent for Auxiliary Building, 2 percent for Shield Building, and 5 percent for Reactor Containment Vessel internal concrete construction).
Revision 25-11/26/14KPS USARB-31Table B.6-6TORNADO-GENERATED MISSILESMissileWeight(lb)Weight toCrossSectionalArea Ratio(lb/sq in.)ExplosiveInjectionHeight(ft)Elevationof OriginAboveTarget(ft)TotalHeightof Drop(ft)VerticalVelocity(ft/sec)VerticalEnergy(ft-lb)HorizontalVelocity(ft/sec)HorizontalEnergy(ft-lb)Wood Plank 4 in x 12 in x 12 ft - 0 in Rough Douglas Fir1501503.1 3.1112NA67NA179NA108NA27,168NANA440NA450,930Steel Girt W 10 in x 11.5 in x 20 ft - 0 in A36 Steel230 2306.8 6.892NA67NA159NA102NA37,157NANA35NA4375Steel Girt W 8 in x 15 in x 20 ft- 0 in A36 Steel300 3006.8 6.835NA67NA102NA81NA30,564NANA35NA5707Steel Pipe 4 in STD. 10.79 lb/ft 4.5 in O.D. x 20 ft - 0 in2162166.8 6.810NA67NA77NA70NA16,435NANA66NA14,610Automobile40000.525NANANANA73330,990Notes:All tornado missiles are assumed to impact end-on, at 90 degrees to surface being impacted. NA = Not Applicable Revision 25-11/26/14KPS USARB-32Table B.6-7INTERNALLY-GENERATED MISSILES INSIDE OF CONTAINMENTMissileWeight to Cross Sectional Area Ratio(lb/sq. in.)Velocity(ft/sec)Impact Point3 in Motor Operator Isolation Valve14.1100Missile Shield Slab6 in x 6 in Valve (Safety Relief Valve)9.375Missile Shield Slab3 in Air operator Relief Valve2.450Missile Shield SlabHousing Plug1.8240Reactor Vessel Missile ShieldDrive Shaft49.5151Reactor Vessel Missile ShieldDrive Shaft and Drive Mech.13514.3Reactor Vessel Missile ShieldNote: All internally generated missiles are assumed to impact at 90 degrees.
Revision 25-11/26/14KPS USARB-33B.7DESIGN CRITERIA FOR COMPONENTSThis section describes the general design criteria for all mechanical, electrical, instrument, and control components used in the plant design.B.7.1 Load CombinationsThe load combinations applicable to Class I, Class I*, Class II, Class III*, and Class III Components are given in detail in subsequent paragraphs and are also listed in Table B.7-1. The term Live loads when used on components consists of thermal and pressure loads.Class I ComponentsClass I Components were analyzed for each of the following conditions of loading:1*Normal Operating ConditionThe load combination consists of Dead and Live loads, together with Environmental loads (wind or snow), wherever applicable.*Normal and OBE ConditionThe load combination consists of Dead and Live loads, together with the greater of the OBE or Wind loads.*Normal and DBE ConditionThe load combination consists of Dead and Live, together with DBE, loads.*Normal and Pipe RuptureThe load combinations consists of Dead, Live, and pipe rupture loads, excluding loads from pipe rupture in the reactor coolant loop.*Normal and DBE and Pipe RuptureThe load combination consists of Dead, Live, DBE loads, and pipe rupture loads, excluding loads from pipe rupture in the reactor coolant loop.1.Replacement steam generator lower units are designed and analyzed to loading combinations defined in Design Specifications 414A03, consistent with ASME Code,Section III, Division 1, Subsection NB, Class 1, 1986 Edition through 1987 Addenda. The original steam domes are analyzed in the same manner as the replacement lower units.
Revision 25-11/26/14KPS USARB-34Class I* ComponentsThose components designated as Class I* were analyzed for each of the following conditions of loading:*Normal Operating ConditionThe load combination consists of Dead and Live loads, together with Environmental loads, if applicable.*Normal and DBE ConditionThe load combination consists of Dead, Live and DBE loads.Class II Components Class II Components were designed for the greater of the following two combination of loads:*Dead, Live and Environmental loads, if applicable, or,*Dead, Live, and Uniform Building Code (UBC) loads specified in Section B.4.5 of this Appendix.Class III* Components Class III* Components were designated for the greater of the two combinations of loads given above for Class II components.Class III ComponentsClass III components were designed for Dead and Live loads, together with Environmental loads, if applicable.As a minimum, Class III components were designed in accordance with the applicable codes as listed in Section B.3. In addition, in accordance with the Wisconsin Public Service Corporation's normal policy for the design of steam-electric generating stations, certain components of the power plant in the Class III category were designed according to the requirements of a higher classification.
B.7.2 Design Criteria1.Deleted2.Deleted Revision 25-11/26/14KPS USARB-351.Deleted2.Deleted3.Deleted4.Deleted5.Deleted 6.Deleted1.Deleted2.Deleted3.Deleted 4.DeletedDesign Criteria for Class I* Components*1.Deleted 2.Deleted3.Deleted Revision 25-11/26/14KPS USARB-36Mechanical and Electrical EquipmentThe following information is HISTORICAL and is not intended or expected to be updated.Westinghouse-Furnished EquipmentThe Standard Westinghouse 2-loop analysis used an envelope of response acceleration spectra which was more conservative than those presented in the Blume Report (Reference 9).The Seismic criteria for Westinghouse furnished equipment were as follows:1.Equipment specifications to vendors required that Westinghouse-supplied Seismic Class I Auxiliary Pumps be designed by the vendor to operate during horizontal and vertical acceleration of 1.0g and 0.67g, respectively and simultaneously. The sum of the primary stresses shall not exceed Section III of the ASME Code for pressure-containing members and other critical components.2.Seismic Class I tanks were designed by Westinghouse PWR to withstand the simultaneous horizontal and vertical forces resulting from the amplified ground acceleration response spectrum curves for the DBE.3.Seismic Class I valves were designed by the vendor to withstand seismic loadings equivalent to 3.0g in the horizontal direction and 2.0g in the vertical direction.To assure that Westinghouse-supplied NSSS Class I mechanical components met the above seismic design criteria, the following procedure was implemented:1.The acceleration factor was included in the Equipment Specification and the vendor had to certify the adequacy of the component to meet this seismic requirement.2.The vendor's drawings and calculations were reviewed by the cognizant engineer responsible for the particular component to determine whether the component met all specification requirements3.Based on engineering judgment and detailed analyses on similar equipment, the cognizant engineer eithera.Accepted the component, orb.Rejected the component as inadequate, or recommended modifications, orc.Requested that the engineering analysis section review the drawing details and perform a detailed analysis, if deemed necessary, using one of the methods described in the following paragraph.To conform to the above, seismic analysis of selected NSSS Seismic Class I components including heat exchangers, pumps, tanks and valves was performed by Westinghouse using one of three methods depending on the relative rigidity of the equipment being analyzed:
Revision 25-11/26/14KPS USARB-37Balance of Plant EquipmentThe seismic design criteria for balance of plant Class I (seismic) mechanical components and electrical equipment are described as follows:1.For the OBE, the mechanical components and electrical equipment shall be designed to be capable of continued safe operation within normal design limits when subjected to the combination of normal loads and OBE loads.2.For the DBE, the mechanical components and electrical equipment are designed so that the deflections or distortions resulting from the combination of normal loads and twice the OBE loads shall not prevent their proper functioning, shall not endanger adjacent or attached equipment, and shall not cause the equipment to operate in an uncontrolled manner.In order to meet these seismic design criteria the following measures were taken for seismic design and restraint:1.Equipment which is rigid and rigidly attached to the supporting structure is analyzed for loading equal to the 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, is analyzed for the peak of the floor response curve with appropriate damping values;3.In some instances, non-rigid equipment is analyzed using a multi-degree-of-freedom modal analysis including the effect of modal participation factors and mode shapes together with the spectral motions of the floor response spectrum defined at the support of the equipment.The inertial forces, moments, and stresses are determined for each mode. They are then summed using the square-root sum-of-the-squares method. A sufficient number of masses are included in the mathematical models to insure that coupling effects of members within the component are properly considered. The results of these analyses indicate that the models contain more masses than necessary. The method of dynamic analysis uses a proprietary computer code called WESTDYN. This code uses as input, inertia values, member sectional properties, elastic characteristics, support restraint data characteristics, and the appropriate seismic response spectrum. Both horizontal and vertical components of the seismic response spectrum are applied simultaneously. The modal participation factors are combined with the mode shapes and the envelope seismic response spectra to give the structural response for each mode. The inertial forces, moments, and stresses are computed for each mode, from which the modal stresses are determined. The stresses are then summed using the square-root sum-of-the-squares method.In general, no additional restraints beyond those normally provided are required to assure seismic adequacy.The following information is HISTORICAL and is not intended or expected to be updated.
Revision 25-11/26/14KPS USARB-381.The mechanical component or electrical equipment and its supports were designed to be sufficiently rigid so that its natural frequency or frequencies will be out of the range of resonance with the building structure where it is located, based on the response acceleration spectrum curves established in the earthquake analysis prepared by John A. Blume and Associates.2.The maximum stresses induced from the combination of normal loads plus OBE loads were maintained below the allowable stress limit of the material as given in the applicable codes.3.The maximum stresses induced from the combination of normal loads plus twice the OBE loads were limited to less than 90 percent of the yield strength of the material under consideration, and the deflections or distortions were so limited that they will not affect proper functioning of the equipment.The analytical or testing methods utilized to verify the adequacy of the above are described as follows:1.Analytical Methodsa.Where practical the natural frequency or frequencies of the component or equipment under consideration were determined by the use of a proper mathematical model.b.For a single-degree-of-freedom model, the natural period was used to determine the horizontal and vertical response accelerations from the structural floor response acceleration spectra. These accelerations were applied at the mass center of the component simultaneously and the system was analyzed statically.c.For a multiple-degree-of-freedom model, where practical, the modal superposition method was used to determine the response of the dynamic system.d.For those components for which the natural frequency could not be determined, the peak value of the structural floor response accelerations for the appropriate mass point multiplied by the maximum torsional acceleration factor at the mass center of the component were applied and the seismic forces were determined.e.For the DBE, the response acceleration values are twice those used for the OBE.2.Testing Methods: (one of the following)a.Continuous TestThe test was executed at frequencies incremented within the range of significant structural response of the applicable structural response spectra. The test consisted of the application of a continuous sinusoidal motion corresponding to the maximum structural acceleration for which the equipment was to be qualified and for an appropriate length of time. The equipment was properly mounted during testing so as to reflect the field-installed condition.
Revision 25-11/26/14KPS USARB-39b.Sine Beat TestNatural or resonant frequencies were detected by scanning from the lowest practical frequency to 25 Hz. The test at resonant frequencies consisted of the application of sine beats of peak acceleration values corresponding to that for which the equipment was to be qualified. The duration of the beat for each particular test frequency was chosen to most nearly produce a magnitude of equipment response equivalent to that produced by the particular floor acceleration with proper damping ratio. The equipment was properly mounted during testing so as to reflect the field-installed condition.Seismic input values used for analysis or testing purposes to verify the adequacy of Class I (seismic) components were obtained from Topical Report JAB-PS-03 prepared by John A. Blume and Associates, Engineers (Reference 9).Instrumentation and Control SystemsThe design bases for protection-grade equipment (Class I) with respect to earthquakes were that for an OBE or DBE, the equipment was designed to ensure that it did not lose its capability to perform its function; i.e., shut the plant down and/or maintain the unit in a safe shutdown condition. For the DBE, the capability of the protection equipment to perform its function wasmaintained.If a seismic disturbance occurs subsequent to an accident, the instrumentation and electrical equipment associated with emergency core cooling will not be interrupted during this disturbance.Initial evaluation of Protection System equipment for its ability to withstand the seismic condition was typically done by actual vibration-type testing of typical protection-grade equipment. Mathematical models derived from empirical tests were not normally used for seismic design evaluation of instrumentation. However, in the absence of empirical test data, such as may be the case for very large equipment (for example, control room panels), evaluation may have beensupported by mathematical analysis or some combination of mathematical analyses and prototype testing. (See Reference 4 for discussion and documentation of some test program results).Design Criteria for Class II and Class III* ComponentsComponents in this class are designed for the conditions of loading specified in Table B.7-1 and in accordance with the design methods and allowable stresses specified in the codes listed in Section B.3. Stresses are combined as for Class I above and reviewed to assure that they are within the limits set forth in the applicable codes.
Revision 25-11/26/14KPS USARB-40Design Criteria for Class III ComponentsComponents in this class are designed for the conditions of loading specified in Table B.7-1 and in accordance with the design methods and allowable stresses specified in the codes listed in Section B.3.
Revision 25-11/26/14KPS USARB-41Table B.7-1LOAD COMBINATIONS FOR COMPONENTS CLASS OF COMPONENTSCondition of LoadingClass I1, 2Class I* 3Classes II and III*Class III1.NormalDead + Live + Environmental Loads (Snow or Wind) If ApplicableDead + Live + Environmental Loads (Snow or Wind) If ApplicableDead + Live + Environmental Loads (Snow or Wind) If ApplicableDead + Live + Environmental Loads (Snow or Wind) If Applicable2.Normal and Operational Basis Earthquake (OBE)Dead + Live + Greater of the OBE or Wind LoadsNADead + Live + UBC LoadsNA3.Normal and Design Basis Earthquake (DBE)Dead + Live +
DBE LoadsDead + Live +
DBE LoadsNANA4.Normal and Pipe RuptureDead + Live + Pipe Rupture Loads Except RCL Pipe BreaksNANANA5.Normal Design Basis Earthquake and Pipe RuptureDead + Live + DBE + Pipe Rupture Loads Except RCL Pipe BreaksNANANANote: NA = Not Applicable1.The replacement steam generator lower units were designed and analyzed to loading combinations defined in Design Specification 414A03, consistent with ASME Code,Section III, Division 1, Subsection NB, Class 1, 1986 Edition through 1987 Addenda. The original steam domes were analyzed in the same manner as the replacement lower units.2.The replacement reactor vessel head was designed and analyzed to loading combinations defined in WCAP-16237-P, Rev 1, Addendum 2, consistent with ASME Code,Section III, Division 1, Subsection N.3, Class 1, 1998 Edition through 2000 Addenda. This methodology was approved by NRC for application to KPS under Letter No. K-04-035, License Amendment 172, dated February 27, 2004.3.The upgraded Auxiliary Building crane is also designed to withstand two-blocking, load hang-up, and broken wire rope without an uncontrolled lowering of the load, in accordance with NUREG-0554.
Revision 25-11/26/14KPS USARB-42Table B.7-2LOADING CONDITIONS AND STRESS LIMITS: PRESSURE VESSELSLoading ConditionsStress Intensity LimitsNote*1.Normal Condition(a) Pm < Sm (b) Pm (or PL) + PB < 1.5 Sm (c) Pm (or PL) + PB + Q < 3.0 Sm122.Upset Condition(a) Pm < Sm (b) Pm (or PL) + PB < 1.5 Sm (c) Pm (or PL) + PB + Q < 3.0 Sm123.Emergency Condition(a) P < 1.2 Sm or Sy, whichever is larger (b) Pm (or PL) + PB < 1.8 Sm, or 1.5 Sy, whichever is larger34.Faulted Condition(a) Stainless Steel Design Limit Curves as given in Figure B.7-2 and Figure B.7-3 (b) Carbon Steel(i) Pm = 1.5 Sm or 1.2 Sy, whichever is larger (ii) Pm (or PL) + PB < 2.25 Sm or 1.875 Sy, whichever is larger4Pm = primary general membrane stress intensity PL = primary local membrane stress intensity PB = primary bending stress intensity Q = secondary stress intensity Sm = stress intensity value from ASME B&PV Code,Section III, Nuclear Vessels Sy = minimum specified material yield strength (ASME B&PV Code,Section III, Table N-424 or equivalent)
- For description of notes, see Notes For Tables B.7-2, B.7-3, And B.7-6.
Revision 25-11/26/14KPS USARB-43Table B.7-3LOADING CONDITIONS AND STRESS LIMITS: PRESSURE PIPINGIN ACCORDANCE WITH USAS B31.1Loading ConditionsStress Limits1.Normal ConditionP < S2.Upset ConditionP < 1.2S3.Emergency ConditionP < 1.5 (1.2S)4.Faulted ConditionFor stainless steelDesign Limit Curves as defined in Figure B.7-3, See Note 4a For carbon steelP < Sy or 1.8S, whichever is higher bWhere:P = Stress S = Allowable stress from USAS B31.1, Code for Power Piping, 1967 Sy = Minimum specified yield strength (ASME B&PV Code,Section III, Table N-424 or equivalent)a.For description of Note 4, see Notes For Tables B.7-2, B.7-3, And B.7-6.b.At some points of high local stress, intensification P may exceed this limit. For such points, local piping deflection will be limited to twice the calculated OBE deflection to ensure no loss of function in the "Faulted Condition."
Revision 25-11/26/14KPS USARB-44Notes For Tables B.7-2, B.7-3, And B.7-6Note 1The limits on local membrane stress intensity (PL < 1.5Sm) and primary membrane plus primary bending stress intensity [PM (or PL) + PB < 1.5SM] need not be satisfied at a specific location if it can be shown by means of limit analysis, or by tests, that the specified loadings do not exceed two-thirds of the lower bound collapse load as per paragraph N-417.6 (b) of the ASME B&PV Code,Section III, Nuclear Vessels.Note 2In lieu of satisfying the specific requirements for the local membrane stress intensity (PL < 1.5Sm), or the primary plus secondary stress intensity (PL + PB + Q < 3SM) at the specific location, the structural action may be calculated on a plastic basis and the design will be considered to be acceptable if shakedown occurs, as opposed to continuing deformation, and if the deformations which occur prior to shakedown do not exceed specified limits, as per paragraph N-417.6 (a) (2) of the ASME B&PV Code,Section III, Nuclear Vessels.Note 3The limits on local membrane stress intensity (PL < 1.5Sm) and primary membrane plus primary bending intensity [Pm (or PL) + PB < 1.5SM] need not be satisfied at a specific location if it can be shown by means of limit analysis, or by test, that the specified loadings do not exceed 120 percent of two thirds of the lower-bound collapse load as per paragraph N417.10 (c) of the ASME B&PV Code,Section III, Nuclear Vessels.Note 4 aAs an alternate to the design limit curves which represent a pseudoplastic instability analysis, a plastic instability analysis may be performed in some specific cases considering the actual strain-hardening characteristic of the material, but with yield strength adjusted to correspond to the tabulated value at the appropriate temperature in Table N-424 or N-425, as per paragraph N-417.11 (c) of the ASME B&PV Code,Section III, Nuclear Vessels. These specific cases will be justified on an individual basis.a.This alternate design procedure was not utilized on this application.
Revision 25-11/26/14KPS USARB-45Table B.7-4LOADING CONDITIONS AND STRESS LIMITS: EQUIPMENT SUPPORTSLoading ConditionsStress Limits1.Normal ConditionWorking stresses or applicable factored load design values2.Upset ConditionWorking stresses or applicable factored load design values 3.Emergency ConditionWithin yield after load redistribution to maintain supported equipment within emergency condition stress limits4.Faulted ConditionPermanent deflection of supports limited to maintain supported equipment within faulted condition stress limitsTable B.7-5LOAD COMBINATION AND STRESS LIMITS FOR CLASS I COMPONENTSLoad CombinationStress Limit1.Normal** (deadweight, thermal and pressure)Normal Condition2.Normal and Operational Basis EarthquakeUpset Condition3.Normal and Design Basis EarthquakeFaulted Condition*4.Normal and Pipe RuptureFaulted Condition5.Normal and Design Basis Earthquake and Pipe RuptureFaulted Condition*This load combination may be evaluated by the emergency condition stress limit.**For Class I piping, stresses due to restrained thermal expansion are treated in accordance with USAS B31.1.0-1967, Power Piping.
Revision 25-11/26/14KPS USARB-46Table B.7-6ALTERNATIVE DESIGN LOADING COMBINATIONS AND STRESS LIMITS:PRESSURE CLASS 1, 2, AND 3 PIPING IN ACCORDANCE WITH ASME SECTION IIIConditionASME Section III Code ClassDesign Loading CombinationsPrimary Stress Pm(P1) + Pb Equation 9Primary + Secondary Stress Pm(P1) + Pb + Q Equation 10Peak Stress Pm(P1) + Pb + Q + F Equation 14Normal and Upset1 (NB-3600)Design pressure, weight, OBE, and other mechanical loads (Equation 9) Pressure, Thermal Expansion and Thermal Gradients (steady-state and transient) (Equation 10, 14)1.8Sm but not greater than 1.5Sy (See Notes 1 & 2) 3Sm (See Notes 1 & 2)Salt = KSp/2 Cumulative usage factor less than 1Emergency1 (NB-3600) Design pressure, weight, DBE, and other mechanical loads (Equation 9)2.25 Sm but not greater than 1.8Sy (See Notes 3 & 4) N/AN/AFaulted1 (NB-3600)Design pressure, weight, DBE, and other mechanical loads (Equation 9) 3Sm but not greater than 2.0 Sy (See Notes 3 & 4)N/AN/ADesign Loading CombinationsNormal and upset2 (NC-3600) and 3 (ND-3600)Design pressure, weight and other sustained loadsDesign pressure, sustained loads, OBE and other occasional mechanical loadsThermal expansionAllowable1.5 Sh 1.8 Sh but not greater than 1.5Sy(1.25Sc + 0.25Sh)f + Sh - (Slp + Sdl)Emergency/ Faulted2 (NC-3600) and 3 (ND-3600)N/AOperating pressure, sustained loads, DBE and other occasional mechanical loadsN/AAllowable2.25 Sh but not greater than 1.8 Sy (Level C) 3Sh but not greater than 2Sy (Level D) Note:The nomenclature, conditions, and applications of the above limits are in accordance (with post-1980 editions approved by NRC) ASME,Section III, Boiler and Pressure Vessel Code, Sub articles NB-3000, NC-3000 and ND-3000. For description of Notes 1, 2, 3, and 4 see Notes for Table B.7-2, Table B.7-3, and Table B.7-4. If the Class 1 NB-3600 allowables are not met the component may be qualified by NB-3200. Plastic Analysis may be performed per ASME NB-3200. Operability when exceeding these requirements may be based on Section III Appendix F criteria.
Revision 25-11/26/14KPS USARB-47Figure B.7-1TYPICAL STRESS STRAIN CURVE Revision 25-11/26/14KPS USARB-48Figure B.7-2COMPARISON BETWEEN DESIGN AND COLLAPSE CONDITIONS HOOP STRESS: 0.90 Sy Revision 25-11/26/14KPS USARB-49Figure B.7-3COMPARISON BETWEEN DESIGN AND COLLAPSE CONDITIONS HOOP STRESS: 0.00 Sy Revision 25-11/26/14KPS USARB-50B.8PROTECTION AGAINST CRANE TOPPLING AND CONTROL OF HEAVY LOADSB.8.1 Protection Against Crane TopplingThe Auxiliary Building crane and the Turbine Building crane are located in areas where they are subject to possible damage from tornado and earthquake. These crane bridges and trolleys are protected against tipping, derailment, and uncontrolled movements that could possibly create damage.To assure stability of the Turbine Building crane, the bridge and trolley are equipped with fixed, fitted rail yokes that allow free rolling movement but prevent the wheels from being lifted or derailed. The Auxiliary Building crane trolley is prevented from derailing by restraints that trap it between the bridge girders. The bridge and trolley wheels are equipped with electrically activated, spring set brakes. Upon loss of power or when the crane or trolley are not under operator control, the springs activate the brakes, locking the wheels firmly in place to prevent rolling out of position. The positive wheel stops and bumpers provided to prevent over-travel of the trolley and bridge will prevent the trolley and bridge from leaving the rails, even in the unlikely event of brake failure.B.8.2 Control of Heavy LoadsAs a result of Generic Task A-36, "Control of Heavy Loads Near Spent Fuel," the NRC issued NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants." NUREG-0612 was to be implemented in two phases. Phase I addressed Section 5.1 of NUREG-0612 and established seven basic guidelines for all nuclear power plants, which detailed provisions for the handling of heavy loads in the area of the reactor vessel near stored spent fuel, in other areas where an accidental load drop could damage equipment required for safe shutdown or decay heat removal. The following cranes are subjected to the seven guidelines of NUREG-0612 Phase I:1.Turbine Building Crane2.Auxiliary Building Fuel Handling Crane3.DeletedThe seven basic guidelines of NUREG-0612, Phase I listed below are satisfied for the above listed cranes.1.Safe Load Paths 2.Load Handling Procedures3.Crane Operator Training 4.Special Lifting Devices Revision 25-11/26/14KPS USARB-515.Lifting Devices (Not Specially Designed) 6.Cranes (Inspection, Testing, and Maintenance)7.Crane DesignThe spent fuel pool bridge and hoist crane has the capability of carrying loads which could, if dropped, fall into the spent fuel pool. However, based on the use of these cranes, they have been excluded from further review against NUREG-0612.The NRC has determined that Kewaunee has adequately addressed NUREG-0612 and has significantly reduced the probability of a heavy load handling accident to an acceptably small value (see NRC Safety Evaluation Report in Reference 16).B.8.3 Design Criteria for Upgraded Auxiliary Building CraneThe Auxiliary Building (AB) crane was upgraded in support of dry spent fuel storage cask loading operations. This upgrade involved the replacement of the original trolley with a single-failure-proof design, replacement of the trolley controls, and an upgrade to the existing AB crane bridge. The upgrade of the AB crane meets the guidance in Section 5.1.6 of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," and NUREG-0554, "Single Failure Proof Cranes for Nuclear Power Plants," as applicable.The AB crane is designated as Class I* per Table B.2-1 and therefore is designed to meet Class I seismic standards. The crane is designed to stay on its rails and not allow an uncontrolled lowering of the load as a result of a seismic event. It is not required to be operational during or after a seismic event. The AB crane is also designed to withstand the crane design basis accident events described in NUREG-0554: two-blocking, load hang-up, and wire rope failure.Because the replacement AB crane trolley is a new component and the crane bridge is an existing component, the construction codes applicable to the two are not identical. The construction codes for the trolley and bridge are as follows:AB Crane Trolley Codes and StandardsConstruction is in accordance with NUREG-0554 and, where NUREG-0554 does not offer specific guidance (e.g., normal condition load combinations and stress acceptance criteria), construction is in accordance with Crane Manufacturers Association of America Specification 70 (CMAA-70), 2004 Edition. Seismic load combinations and stress analysis acceptance criteria, as well as guidance used to address two-blocking, load hang-up, and wire rope failure are taken from ASME NOG-1-2004.AB Crane Bridge Codes and Standards Revision 25-11/26/14KPS USARB-52Construction is in accordance with NUREG-0554 and Electrical Overhead Crane Institute Standard 61 (EOCI-61), CMAA-70 (2004), and ASME NOG-1-2004 in that hierarchy, where NUREG-0554, and EOCI-61 do not offer specific construction guidance.A-B Crane Seismic Response Spectra, Damping and AccelerationsThe seismic analysis of the AB crane considers trolley and bridge drive wheel rolling when the seismic forces exceed the drive wheel brake resisting force. This nonlinear boundary condition required seismic time history inputs to be developed consistent with Standard Review Plan (SRP), NUREG-0800, Section 3.7.1, Revision 3, Option II. With the exception of the nonlinear boundary condition at the trolley and bridge drive wheels, the seismic analysis of the upgraded AB crane is consistent with ASME NOG-1-2004.The Blume Report, which forms the basis for seismic analyses at the Kewaunee Power Station, does not include horizontal response spectra data for a mass point at the location of the AB crane rail appropriate for use in analyzing the upgraded crane. Therefore, a lumped-mass stick model of the AB steel structure was used to generate additional horizontal response spectra applicable for use at the AB crane rail. Two percent damping for the Safe Shutdown Earthquake condition was applied to both the vertical and horizontal spectra at the crane rail elevation.Five sets of seismic acceleration time histories were then developed representing the response of the AB crane at the base of the crane bridge rails (Reference 44). Each set of time histories contains two horizontal and one vertical time history, for a total of fifteen time histories. The time histories were used in conjunction with a 3-D model of the crane to perform the nonlinear seismic analysis. The methodology for analyzing the response of the crane during a seismic event was based on the application of the commercially available finite element analysis computer program. SAP 2000, Version 11. The use of SAP 2000 was reviewed pursuant to SRP Section 3.9.1, "Special Topics for Mechanical Components," which provides applicable criteria for evaluating computer programs for mechanical and structural design and analysis.The NRC approval of the seismic methodology for the AB crane (Reference 45) is subject to the following limitations:1.The analyses are based on the seismic acceleration time histories reported in the license amendment request submittal dated July 7, 2008 (Reference 44).2.The calculated critical wheel tractions should be increased by 25 percent for the crane drive wheels and 100 percent for the trolley wheels.3.The seismic methodology use is limited to the AB crane.
Revision 25-11/26/14KPS USARB-53B.9DELETEDB.9.1 DeletedB.9.2 Deleted B.9.3 DeletedB.9.4 Deleted Revision 25-11/26/14KPS USARB-54Table B.9-1DELETED Revision 25-11/26/14KPS USARB-55B.10DELETEDB.10.1 DeletedB.10.2 Deleted B.10.3 DeletedB.10.4 DeletedB.11INTERNAL FLOODING B.11.1 DeletedB.11.2 Flooding Design CriteriaThe plant must withstand the consequences of an internal flooding event in such a manner that it retains the capability to achieve and maintain the reactor in a safe shutdown condition and to limit the consequences of a design basis accident.a.Deletedb.Deletedc.Deletedd.Deletede.Deletedf.DeletedB.11.3 DeletedB.11.4 DeletedB.11.5 ConclusionOn May 7, 2013, Dominion Energy Kewaunee, Inc. (DEK) submitted the second of two letters required, pursuant to 10 CFR 50.82(a)(1)(i) and 10 CFR 50.82(a)(1)(ii), to certify that it has permanently ceased power operation of KPS, and that the reactor was permanently defueled. Therefore, as specified in 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for Kewaunee Power Station no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel.
Revision 25-11/26/14KPS USARB-56Additionally, the current design basis of the station includes no SSCs that are credited with functions necessary to mitigate a design basis accident and are vulnerable to damage from the rupture of a tank or pipe.As stated previously, the design criteria established for appropriately addressing the consequences of internal flooding (i.e., the effects of the rupture of a tank or pipe) are limited to ensuring that the station design retain the capability to:*Achieve and maintain the reactor in a safe shutdown condition and*Limit the consequences of a design basis accidentWith the permanent cessation of plant operation, and the prohibition of emplacement of fuel within the reactor vessel, the capability to "achieve and maintain the reactor in a safe shutdown condition has been permanently achieved.Similarly, with no SSCs that are credited with design basis accident mitigation being vulnerable to damage from internal flooding, that design criterion is also permanently met.Therefore, no design features of the station need be retained based solely upon their contribution to meeting the internal flood mitigation design criteria.
Revision 25-11/26/14KPS USARB-57Figure B.11-1DELETED Revision 25-11/26/14KPS USARB-58Figure B.11-2DELETED Revision 25-11/26/14KPS USARB-59B.12DELETEDB.12.1 DeletedB.12.2 Deleted B.12.3 DeletedREFERENCES1.Morris, Hansen, Holley, Biggs, Namyet, and Minami, Structural Design for Dynamic Loads, McGraw-Hill Co., Inc., New York, 1959.2.Deleted3.Housner, George W., Vibration of Structures Induced by Seismic Waves, Shock and Vibration Handbook, Volume III, McGraw-Hill, Inc., New York, 1961.4.Vogeding, E. L., Topical Report, Seismic Testing of Electrical and Control Equipment, WCAP 7817, December 1971.5.Deleted 6.Deleted7.Deleted8.John A. Blume & Associates, Engineers, Kewaunee Nuclear Power Plant-Earthquake Analysis of the Reactor-Auxiliary-Turbine Building, JAB-PS-01, February 16, 1971, (submitted as part of Amendment No. 9 to this license application).9.John A. Blume & Associates, Engineers, Kewaunee Nuclear Power Plant-Earthquake Analysis: Reactor-Auxiliary-Turbine Building Response Acceleration Spectra, JAB-PS-03, February 16, 1971 (submitted as Amendment No. 9 to this license application).10.Deleted11.Deleted12.Deleted 13.Deleted14.Supplement No. 1 to Generic Letter (GL) 87-02 which transmits Supplemental Safety Evaluation Report No. 2 (SSER No. 2) on SQUG Generic Implementation Procedure, Revision 2 as corrected on February 14, 1992 (GIP-2), May 22, 1992.15.Deleted Revision 25-11/26/14KPS USARB-6016.NRC Safety Evaluation Report, S. A. Varga (NRC) to C.W. Giesler (WPS), Letter No. K-84-61, March 16, 1984.17.Letter from C. R. Steinhardt (WPSC) to the NRC Document Control Desk, September 17, 1992.18.Letter from C.R. Steinhardt (WPSC) to the NRC Document Control Desk, February 18, 1993.19.Seismic Qualification Utility Group (SQUG), Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Power Plant Equipment, Revision 2 as corrected February 14, 1992.20.Deleted21.Deleted22.Deleted23.Deleted 24.Letter from W. O. Long (NRC) to M. L. Marchi (WPSC), Kewaunee Nuclear Power Plant-Safety Evaluation Report for USI A-46 Program Implementation, Letter No. K-98-47, April 14, 1998.25.Seismic Qualification Utility Group (SQUG), Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Power Plant Equipment, Revision 3, May 16, 1997.26.Supplemental Safety Evaluation Report No. 3 (SSER No. 3) on the Review of Revision 3 to the Generic Implementation Procedure for Seismic Verification of Nuclear Power Plant Equipment updated May 16, 1977, (GIP-3), (TAC No. M93624).27.License Application Amendment 17 dated May 12, 1972 from E. W. James (WPS) to P.A. Morris (AEC).28.License Application Amendment 24 dated January 24, 1973 from E. W. James (WPS) to J. F. O'Leary (AEC).29.License Application Amendment 27 dated March 16, 1973 from E. W. James (WPS) to J. F. O'Leary (AEC).30.License Application Amendment 28 dated April 13, 1973 from E. W. James (WPS) to J. F. O'Leary (AEC).31.Safety Evaluation of Kewaunee Nuclear Power Plant, Supplement 2 dated July 24, 1972.32.NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Reactors (LWR Edition) dated July 1981.33.Letter October 31, 1972 to R. C. DeYoung (NRC) from E. W. James (WPS).
Revision 25-11/26/14KPS USARB-6134.Deleted35.NRC Safety Evaluation Report, S.A. Varga (NRC) to C.W. Giesler (WPSC), Letter No. K-84-204, September 21, 1984.36.Missile Protection Criteria Prairie Island Nuclear Generating Plant and Kewaunee Nuclear Power Plant Westinghouse Letter PIW-P-16, KW-P-20 dated July 31, 1967.37.Operability Determination Closure Request for OBD 135-EDG Exhaust Ducts dated April 27, 2008.38.MEMO FPE 2007-0100 Evaluation of KPS Main Steam Safety Valves and Steam Generator Power Operated Relief Valves in a Design Basis Tornado Event dated January 9 2008.39.McDonald - Mehta Engineers Letter Report, Tornado Effects on Turbine Building and Diesel Generator Exhaust Lines dated April 28, 2005.40.Letter from Steven A. Varga (NRC) to CW Giesler (WPSC) Subject Control of Heavy Loads -NUREG-0612-Phase II dated June 13, 1984.41.Deleted 42.Deleted 43.Deleted44.License Amendment Request 239, Request for Review and Approval of Seismic Analysis Methodology for Auxiliary Building Crane, July 7, 2008 (includes seismic time histories).45.NRC Safety Evaluation Report, P.S. Tam (NRC) to D.A. Christian (Dominion), Kewaunee Power Station, Issuance of Amendment Re: Seismic Analysis Methodology for the Auxiliary Building Crane, April 30, 2009.46.License Application Amendment 32 dated August 31, 1973, from E.W. James (WPS) to J.F O'Leary (AEC).47.NRC Generic Letter 89-10: Safety-Related Motor-Operated Valve Testing and Surveillance, dated June 28, 1989, including Supplements 1 through 7.48.NRC letter to WPSC: Close-Out of Generic Letter (GL) 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance, dated January 4, 1996.49.Letter from Thomas T. Martin (NRC) to All Holders of Operating Licenses, NRC Generic Letter 96-05: Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves, September 18, 1996.50.Joint BWR, Westinghouse and Combustion Engineering Owners' Group Program on Motor-Operated Valve Periodic Verification, MPR-1807, Revision 2, July 1997.
Revision 25-11/26/14KPS USARB-6251.Safety Evaluation on Joint Owners' Group Program on Periodic Verification of Motor-Operated Valves Described in Topical Report NEDC-32719, Revision 2 (MPR-1807, Revision 2), October 30, 1997.52.Tae Kim (NRC) to Mark L. Marchi (WPSC), Kewaunee Nuclear Power Plant - Closure of Generic Letter 96-05, Periodic Verification of Design-Basis Capability of Safety Related Motor-Operated Valves and Safety Evaluation by the Office of Nuclear Reactor Regulation Relating to Response to Generic Letter 96-05, Periodic Verification of Design-Basis Capability of Safety Related Motor-Operated Valves, December 16, 1999.53.Joint Owners' Group (JOG) Motor-Operated Valve Periodic Verification Program Summary, MPR-2524-A, Revision 1, September 2010.54.Final Safety Evaluation on Joint Owners' Group Program on Motor-Operated Valve Periodic Verification, September 25, 2006.55.Deleted56.Deleted