ML14339A653

Updated Safety Analysis Report (Usar), Rev 25 - Appendix B - Special Design Procedures

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: ML14339A613 ML14339A614 ML14339A615 ML14339A616 ML14339A618 ML14339A621 ML14339A622 ML14339A624 ML14339A648 ML14339A649 ML14339A650 ML14339A651 ML14339A652 ML14339A653 ML14339A654 ML14339A655 ML14339A656 ML14339A674 ML14339A676
References
14-572
Download: ML14339A653 (68)

Text

Appendix B Special Design Procedures Intentionally Blank Table of Contents tion Title Page 1 Scope . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 2 Classification of Structures and Components. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 B.2.1 Definition of Nuclear Safety Design Classifications (NSDC) . . . . . . . . . . . . B-2 3 Design Codes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-10 4 Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-11 B.4.1 Environmental Loads. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12 B.4.2 Tornado Loads. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12 B.4.3 Live Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12 B.4.4 Dead Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-12 B.4.5 Seismic Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-13 B.4.6 Design Basis Accident (DBA) Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-13 B.4.7 Other Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-13 B.4.8 Seismic Design and Verification of Modified, New and Replacement Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-13 5 Protection of Class I Items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-14 6 Design Criteria for Structures. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-15 B.6.1 Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-15 B.6.2 Stress Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-17 B.6.3 Structural Design Basis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-17 7 Design Criteria for Components. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-33 B.7.1 Load Combinations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-33 B.7.2 Design Criteria. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-34 8 Protection Against Crane Toppling and Control of Heavy Loads . . . . . . . . . . . . . . . B-50 B.8.1 Protection Against Crane Toppling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-50 B.8.2 Control of Heavy Loads . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-50 B.8.3 Design Criteria for Upgraded Auxiliary Building Crane . . . . . . . . . . . . . . . B-51 9 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53 B.9.1 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53 B.9.2 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53 B.9.3 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53 B.9.4 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-53

Table of Contents (continued) tion Title Page 0 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.10.1 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.10.2 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.10.3 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.10.4 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 1 Internal Flooding . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.11.1 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.11.2 Flooding Design Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.11.3 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.11.4 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 B.11.5 Conclusion. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-55 2 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59 B.12.1 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59 B.12.2 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59 B.12.3 Deleted. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59 References . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-59

List of Tables le Title Page

-1 Classification of Structures, Systems and Components . . . . . . . . . . . . . . . . . . B-4

-1 Load Combinations for Structures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-25

-2 Applicable Code Stresses Class I Structures:

Reinforced Concrete - Structural Steel . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-27

-3 Applicable Code Stresses: Class I Structures . . . . . . . . . . . . . . . . . . . . . . . . . . B-28

-4 Allowable Stresses: Class I*, II, III*, III and IV Structures . . . . . . . . . . . . . . . B-29

-5 Damping Factors. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-30

-6 Tornado-Generated Missiles. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-31

-7 Internally-Generated Missiles Inside Of Containment . . . . . . . . . . . . . . . . . . . B-32

-1 Load Combinations For Components Class Of Components . . . . . . . . . . . . . . B-41

-2 Loading Conditions and Stress Limits: Pressure Vessels . . . . . . . . . . . . . . . . . B-42

-3 Loading Conditions and Stress Limits: Pressure Piping in Accordance with USAS B31.1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-43

-4 Loading Conditions and Stress Limits: Equipment Supports . . . . . . . . . . . . . . B-45

-5 Load Combination and Stress Limits for Class I Components. . . . . . . . . . . . . B-45

-6 Alternative Design Loading Combinations and Stress Limits:

Pressure Class 1, 2, and 3 Piping In Accordance With ASME Section III . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-46

-1 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-54

List of Figures ure Title Page

-1 Typical Stress Strain Curve . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-47

-2 Comparison Between Design and Collapse Conditions Hoop Stress: 0.90 Sy B-48

-3 Comparison Between Design and Collapse Conditions Hoop Stress: 0.00 Sy B-49 1-1 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-57 1-2 Deleted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-58

Special Design Procedures SCOPE special design procedures contained in this appendix apply to all structures, systems luding instruments and controls), and all components.

CLASSIFICATION OF STRUCTURES AND COMPONENTS structures, systems (including instruments and controls), and components are classified as ss I, I*, II, III, III* or IV according to their function and importance in relation to the safe ommissioning of the facility, with emphasis on the degree of integrity required to protect the lic. These are listed in Table B.2-1.

Turbine Building, Administration Building, Auxiliary Building and Shield Building ctures are constructed as a contiguous complex. In general, these structures are identified as er Class I or Class III by placing emphasis on the predominant use of the structure in its tion to the safe decommissioning of the station.

ome instances there may be more than one classification applicable within a building or cture. This situation is treated as a mixed classification.

ividual components or portions of a system may be determined to have a different sification than the system as a whole. This determination would be accomplished by s i d e r i n g d e s i g n a n d f u n c t i o n a l i t y r e q ui r e m e n t s o f b o t h t h e s y s t e m a n d t h e ponents/sub-components, consistent with the 10 CFR 50, Appendix B program for the waunee Power (KPS).

definition of the nuclear safety design classifications is given in the following paragraphs1:

Class I Those 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.

Class II Those 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.

Class III Those 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.

Mixed Classification This 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.

Class I - Part Class III The 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.

or clarity and continuity, the NSDC definitions have not been revised to reflect the permanent shutdown of the tation.

A substantial amount of radioactivity is defined as that amount of radioactive material, which would produce adiation levels at the site boundary in excess of 1.0 percent of 10 CFR 100 guidelines.

eactor operation is defined as the condition where the reactor is producing only that power required to maintain he Reactor Coolant System (RCS) at normal operating pressure and temperature.

The 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 Compressors

b. Mezzanine Floor

• Batteries Class I designation applies to the walls, floors, ceilings, structural support and foundations of ctures that isolate, support, or are associated with the protection of Class I equipment.

esponse to Bulletin 80-11, which identified NRC concerns regarding the structural integrity of ty-related masonry walls a detailed study was performed to provide a technical evaluation of plants masonry walls, that at the time were classified as safety-related. The basic documents guidance in this review were the criteria developed by the Structural and Geo-Technical ineering Branch of the NRC. The review concluded that the safety-related masonry walls in plant could withstand the loads and load combinations, as specified in the USAR, without eeding allowable stress limits (see NRC Safety Evaluation Report in Reference 35).

or 142 and Door 143 were modified per DCR 3594 to allow venting of Room 302 and ridor 304 in the Auxiliary Building to accommodate atmospheric pressure changes due to a ado. Room 302 and Corridor 304 are partially enclosed with Class I masonry block walls that not designed to withstand differential pressure loads from a tornado. Venting of Room 302 and ridor 304 will minimize the pressure loads on the affected masonry block walls and help ure the structural integrity of the walls during a tornado.

or 49 was modified with breakaway pins per DCR 3597 to provide relay room block wall tection from Atmospheric Pressure Change (APC) which could result from design tornado s as specified in USAR Appendix B, Section B.4.2, Item 1.

CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTS e: 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.

e: Some abandoned systems, structures, or components (SSC) will continue to be listed in this table.

m Class Classification of Buildings and Structures actor containment vessel (including all penetrations, air locks, isolation valves, I*

cuum relief devices, and internal containment structures performing Class I function) ield Building (including vent and all penetrations) I*

ent Fuel Pool Structure (including fuel transfer tubes and valves) I ntrol Room I reenhouse (including Access Tunnel and areas housing Service Water, Turbine I ilding Ventilation and Screenhouse Ventilation System Components) ncrete Encased Electrical (Class 1) Screenhouse Conduit Structure (5980) I rculating Water Intake and Discharge Structures I xiliary Building (areas housing Auxiliary Building Special Ventilation System, I waste storage, and Engineered Safety Features) xiliary Building Support System for cranea I*

xiliary Building (except Class I or I*)b III*

rbine Building (areas housing safeguard batteries, safety significant 480V switchgear, I compressor) rbine Building Support System for Turbine Building crane I*

rbine Building (Except Class I or I*) III*

ministrative Building basement (586 feet 0 inch), includes diesel generator room I ergency Diesel Generation Room Air Inlet and Outlet Structures I ministrative Building (first and second floors, 606 feet 0 inch and 626 feet 0 inch) III C basement (586 feet 0 inch) I C upper floors (606 feet 0 inch and 626 feet 0 inch) III*

scellaneous structures III curity Building III fice/Warehouse Annex IV ministration & Training Facility IV Loading Dock IV gmented Water System Building IV

CLASSIFICATION OF STRUCTURES, SYSTEMS AND COMPONENTS e: 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.

e: Some abandoned SSCs will continue to be listed in this table.

m Class Classification of Systems and Components actor Plant Equipment Reactor

- Reactor pressure vessel and its supports I*

- Vessel internals I*

- Fuel assemblies I*

- Rod Cluster Control Assemblies (RCCAs) and drive mechanisms I*

- In-core instrumentation structures I*

Reactor Coolant System

- Piping and valves containing full system pressure (including safety and relief I*

valves)

- Steam generators I*

- Pressurizer (excluding pressurizer relief tank, piping downstream of pressurizer I*

relief and safety valves)

- Reactor coolant pumps I*

- Supporting and positioning members I*

- Primary Sampling System (up to second isolation valve) I*

- Pressurizer relief tank and piping (downstream of pressurizer relief valves) II ergency Core Cooling System

- Safety Injection System (including Accumulator Tanks, Safety Injection Pumps, I*

Residual Heat Removal Pumps (RHR), Refueling Water Storage Tank, RHR Heat Exchangers (RHR), and Primary Connecting Piping and Valving) sidual Heat Removal System I*

ernal Containment Spray System (including spray pumps, spray ring headers, and I*

mary connecting piping and valving) mary Sampling System (beyond second isolation valve) III mponent Cooling System I*

actor Control and Protection System I*

diation Monitoring System (to the extent that it must function in support of Class I I*

uipment)

e: 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.

e: Some abandoned SSCs will continue to be listed in this table.

m Class Classification of Systems and Components (continued) ergency Power Supply System

- Diesel Generators I*

- Fuel Oil Storage Tank I*

- Diesel Generator Cooling System I*

- Safety features buses I*

- Emergency Load Distribution System I*

- DC power supply, batteries, cable I*

- Diesel Generator Fuel Oil Vent Lines I*

- Fuel Oil Supply Lines to Day Tanks I*

trumentation

- 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 III Protection System, which is Class I) clear Fuel Handling and Storage

- New Fuel Storage Racks I*

- Spent Fuel Storage I

- Spent Fuel Pool Liner I*

- Fuel Transfer System (Including Fuel Transfer Carriage, Containment Upender III and Auxiliary Building Upender)

- Spent Fuel Pool Cooling System (Piping and valving whose failure could result in I*

significant release of pool water)

- Spent Fuel Pool Cooling System (portions not Class I) III ntilation Systems Shield Building Ventilation System I*

Auxiliary Building Special Ventilation System (includes Zone SV isolation dampers I*

and boundary ductwork)

Auxiliary Building Air Conditioning System III Auxiliary Building Ventilation System III

e: 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.

e: Some abandoned SSCs will continue to be listed in this table.

m Class Classification of Systems and Components (continued)

- 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 III Class I)

- 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 II segment in the reactor cavity)

- Reactor Support Cooling System II Control Room Air Conditioning System with Service Water System cooling water I*

supply. (Includes Relay Room and Mechanical Equipment Room)

- Control Room Chillers for normal operation III Turbine 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 II emical and Volume Control System

- All items except those listed below. I*

- Boric acid transfer pumps II

- Boric acid filter II

- Boric acid heat tracing II

- Batch tank III*

- Evaporator condensate demineralizers III*

- Condensate filter III*

- Monitor tanks III*

e: 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.

e: Some abandoned SSCs will continue to be listed in this table.

m Class Classification of Systems and Components (continued)

- Monitor tank pumps III*

- Deborating demineralizers III*

- Concentrates holding tank III*

- Concentrates holding tank transfer pumps III

- Chemical mixing tank III

- Resin fill tank III aste Disposal System

- Waste Hold-Up Tank III*

- Sump Tank III*

- Gas Decay Tanks I*

- Reactor Coolant Drain Tank and Pumps II

- Waste Gas Compressor Package I*

- Waste Evaporator Feed Pump III*

- Sump Tank Pumps III*

- Interconnecting Piping and Valves Between Class I Equipment I*

- Waste Evaporatorc III*

- Waste Evaporator Condensate Tanks III*

- Laundry and Hot Shower Tanks III tomatic Gas Analyzer H2 and O2 III*

trogen Supply Manifold III drogen Supply Manifold III scellaneous Reactor Plant Equipment

- Steam Generator Blowdown System upstream of Isolation Valves BT3A and I*

BT3B outside of containment

- Steam Generator Blowdown System downstream of Isolation Valves BT3A III and BT3B

- Polar Crane I*

- Manipulator Crane III

- Fuel Pool Bridge Crane I*

- Auxiliary Building Crane I*

e: 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.

e: Some abandoned SSCs will continue to be listed in this table.

m Class Classification of Systems and Components (continued)

- Turbine Building Crane I*

- All Other Cranes III

- Conventional Equipment, Tanks, Piping (other than Class I and II) III rbine Plant

- Turbine, Generator, Foundations, Exciter, Oil Purification, Turbine Gland Seal III System, Reheaters and Moisture Separators, Hydrogen and CO2 Systems rvice Water System

- Serving Class I equipment I*

- All that is not Class I III ake-Up Water Systems III

- Reactor Make-Up Water Storage Tank III rculating Water System

- Circulating water pumps III

- Intake piping to Screenhouse I*

- Circulating water pump discharge piping III

- Condenser discharge piping III ndensate and Feedwater Systems

- Main Condenser III

- Condensate System III

- Main Feedwater System (excluding Class I piping and isolation valves) III r Removal System III xiliary Feedwater System I*

ain Steam System

- Main Steam System (portions not Class I) III*

- Main steam, safety, relief, and isolation valves I*

- Main steam up to isolation valves including steam piping to Turbine Driven I*

AFW Pump am Dump System III*

e: 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.

e: Some abandoned SSCs will continue to be listed in this table.

m Class Classification of Systems and Components (continued) ating Steam System (those portions in diesel generator rooms, battery rooms, I*

eenhouse, and auxiliary building steam exclusion zones) ater and Moisture Separator Drain System III eed Steam System III condary Sample System III scellaneous Power Systems and Plant Equipment

- Station and Instrument Air System III

- Instrument Air System - Portions required for safe shutdown I*

- 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 System III ansformers

- Main Auxiliary Transformer II

- Reserve Auxiliary Transformer II

- Tertiary Auxiliary Transformer II

- 4.16-0.480 kV safety features transformers I*

- 4.16-0.480 kV auxiliary transformer (other than Class I) III

- Transformer serving pressurizer heaters from safety features bus II For definition of Class I*, refer to Section B.2.1.

For definition of Class III*, refer to Section B.2.1 .

No longer in service.

DESIGN CODES design and construction of this plant has been in accordance with the following codes, as licable:

American Institute of Steel Construction Specification for the Design, Fabrication and Erection of Structural Steel Buildings, 1963 Edition1 American Welding Society Code D 1.0 Standards for Arc and Gas Welding in Building Construction International Conference of Building Officials Uniform Building Code, 1967 Edition Atomic Energy Commission publication TID 7024 Nuclear Reactors and Earthquakes American Society of Mechanical Engineers Boiler and Pressure Vessel Code Piping Code, USAS B31.1.0-1967 with applicable N-code cases to ASA B31.1-19552, 3 Welding Research Council Bulletin No. 107, 1965 Edition Wisconsin Administrative Code: Rules of Department of Industry, Labor & Human Relations Crane Manufacturers Association of America Specification 70, Specifications for Top Running Bridge and Gantry Type Multiple Girder Electric Overhead Traveling Cranes, 2004 Edition.

ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), 2004 Edition.

Electrical Overhead Crane Institute (EOCI) Standard 61.

NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, dated July 1980.

NUREG-0554, Single Failure Proof Cranes for Nuclear Power Plants, dated May 1979.

LOADS Structures and Components in this plant are designed to withstand various kinds and binations of loads.

different kinds of loads treated in the design are described in the subsequent paragraphs.

later edition may be used for plant physical changes provided appropriate reconciliation is documented.

An alternative Design Code to USAS B31.1 is ASME Section III (Post 1980 Editions Approved by NRC, eference Table B.7-6).

uring RFO 28 tubing for penetrations 1, 3, 21, 27E, 27EN, 27N, 27NE, and 36, located between containment nd the shield building, was analyzed to ASME Section III, reference Table B.7-6. Analyses were performed to econcile thermal stresses that may occur during sampling and differences in displacement of the containment and hield buildings due to annual temperature variations and periodic ILRT testing.

mponents.

.1 Environmental Loads se consist of wind and snow loads.

Snow Load A 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 Load The 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.

.2 Tornado Loads nado loadings used in design consist of the following:

A differential pressure equal to 3 psi. This pressure is assumed to build up from normal atmospheric pressure in 3 seconds.

A lateral force caused by a funnel of wind having a peripheral tangential velocity of 300 mph and a forward progression of 60 mph.

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.

plant site was examined for possible sources of other missiles including building and ipment parts which were evaluated to determine the potentially most damaging missile. The lt of this study is reported in Section B.6.3 of this appendix. The conclusion of this study is no other missiles are as damaging as the design missiles given above.

.3 Live Loads ipment loads are specified from manufacturers drawings and floor loads are based upon the nded use of the floor.

.4 Dead Loads d loads consist of the weight of structural steel, concrete, dead weight of the component, etc.,

omputed for each case.

eral different seismic loads were used in the design of this plant.

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.

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.

Uniform Building Code Earthquake Loads The 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.

.6 Design Basis Accident (DBA) Loads DBA for this plant was the instantaneous double-ended rupture of the cold leg of the RCS.

s accident transmits loads to structures and equipment, which were designated as DBA loads.

he permanently shutdown and defueled condition, DBA loads, as defined above, are reduced ero.

.7 Other Loads ddition to all the above loads listed, other loads were used in the design wherever applicable.

ong these were ice loads, jet forces, other pipe rupture loads, etc.

.8 Seismic Design and Verification of Modified, New and Replacement Equipment February 19, 1987 the NRC issued Generic Letter (GL) 87-02, Verification of Seismic quacy of Mechanical and Electrical Equipment in Operating Reactors, Unresolved Safety e (USI) A-46.

Seismic Qualification Utility Group (SQUG) developed Generic Implementation Procedure P) For Seismic Evaluation of Nuclear Plant Equipment (Reference 19 and Reference 25), as dified and supplemented by the U.S. Nuclear Regulatory Commission Supplemental Safety luation Report Nos. 2 and 3 (Reference 14 and Reference 26), which may be used as an rnative to existing methodologies for the seismic design and verification of modified, new or

d provided they have received USNRC review and approval.

SQUG methodology may only be applied to certain classes of active mechanical and trical equipment as specified in the SQUG GIP (Reference 19 and Reference 25), electrical ys, new and replacement cable and conduit raceway systems, tanks and heat exchangers. The criteria and procedures may be applied to modification or repair of existing anchorage chor bolts or welds) including one-for-one component replacements and new anchorage gns. However, for new installations and new anchorage designs, the factor of safety currently mmended for new nuclear power plants shall be met when determining allowable anchorage s.

specified in the GIP, and accepted by the NRC in Reference 14 and Reference 26, the use of ercent damped response spectra may be used when performing seismic evaluations in ordance with the GIP. However, as stated in Reference 18, if it is determined that the ipment natural frequency is within +/- 25 percent of the frequency associated with the peak eleration, the peak acceleration will be used as the input motion for that piece of equipment.

sequently, in 1998 the NRC accepted Kewaunees USI A-46 program implementation as cribed in the NRC Safety Evaluation Report in Reference 24.

PROTECTION OF CLASS I ITEMS 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. Deleted

c. 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.

turbines.)

(iii.) Deleted protection is required if the factors described under item a (non-HELB), item f, and item g not affect any Class I systems, or if redundant systems are provided and the physical aration of these systems is sufficient to prevent these factors from damaging both systems.

er item c and item d, redundancy and physical separation may decrease the requirements for ection. If redundancy and physical separation are not used, and if the surrounding building is designed as a missile barrier, missile protection by shielding is necessary, either by shielding source itself or by shielding the system.

DESIGN CRITERIA FOR STRUCTURES s section describes the general Design Criteria for Structures used in the plant design. Special ations like protection against crane toppling, etc., are given towards the end of this appendix in arate paragraphs.

.1 Load Combinations load combinations applicable to Class I, Class I*, Class II, Class III*, Class III, and Class IV ctures are given in detail in subsequent paragraphs and are also listed in Table B.6-1.

ss I Structures ss I Structures are analyzed for each of the following conditions of loading:

Normal Operating Conditions The load combinations consist of Dead and Live loads together with the environmental loads (wind and snow) as specified in Section B.4.1.

Operational Basis Earthquake Conditions The combination consists of Dead, Live, DBA, and snow loads together with the greater of the OBE or Wind loads.

Design Basis Earthquake Conditions The load combination consists of Dead, Live, Snow, and DBA loads together with the DBE loads.

Tornado Condition

design tornado and tornado missile loads, if any. These loads are assumed non-coincident with DBA or Seismic loads.

ddition to the above four conditions, windrowed ice loading was considered for the following ss I structures:

The screenhouse water intake structure is located inland from the shore of the Lake Michigan and is not therefore subjected to ice loading.

The circulating water intake structure is designed for ice loading.

ss I* Structures se structures designated as Class I* were analyzed for each of the following conditions of ing:

Normal Operating Conditions The load combination consists of Dead and Live loads together with environmental loads (wind and snow).

Design Basis Earthquake Conditions The load combination consists of Dead, Live, Snow and the DBE loads.

ss II Structures ss II Structures were designed for the greater of the following two combinations of loads:

Dead, Live, and Environmental loads (wind and snow), or Dead, Live, and Uniform Building Code earthquake loads specified in Section B.4.5 of this Appendix.

ss III* Structures ss III* Structures were designed for the greater of the two combinations of loads given above Class II Structures.

ss III Structures ss III Structures were designed for Dead and Live loads together with environmental loads nd and snow).

a minimum, Class III Structures were designed in accordance with the applicable codes as ed in Section B.3. In accordance with the Wisconsin Public Service Corporations normal cy for the design of steam-electric generating stations, certain items of power plant structures he Class III category were designed according to the requirements of a higher classification.

ss IV structures were designed for Dead and Live loads together with environmental loads nd and snow) in accordance with the State of Wisconsin Administrative Code.

.2 Stress Design Criteria mal Operation allowable stress design criteria that were applied for normal operating conditions were in ordance with the applicable code(s) listed in Section B.3. These code allowable stresses are marized in Table B.6-2, Table B.6-3, and Table B.6-4.

ign Basis Accident and Operational Basis Earthquake allowable stress design criteria applied for the DBA condition in combination with the OBE e that stresses remain within the allowable limits specified by the applicable code(s) listed ein, except that allowable stresses were not increased for the earthquake condition as is mitted by some codes. These code allowable stresses are summarized in Table B.6-2, le B.6-3, and Table B.6-4.

e Shutdown design criteria for tornado missiles, the 300-mph design tornado condition, and also for the A in combination with the DBE were that the reactor can be safely shut down and that there be ncontrolled release of radioactivity.

meet these criteria, structures or components were examined for their function in the total em to assure a safe and orderly shut down.

se criteria, as applied to tornado winds, and to the DBA condition in combination with DBE s, will permit some permanent deformation but will not permit loss of structural function. In sense, structural function is defined to mean that structures will remain intact and continue to port their normal operating loads after an earthquake and/or DBA, but may require repair or acement for future continued use.

nado missiles may result in large local deformations, but the criteria will not permit the missile reach the barrier so that essential safety features functions are jeopardized.

.3 Structural Design Basis ss I Structures designs of Class I structures for seismic, tornado winds, tornado missiles, etc., are given in sequent paragraphs.

dynamic analysis, an equivalent multi-mass mathematical model was constructed to roximate the structural system. The effect of the foundation soils was included in the model by ns of equivalent springs. The spectral method was then used to determine the maximum onse of each mass point for each node, using as input the OBE (Plate 8 in Appendix A) and damping factors given in Table B.6-5. The total response for each point was determined by the t-mean-square method. From this, a set of curves was developed which show the variation h height of the maximum translational accelerations, displacements, shears and moments in structure. All of the above work was performed by John A. Blume and Associates and is orted in detail in a separately submitted Topical Report JAB-PS-01(s) (hereinafter referred to The Blume Report). Vertical acceleration equal to two-thirds the horizontal ground eleration was applied to the structure.

erational Basis Earthquake ng the data presented in the Blume Report, stresses were computed for the various parts of the nt structures. The stresses resulting from both horizontal and vertical acceleration were bined to obtain the total earthquake stresses. Earthquake stresses were then added linearly and ctly to stresses caused by DBA, snow, dead loads, and the appropriate operating loads to ain the total stresses. The total stresses were reviewed to ensure that they were within the ximum stress limits as established in Table B.6-2 and Table B.6-3. Direct superposition of sses has been used for all loads except missile impact and contact points of pipe rupture raints. For these loads the material is stressed beyond the elastic range. Design procedures for sile impacts are given in the section entitled Tornado Missiles of this appendix.

ign Basis Earthquake forces for the DBE were taken to be two times the forces as determined by the spectral lyses for the OBE. Stresses were combined as before and it was established that they were hin limits as indicated in Table B.6-2 and Table B.6-3.

nado Winds ctures were analyzed for stresses due to tornado and missile loads. Stresses due to tornado ding were combined with stresses due to dead loads and the appropriate operating loads to in a total stress. Maximum stresses were limited to those specified in Table B.6-2.

nado Missiles ctrum of Missiles Considered ny missiles were considered, but only the most damaging missiles were used for design.

siles were assumed to be generated by explosive injection due to pressure differential, by lding component failure, and by aerodynamic lifting, each resulting in an airborne or

ugh to attain high horizontal velocities. Table B.6-6 lists the missiles considered and the imum velocities that would be attained by each.

ign for Missiles design basis missile protection criteria (Reference 36) states systems required to shut the nt down and to keep the plant in a safe shutdown condition shall not be prevented from orming their function by external missiles. It also states: protection of the equipment relied n to provide reasonable assurance of safe plant operation can be achieved by either housing or ing them part of redundant systems with such physical separation that sufficient back-up is vided to assure no loss-of-function of them.

tems, structures, and components (SSC) were designed to meet the design basis missile tection criteria, considering the spectrum of design basis missiles and the limiting, most aging design basis missile for that specific SSC. Systems relied upon for safe shutdown and plant operation in a tornado event were either placed (housed) in a class 1 structure for ection or designed as a redundant system with sufficient physical separation to ensure that the le limiting, most damaging DB missile would not cause a loss of function of that system.

tems with peripheral, unprotected SSC were evaluated for tornado effects to address the ential vulnerabilities in these systems (References 37, 38, 39). Based on these evaluations it concluded that: 1) the system design is in compliance with the original design basis tornado sile protection criteria 2) there is adequate physical separation and redundancy in the design to ure that the system is capable of performing its design function required in a tornado event for shutdown 3) the plant will be able to withstand the consequences of the tornado, will retain capability to achieve and maintain the reactor in a safe shutdown condition, and there will be uncontrolled release of radioactivity as a result of the tornado event 4) the unprotected pheral SSC in these systems can reasonably be exempt from the requirement for specific sile barriers without jeopardizing the health and safety of the public.

concept in analysis and design considered impact to be a plastic collision between the missile the structure.

nado missiles generally are of an intermediate energy level. Their total kinetic energy is ipated by energy absorption of the affected structure as a whole. This results from the elastic plastic response of the structure to the impact force, energy absorption by the missile itself to plastic deformation of the missile, and by the building structure missile barrier member due ocal plastic deformation.

missile barrier of reinforced concrete will react to missile impact as a combination of

-ductile concrete and ductile reinforcing steel. The mode of concrete failure will be brittle ture such as might result from punching shear. Shear cracks will occur at the impact area

mber will respond elastically and plastically as a moment-resisting reinforced concrete ment up to the point of brittle fracture of the concrete, and then the reinforcement will respond ensile strands in membrane actions, elongating plastically to absorb the kinetic energy.

problem of establishing a missile barrier can be subdivided according to the behavioral onse of the characteristic structural element, i.e., slab, wall, beam, and column.

b and walls can respond by perforating or shear failure, plastic bending, and finally forming a ile membrane as described above.

omparison was made of various penetration formulas such as the Army Corps of Engineers, listic Research Laboratory, and Modified Petry before selection of the Modified Petry formula he most commonly used and best fit to the controlling conditions. None of the available mulas developed from empirical ballistic information were particularly suited to tornado sile problem solutions.

sing the Petry formula, the usual rule is to make a slab or wall of a thickness at least twice the etration determined by the second Modified Petry formula for concrete of finite thickness.

s was done assuming all deformation to occur in the concrete (indestructible missile). A ection factor was applied to steel missiles of non-circular or open cross-section, such as steel s and steel pipe, so that the area used in the Petry formula to determine the theoretical etration of an indestructible missile was three times the net cross-sectional area of the steel.

umption of an indestructible missile leads to very high peak loads and shear stresses when ing an analysis for impulse loading, therefore, experiments of limited scope were performed ch verified that almost all of the local plastic deformation would occur in the wood (for a d missile) impacting on concrete, and that steel missiles would enter a plastic range while etrating concrete.

provide a workable solution for applying the Petry formula to a wood missile, a K value dicted on plastic deformation (or destruction) of the wood was used to determine the netration or deceleration path, and from this a peak load was obtained. In the case of the steel sile, the peak load is limited by the short-duration yield strength of the steel.

le B.6-6 is a tabulation of tornado-generated missiles, which shows the weight to cross tional area ratio and gives the impact velocity of the worst case for these missiles. All ado-generated missiles were assumed to impact end on at 90 degrees to the surface being acted, and all areas of Class I structures exposed to either falling or horizontally flying ado missiles are investigated. The tornado missiles were assumed to come from stored erial, destruction of lower class structures, off-site construction, etc. The peak loads associated h the various missiles are as follows:

Vertical falling wood plank 288 kips Steel girts 197 and 257 kips Steel pipe 180 kips Automobile 182 kips ng the peak load, slabs and walls were analyzed for their response to shear (approximately at ultimate strength of the concrete, in shear) and ability to develop plastic hinges and a tensile mbrane of reinforcing steel. After the shear failure of the slab or wall, the plastic deformation he longitudinal reinforcing is calculated not to exceed a strain of 5 percent.

nforced concrete beams in a horizontal plane were analyzed for impulse loading. A angular force-time curve was assumed so that the methods contained in Reference 1 could be

d. The dynamic system was established, including boundary conditions, size of member, mber characteristics, reinforcing, loading, span, etc., to determine the natural frequency and tic strength of the member. From the peak load previously found and the plastic resistance, the tility factor was determined and this was conservatively limited to 6. If this is exceeded, the m is redesigned to limit the ductility factor to 6. The dynamic reactions were calculated for the tic or plastic strain range, as required, and combined with other loads (Dead loads, etc.). A imum value of missile impact reaction of 300 kips was used in order to provide a minimum ar strength capability for missiles impacting near a support.

allowable shear stresses used were:

f 'cd for reinforced beam webs, f 'cd for d/2 stirrup spacing, 0 f 'cd for d/4 stirrup spacing re:

0.85, and d = 1.25, ultimate compressive strength of concrete, using a minimum web reinforcement of 0.15 percent bs (beam width, b x bar spacing, s).

rup stress was limited to 0.85 times 1.25 fy (yield strength of reinforcing bars) and bond stress limited to 0.15f 'c with 0.85 of the summation of the perimeter of bars.

ms designed by this procedure will have very minimal plastic deflection under tornado missile act. Beams, which were too small to comply with the above requirements were investigated the capability to hang from adjacent slabs as a thickened portion of the slab.

er applicable loading. The 300-kip load was chosen to establish a minimum strength in mns subject to missile impact and exceeds the dynamic reaction from the beams.

stress level in columns under the above loading was limited to 1.5 times the ACI code wable stress to provide a higher factor of safety in the columns than that used in beam and slab gn.

listed procedures were conservative and provide for missile barriers that can absorb sufficient sile energy to reduce the missile velocity to zero without physical breach of the barrier, and p cracking and plastic deformation within acceptable levels.

ss I* Structures design of Class I* Structures is similar to the design of Class I Structures for seismic loads

, as detailed previously in this section.

ll other respects the design requirements of Class I* Structures are identical to the design uirements of Class III Structures, as detailed below.

ss II Structures ctures in this class were designed for the conditions of loading specified in Section B.6.1 and le B.6-1 and in accordance with the design methods and allowable stresses specified in the es listed in Section B.3. Stresses were combined as before and reviewed to assure that they e within the limits set forth in Table B.6-4.

ss III* Structures design of Class III* Structures is similar to the design of Class II Structures for the condition oading specified in Section B.6.1 and Table B.6-1.

ll other respects the design of Class III* Structures is identical to the design of Class III ctures as detailed below.

ss III Structures ctures in this class were designed for the conditions of loading specified in Section B.6.1 and le B.6-1, and in accordance with the design methods and allowable stresses specified in the es listed in Section B.3. Stresses were combined as before and reviewed to assure that they e within the limits set forth in Table B.6-4.

ctures in this class were designed for the conditions of loading specified in Section B.6.1 and le B.6-1 and comply with the requirements of the State of Wisconsin Administrative Code.

se facilities provide staff working space, employee facilities, and material and record storage ce. These structures are designed to be independent of other plant structures except for minor s imposed at the interconnection to existing facilities. These structures connect to Class III or ss III* structures only.

ed Classification Structures lass I area located in a lower class structure was treated as a Class I structural system within lower class structure.

mponents of the Class I structural system which were required to meet the total structural ction of this system may extend into the lower class area and were analyzed for their Class I ction. These components include related foundations, supporting structures and overhead ctures.

design provisions made where a structure of a lower seismic design classification is adjacent structure of a higher classification to prevent damage to the higher classification structure er 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.

Class I concrete structures were analyzed and designed to withstand the effects of a tornado ccordance with the parameters as established in this appendix.

l-framed structures are enclosed with metal siding and roof decking. The siding and a portion he roof decking have been attached with pressure relief fasteners to vent the building from ado pressures and forces. This will prevent the stresses in the main structural frames from eeding the allowable limits established in this appendix, and thus prevent their collapse onto ss I structures.

ds or missiles will not cause a loss of function to the Class I structure by direct or indirect ure of structural components.

LOAD COMBINATIONS FOR STRUCTURES Class of Structures Conditions of Loading Class I Class I* Class II Class III* Class III Class IV Normal Operating Dead + Live + Dead + Live + Dead + Live + Dead + Live + Dead + Live + Dead + Live Wind + Snow Wind + Snow Wind + Snow Wind + Snow Wind + Snow Wind + Sno Operational Basis Dead + Live + NA NA NA NA NA Earthquake (OBE) DBA + Snow +

Greater of the OBE or Wind Design Basis Dead + Live + Dead + Live + NA NA NA NA Earthquake (DBE) Snow + DBA + Snow + DBE DBE Tornado Dead + Live + NA NA NA NA NA 300 mph Design Tornado +

Tornado Missile, if any

LOAD COMBINATIONS FOR STRUCTURES Class of Structures Conditions of Loading Class I Class I* Class II Class III* Class III Class IV Other In addition to NA Dead + Live + Dead + Live + NA Wind + Snow above, jet forces Uniform Uniform Loads are ice loads, pipe Building Code Building Code specified in t rupture loads, etc., Zone 1 Zone 1 State of whichever is earthquake loads earthquake loads Wisconsin applicable (see (see Administrati Section B.4.5) Section B.4.5) Code as 80 m and 30 lb/ft2 Note: N/A = Not Applicable

APPLICABLE CODE STRESSES CLASS I STRUCTURES:

REINFORCED CONCRETE - STRUCTURAL STEEL Loading Condition Reinforced Concrete Structural Steel

1. Normal Operating Condition: ACI 318-63 allowable values AISC allowable values Dead and Live Loads + Environmental Loads (Wind + Snow)

2. Operational Basis Earthquake Condition: ACI 318-63 allowable values AISC allowable values Dead + Live + DBA + Snow + Greater of the OBE OR Wind

3. Design Basis Earthquake Condition: 1 1/2 times ACI 318-63 allowable 1 1/2 times AISC allowable values Dead Loads + Live + Snow Loads + DBA + values DBE Load

4. Tornado Condition: fc = 0.75 f 'c fs = 0.90 Y.S.

Dead Loads + Live Loads + 300 mph fs = 0.90 Y.S.

Design Tornado (Does not include Tornado Missile)

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

APPLICABLE CODE STRESSES: CLASS I STRUCTURES Reinforced Steel Stresses 615 Grade 40 A 615 Grade 60 Allowable Allowable Percent of Concrete Working Stress Percent of Min. Working Stress Min. Spec.

Loading Condition Criteria Stresses fc psi Spec. Yield 1 psi Yield 1 Normal Operating ACI 318-63 0.45 f 'c 20,000 50 24,000 40 Condition Allowable values Operational Basis ACI 318-63 0.45 f 'c 20,000 50 24,000 40 Earthquake Allowable Values Design Basis 11/2 times 0.675 f 'c 30,000 75 36,000 60 Earthquake ACI 318-63 allowable values

1. Minimum specified yield points of steel reinforcements are as follows:

A615 Grade 40 40,000 psi A615 Grade 60 60,000 psi

ALLOWABLE STRESSES: CLASS I*, II, III*, III AND IV STRUCTURES ass Loading Condition Criteria Item (3), Table B.6-2 and Table B.6-3 Item (3), Table B.6-2 and Table B.6-3 Dead load plus live loads, plus greater ACI 318-63 and AISC allowable of wind plus snow or Zone I stresses with no increase in earthquake stresses for earthquake condition

• Same as for Class II above ACI 318-63 and AISC allowable stresses with no increase in stresses for earthquake condition Item (1), Table B.6-2 ACI 318-63 and AISC allowable stresses Item (1), Table B.6-2 State of Wisconsin Administrative Code

DAMPING FACTORS Percent of Critical m Damping*

eactor Containment vessel 1.0 ield Building 2.0 eactor containment vessel internal concrete 5.0 eel frame structures 2.0 einforced concrete construction 2.0 ping systems 0.5 ectrical and mechanical equipment evaluated in accordance with the 1.0 ume Report (Reference 9) undation soils 5.0 ectrical and mechanical equipment evaluated in accordance with the SQUG 5.0 IP**

• 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).

TORNADO-GENERATED MISSILES Weight to Elevation Cross Explosive of Origin Total Sectional Injection Above Height Vertical Vertical Horizontal Horizont Weight Area Ratio Height Target of Drop Velocity Energy Velocity Energy Missile (lb) (lb/sq in.) (ft) (ft) (ft) (ft/sec) (ft-lb) (ft/sec) (ft-lb)

Wood Plank 4 in x 12 in x 12 ft - 0 in 150 3.1 112 67 179 108 27,168 NA NA Rough Douglas Fir 150 3.1 NA NA NA NA NA 440 450,93 Steel Girt W 10 in x 11.5 in x 20 ft - 0 in 230 6.8 92 67 159 102 37,157 NA NA A36 Steel 230 6.8 NA NA NA NA NA 35 4375 Steel Girt W 8 in x 15 in x 20 ft- 0 in 300 6.8 35 67 102 81 30,564 NA NA A36 Steel 300 6.8 NA NA NA NA NA 35 5707 Steel Pipe 4 in STD. 10.79 lb/ft 216 6.8 10 67 77 70 16,435 NA NA 4.5 in O.D. x 20 ft - 0 in 216 6.8 NA NA NA NA NA 66 14,610 Automobile 4000 0.5 25 NA NA NA NA 73 330,99 Notes: All tornado missiles are assumed to impact end-on, at 90 degrees to surface being impacted.

NA = Not Applicable

INTERNALLY-GENERATED MISSILES INSIDE OF CONTAINMENT Weight to Cross Sectional Area Ratio Velocity issile (lb/sq. in.) (ft/sec) Impact Point n Motor Operator 14.1 100 Missile Shield lation Valve Slab n x 6 in Valve 9.3 75 Missile Shield afety Relief Valve) Slab n Air operator 2.4 50 Missile Shield lief Valve Slab using Plug 1.8 240 Reactor Vessel Missile Shield ive Shaft 49.5 151 Reactor Vessel Missile Shield ive Shaft and 135 14.3 Reactor Vessel ive Mech. Missile Shield te: All internally generated missiles are assumed to impact at 90 degrees.

DESIGN CRITERIA FOR COMPONENTS s section describes the general design criteria for all mechanical, electrical, instrument, and trol components used in the plant design.

.1 Load Combinations load combinations applicable to Class I, Class I*, Class II, Class III*, and Class III mponents are given in detail in subsequent paragraphs and are also listed in Table B.7-1. The Live loads when used on components consists of thermal and pressure loads.

ss I Components ss I Components were analyzed for each of the following conditions of loading:1 Normal Operating Condition The load combination consists of Dead and Live loads, together with Environmental loads (wind or snow), wherever applicable.

Normal and OBE Condition The load combination consists of Dead and Live loads, together with the greater of the OBE or Wind loads.

Normal and DBE Condition The load combination consists of Dead and Live, together with DBE, loads.

Normal and Pipe Rupture The 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 Rupture The load combination consists of Dead, Live, DBE loads, and pipe rupture loads, excluding loads from pipe rupture in the reactor coolant loop.

eplacement steam generator lower units are designed and analyzed to loading combinations defined in Design pecifications 414A03, consistent with ASME Code,Section III, Division 1, Subsection NB, Class 1, 1986 dition through 1987 Addenda. The original steam domes are analyzed in the same manner as the replacement ower units.

se components designated as Class I* were analyzed for each of the following conditions of ing:

Normal Operating Condition The load combination consists of Dead and Live loads, together with Environmental loads, if applicable.

Normal and DBE Condition The load combination consists of Dead, Live and DBE loads.

ss II Components ss 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.

ss III* Components ss III* Components were designated for the greater of the two combinations of loads given ve for Class II components.

ss III Components ss III components were designed for Dead and Live loads, together with Environmental loads, pplicable.

a minimum, Class III components were designed in accordance with the applicable codes as d in Section B.3. In addition, in accordance with the Wisconsin Public Service Corporations mal policy for the design of steam-electric generating stations, certain components of the er plant in the Class III category were designed according to the requirements of a higher sification.

.2 Design Criteria Deleted Deleted

Deleted Deleted Deleted Deleted Deleted Deleted Deleted Deleted Deleted ign Criteria for Class I* Components Deleted Deleted Deleted

e following information is HISTORICAL and is not intended or expected to be updated.

estinghouse-Furnished Equipment e Standard Westinghouse 2-loop analysis used an envelope of response acceleration spectra ich was more conservative than those presented in the Blume Report (Reference 9).

e Seismic criteria for Westinghouse furnished equipment were as follows:

. 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.

. 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.

. 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.

assure that Westinghouse-supplied NSSS Class I mechanical components met the above smic design criteria, the following procedure was implemented:

. 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.

. The vendors drawings and calculations were reviewed by the cognizant engineer responsible for the particular component to determine whether the component met all specification requirements

. Based on engineering judgment and detailed analyses on similar equipment, the cognizant engineer either

a. Accepted the component, or

b. Rejected the component as inadequate, or recommended modifications, or

c. 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.

conform to the above, seismic analysis of selected NSSS Seismic Class I components luding heat exchangers, pumps, tanks and valves was performed by Westinghouse using one three methods depending on the relative rigidity of the equipment being analyzed:

. 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;

. 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;

. 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.

e inertial forces, moments, and stresses are determined for each mode. They are then mmed using the square-root sum-of-the-squares method. A sufficient number of masses are cluded in the mathematical models to insure that coupling effects of members within the mponent are properly considered. The results of these analyses indicate that the models ntain more masses than necessary. The method of dynamic analysis uses a proprietary mputer code called WESTDYN. This code uses as input, inertia values, member sectional operties, elastic characteristics, support restraint data characteristics, and the appropriate smic response spectrum. Both horizontal and vertical components of the seismic response ectrum are applied simultaneously. The modal participation factors are combined with the ode shapes and the envelope seismic response spectra to give the structural response for each ode. The inertial forces, moments, and stresses are computed for each mode, from which the odal stresses are determined. The stresses are then summed using the square-root m-of-the-squares method.

general, no additional restraints beyond those normally provided are required to assure smic adequacy.

ance of Plant Equipment seismic design criteria for balance of plant Class I (seismic) mechanical components and trical equipment are described as follows:

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.

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.

rder to meet these seismic design criteria the following measures were taken for seismic gn and restraint:

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.

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.

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.

analytical or testing methods utilized to verify the adequacy of the above are described as ows:

Analytical Methods

a. 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.

Testing Methods: (one of the following)

a. Continuous Test The 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.

Natural 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.

smic input values used for analysis or testing purposes to verify the adequacy of Class I smic) components were obtained from Topical Report JAB-PS-03 prepared by John A. Blume Associates, Engineers (Reference 9).

rumentation and Control Systems design bases for protection-grade equipment (Class I) with respect to earthquakes were that an OBE or DBE, the equipment was designed to ensure that it did not lose its capability to form its function; i.e., shut the plant down and/or maintain the unit in a safe shutdown dition. For the DBE, the capability of the protection equipment to perform its function was ntained.

seismic disturbance occurs subsequent to an accident, the instrumentation and electrical ipment associated with emergency core cooling will not be interrupted during this disturbance.

ial evaluation of Protection System equipment for its ability to withstand the seismic condition typically done by actual vibration-type testing of typical protection-grade equipment.

hematical models derived from empirical tests were not normally used for seismic design luation of instrumentation. However, in the absence of empirical test data, such as may be the e for very large equipment (for example, control room panels), evaluation may have been ported by mathematical analysis or some combination of mathematical analyses and prototype ing. (See Reference 4 for discussion and documentation of some test program results).

ign Criteria for Class II and Class III* Components mponents in this class are designed for the conditions of loading specified in Table B.7-1 and ccordance with the design methods and allowable stresses specified in the codes listed in tion B.3. Stresses are combined as for Class I above and reviewed to assure that they are hin the limits set forth in the applicable codes.

mponents in this class are designed for the conditions of loading specified in Table B.7-1 and ccordance with the design methods and allowable stresses specified in the codes listed in tion B.3.

LOAD COMBINATIONS FOR COMPONENTS CLASS OF COMPONENTS Condition of Loading Class I1, 2 Class I* 3 Classes II and III* Class III

1. Normal Dead + Live + Dead + Live + Dead + Live + Dead + Live +

Environmental Loads Environmental Loads Environmental Loads Environmental Lo (Snow or Wind) If (Snow or Wind) If (Snow or Wind) If (Snow or Wind) If Applicable Applicable Applicable Applicable

2. Normal and Operational Dead + Live + Greater NA Dead + Live + UBC NA Basis Earthquake (OBE) of the OBE or Wind Loads Loads

3. Normal and Design Basis Dead + Live + Dead + Live + NA NA Earthquake (DBE) DBE Loads DBE Loads

4. Normal and Pipe Rupture Dead + Live + Pipe NA NA NA Rupture Loads Except RCL Pipe Breaks

5. Normal Design Basis Dead + Live + DBE + NA NA NA Earthquake and Pipe Pipe Rupture Loads Rupture Except RCL Pipe Breaks Note: NA = Not Applicable

1. 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 man 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 AS Code,Section III, Division 1, Subsection N.3, Class 1, 1998 Edition through 2000 Addenda. This methodology was approved by NRC for application to KP 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 load, in accordance with NUREG-0554.

LOADING CONDITIONS AND STRESS LIMITS: PRESSURE VESSELS Loading Conditions Stress Intensity Limits Note*

1. Normal Condition (a) Pm < Sm 1 (b) Pm (or PL) + PB < 1.5 Sm 2 (c) Pm (or PL) + PB + Q < 3.0 Sm

2. Upset Condition (a) Pm < Sm 1 (b) Pm (or PL) + PB < 1.5 Sm 2 (c) Pm (or PL) + PB + Q < 3.0 Sm

3. Emergency Condition (a) P < 1.2 Sm or Sy, whichever is larger 3 (b) Pm (or PL) + PB < 1.8 Sm, or 1.5 Sy, whichever is larger

4. Faulted Condition (a) Stainless Steel 4 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 larger Pm = 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.

LOADING CONDITIONS AND STRESS LIMITS: PRESSURE PIPING IN ACCORDANCE WITH USAS B31.1 ading Conditions Stress Limits

. Normal Condition P<S

. Upset Condition P < 1.2S

. Emergency Condition P < 1.5 (1.2S)

. Faulted Condition For stainless steel Design Limit Curves as defined in Figure B.7-3, See Note 4a For carbon steel P < Sy or 1.8S, whichever is higher b here:

= Stress

= Allowable stress from USAS B31.1, Code for Power Piping, 1967

= Minimum specified yield strength (ASME B&PV Code,Section III, Table N-424 or uivalent)

For description of Note 4, see Notes For Tables B.7-2, B.7-3, And B.7-6.

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.

te 1 The 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.

te 2 In 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.

te 3 The 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.

te 4 a As 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.

This alternate design procedure was not utilized on this application.

LOADING CONDITIONS AND STRESS LIMITS: EQUIPMENT SUPPORTS ading Conditions Stress Limits

. Normal Condition Working stresses or applicable factored load design values

. Upset Condition Working stresses or applicable factored load design values

. Emergency Condition Within yield after load redistribution to maintain supported equipment within emergency condition stress limits

. Faulted Condition Permanent deflection of supports limited to maintain supported equipment within faulted condition stress limits Table B.7-5 LOAD COMBINATION AND STRESS LIMITS FOR CLASS I COMPONENTS ad Combination Stress Limit

. Normal** (deadweight, thermal and pressure) Normal Condition

. Normal and Operational Basis Earthquake Upset Condition

. Normal and Design Basis Earthquake Faulted Condition*

. Normal and Pipe Rupture Faulted Condition

. Normal and Design Basis Earthquake and Pipe Rupture Faulted 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.

ALTERNATIVE DESIGN LOADING COMBINATIONS AND STRESS LIMITS:

PRESSURE CLASS 1, 2, AND 3 PIPING IN ACCORDANCE WITH ASME SECTION III ASME Primary Stress Primary + Secondary Section III Code Pm(P1) + Pb Stress Pm(P1) + Pb + Q Peak Stress Pm(P1) + Pb + Q + F Condition Class Design Loading Combinations Equation 9 Equation 10 Equation 14 Normal and Upset 1 (NB-3600) Design pressure, weight, OBE, and 1.8Sm but not 3Sm (See Notes 1 & 2) Salt = KSp/2 other mechanical loads (Equation 9) greater than 1.5Sy Cumulative usage factor less than Pressure, Thermal Expansion and (See Notes 1 & 2)

Thermal Gradients (steady-state and transient) (Equation 10, 14)

Emergency 1 (NB-3600) Design pressure, weight, DBE, and 2.25 Sm but not N/A N/A other mechanical loads (Equation 9) greater than 1.8Sy (See Notes 3 & 4)

Faulted 1 (NB-3600) Design pressure, weight, DBE, and 3Sm but not N/A N/A other mechanical loads (Equation 9) greater than 2.0 Sy (See Notes 3

& 4)

Design Loading Combinations Normal and upset 2 (NC-3600) Design pressure, weight and other Design pressure, sustained loads, OBE and Thermal expansion and sustained loads other occasional mechanical loads 3 (ND-3600)

Allowable 1.5 Sh 1.8 Sh but not greater than 1.5Sy (1.25Sc + 0.25Sh)f + Sh - (Slp + S Emergency/ 2 (NC-3600) N/A Operating pressure, sustained loads, DBE and N/A Faulted and other occasional mechanical loads 3 (ND-3600)

Allowable 2.25 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.

TYPICAL STRESS STRAIN CURVE COMPARISON BETWEEN DESIGN AND COLLAPSE CONDITIONS HOOP STRESS: 0.90 Sy COMPARISON BETWEEN DESIGN AND COLLAPSE CONDITIONS HOOP STRESS: 0.00 Sy LOADS

.1 Protection Against Crane Toppling Auxiliary Building crane and the Turbine Building crane are located in areas where they are ject to possible damage from tornado and earthquake. These crane bridges and trolleys are tected against tipping, derailment, and uncontrolled movements that could possibly create age.

assure stability of the Turbine Building crane, the bridge and trolley are equipped with fixed, d rail yokes that allow free rolling movement but prevent the wheels from being lifted or iled. The Auxiliary Building crane trolley is prevented from derailing by restraints that trap it ween the bridge girders. The bridge and trolley wheels are equipped with electrically activated, ng set brakes. Upon loss of power or when the crane or trolley are not under operator control, springs activate the brakes, locking the wheels firmly in place to prevent rolling out of ition. The positive wheel stops and bumpers provided to prevent over-travel of the trolley and ge will prevent the trolley and bridge from leaving the rails, even in the unlikely event of ke failure.

.2 Control of Heavy Loads a result of Generic Task A-36, Control of Heavy Loads Near Spent Fuel, the NRC issued REG-0612, Control of Heavy Loads at Nuclear Power Plants. NUREG-0612 was to be lemented in two phases. Phase I addressed Section 5.1 of NUREG-0612 and established en basic guidelines for all nuclear power plants, which detailed provisions for the handling of vy loads in the area of the reactor vessel near stored spent fuel, in other areas where an dental load drop could damage equipment required for safe shutdown or decay heat removal.

following cranes are subjected to the seven guidelines of NUREG-0612 Phase I:

Turbine Building Crane Auxiliary Building Fuel Handling Crane Deleted seven basic guidelines of NUREG-0612, Phase I listed below are satisfied for the above d cranes.

Safe Load Paths Load Handling Procedures Crane Operator Training Special Lifting Devices

Cranes (Inspection, Testing, and Maintenance)

Crane Design spent fuel pool bridge and hoist crane has the capability of carrying loads which could, if pped, fall into the spent fuel pool. However, based on the use of these cranes, they have been luded from further review against NUREG-0612.

NRC has determined that Kewaunee has adequately addressed NUREG-0612 and has ificantly reduced the probability of a heavy load handling accident to an acceptably small e (see NRC Safety Evaluation Report in Reference 16).

.3 Design Criteria for Upgraded Auxiliary Building Crane Auxiliary Building (AB) crane was upgraded in support of dry spent fuel storage cask ding operations. This upgrade involved the replacement of the original trolley with a le-failure-proof design, replacement of the trolley controls, and an upgrade to the existing AB e bridge. The upgrade of the AB crane meets the guidance in Section 5.1.6 of NUREG-0612, ntrol of Heavy Loads at Nuclear Power Plants, and NUREG-0554, Single Failure Proof nes for Nuclear Power Plants, as applicable.

AB crane is designated as Class I* per Table B.2-1 and therefore is designed to meet Class I mic standards. The crane is designed to stay on its rails and not allow an uncontrolled ering of the load as a result of a seismic event. It is not required to be operational during or r a seismic event. The AB crane is also designed to withstand the crane design basis accident nts described in NUREG-0554: two-blocking, load hang-up, and wire rope failure.

ause the replacement AB crane trolley is a new component and the crane bridge is an existing ponent, the construction codes applicable to the two are not identical. The construction codes the trolley and bridge are as follows:

Crane Trolley Codes and Standards struction is in accordance with NUREG-0554 and, where NUREG-0554 does not offer cific guidance (e.g., normal condition load combinations and stress acceptance criteria),

struction is in accordance with Crane Manufacturers Association of America Specification 70 AA-70), 2004 Edition. Seismic load combinations and stress analysis acceptance criteria, as l as guidance used to address two-blocking, load hang-up, and wire rope failure are taken from ME NOG-1-2004.

Crane Bridge Codes and Standards

ndard 61 (EOCI-61), CMAA-70 (2004), and ASME NOG-1-2004 in that hierarchy, where REG-0554, and EOCI-61 do not offer specific construction guidance.

Crane Seismic Response Spectra, Damping and Accelerations seismic analysis of the AB crane considers trolley and bridge drive wheel rolling when the mic forces exceed the drive wheel brake resisting force. This nonlinear boundary condition uired seismic time history inputs to be developed consistent with Standard Review Plan (SRP),

REG-0800, Section 3.7.1, Revision 3, Option II. With the exception of the nonlinear boundary dition at the trolley and bridge drive wheels, the seismic analysis of the upgraded AB crane is sistent with ASME NOG-1-2004.

Blume Report, which forms the basis for seismic analyses at the Kewaunee Power Station, s not include horizontal response spectra data for a mass point at the location of the AB crane appropriate for use in analyzing the upgraded crane. Therefore, a lumped-mass stick model of AB steel structure was used to generate additional horizontal response spectra applicable for at the AB crane rail. Two percent damping for the Safe Shutdown Earthquake condition was lied to both the vertical and horizontal spectra at the crane rail elevation.

e sets of seismic acceleration time histories were then developed representing the response of AB crane at the base of the crane bridge rails (Reference 44). Each set of time histories tains two horizontal and one vertical time history, for a total of fifteen time histories. The time ories were used in conjunction with a 3-D model of the crane to perform the nonlinear seismic lysis. The methodology for analyzing the response of the crane during a seismic event was ed on the application of the commercially available finite element analysis computer program.

P 2000, Version 11. The use of SAP 2000 was reviewed pursuant to SRP Section 3.9.1, ecial Topics for Mechanical Components, which provides applicable criteria for evaluating puter programs for mechanical and structural design and analysis.

NRC approval of the seismic methodology for the AB crane (Reference 45) is subject to the owing limitations:

The analyses are based on the seismic acceleration time histories reported in the license amendment request submittal dated July 7, 2008 (Reference 44).

The calculated critical wheel tractions should be increased by 25 percent for the crane drive wheels and 100 percent for the trolley wheels.

The seismic methodology use is limited to the AB crane.

.1 Deleted

.2 Deleted

.3 Deleted

.4 Deleted

DELETED 0 DELETED 0.1 Deleted 0.2 Deleted 0.3 Deleted 0.4 Deleted 1 INTERNAL FLOODING 1.1 Deleted 1.2 Flooding Design Criteria plant must withstand the consequences of an internal flooding event in such a manner that it ins the capability to achieve and maintain the reactor in a safe shutdown condition and to limit consequences of a design basis accident.

a. Deleted

b. Deleted

c. Deleted

d. Deleted

e. Deleted

f. Deleted 1.3 Deleted 1.4 Deleted 1.5 Conclusion May 7, 2013, Dominion Energy Kewaunee, Inc. (DEK) submitted the second of two letters uired, pursuant to 10 CFR 50.82(a)(1)(i) and 10 CFR 50.82(a)(1)(ii), to certify that it has manently ceased power operation of KPS, and that the reactor was permanently defueled.

refore, as specified in 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for Kewaunee Power ion no longer authorizes operation of the reactor or emplacement or retention of fuel into the tor vessel.

ctions necessary to mitigate a design basis accident and are vulnerable to damage from the ure of a tank or pipe.

stated previously, the design criteria established for appropriately addressing the consequences nternal flooding (i.e., the effects of the rupture of a tank or pipe) are limited to ensuring that station design retain the capability to:

Achieve and maintain the reactor in a safe shutdown condition and Limit the consequences of a design basis accident h the permanent cessation of plant operation, and the prohibition of emplacement of fuel hin the reactor vessel, the capability to achieve and maintain the reactor in a safe shutdown dition has been permanently achieved.

ilarly, with no SSCs that are credited with design basis accident mitigation being vulnerable to age from internal flooding, that design criterion is also permanently met.

refore, no design features of the station need be retained based solely upon their contribution eeting the internal flood mitigation design criteria.

DELETED DELETED 2 DELETED 2.1 Deleted 2.2 Deleted 2.3 Deleted FERENCES Morris, Hansen, Holley, Biggs, Namyet, and Minami, Structural Design for Dynamic Loads, McGraw-Hill Co., Inc., New York, 1959.

Deleted Housner, George W., Vibration of Structures Induced by Seismic Waves, Shock and Vibration Handbook, Volume III, McGraw-Hill, Inc., New York, 1961.

Vogeding, E. L., Topical Report, Seismic Testing of Electrical and Control Equipment, WCAP 7817, December 1971.

Deleted Deleted Deleted 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).

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).

Deleted Deleted Deleted Deleted 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.

Deleted

No. K-84-61, March 16, 1984.

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Seismic Qualification Utility Group (SQUG), Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Power Plant Equipment, Revision 2 as corrected February 14, 1992.

Deleted Deleted Deleted Deleted 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.

Seismic Qualification Utility Group (SQUG), Generic Implementation Procedure (GIP) for Seismic Verification of Nuclear Power Plant Equipment, Revision 3, May 16, 1997.

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).

License Application Amendment 17 dated May 12, 1972 from E. W. James (WPS) to P.A.

Morris (AEC).

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O'Leary (AEC).

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O'Leary (AEC).

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O'Leary (AEC).

Safety Evaluation of Kewaunee Nuclear Power Plant, Supplement 2 dated July 24, 1972.

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Reactors (LWR Edition) dated July 1981.

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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.

Operability Determination Closure Request for OBD 135-EDG Exhaust Ducts dated April 27, 2008.

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.

McDonald - Mehta Engineers Letter Report, Tornado Effects on Turbine Building and Diesel Generator Exhaust Lines dated April 28, 2005.

Letter from Steven A. Varga (NRC) to CW Giesler (WPSC) Subject Control of Heavy Loads

-NUREG-0612-Phase II dated June 13, 1984.

Deleted Deleted Deleted License Amendment Request 239, Request for Review and Approval of Seismic Analysis Methodology for Auxiliary Building Crane, July 7, 2008 (includes seismic time histories).

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.

License Application Amendment 32 dated August 31, 1973, from E.W. James (WPS) to J.F OLeary (AEC).

NRC Generic Letter 89-10: Safety-Related Motor-Operated Valve Testing and Surveillance, dated June 28, 1989, including Supplements 1 through 7.

NRC letter to WPSC: Close-Out of Generic Letter (GL) 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance, dated January 4, 1996.

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.

Joint BWR, Westinghouse and Combustion Engineering Owners' Group Program on Motor-Operated Valve Periodic Verification, MPR-1807, Revision 2, July 1997.

Motor-Operated Valves Described in Topical Report NEDC-32719, Revision 2 (MPR-1807, Revision 2), October 30, 1997.

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.

Joint Owners' Group (JOG) Motor-Operated Valve Periodic Verification Program Summary, MPR-2524-A, Revision 1, September 2010.

Final Safety Evaluation on Joint Owners' Group Program on Motor-Operated Valve Periodic Verification, September 25, 2006.

Deleted Deleted