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{{#Wiki_filter:WOLF CREEK UPDATED SAFETY ANALYSIS REPORT (USAR)
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Revision - 35 Release Date - DC65 3/10/2022
 
WOLF CREEK TABLE OF CONTENTS CHAPTER
 
==1.0 INTRODUCTION==
AND GENERAL DESCRIPTION OF THE PLANT Section                                                  Page
 
==1.1    INTRODUCTION==
1.1-1 1.1.1  LICENSES REQUESTED                                1.1-1 1.1.2  PLANT UNITS                                      1.1-1 1.1.3  PLANT LOCATION                                    1.1-2 1.1.4  CONTAINMENT STRUCTURE                            1.1-3 1.1.5  NUCLEAR STEAM SUPPLY AND TURBINE-GENERATOR                                1.1-3 1.1.6  SCHEDULE FOR FUEL LOADING AND OPERATION          1.1-3 1.1.7  DESIGN BASES                                      1.1-3 1.2    GENERAL PLANT DESCRIPTION                        1.2-1 1.2.1  PLANT SITE DESCRIPTION                            1.2-1 1.2.1.1 Site Location                                    1.2-1 1.2.1.2 Site Ownership                                    1.2-1 1.2.1.3 Access to the Site                                1.2-1 1.2.1.4 Environs                                          1.2-1 1.2.1.5 Geology                                          1.2-2 1.2.1.6 Seismology                                        1.2-3 1.2.1.7 Hydrology                                        1.2-3 1.2.1.8 Meteorology                                      1.2-5 1.2.2  GENERAL ARRANGEMENTS OF STRUCTURES                1.2-5 1.2.3  PRINCIPAL DESIGN CRITERIA                        1.2-8 1.2.3.1 SNUPPS Design Envelope                            1.2-8 1.2.4  NUCLEAR STEAM SUPPLY SYSTEM                      1.2-8 1.2.5  ENGINEERED SAFETY FEATURES AND EMERGENCY SYSTEMS                                          1.2-10 1.2.5.1 Containment                                      1.2-10 1.2.5.2 Emergency Core Cooling System                    1.2-12 1.2.5.3 Auxiliary Feedwater System                        1.2-12 1.0-i                          Rev. 29
 
WOLF CREEK TABLE OF CONTENTS (CONTINUED)
Section  Page 1.2.6    PLANT INSTRUMENTATION AND CONTROL SYSTEMS  1.2-13 1.2.6.1  Protection System                          1.2-14 1.2.6.2  Reactor Instrumentation and Control System 1.2-14 1.2.6.3  Radiation Monitoring System                1.2-15 1.2.6.4  Balance-of-Plant Instrumentation and Control Systems                            1.2-15 1.2.7    PLANT ELECTRIC POWER SYSTEM                1.2-15 1.2.7.1  Transmission and Generation Systems        1.2-15 1.2.7.2  Electric Power Distribution System        1.2-16 1.2.8    POWER CONVERSION SYSTEM                    1.2-17 1.2.8.1  Main Steam Supply System                  1.2-17 1.2.8.2  Main Condenser Evacuation System          1.2-17 1.2.8.3  Turbine Gland Sealing System              1.2-18 1.2.8.4  Turbine Bypass System                      1.2-18 1.2.8.5  Circulating Water System                  1.2-18 1.2.8.6  Condensate Cleanup System                  1.2-18 1.2.8.7  Condensate and Feedwater System            1.2-19 1.2.8.8  Steam Generator Blowdown System            1.2-19 1.2.8.9  Secondary Liquid Waste System              1.2-19 1.2.8.10 Wastewater Treatment System                1.2-19 1.2.9    AUXILIARY SYSTEMS                          1.2-20 1.2.9.1  Chemical and Volume Control System        1.2-20 1.2.9.2  Residual Heat Removal System              1.2-20 1.2.9.3  Fuel Handling and Storage System          1.2-21 1.2.9.4  Service Water Systems                      1.2-21 1.2.9.5  Component Cooling Water System            1.2-22 1.2.9.6  Compressed Air Systems                    1.2-23 1.2.9.7  Fire Protection Systems                    1.2-23 1.2.9.8  Heating, Ventilating, and Air-Conditioning Systems                                    1.2-24 1.2.9.9  Sampling Systems                          1.2-25 1.2.9.10 Service Gas System                        1.2-25 1.2.9.11 Communications System                      1.2-25 1.2.9.12 Diesel Generator Support Systems          1.2-26 1.2.10  WASTE PROCESSING SYSTEMS                  1.2-26 1.2.11  SHARED FACILITIES AND COMPONENTS          1.2-27 1.2.12  REFERENCES                                1.2-27 1.0-ii                Rev. 29
 
WOLF CREEK TABLE OF CONTENTS (CONTINUED)
Section Page 1.3    COMPARISON TABLES                        1.3-1 1.3.1  COMPARISON WITH SIMILAR FACILITY DESIGNS  1.3-1 1.3.2  COMPARISON OF FINAL AND PRELIMINARY INFORMATION                              1.3-1 1.3.3  COMPLIANCE WITH NRC REGULATIONS          1.3-1 1.4    IDENTIFICATION OF AGENTS AND CONTRACTORS  1.4-1 1.4.1  APPLICANTS                                1.4-1 1.4.2  SNUPPS                                    1.4-1 1.4.3  NUCLEAR STEAM SUPPLY SYSTEM MANUFACTURER  1.4-3 1.4.4  STANDARD PLANT (LEAD) ARCHITECT/ENGINEER  1.4-4 1.4.5  TURBINE-GENERATOR MANUFACTURER            1.4-4 1.4.6  SITE ARCHITECT/ENGINEER                  1.4-5 1.4.7  CONSULTANT FIRMS                          1.4-5 1.4.7.1 SNUPPS Consultants                        1.4-5 1.4.7.2 WCGS Specific Consultants                1.4-7 1.4.8  CONSTRUCTOR                              1.4-10 1.4.9  DIVISION OF RESPONSIBILITIES              1.4-11 1.4.9.1 Utility Company                          1.4-11 1.4.9.2 Standard Plant Architect/Engineer        1.4-11 1.4.9.3 SNUPPS Staff                              1.4-12 1.4.9.4 Site Architect/Engineer                  1.4-12 1.4.9.5 Security Consultant                      1.4-12 1.5    REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION                          1.5-1 1.5.1  17 x 17 FUEL ASSEMBLY                    1.5-1 1.5.1.1 Rod Cluster Control Spider Tests          1.5-1 1.5.1.2 Grid Tests                                1.5-1 1.5.1.3 Fuel Assembly Structural Tests            1.5-1 1.5.1.4 Guide Tube Tests                          1.5-1 1.5.1.5 Prototype Assembly Tests                  1.5-2 1.5.1.6 Departure from Nucleate Boiling Tests    1.5-2 1.5.1.7 Incore Flow Mixing                        1.5-2 1.0-iii                  Rev. 29
 
WOLF CREEK TABLE OF CONTENTS (CONTINUED)
Section                                                Page 1.5.2  FIRE STOPS                                      1.5-2 1.5.3  OTHER PROGRAMS                                  1.5-2 1.5.3.1 Generic Programs of Westinghouse                1.5-2 1.5.3.2 Generic Programs of Bechtel                    1.5-3 1.5.3.3 Test of a Wolf Creek Steam Generator            1.5-3 1.
 
==5.4  REFERENCES==
1.5-3 1.6    MATERIAL INCORPORATED BY REFERENCE              1.6-1 1.7    DRAWINGS AND OTHER DETAILED INFORMATION        1.7-1 1.7.1  ELECTRICAL, INSTRUMENTATION AND CONTROL DRAWINGS                                        1.7-1 1.7.2  PIPING AND INSTRUMENTATION DIAGRAMS            1.7-1 1.7.3  MISCELLANEOUS CONTROLLED DRAWINGS              1.7-1 1.8    CONFORMANCE TO NRC REGULATORY GUIDES            1.8-1 1.9    NRC REGULATORY REQUIREMENTS REVIEW COMMITTEE CATEGORY 2, 3, AND 4 MATTERS                    1.9-1 1.0-iv                      Rev. 30
 
WOLF CREEK TABLE OF CONTENTS (CONTINUED)
LIST OF TABLES Table no.                        Title 1.1-1    Acronyms Used in the USAR 1.2-1    Design Envelope 1.3-1    Design Comparison 1.3-2    Major Analyses Not Included in Topical Reports 1.3-3    Significant Design Changes from the PSAR 1.3-4    Compliance with NRC Regulations, 10 CFR 1.4-1    Nuclear Power Plants Completed or Currently Under Design by Sargent & Lundy 1.4-2    Other Nuclear Power Plants with Partial Sargent & Lundy Design Responsibility 1.6-1    Bechtel Topical Reports Incorporated by Reference 1.6-2    Westinghouse Topical Reports Incorporated by Reference 1.6-3    USAR Figure/Controlled Drawing Cross-Reference 1.6-4    Incorporated by Reference USAR Section.Controlled Document Cross-Reference 1.7-1    Electrical, Instrumentation, and Control Drawings 1.7-2    Piping and Instrumentation Diagrams 1.7-3    Additional Controlled Drawings Used in the USAR 1.9-1    Category 2, 3 and 4 Regulatory Guides 1.9-2    Category 2, 3, and 4 Branch Technical Positions 1.9-3    Category 4 SRP Criteria 1.9-4    Other Category 4 Positions 1.0-v                        Rev. 30
 
WOLF CREEK CHAPTER 1 - LIST OF FIGURES
*Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure #  Sheet                        Title                      Drawing #*
1.1-1    1    Symbols and Legend for System Flow and Piping      M-120101 and Instrumentation Diagrams 1.1-1    2    Symbols and Legend for System Flow and Piping      M-120102 and Instrumentation Diagrams 1.1-1    3    Symbols and Legend for System Flow and Piping      M-020103 and Instrumentation Diagrams 1.1-1    4    Symbols and Legend for System Flow and Piping      M-020104 and Instrumentation Diagrams 1.2-1    0    Peninsular Plant Arrangement Standard Power        M-1G001 Systems & Structure Interface 1.2-2    0    Equipment Location Radwaste Building Plan El.      M-1G010 1976'-0" 1.2-3    0    Equipment Location Radwaste Building Plan El.      M-1G011 2000'-0" 1.2-4    0    Equipment Location Radwaste Building Plan El.      M-0G012 2022'-0" 1.2-5    0    Equipment Location Radwaste Building El. 2031'-    M-1G013 6" & Roof Plan 1.2-6    0    Equipment Location Radwaste Building Sections A    M-1G014
                & B 1.2-7    0    Equipment Location Radwaste Building Sections C    M-1G015
                & E 1.2-8    0    Equipment Location Radwaste Building Sections D    M-1G016
                & F 1.2-9    0    Equipment Location Reactor and Auxiliary Bldgs      M-1G020 Plan - Basement El. 1974'-0" 1.2-10    0    Equipment Location Auxiliary Building Partial      M-1G021 Plan El. 1988'-0" & El. 2013'-6" 1.2-11    0    Equipment Location Reactor and Auxiliary            M-1G022 Building Plan Ground Floor Elevation 2000'-0" 1.2-12    0    Equipment Location Reactor and Auxiliary            M-1G023 Building Plan El. 2026'-0" 1.2-13    0    Equipment Location Reactor and Auxiliary            M-1G024 Buildings Plan Operating Floor El. 2047'-6" 1.2-14    0    Equipment Locations Reactor and Auxiliary          M-1G025 Buildings Plan El. 2068'-8" 1.2-15    0    Equipment Location Reactor and Auxiliary            M-1G026 Building Section A 1.2-16    0    Equipment Locations Reactor and Auxiliary          M-1G027 Buildings Section B 1.2-17    0    Equipment Location Reactor and Auxiliary            M-1G028 Building Section C 1.2-18    0    Equipment Location Reactor and Auxiliary            M-1G029 Building Section D 1.0-vi                            Rev. 17
 
WOLF CREEK CHAPTER 1 - LIST OF FIGURES
*Refer to Section 1.6 and Table 1.6-3. Controlled drawings were removed from the USAR at Revision 17 and are considered incorporated by reference.
Figure #  Sheet                          Title                        Drawing 1.2-19    0    Equipment Location Auxiliary Building Sections E,    M-1G030 F, & G 1.2-20    0    Equipment Location Fuel Building Plan Elevation      M-1G040 2000'- 0", 2026'-0" and 2047'-6" 1.2-21    0    Equipment Location Fuel Building Sections A, B, &    M-1G041 C
1.2-22    0    Equipment Location Fuel Building Sections D, E, &    M-1G042 F
1.2-23    0    Equipment Location Control Building &                M-1G050 Communication Corridor Plan Elevation 1974'- 0" &
1984'-0" 1.2-24    0    Equipment Location Control & Diesel Generator        M-1G051 Buildings & Communication Corridor Plan Elevation 2000'-0" & 2016'-0" 1.2-25    0    Equipment Location Control & Diesel Generator        M-1G052 Buildings & Communication Corridor Plan Elevation 2032'-0" & 2047'-6" 1.2-26    0    Equipment Location Control & Diesel Generator        M-1G053 Buildings & Corridor Plan Elevation 2061'- 6",
2066'-0" & 2073'-6" & Section D.
1.2-27    0    Equipment Location Control & Diesel Generator        M-1G054 Buildings & Communication Corridor Section A 1.2-28    0    Equipment Location Control & Diesel Generator        M-1G055 Buildings Sections B & C 1.2-29    0    Equipment Location Turbine Building Condenser Pit    M-1G060 Plan Elevation 1983'-0" 1.2-30    0    Equipment Location Turbine Building Ground Floor    M-1G061 Plan Elevation 2000'-0" 1.2-31    0    Equipment Location Turbine Building Partial Plan    M-1G062 Elevation 2015'-4" 1.2-32    0    Equipment Location Turbine Building Mezzanine        M-1G063 Floor Plan Elevation 2033'-0" 1.2-33    0    Equipment Location Turbine Building Operating        M-1G064 Floor Plan Elevation 2065'-0" 1.2-34    0    Equipment Location Turbine Building Section A        M-1G065 1.2-35    0    Equipment Location Turbine Building Section B        M-1G066 1.2-36    0    Equipment Location Turbine Building Section C        M-1G067 1.2-37    0    Equipment Location Turbine Building Section D        M-0G068 1.2-38    0    Equipment Location Turbine Building Section E        M-1G069 1.2-39    0    Equipment Location Turbine Building Section F        M-1G070 1.2-40    0    Equipment Location Turbine Building Section G        M-0G071 1.2-41    0    Equipment Location Turbine Building Section H        M-1G072 1.2-42    0    Turbine Component Laydown Area, Elevation 2065'-    M-1G073 0"
1.2-43    0    Site Area Layout 1.2-44    0    Site Plan                                            8025-C-KG1202 1.0-vii                        Rev. 29
 
WOLF CREEK CHAPTER
 
==1.0 INTRODUCTION==
AND GENERAL DESCRIPTION OF THE PLANT
 
==1.1  INTRODUCTION==
 
Kansas City Power & Light Company, Kansas Gas and Electric Company (KG&E) and Union Electric Company joined together to design, purchase, and license a nuclear block for a generating station acceptable at any of several sites, under the acronym of SNUPPS, Standardized Nuclear Unit Power Plant System. The terminology "the Operating Agent" is used throughout this report to identify the managing corporation for WCGS. At this time the Operating Agent is Wolf Creek Nuclear Operating Corporation (WCNOC).
1.1.1 LICENSE REQUESTED The Safety Analysis Report was submitted to the Nuclear Regulatory Commission (NRC) in support of the application by the Operating Agent for a Class 103 license to operate a nuclear power facility.
The participants in the Wolf Creek project and their portions of ownership were: Kansas City Power & Light Company (47 percent), Kansas Electric Power Cooperative, Incorporated (6 percent), and Kansas Gas and Electric Company (47 percent) which would eventually merge with Kansas Power & Light to become Western Resources and eventually Westar. Great Plains and Westar merged. The Great Plains and Westar merger was finalized June 4, 2018, and formed a new company called Evergy. Evergy owns 94% of WCGS. The remaining 6% ownership interest is held by Kansas Electric Power Cooperative, Inc. (KEPCO). See Section 1.4.1 for additional discussion of plant ownership.
This report was originally submitted in two parts, the SNUPPS FSAR and the Wolf Creek Site Addendum. It was combined into one report, the Wolf Creek Updated Safety Analysis Report, in the first update after receipt of the Operating License. This report follows the format recommended by Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants. Sufficiently detailed design information is provided in this report to make a definitive evaluation that the Wolf Creek Generating Station (WCGS) can be operated without undue risk to the health and safety of the public.
The Licensees received a low power (less than five percent) license to operate the Wolf Creek Generating Station on March 11, 1985. The full power license was issued on June 4, 1985.
1.1.2 PLANT UNITS The application was for a single pressurized water reactor nuclear unit. The power block was built to the SNUPPS duplicate plant 1.1-1                            Rev. 33
 
WOLF CREEK design. The ESW Vertical Loop Chase design is not included in the SNUPPS duplicate plant design. Evaluations of the site characteristics and the design of the cooling system and other site-related systems and facilities have considered the installation of a second nuclear unit at a later date.
The WCGS power block, consists of these structures, including enclosed systems and components:
: a. Reactor building (containment)
: b. Turbine building
: c. Control building
: d. Auxiliary building
: e. Diesel generator building
: f. Fuel building
: g. Radwaste building
: h. Storage tanks (refueling water, condensate, demineralized water, reactor makeup water, and emergency fuel oil)
: i. Transformers (main, unit auxiliary, ESF, and station service) and vaults
: j. ESW Vertical Loop Chase Due to the use of the SNUPPS standard design for these items, design envelopes were developed by use of the most restrictive site conditions imposed by any one of the four original sites or by generic design criteria which were conservative for each of the sites. With the cancellation of the Tyrone plant, however, the four-site enveloping approach was modified in the seismic design area (e.g. development of spectra) for work not yet completed to include only the three remaining sites. Refer to Sections 2.5 and 3.7(B) for details. The design envelopes were not revised to reflect the cancellation of Sterling.
The ESW Vertical Loop Chase design was analyzed using the design envelopes applied to the original SNUPPS standard plant design.
1.1.3 PLANT LOCATION The site for the Wolf Creek Generating Station, Unit No. 1, is located approximately 3.5 miles northeast of the town of Burlington, in Coffey County, Kansas. The site is situated approximately 3.5 miles east of the Neosho River and the John Redmond Reservoir. The nearest population center is Emporia, Kansas, located 28 miles west-northwest of the site. It is approximately 75 miles southwest of Kansas City, Kansas.
1.1-2                            Rev. 32
 
WOLF CREEK 1.1.4 CONTAINMENT STRUCTURE The containment, which was designed by the Bechtel Power Corporation, is a carbon steel-lined, concrete structure. The walls and dome are post-tensioned, prestressed concrete, and the base slab is reinforced concrete.
1.1.5 NUCLEAR STEAM SUPPLY SYSTEM AND TURBINE-GENERATOR The nuclear steam supply system (NSSS) for Wolf Creek is a pressurized water reactor (PWR) which was designed and supplied by the Westinghouse Electric Corporation.
The reactor core was designed for an output of 3,411 MWt. When the reactor coolant pump input of 14 MWt was added to the core output, the warranted nuclear steam supply system output was 3,425 MWt, which was defined as the rated power in the license application. The engineered safety features were designed for a core power of 3,565 MWt. An additional 2 percent conservatism was added for some analyses to give a maximum accident analysis power of 3,636 MWt. Analyses were performed in 1992 to uprate the reactor core power to 3565 MWt.
The turbine generator is rated for operation at the NSSS output of 3,425 MWt.
The corresponding turbine generator electrical output is 1,186 MWe. The turbine generator has a valve wide open capability of 1,234 MWe, assuming an NSSS output of approximately 105 percent of the rated steam flow. The turbine generator was designed and supplied by the General Electric Company.
The Wolf Creek Power Rerate Program increases the licensed reactor core power level from 3411 MW(th) to 3565 MW(th). The estimated turbine-generator output is 1228 MW(e) at the Power Rerate condition, which is based on an NSSS output of 3579 MW(th) and a Reactor Coolant System hot leg (Thot) temperature of 618.2 °F.
The turbine has been upgraded with new monoblock rotors on both the high pressure (HP) and low pressure (LP) turbines. The new HP turbine is a dense pack design (increase in stages from 7 to 9). The last stage buckets in the LP turbines have been increased from 38 to 43. These efficiency enhancements resulted in an expected turbine-generator output of 1268 MWe.
1.1.6 SCHEDULE FOR FUEL LOADING AND OPERATION A low power operating license was issued for WCGS on March 11, 1985. WCGS first entered commercial operation on September 3, 1985.
1.1.7 DESIGN BASES As used within this USAR, the design bases are a list of requirements that the system must meet in order to:
: a. Perform directly a specified safety or power generation function including support of another function (e.g.,
provide cooling water flow for other components, maintain a given compartment temperature).
: b. Comply with a regulatory or statutory requirement or guideline (e.g., a jurisdictional building code).
1.1-3                          Rev. 25
 
WOLF CREEK
: c. Meet a specific operator interface, startup, or specific testing requirement.
: d. Meet a design classification or code requirement (e.g.,
be designed to withstand the safe shutdown earthquake).
Items implicit in contemporary design practices (e.g., use of the English system of weights and measures or the exercise of good engineering practice) are not specified as design bases.
Safety design bases are engineering objectives which must be met by safety-related structures, systems, or components.
Safety-related items are defined as those plant features necessary to ensure the following:
: a. The integrity of the reactor coolant pressure boundary.
: b. The capability to shut down the reactor after a design basis accident and maintain it in a post-accident safe shutdown condition.
: c. The capability to prevent or mitigate the consequences of accidents that could potentially result in offsite exposures approaching the guideline exposures of 10 CFR 50.67.
Items which are associated with safety-related equipment, but which in themselves are not absolutely essential to the safety function of the equipment, are not considered safety-related.
Power generation design bases support, either directly or indirectly,the major electrical power generation function of the station. Examples of power generation design bases are the requirements to provide adequate radiation shielding and domestic water for plant personnel.
Sections describing Westinghouse-supplied systems and components do not provide safety design bases or power generation design bases as such. These sections do give functional descriptions and are in compliance with Regulatory Guide 1.70.
1.1-4                        Rev. 34
 
WOLF CREEK TABLE 1.1-1 ACRONYMS USED IN THE USAR AC    Alternating Current ACI    American Concrete Institute ACRS  Advisory Committee on Reactor Safeguards A/E    Architect/Engineer AFAS  Auxiliary Feedwater Actuation System AFS    Auxiliary Feedwater System AISC  American Institute of Steel Construction ALARA  As Low as Reasonably Achievable AMSAC (ATWS) Mitigation System Activation Circuitry ANSI  American National Standards Institute APRM  Average Power Range Monitor ARM    Area Radiation Monitor ARW    Chemical Waste ASCE  American Society of Civil Engineers ASME  American Society of Mechanical Engineers ASTM  American Society for Testing and Materials ATWS  Anticipated Transients Without Scram AVT    All Volatile Treatment AWS    American Welding Society BOP    Balance of Plant B&PVC  Boiler and Pressure Vessel Codes BRS    Boron Recycle System BTP    Branch Technical Position CAS    Compressed Air System CCS    Condensate Cleanup System CCWS  Component Cooling Water System CDS    Condensate Demineralizer System CeCWS  Central Chilled Water System CFR    Code of Federal Regulations CFS    Condensate and Feedwater System CGCS  Combustible Gas Control System CHF    Critical Heat Flux CIS    Containment Isolation Signal ClCWS  Closed Cooling Water System CM    Center of Mass CMAA  Crane Manufacturing Association of America CP    Construction Permit CPR    Critical Power Ratio CPIS  Containment Purge Isolation System/Signal CR    Center of Rigidity CRD    Control Rod Drive CRDA  Control Rod Drop Accident CRDM  Control Rod Drive Mechanism CREA  Control Rod Ejection Accident CRVIS  Control Room Ventilation Isolation System/Signal CRW    Tritiated Waste Rev. 4
 
WOLF CREEK TABLE 1.1-1  (Sheet 2)
CSD      Cold Shutdown CST      Condensate Storage Tank CSTS    Condensate Storage and Transfer System CtCS    Containment Cooling System CVCS    Chemical and Volume Control System CWS      Circulating Water System DAC      Derived Air Concentration DBA      Design Basis Accident DBE      Design Basis Event DC      Direct Current DEPSG    Double Ended Pump Suction Guillotine DG      Diesel Generator DGB      Diesel Generator Building DoWS    Domestic Water System DNB      Departure From Nucleate Boiling DNBR    Departure From Nucleate Boiling Ratio DRW      Potentially Radioactive Nontritiated Waste DSC      Dry Shielded Canister DWMS    Demineralized Water Makeup System DWST    Demineralized Water Storage Tank DWSTS    Demineralized Water Storage and Transfer System DWT      Dead Weight Test ECCS    Emergency Core Cooling System EHC      Electrohydraulic Control EOL      End of Life EDECAIES Emergency Diesel Engine Combustion Air Intake and Exhaust System EDECWS  Emergency Diesel Engine Cooling Water System EDEFSTS  Emergency Diesel Engine Fuel Oil Storage and Transfer System EDELS    Emergency Diesel Engine Lubrication System EDESS    Emergency Diesel Engine Start System EFOST    Emergency Fuel Oil Storage Tank ER      Environmental Report ESFS    Engineered Safety Feature System ESFAS    Engineered Safety Feature Actuation System ESWS    Essential Service Water System ESWVLC  Essential Service Water Vertical Loop Chase FBIS    Fuel Building Isolation Signal FED      Floor and Equipment Drainage FDDR    Field Deviation Disposition Request FHA      Fuel Handling Accident FHS      Fuel Handling System FMEA    Failure Modes and Effects Analysis FPCC    Fuel Pool Cooling and Cleanup FPRCS    Fission Product Removal and Control System FPS      Fire Protection System FRS      Floor Response Spectra FSAR    Final Safety Analysis Report FSF      Fuel Storage Facility GDC      General Design Criteria Rev. 35
 
WOLF CREEK TABLE 1.1-1 (Sheet 3)
GRWS  Gaseous Radwaste System HELB  High Energy Line Break HEPA  High Efficiency Particulate Air (filter)
HEX  Heat Exchanger Hga  Inches of Mercury Absolute HSM  Horizontal Storage Module HSST  Heavy Section Steel Technology HVAC  Heating, Ventilation and Air Conditioning IAC  Interim Acceptance Criteria IEEE  Institute of Electrical and Electronics Engineers ILRT  Integrated Leakage Rate Test ISFSI Independent Spent Fuel Storage Installation ISI  Inservice Inspection LCO  Limiting Condition of Operation LEFM  Linear Elastic Fracture Mechanics LOCA  Loss-of-Coolant Accident LPRM  Local Power Range Monitor LPZ  Low Population Zone LRWS  Liquid Radwaste System LRW  Potentially Radioactive Secondary Liquid Waste LSP  Low Suction Pressure MCARS Main Condenser Air Removal System MCC  Motor Control Center MCES  Main Condenser Evacuation System MCPR  Minimum Critical Power Ratio MFIV  Main Feedwater Isolation Valve MG    Motor Generator Set MOV  Motor-Operated Valve MPC  Maximum Permissible Concentration MS    Manufacturer's Standard MSIV  Main Steam Isolation Valve MSL  Mean Sea Level MSLB  Main Steam Line Break MSSS  Main Steam Supply System NDT  Nondestructive Testing NDTT  Nil-Ductility Transition Temperature NFSF  New Fuel Storage Facility NPSH  Net Positive Suction Head NRC  Nuclear Regulatory Commission NSSS  Nuclear Steam Supply System OBE  Operating Basis Earthquake OL    Operating License OPS  Offsite Power Systems ORE  Occupational Radiation Exposures PA    Public Address PAMS  Post-Accident Monitoring System PCT  Peak Cladding Temperature PHS  Plant Heating System Rev. 35
 
WOLF CREEK TABLE 1.1-1 (Sheet 4)
P&ID  Piping and Instrumentation Diagram PLS    Precautions, Limitations, and Setpoints PMF    Probable Maximum Flood PRA    Peak Recording Accelograph PRM    Process Radiation Monitoring PSAR  Preliminary Safety Analysis Report PSS    Process Sampling System PWR    Pressurized Water Reactor RCP    Reactor Coolant Pumps RCPB  Reactor Coolant Pressure Boundary RCS    Reactor Coolant System RHR    Residual Heat Removal RMWCS  Reactor Makeup Water Control System RMWS  Reactor Makeup Water System RMWST  Reactor Makeup Water Storage Tank RO    Reactor Operator RPV    Reactor Pressure Vessel RRS    Required Response Spectrum RWB    Radwaste Building RWST  Refueling Water Storage Tank SACF  Single Active Component Failure SAR    Safety Analysis Report SAS    Secondary Alarm Station SGB    Steam Generator Blowdown SGBIS  Steam Generator Blowdown Isolation System/Signal SGBS  Steam Generator Blowdown System SIS    Safety Injection Signal SIT    Structural Integrity Test SJAE  Steam Jet Air Ejectors SLWS  Secondary Liquid Waste System SMA    Strong Motion Accelerometer SNUPPS Standard Nuclear Unit Power Plant System SRO    Senior Reactor Operator SRP    Standard Review Plan SRS    Solid Radwaste System SRSS  Square Root of the Sum of the Squares SRW    Detergent Waste SSE    Safe Shutdown Earthquake SWS    Service Water System TBS    Turbine Bypass System TG    Turbine Generator TGSS  Turbine Gland Sealing System TRS    Test Response Spectrum UHS    Ultimate Heat Sink USAR  Updated Safety Analysis Report USGS  U.S. Geological Survey Rev. 14
 
WOLF CREEK TABLE 1.1-1 (Sheet 5)
UT    Ultrasonic Testing VWO  Valves Wide Open W    Westinghouse WCGS  Wolf Creek Generating Station WCNOC Wolf Creek Nuclear Operating Corporation Rev. 2
 
WOLF CREEK 1.2  GENERAL PLANT DESCRIPTION This section describes the plant site, general arrangement of plant structures, design criteria and general design of major systems. Comparisons between the WCGS design and to the design at other plants were made at the time of application for the Operating License. These comparisons are considered historic references and will not be updated to reflect changes in other utilities designs.
1.2.1 PLANT SITE DESCRIPTION 1.2.1.1  Site Location The WCGS site is located approximately 3.5 miles northeast of the town of Burlington in Coffey County, Kansas. The site is situated approximately 3.5 miles east of the Neosho River and the John Redmond Reservoir. It is approximately 75 miles southwest of Kansas City, Kansas. Site location is discussed in more detail in Section 2.1.1.
1.2.1.2  Site Ownership The Licensees have either purchased or have obtained easements on the necessary land within the site boundary. Full ownership control of the exclusion area is presently exercised by the Operating Agent and will continue including all mineral rights with full authority to determine all activities within the exclusion area including exclusion or removal of personnel and property from the area. This is in accordance with the exclusion area requirements of 10 CFR 100.3(a).
1.2.1.3  Access to the Site There are no public highways, county and/or township roads, public waterways, or public railroads that traverse the exclusion area. There are no persons living in the exclusion area. There is to be no one working in the exclusion area except employees of the Applicants and their authorized agents. The exclusion area is patrolled periodically by plant guards to ensure awareness of access to the area by individuals. See Section 2.1.2 for further details on access to the exclusion area and details concerning lake use. Controlled access to the protected area is monitored by guards on a 24 hour per day basis.
1.2.1.4  Environs The area within 10 miles of the site is rural and of low population; in 1980 the population within the 10-mile radius was 6,652. The only incorporated 1.2-1                        Rev. 0
 
WOLF CREEK places,as defined by the U.S. Bureau of Census, within 10 miles of the site are Burlington and New Strawn. In 1980, Burlington recorded a population of 2,901 and New Strawn, which was incorporated in 1971, had a 1980 population of 457. Topeka, located 53 miles north of the site, recorded a metropolitan population of 115,266 in 1980 and Emporia, 28 miles west-northwest of the site, in 1980 recorded a population of 25,287. Most of the land surrounding the site is used for agricultural purposes with the exception of such rural service centers as Burlington. All recreational facilities are city-owned parks with the exception of the John Redmond Reservoir area, the main recreational facility in the area which is federally operated. Use of the Wolf Creek lake for recreation is discussed in Section 2.1.2. The region is expected to retain its distinctly rural character.
The Low Population Zone is chosen as the area within a 2.5-mile radius of the plant site.
1.2.1.5  Geology The WCGS site is located within the Central Stable Region of the North American Continent. This region was subjected to gentle structural uparching and down-warping during Mesozoic and Paleozoic time. These structural movements resulted in the formation of broad-scale basins and arches which have been modified locally by folding and faulting. Geotechnical investigations at the site during construction excavation have identified the presence of localized zones of penecontemporaneous deformation in the bedrock. However, the investigations have established the last age of deformation as Pennsylvanian, and there is no known macroseismic activity associated with these zones and no structural association with capable faults (Reference 1). The faulting, shearing and deformation, therefore, are noncapable as defined by Appendix A to 10 CFR 100.
The surface bedrock in the site area consists of alternating layers of Pennsylvanian age shales, limestones, sandstones, and a few thin coal seams.
These bedrock units dip gently to the west and northwest and have been folded locally into small-scale plunging anticlines and synclines. At the site, the Precambrian basement is present at a depth of approximately 2,500 feet. The Precambrian rocks consist of approximately 1,000 feet of sedimentary deposits which rest on a granitic basement complex.
The site area has been submaturely to maturely dissected by the Neosho River and its tributaries to form flat to gently rolling uplands with a maximum topographic relief of 100 feet or less from the uplands to the valley floors.
Residual soils ranging in thickness from 0 to 16 feet are developed 1.2-2                        Rev. 10
 
WOLF CREEK on the Pennsylvanian strata. Quaternary alluvium reaches a thickness of approximately 25 feet in the Wolf Creek valley. Scattered Tertiary age deposits of clayey gravel cap some of the higher hills in the site area.
Glacial deposits are not present at the site. The alternating Pennsylvanian strata forming the bedrock surface consist of competent rock units with a low amount of structural discontinuities in the rock mass. No major geologic features have been identified which could adversely affect the stability of subsurface materials at seismic Category I facilities. Minor geologic features, such as jointing, the zones of penecontemporaneous deformation, and the weathering profile in the rock, were considered during design and construction of facilities. Comprehensive geotechnical investigations of the site have determined the subsurface conditions in adequate detail to provide design criteria for foundation support of safety-related facilities. Major seismic Category I structures are supported on competent rock. Only minor, localized modifications to foundation materials were required in design and construction to provide uniform support of safety-related facilities.
1.2.1.6  Seismology The plant site is located in a relatively seismic stable region of the central United States. No earthquake epicenter has been reported closer than 40 miles to the site, and the nearest shocks have had epicentral intensities no greater than Intensity III. At distances of 85 and 105 miles from the site, earthquakes of Intensity VII to VII-VIII have been recorded. Since 1800, only seven earthquakes of Intensity V or greater have occurred within 100 miles of the site, and 16 events of Intensity VI or greater have been recorded within 200 miles. Previously recorded earthquakes probably have not generated intensities greater than VI at the site, and none of the buildings in the vicinity of the site have sustained any known structural damage due to earthquakes.
An Operating Basis Earthquake corresponding to a horizontal acceleration of six percent of gravity and a Safe Shutdown Earthquake corresponding to a horizontal acceleration of 12 percent of gravity was selected for the site. However, a seismic evaluation of these structures, systems, and components using the Lawrence Livermore Laboratories spectrum anchored at 0.15g for structures supported on bedrock is contained in Appendix 3C.
1.2.1.7  Hydrology 1.2.1.7.1  Surface Water Hydrology The plant site is located within the Wolf Creek watershed northeast of Burlington, Kansas. The topography within the watershed varies from 1.2-3                        Rev. 14
 
WOLF CREEK undulating hills upstream of the plant site to a floodplain area shared with the Neosho River with a drainage area within Kansas of 6,300 sq. miles near the mouth of Wolf Creek with a drainage area of 35 sq. miles. The cooling lake alters the draining pattern of the watershed, but safety-related facilities are protected from severe hydrological events.
The cooling lake is designed to supply adequate cooling water to the plant during a one in fifty year drought. Makeup water is supplied to the cooling lake from the Wolf Creek watershed runoff and from makeup water pumped from John Redmond Reservoir. The region surrounding the site is not characterized by events such as tsunamis, surge activity, or severe ice flooding. Major dam failures on the Neosho River above Wolf Creek watershed will not affect safety-related facilities.
The flow of the Neosho River is controlled by three reservoirs above the site.
The Maximum flood design elevation resulting from the probable maximum flood routed through the cooling lake with coincident wave activity, is below the plant site grade. Reference Tables 2.4-16 and 3.4-1.
1.2.1.7.2  Groundwater Hydrology Only small quantities of groundwater are available within a 50-mile radius of the plant site. The groundwater is produced from three types of aquifers: the alluvial deposits in the river valleys, the weathered bedrock including the shallow soil, and the unweathered bedrock.
The alluvial aquifers are composed of silts, sands, and gravel. Yields from wells in the alluvial aquifers are up to 100 gallons per minute. Recharge to such aquifers occurs from precipitation and from rivers during periods of high flow. Regionally, discharge from the alluvial aquifers normally flows into the rivers.
The weathered bedrock aquifer consists of weathered shales, siltstones, sandstones, and limestones. Pressure tests indicate that this aquifer is sufficiently permeable to yield up to 10 gallons per minute for livestock and domestic wells. Recharge occurs from precipitation and locally from downward percolation through the overlying alluvium. Discharge occurs into both alluvium and streams.
The consolidated bedrock aquifers are composed of sandstones and limestones which are limited to yields ranging from about l to 10 gallons per minute.
Recharge to such aquifers occurs by precipitation and infiltration of 1.2-4                        Rev. 33
 
WOLF CREEK surface water at the outcrops. Where overlain by shales and siltstones, which act as aquitards and aquicludes, vertical recharge to the limestones and sandstones is minimal.
There is no anticipated use of groundwater at the plant site. The operation of the plant will not have any detrimental effect on the groundwater environment, nor will local groundwater use affect the operation of the plant.
1.2.1.8  Meteorology The continental location of the site ensures a wide seasonal range of temperature and frequent day to day temperature changes due to frequent passage of cyclonic systems through the vicinity. The maximum temperature was 117 degrees Fahrenheit recorded at Burlington, Kansas. The lowest extreme temperature was -26 degrees Fahrenheit. The prevailing winds are from the south to southeast except during the winter when north to northwest winds prevail. There are no meteorologically significant terrain features or bodies of water within 50 miles of the site.
The site vicinity is subject to occasional severe thunderstorms and the possibility of a tornado from early spring until autumn. The world record 42 minute rainfall of 12 inches occurred at Holt, Missouri, approximately 120 miles from the Wolf Creek Site. However, precipitation is generally moderate throughout the year and snowfall ranges from very little during some winters to substantial during others.
The fastest wind, excluding tornadoes was 86 mph.
The diffusion climatology is generally favorable due to the frequent passage of cyclonic storm systems. The poorest diffusion conditions occur during (1) nighttime inversions which become most developed during winters and (2) dominance of the site area by stagnant anticyclonic systems which may persist for several consecutive days, especially during late summer and autumn.
1.2.2 GENERAL ARRANGEMENT OF STRUCTURES The principal structures located on the Wolf Creek Generating Station site are listed below.
: a. Reactor Building - houses the reactor, reactor coolant piping, steam generators, pressurizer, reactor coolant pumps, accumulators, and the containment air coolers; 1.2-5                        Rev. 27
 
WOLF CREEK
: b. Auxiliary Building - houses the engineered safety features and nuclear auxiliary systems equipment;
: c. Turbine Building - houses the turbine generator, condensers, main feed pumps, and other power-conversion equipment;
: d. Fuel Building - houses the new fuel storage vault, the fuel storage pool, the fuel handling system, and a portion of the spent fuel pool cooling and cleanup system;
: e. Radwaste Building - houses the radioactive waste treatment facilities and boron recycle system components;
: f. Control Building - houses the main control room, the computer,the Class IE switchgear, the Class IE battery rooms, the access control area, cable spreading rooms, and portions of the main control room emergency ventilation systems;
: g. Storage Tanks - include the condensate storage tank, the refueling water storage tank, the reactor makeup water storage tank, the demineralized water tank, and the emergency fuel oil storage tanks;
: h. Diesel Generator Building - houses the diesel generators and associated equipment;
: i. Transformer Vaults - house oil retaining pits for the main transformers, startup transformer, station service transformer, unit auxiliary transformer, and ESF transformers;
: j. Communication Corridor;
: k. Deleted
: l. Cooling lake and ultimate heat sink;
: m. Circulating Water Screenhouse - houses traveling screens, service water pumps and strainers, circulating water pumps, fire protection pumps, and chemical injection systems for raw water treatment;
: n. Essential Service Water Pumphouse - houses pumps and strainers for the essential service water system;
: o. Deleted
: p. Hot Machine Shop;
: q. Administration Building;
: r. Shop building;
: s. Materials Center (Warehouse);
: t. Technical Support Center; 1.2-6                        Rev. 30
 
WOLF CREEK
: u. Screening and Security Building(s);
: v. Covered Walkway;
: w. Education Center simulator/training complex);
: x. Non Discharging Sewage Lagoon;
: y. Switchyard;
: z. Make-up Water Screenhouse (located below John Redmond Dam);
aa. Make-up Discharge Structure; bb. Outage Processing Center; cc. Support Building West; dd. General Office Building ee. Waste Water Treatment Facility; ff. Waste Water Treatment Facility -- houses the recirculation pumps, chemical reagent storage tanks and feedpumps for the wastewater treatment system; gg. Waste Water Retention Basins -- two 300,000 gallon open top concrete basins used for retaining and neutralizing secondary regenerative wastewaters prior to discharging to the WCGS cooling lake.
hh. Owens Corning Building ii. Cathodic Protection Building (Rectifier Shelter #1) jj. Cable Reel Yard Building kk. X-Ray Building ll. Water Treatment Building North mm. Chemical Addition Building nn. Station Blackout Diesel Generator Missile Barrier oo. Cathodic Protection Building (Rectifier Shelter #2, near Firing Range) pp. ESW Vertical Loop Chase - houses both trains of the ESW vertical loops.
qq. Primary Flex Storage Building rr. Emergency Operations Facility ss. Independent Spent Fuel Storage Installation (ISFSI) - Includes ISFSI pad, approach apron, storage modules for spent fuel canisters, and haul path from fuel building to ISFSI pad.
1.2-7                        Rev. 35
 
WOLF CREEK NOTE: The above list gives a general use description of the principal structures. The actual name of the structure may differ.
The general arrangement of these and other structures and equipment is shown in Figures 1.2-1 through 1.2-43. The site area layout is shown in Figure 1.2-44.
1.2.3 PRINCIPAL DESIGN CRITERIA The plant was designed so that it could be constructed and operated to produce electric power in a safe and reliable manner. Plant design conforms to applicable codes, standards, and regulations identified in appropriate sections of the USAR.
The plant was designed, fabricated, constructed, and is operated in such a way that the release of radioactive materials to the environment is limited to values less than the limits and guideline values of applicable federal regulations pertaining to the release of radioactive materials for normal operations, abnormal events, and design basis accidents.
The plant was designed in accordance with 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, as described in Section 3.1.
The WCGS Independent Spent Fuel Storage Installation (ISFSI) was designed in accordance with 10 CFR 72, Subpart F, General Design Criteria, as described in the NUHOMS EOS System UFSAR.
1.2.3.1  SNUPPS Design Envelope The Wolf Creek power block was designed and evaluated to the SNUPPS design envelope which was established by:
: a. A design criterion which is conservative for all of the sites, or
: b. The limiting site condition existing at any site for the condition of interest.
A tabulation of the SNUPPS design envelope is presented in Table 1.2-1.
1.2.4 NUCLEAR STEAM SUPPLY SYSTEM The nuclear steam supply system (NSSS) consists of a reactor and four closed reactor coolant loops connected in parallel to the reactor vessel, each loop containing a reactor coolant pump and a steam generator. The NSSS also contains an electrically heated pressurizer and various other auxiliary systems.
High pressure water circulates through the reactor core to remove the heat generated by the nuclear chain reaction. The heated water exits from the reactor vessel and passes via the coolant loop piping to the steam generators.
Here it gives up its thermal energy to the feedwater to generate steam for the turbine generator. The cycle is completed when the water is pumped back to the reactor vessel. The entire reactor coolant system is composed of leaktight components to ensure that all fluids are confined to the system.
1.2-8                        Rev. 35
 
WOLF CREEK The core is of the multiregion type. All fuel assemblies are mechanically compatible, although the fuel enrichment is not the same in all the assemblies.
The initial reactor core design for WCGS is essentially identical to the design for the Comanche Peak units (Docket Nos. 50-445 and 50-446).
In the initial core loading, three fuel enrichments are used. Fuel assemblies with the highest enrichments are placed in the core periphery, or outer region, and the two groups of lower enrichment fuel assemblies are arranged in a selected pattern in the central region. In subsequent refuelings, some of the fuel assemblies (e.g., one-third) are discharged. The remaining fuel assemblies are, in general, moved to other positions in the core, and fresh fuel assemblies are added to fill out the core.
Rod cluster control assemblies, which consist of clusters of cylindrical absorber rods, are used for controlling core reactivity. The absorber rods move within guide tubes in certain fuel assemblies. Each absorber rod is attached to a spider connector above the core. The spider connector is attached to a drive shaft, which may be raised and lowered by a drive mechanism mounted on the reactor vessel head. The rod cluster control assemblies drop into the core under the effect of gravity when a reactor trip (SCRAM) occurs.
Supplementary reactivity control is provided by boric acid dissolved in the reactor coolant water.
The reactor coolant pumps are Westinghouse Model 93Al vertical, single stage, centrifugal pumps of the shaft-seal type. WCGS was one of the first domestic operating units to utilize Model 93Al reactor coolant pumps. However, Westinghouse Model 93Al pumps were previously reviewed by the NRC in conjunction with the RESAR-3S application (Docket No. STN-50-545). A Safety Evaluation Report (NUREG-0104) and Preliminary Design Approval (PDA-7) were issued on RESAR-3S in December 1976. In addition, the Model 93Al reactor coolant pumps are similar to the Model 93A pumps used in the Comanche Peak units (the major difference is the flow capacity as indicated in Table 1.3-1).
The pumps utilized at WCGS are identical to those used at Callaway.
The steam generators are Westinghouse Model F vertical U-tube units, which contain thermally treated Inconel tubes. The Model F steam generator includes features (discussed in detail in Section 5.4.2) such as improved tube support plate design and high circulation ratio, which are designed to minimize most forms of corrosion, sludge buildup, and chemical attack. Integral moisture separation equipment reduces the moisture content of the existing steam to one-quarter percent or less. WCGS was the second domestic operating unit to utilize Model F steam generators, (Callaway was the first domestic operating unit to utilize the Westinghouse Model F steam Generator). Westinghouse Model F steam generators were previously reviewed by the NRC in conjunction with the Sundesert PSAR application (Docket Nos. 50-582 and 50-583). An interim Safety Evaluation Report (NUREG-0469) was issued on the Sundesert PSAR application in October 1978.
1.2-9                        Rev. 35
 
WOLF CREEK Essentially all of the metal surfaces in contact with the reactor water are stainless steel, except the steam generator tubes (Inconel), fuel assembly skeleton (stainless steel, Inconel and zirconium ally) and fuel rod cladding (zirconium ally).
An electrically heated pressurizer connected to one reactor coolant loop maintains reactor coolant system pressure during normal operation and limits pressure variations during plant load transients. The pressurizer utilized at WCGS is essentially identical to those utilized in the Comanche Peak units, Callaway and many other facilities that are currently in operation.
Auxiliary system components are provided to charge makeup water into the reactor coolant system, purify reactor coolant water, provide chemicals for corrosion inhibition and reactivity control, cool system components, remove decay heat when the reactor is shut down, and provide for emergency safety injection.
1.2.5 ENGINEERED SAFETY FEATURES AND EMERGENCY SYSTEMS 1.2.5.1  Containment 1.2.5.1.1  Containment Structure The containment structure is a prestressed, post-tensioned concrete structure with a cylindrical wall, a hemispherical dome, and a flat foundation slab. The wall and dome form a prestressed, post-tensioned system consisting of horizontal tendons in the wall and inverted U-shaped vertical tendons in the wall and dome. The foundation slab is reinforced with carbon steel. The inside surface of the structure is lined with a carbon steel liner to ensure a high degree of leaktightness. The containment structure completely encloses the reactor and reactor coolant system, i.e, the reactor pressure vessel, the steam generators, the reactor coolant loops and portions of the associated auxiliary systems, the pressurizer, accumulator tanks, and associated piping, described in Section 1.2.4. The design ensures that the containment structure is protected against postulated missiles from both equipment failures and external sources. The containment design provides means for the integrated leak rate testing of the containment structure and for local leak rate testing of individual piping, electrical, and access penetrations of the containment.
For details, refer to Chapters 3.0 and 6.0.
The containment structure design is the same standard state-of-the-art design that has been applied to several other Bechtel-designed plants. The basic structure is similar to the containment structures at Farley, Palo Verde, and Turkey Point and identical to the structure at Callaway.
1.2-10                        Rev. 18
 
WOLF CREEK 1.2.5.1.2  Containment Spray System The containment spray system is one of the two independent pressure-reducing systems and is operated in conjunction with the containment fan coolers to provide adequate cooling of the containment atmosphere following a LOCA. Two of the four fan coolers and one of the two containment spray pumps operating are sufficient to cool the containment atmosphere. This reduces the pressure inside the containment, thus minimizing the release of radioactivity from the structure.
The containment spray system supplies borated water to cool the containment atmosphere. The pumps take suction from the refueling water storage tank.
When the storage tank supply is depleted, suction of the pumps is aligned to pump water from the containment sump directly into the containment during the recirculation mode of operation. Sodium hydroxide is added to the spray to remove iodine from the containment atmosphere in the post-LOCA condition.
The containment spray system is similar in design and function to the reactor building spray system at the Midland Plant, Units 1 and 2. Although the WCGS containment spray system utilizes two containment sumps, versus one for Midland, both systems function under equivalent conditions. The WCGS containment spray system is identical to the system used at Callaway.
1.2.5.1.3  Containment Cooling System The containment fan cooler system is the second of the two independent pressure-reducing systems. The system consists of four fan cooling units. The operation of two of these units and one of the containment spray pumps is sufficient to meet the design requirements for containment depressurization after a postulated LOCA. Containment atmosphere is drawn through the fan cooler units to cool the air and condense steam from the containment atmosphere after a LOCA. During normal plant operation, three fan cooler units are required to remove sensible heat generated from equipment inside the containment and maintain the containment atmosphere below 120 F.
The containment cooling system design is a state-of-the-art design used throughout the nuclear industry. Components of the containment cooling system are similar in design and function to individual components that are used in the containment cooling systems at Farley, Palo Verde, and Midland and are identical to components used in the system at Callaway.
1.2-11                        Rev. 1
 
WOLF CREEK 1.2.5.1.4  Combustible Gas Control Systems Control of combustible gases in the containment following a LOCA is provided by two 100-percent-capacity electric (thermal) hydrogen recombiners located within the containment, which maintain the post-LOCA hydrogen concentration in the containment atmosphere below the lower flammability limit.
A hydrogen purge subsystem is also provided for combustible gas control as a backup system.
The combustible gas control system including the hydrogen recombiner system, hydrogen monitoring system, and the hydrogen purge system has components that are similar in design and function to the combustible gas control system used at Midland Plant, Units 1 and and 2.
The hydrogen recombiners utilized at WCGS are essentially identical to those utilized in the Comanche Peak units, Callaway, and many other facilities that are currently in operation.
1.2.5.1.5  Containment Isolation System The containment isolation system preserves the ability of the containment boundary to minimize the release of fission products to the environment while at the same time allowing the normal and emergency passage of fluids through the containment boundary. System components include isolation valves that satisfy the containment isolation criteria. The containment isolation system is similar in design and function to the standard design that has been applied to several other Bechtel-designed plants. The containment isolation system used at WCGS is essentially identical to that utilized at Callaway.
1.2.5.2  Emergency Core Cooling System The emergency core cooling system (ECCS) injects borated water into the reactor coolant system following a LOCA. This limits damage to the fuel assemblies and limits metal-water reactions and fission product release. The ECCS also provides continuous long-term post-LOCA cooling of the core by recirculating borated water between the containment sumps and the reactor core. The ECCS design at WCGS is functionally identical to the ECCS design for the Comanche Peak units and Callaway.
1.2.5.3  Auxiliary Feedwater System When the main feedwater system is not in operation and the reactor coolant temperature is greater than 350° degrees F, the auxiliary 1.2-12                      Rev. 1
 
WOLF CREEK feedwater system is used to supply water to the secondary side of the steam generators.
The system consists of two motor-driven pumps which are powered by the emergency diesel generators if there is loss of offsite power and one steam-turbine-driven pump. During normal plant cooldown, the auxiliary feedwater system can, if necessary, be used to supply feedwater to the steam generators for removal of decay and sensible heat from the reactor coolant system. See section 7.3.6.1.1 for a description of this operation.
The auxiliary feedwater system has a design that is similar to the auxiliary feedwater system design on the Midland Plant, Units 1 and 2. Both designs utilize steam-driven and ac-powered motor-driven auxiliary feedwater pumps.
However, the WCGS design utilizes an additional motor-driven pump for reliability. The auxiliary feedwater system utilized at WCGS is identical to that used at Callaway.
1.2.6 PLANT INSTRUMENTATION AND CONTROL SYSTEMS The plant instrumentation and control systems ensure safe and orderly operation of all systems and processes over the full operating range of the plant. The control room is designed to enable operators to start up, operate, and shutdown the plant. Supervision of both the nuclear and turbine generator systems is accomplished from the control room. Additional controls at appropriate locations outside the control room (in particular, an auxiliary shutdown panel in the auxiliary building) ensure the capability of reaching and maintaining a post-accident or post-fire shutdown condition in the unlikely event that the control room becomes uninhabitable. (Note that the control room is protected from fire, breach of security, and missiles, and contains a redundant ventilation system filtered to remove iodine.)
The WCGS instrumentation and control systems summarized below and discussed in detail in Chapter 7.0 are functionally similar to those systems utilized in the Comanche Peak units and are essentially identical to those systems utilized at Callaway.
1.2-13                        Rev. 14
 
WOLF CREEK 1.2.6.1  Protection System The plant protection system monitors selected plant parameters in order to initiate reactor trip and/or engineered safety features actuation. Multiple independent channels monitor each of the selected plant parameters. The plant protection system logic was designed to initiate automatically protective action whenever the monitored parameters reach a limiting safety system setting. Redundancy was provided to assure that no single failure would prevent protective action when it was required. The plant protection system was designed in conformance to IEEE Standard 279 "Criteria for Protection Systems for Nuclear Power Generating Stations."
1.2.6.1.1  Reactor Trip System The reactor trip system shuts down the reactor whenever the limits of safe operation are approached. The safe operating region was defined by such considerations as mechanical/hydraulic limitations on equipment and heat transfer phenomena. Therefore, the reactor trip system keeps surveillance on process or calculated variables which are directly related to those equipment limitations. Whenever a direct process or calculated variable exceeds a setpoint, the reactor would automatically be shut down.
1.2.6.1.2  Engineered Safety Features Actuation System The engineered safety features actuation system was designed to detect symptoms of a loss-of-coolant accident, a steam-line break, a feedwater-line break, loss of offsite power, or a fuel handling accident and to actuate the appropriate engineered safety features systems as certain threshold levels of each indicator symptom are passed.
1.2.6.2  Reactor Instrumentation and Control System The reactor is controlled (1) by taking advantage of inherent neutronic characteristics, e.g., temperature coefficients of reactivity; (2) by control rod cluster motion, which is used for load transients and for startup and shutdown; (3) and by a soluble neutron absorber, boron, in the form of boric acid inserted during cold shutdown, partially removed at startup, and adjusted in concentration during core lifetime to compensate for fuel consumption and accumulation of fission products. The control system allows the plant to accept step load increases of up to 10 percent and ramp load increases of up to 5 percent per minute over the load range of 15 to 100 percent of full power.
Equal step and ramp load reductions are possible, over the range of 100 to 15 percent of full power.
The control system utilizes an Ovation-based distributed control system (DCS) with redundant controllers and power supplies. The DCS architecture is based on functional and hardware redundancy to create a robust and reliable system.
1.2-14                        Rev. 35
 
WOLF CREEK 1.2.6.3  Radiation Monitoring System The liquid and gaseous effluents from the plant are continuously monitored for radioactivity. Release rates are monitored and recorded. The process radiation monitoring system detects radioactivity in fluid systems which is indicative of fuel clad defects and/or fluid leakage between process systems.
Area monitoring stations are provided to measure gamma radiation at selected locations in the plant. Radiation levels, as detected by these monitors, are indicated in the control room, and above normal values are annunciated.
1.2.6.4  Balance-of-Plant Instrumentation and Control Systems The turbine and generator control systems are designed to regulate generator load. The turbine-generator protection system is designed to ensure safe operation of the unit. The analog Electro-Hydraulic Control (EHC) system has been replaced with a new digital Turbine Control System (TCS). The new system utilizes an Ovation-Based Distributed Control System (DCS). Two redundant sets of controllers are used in the turbine control system. The TCS architecture is based on combined functional and hardware redundancy to create a robust and reliable system.
Additional instrumentation and controls allow manual or automatic control of various temperatures, pressures, flows, and liquid levels throughout the plant.
Indicators, recorders, annunciators, and the plant computer inform the operating personnel at the equipment location and/or the control room of plant conditions and performance.
1.2.7 PLANT ELECTRIC POWER SYSTEM 1.2.7.1  Transmission and Generation Systems The generating units are connected to the respective utility transmission systems. The transmission system nominal voltage is 345 kV for Kansas Gas and Electric Company (KG&E) and Great Plains. The utilities have integrated transmission networks and interconnections with neighboring systems. A description of system network and interconnection for each utility is given in Chapter 8.0.
The main generator is a General Electric 1,800 rpm, three-phase, 60-cycle synchronous unit. The generator is connected directly to the turbine shaft and is equipped with an excitation system coupled directly to the generator shaft.
Power from the generators is stepped up from 25 kV by the unit main transformers and supplied by overhead lines to the switchyard. A unit auxiliary transformer is connected to the main generator through an isolated phase bus duct to supply the auxiliary loads of the unit during power generation.
1.2-15                        Rev. 35
 
WOLF CREEK 1.2.7.2  Electric Power Distribution System Electric power is supplied from the switchyard to the onsite power system for the electrical auxiliaries of each unit through two independent circuits. One circuit supplies power through a startup transformer and the other through an engineered safety features (ESF) transformer. The startup transformer feeds two 13.8-kV buses and a second ESF transformer. Power is supplied to auxiliaries at 13.8 kV, 4.16 kV, 480 V, 480/277 V, and 208/120 V ac. Refer to Figure 8.3-1.
The power distribution system includes the Class IE and non-Class IE ac and dc power systems. The Class IE power system supplies equipment used to shut down the reactor and limit the release of radioactive material following a design basis event.
The Class IE ac system for each unit consists of two independent and redundant load groups and four independent 120-V vital ac instrumentation and control power supply systems. The load groups include 4.16-kV switchgear, 480-V load centers, and motor control centers.
Two diesel generators are provided as a standby power source for each unit, one for each of the two Class IE load groups. Each generator has sufficient capacity to operate all the equipment of one load group, which is necessary to prevent undue risk to public health and safety in the event of a design basis accident.
The non-Class IE ac system includes 13.8-kV switchgear, 4.16-kV switchgear, 480-V load centers, and motor control centers.
The vital ac instrumentation and control power supply systems include battery systems, static inverters, and distribution panels. All voltages listed are nominal values, and all electrical Class IE equipment is designed to accept the expected range in voltage.
The Class IE electrical systems are similar to Class IE systems utilized on many other Bechtel-designed plants since designs that meet the requirements and standards of the nuclear industry develop in a similar manner. For instance, Class IE system and components have a similar design and function to the ac systems and components at the Midland Plant, Units 1 and 2. In addition, the Class IE dc systems and components are similar to the dc systems and components at Palo Verde. The Class IE system used at WCGS is similar to the system used at Callaway.
Direct current power for the Class IE dc loads of each unit is supplied by four independent Class IE 125-V dc batteries and 1.2-16                        Rev. 0
 
WOLF CREEK associated battery chargers. One 250-V and four 125-V non-Class 1E batteries and associated battery chargers are also provided to supply 250-V and 125-V dc power for the non-Class 1E dc system loads.
The Station Blackout Diesel Generator (SBO DG) system consists of three (3) non-safety related diesel generators that are capable of supplying essential loads on bus NB001 or bus NB002 required to reliably and safely shut down the unit following a station blackout event. The SBO DG system is also capable of supplying power to the non-safety auxiliary feedwater pump (NSAFP). Station blackout means the complete loss of alternating current (ac) electric power to the essential and nonessential switchgear buses in a nuclear power plant (i.e.,
loss of offsite electric power system concurrent with turbine trip and unavailability of the onsite emergency ac power system).
The SBO DG system is not credited for coping with a station blackout per NRC Regulatory Guide 1.155 and NUMARC 87-00, but is instead installed to provide plant MSPI/PRA margin.
The SBO DGs are located within a missile barrier designed to limit the average wind speed downstream of the barrier entrance to less than or equal to 150 mph during a 230 mph tornado event in accordance with NRC Regulatory Guide 1.76, Rev. 1.
1.2.8 POWER CONVERSION SYSTEM Thermal energy that is generated by the NSSS is converted into electrical energy through the steam cycle process by the turbine generator.
The turbine is a tandem-compound, six-flow, four-element, 1,800-rpm unit, having one high-pressure and three low-pressure elements. Combination moisture separator-reheaters are employed to dry and reheat the steam between the high-and low-pressure turbines. The auxiliaries include deaerating surface condensers, condenser evacuation system, turbine-driven main feedwater pumps, motor-driven condensate pumps, and seven stages of feedwater heating. The steam and turbine systems were designed to receive the heat energy produced in the reactor during normal operation as well as a 50-percent load reduction of the turbine-generator. Heat dissipation under the latter condition is accomplished by steam dump to the condenser (40-percent full load). The steam dump enables the plant to accept a loss of 50-percent external load without reactor or turbine trip. The condensers are cooled by the circulating water system.
1.2.8.1  Main Steam Supply System The main steam supply system consists of the piping and valves that are necessary to deliver saturated steam from the steam generators to the turbine generator. Four 28-inch main steam lines carry steam from the top of the steam generators, through four main steam isolation valves, one in each line, to the turbine stop valves at the inlet to the turbine generator. Each main steam line is also equipped with five code safety valves and one atmospheric relief valve. The main steam supply system is similar to the main steam supply system at Palo Verde and identical to the main steam supply system at Callaway.
1.2.8.2  Main Condenser Evacuation System The main condenser evacuation system provides a means for removing air and noncondensible gases from the main condenser. The main condenser evacuation system uses three vacuum pumps to perform this function; two for normal operation and the third that is started to help draw a vacuum during the startup mode.
1.2-17                        Rev. 30
 
WOLF CREEK This system is similar to the main condenser evacuation system that is utilized on Palo Verde with the exception that the Palo Verde design includes a fourth vacuum pump. The system used at WCGS is identical to the system used at Callaway.
1.2.8.3  Turbine Gland Sealing System The turbine gland sealing system seals the turbine shaft penetrations and the turbine valve stems to prevent the escape of steam or the introduction of air at these places in the steam areas of the turbine. This system utilizes standard industry components and is similar in design and function to the system utilized at San Onofre and identical to the system used at Callaway.
1.2.8.4  Turbine Bypass System The turbine bypass system, more commonly known as the steam dump system, is provided to reduce the transient effects of plant startup, hot shutdown, cooldown, and step reductions in generator loadings on the reactor coolant system. The steam dump system has the capability to bypass up to 40 percent of full steam flow from the steam generators to the main condenser. This system uses air-operated globe valves to perform its function and is similar in design and function to the steam dump system at Palo Verde and identical to the system at Callaway.
1.2.8.5  Circulating Water System The circulating water system supplies cooling water from the plant's cooling water source to the main condenser to condense the steam that discharges from the exhaust of the turbine or the turbine bypass system. The Wolf Creek site utilizes a large cooling lake for its source of circulating water and cooling mechanism as does Comanche Peak.
1.2.8.6  Condensate Cleanup System The condensate cleanup system, more commonly known as the condensate demineralizer system, contains demineralizers that are utilized to maintain the required purity of the feedwater which supplies the steam generators. The condensate demineralizer system is similar in design and function to the cleanup system utilized at Midland Plant, Units 1 and 2. The WCGS design includes additional components, such as a waste collection tank and sluice water pumps. The system used at WCGS is identical to the system used at Callaway.
1.2-18                        Rev.1
 
WOLF CREEK 1.2.8.7  Condensate and Feedwater System The condensate and feedwater system receives condensate from the condenser hotwells and delivers feedwater to the steam generators at a temperature that provides maximum steam plant efficiency.
The condensate and feedwater system includes seven stages of feedwater heaters, six demineralizer vessels, and various pumps, valves, and piping to perform its intended function. This system is similar in design and function to the condensate and feedwater system at Trojan and identical to the system used at Callaway.
During normal plant cooldown and startup, the feedwater system is used to supply feedwater to the steam generators for removal of decay and sensible heat from the reactor coolant system.
1.2.8.8  Steam Generator Blowdown System The steam generator blowdown system functions to maintain the secondary side water within the NSSS supplier's water chemistry specifications. This system includes a flash tank, filters, demineralizers, containment isolation valves, and various piping, all of which are common to most plant designs. This system, however, employs an improved design which provides much larger flow rates, four to five times larger, which enhances the blow-down function.
1.2.8.9  Secondary Liquid Waste System The secondary liquid waste system processes condensate demineralizer regeneration wastes and either directs these wastes for processing and re-use or discharge. This system also processes potentially radioactive liquid wastes. Secondary liquid waste systems are usually plant specific and depend upon the design of the systems they serve. The secondary liquid waste system is similar in design and function to most other Bechtel-designed projects.
1.2.8.10  Wastewater Treatment System The wastewater treatment facility processes wastewater discharges from the makeup demineralized water system (WM), the condensate demineralized system (AK), the secondary liquid waste system (HF), water treatment plant acid and caustic skid drains, the water treatment plant chemical spill catch basin, and the wastewater treatment facility chemical spill catch basin prior to discharge to the WCGS cooling lake. This system is designed to ensure that the above mentioned wastewater are in compliance with the pH limitations set forth in the NPDES permit. This system shall also process wastewater as allowed by the Offsite Dose Calculation Manual.
1.2-19                      Rev. 19
 
WOLF CREEK 1.2.9 AUXILIARY SYSTEMS 1.2.9.1  Chemical and Volume Control System The chemical and volume control system (CVCS) performs the following functions:
: a. Reactivity control
: b. Regulation of reactor coolant inventory
: c. Reactor coolant purification
: d. Chemical additions for corrosion control
: e. Seal water injection to reactor coolant pump seals Reactor coolant system is continuously purified by removing a small fraction of the reactor coolant flow through the letdown system. Letdown water is cooled in the regenerative heat exchanger. From there, the coolant flows to a letdown heat exchanger and through a demineralizer where corrosion and fission products are removed. The coolant then passes through a filter and is sprayed into the volume control tank, from which it is returned to the reactor coolant system by the charging system.
The CVCS automatically adjusts the amount of reactor coolant to compensate for changes in specific volume due to coolant temperature changes and reactor coolant pump shaft seal leakage in order to maintain a programmed level in the pressurizer.
The CVCS design for WCGS is similar to the CVCS design for the Comanche Peak units. The major difference is that WCGS includes provisions in the CVCS and residual heat removal system (see Section 7.4) to improve the capability to achieve and maintain cold shutdown. The CVCS system at WCGS is identical to the system used by Callaway.
1.2.9.2  Residual Heat Removal System The residual heat removal system (RHRS) is used to remove heat from the reactor coolant at a controlled rate when the reactor coolant pressure is less than approximately 425 psig and the temperature is from 350 degrees F to 140 degrees F, and to maintain the proper reactor coolant temperature during refueling.
The design of the RHRS includes two motor-operated isolation valves that are closed during normal operations. They are provided with both a prevent-open interlock and RHRS-Iso-Valve-Open alarm which are designed to prevent possible exposure of the RHRS to normal RCS operating pressure.
The isolation valves are opened for residual heat removal during a plant cooldown after the RCS temperature is reduced to approximately 350°F and RCS pressure is less than approximately 360 psig. During a plant startup, the inlet isolation valves are shut after drawing a bubble in the pressurizer and prior to increasing RCS pressure above approximately 425 psig (alarm setpoint).
1.2-20                      Rev. 14
 
WOLF CREEK The residual heat removal pumps are used to circulate the reactor coolant through two residual heat removal heat exchangers, returning it to the reactor coolant system through the lowpressure injection header.
The RHRS design for WCGS is similar to the RHRS design for the Comanche Peak units. The major difference is that at Wolf Creek, provisions are included in the CVCS and RHRS (see Section 7.4) to improve the capability to achieve and maintain cold shutdown. The RHRS used at WCGS is identical to the system used by Callaway.
1.2.9.3  Fuel Handling and Storage System The reactor is refueled using equipment designed to handle and store spent fuel under water from the time it leaves the reactor vessel until it is placed in a cask for transfer to the ISFSI or a cask for shipment from the site. Underwater transfer of spent fuel provides an optically transparent radiation shield, as well as a reliable source of coolant for removal of decay heat. This system also provides capability for receiving, handling, and storing new fuel.
The fuel handling system is divided into two areas: the reactor cavity, which is flooded for refueling, and the fuel storage pool, which is outside the reactor containment and is accessible to plant personnel. The fuel storage pool consists of the spent fuel pool and the cask loading pool (with fuel storage racks installed). The reactor cavity and the fuel storage pool are connected by a fuel transfer system which carries the fuel through an opening in the reactor containment. The fuel pool cooling and cleanup system removes decay heat from fuel stored in the spent fuel pool and maintains the purity of the fuel pool water.
Spent fuel is removed from the reactor vessel by a refueling machine and placed in the fuel transfer system. In the spent fuel pool, the fuel is removed from the transfer system and placed into storage racks. After a suitable decay period, the fuel is removed from storage and loaded into a shipping cask for off-site transport or a transfer cask when storing spent fuel at the ISFSI.
The fuel handling system and new fuel storage racks utilized at WCGS are similar to those utilized in the Comanche Peak units and many other facilities that are currently in operation.
1.2.9.4  Service Water Systems 1.2.9.4.1  Service Water System During normal plant operation, the service water system (SWS) supplies cooling water to the turbine building closed cooling water heat exchangers, central chiller condensers, and pump out units, condenser vacuum pump seal water coolers, steam packing exhauster, generator stator coolers, generator hydrogen coolers, turbine-generator lube oil coolers, steam generator blowdown non-regenerative heat exchanger, CVCS chiller unit, waterbox venting pump seal water coolers, motor driven feed pump, and air compressors. The system also supplies cooling water to the essential service water system during normal operation.
The system draws water from the cooling lake, pumps the coolant through the heat exchangers, and discharges it into the circulating water discharge, where it is directed back to the cooling 1.2-21                        Rev. 35
 
WOLF CREEK lake. Water returning from the essential service water system is returned to the Ultimate Heat Sink and/or the cooling lake. Makeup water for the cooling lake is provided by pumps in the makeup water screenhouse. The SWS is described in detail in Section 9.2.1.
The system is similar in design and function to the service water system utilized at Midland Plant, Units 1 and 2.
1.2.9.4.2  Essential Service Water System The essential service water system (ESWS) provides cooling water from the ultimate heat sink (cooling lake) for plant components which require cooling for safe shutdown of the reactor following an accident and/or loss of offsite power. These components are the component cooling water heat exchangers, containment air coolers, diesel generator coolers, safety injection pump room coolers, RHR pump room coolers, containment spray pump room coolers, centrifugal charging pump room coolers, component cooling water pump room coolers, auxiliary feedwater pump room coolers, control room air-conditioning condensers, Class 1E switchgear air-conditioning condensers, penetration room coolers, fuel pool cooling pump room cooler, and air compressor and after cooler.
The ESW cooling water is discharged to the ultimate heat sink. The essential service water pumps, prelube storage tanks, self-cleaning strainers, and traveling water screens are located in a seismic Category I pumphouse. Other parts of the system located outside the power block are either buried underground or located in seismic Category I structures. The ESWS is described in detail in Section 9.2.1.
The essential service water system provides emergency makeup to the spent fuel pool and component cooling water systems. It is also the backup water supply to the auxiliary feedwater system.
The essential service water system is similar in design and function to the essential service water system utilized at the Midland Plant, Units 1 and 2.
1.2.9.5  Component Cooling Water System The component cooling water system is a closed loop circulating water system serving heat exchangers whose operation is required for the safe shutdown of the reactor. Heat is removed from the closed loop by the essential service water system. Radiation monitors are provided to detect any radioactive leakage into the component cooling system.
1.2-22                        Rev. 14
 
WOLF CREEK The component cooling water system is similar in design and function to the component cooling water system that is utilized at the Midland Plant, Units 1 and 2 and is identical to the system utilized at Callaway.
1.2.9.6  Compressed Air Systems Four nonlubricated air compressors, with separate aftercoolers, discharge compressed air to three air receivers that supply compressed air to a common header. This header furnishes compressed air for both the plant air system and the instrument air system. Instrument air is dried and filtered downstream of the common supply header.
The plant air system provides compressed air for normal maintenance service at various stations throughout the plant. The instrument air system provides compressed air for the operation of all air-operated instruments and valves.
The compressed air system is similar in design and function to the compressed air systems that are utilized at the Trojan and Midland Plant, Units 1 and 2 and is identical to the system utilized at Callaway.
1.2.9.7  Fire Protection Systems The fire protection system was designed to provide water to the plant fire protection system and site fire protection facilities. An outside underground yard loop surrounds the power block and provides water to all buildings and hydrants spaced around the plant site. Water for the fire protection system is provided from the circulating water screenhouse intake bay. The system is described in detail in Section 9.5.1.
The plant fire protection system consists of the following materials, structures, detection devices, alarms, and suppression and extinguishing facilities, selected and designed to minimize fire hazards and fire damage:
: a. Automatic wet-pipe sprinklers;
: b. Automatic pre-action systems;
: c. Water spray systems;
: d. Halon 1301 systems;
: e. Standpipe and hoserack assemblies; 1.2-23                      Rev. 14
 
WOLF CREEK
: f. Portable extinguishers;
: g. Fire and smoke detection and alarm systems;
: h. Fire walls and barriers;
: i. Fire resistant and noncombustible materials of construction; and
: j. Smoke and heat vents; Portions of the fire protection system that protect or pass through areas containing equipment required for safe shutdown of the plant during and after an earthquake are seismically analyzed and supported to prevent damage to this equipment. The system is designed to preclude flooding of safety-related equipment under seismic conditions.
The fire protection system provides an adequate supply of water to hydrants, hose stations, sprinklers, and deluge systems, based on the maximum automatic sprinkler or fixed water spray system demand with the simultaneous flow for hose streams outside the power block.
Noncombustible and heat-resistant materials are used throughout the plant.
Plant fire walls are provided and rated according to their particular location in the plant, and penetrations through fire barriers are fitted with fire stops having, as a minimum, the same rating as the barrier.
Instrumentation and controls are provided for the proper operation of the fire-fighting systems and for fire detection and annunciation.
The fire protection system was designed with components and systems that are common to many plants throughout the nuclear industry and, therefore, is comparable to most fire protection systems utilized at other plants. This system is most similar in design and function to the fire protection system utilized at San Onofre.
1.2.9.8  Heating, Ventilating, and Air-Conditioning Systems The heating, ventilating, and air-conditioning (HVAC) systems are designed to provide a suitable environment for equipment and to ensure the safety of personnel.
Redundant cooling and ventilating systems serving engineered safety features equipment rooms and the main control room meet 1.2-24                      Rev. 0
 
WOLF CREEK seismic Category I requirements and are each supplied from separate Class IE electrical buses. These systems satisfy the single failure criterion.
The HVAC systems serving the control room areas are similar in design and function to the HVAC systems at the Midland Plant, Units 1 and 2 and Callaway.
The nonsafety-related HVAC systems are specifically tailored to suit the design of other portions of the plant but are similar in design and function to that of other Bechtel-designed projects.
1.2.9.9  Sampling Systems The sampling systems collect representative samples of the various process fluids. The systems include a primary sampling system, secondary sampling system, radwaste sampling system, and local grab sample provisions. Samples are routed to centralized sampling stations or local sample stations, all of which are located outside the reactor containment. Both liquid and gaseous samples are taken. Automatic and "on-line" analyses are made for some samples.
Chemical and radiochemical laboratory analyses are performed on other samples to determine chemical composition, boron concentration, fission and corrosion product activity levels, dissolved gas concentration, gross radioactivity, and specific isotopic analyses. The results are used to regulate boron control adjustments, monitor fuel rod integrity, evaluate demineralizer performance, control effluent releases, and maintain correct water chemistry.
The sampling system is specifically tailored to suit the design of other portions of the plant but is similar in design and function to sampling systems utilized throughout the nuclear industry.
1.2.9.10  Service Gas System The service gas system provides for the handling and storage of commonly used service gases. The service gas system has provisions to protect against nitrogen and hydrogen gas ruptures and is comparable in design and function to the service gas system at Palo Verde. The service gas system also provides its function for several other gases, e.g. carbon dioxide and oxygen.
1.2.9.11  Communications System The communications system provides components and distribution for the total communications network of the plant including intercom systems and remote communications devices. The communication system is similar in design and function to the communications system at Arkansas Nuclear One - Unit 2 and Callaway.
1.2-25                        Rev. 0
 
WOLF CREEK 1.2.9.12  Diesel Generator Support Systems The emergency diesel engine fuel oil storage and transfer system provides onsite fuel oil storage and transfer of fuel oil to the diesel engines. The storage capacity of this system is somewhat larger than the storage capacity at other plants with a similar design.
The emergency diesel engine cooling water system is a closed cycle system that provides a source of cooling water to the diesel engines. The emergency diesel engine cooling water system is the intermediate system that transfers heat between the diesel engine and the essential service water system and is similar in design and function to the typical nuclear industry design.
The emergency diesel engine starting system provides startup air to the diesel engines with two independent, redundant starting air trains per engine. The emergency diesel engine lubricating system consists of two major subsystems; 1) the main oil system, and 2) the rocker oil system. Each engine has its own independent redundant lubricating system. The emergency diesel engine combustion air intake and exhaust system provides filtered combustion air to the diesel engines and discharges the exhaust via silencers in the discharge stacks.
The diesel generator support systems are tailored to the specific design of the diesel engines and are similar in design and function to the diesel generator support systems at San Onofre and Midland Plant, Units 1 and 2 and identical to the system at Callaway.
1.2.10      WASTE PROCESSING SYSTEMS The waste processing systems provide all the equipment necessary for controlled treatment and preparation for retention or disposal of all liquid, gaseous, and solid wastes produced as a result of reactor operation. The liquid waste processing system collects, processes, and removes or concentrates radioactive constituents, and processes them until suitable for processing in the solid radwaste system. The gaseous waste processing system removes fission product gases from the reactor coolant and contains these gases during normal plant operation. The solid radwaste system receives, processes, packages, and stores all radioactive wastes generated until shipment offsite.
1.2-26                      Rev. 19
 
WOLF CREEK 1.2.11      SHARED FACILITIES AND COMPONENTS WCGS utilized the SNUPPS standard plant or power block design which was developed to be acceptable for installation at any one of several sites. Wolf Creek is a single unit site and has no shared safety-related facilities and components.
1.2.12      REFERENCES SECTION 1.2
: 1. Dames & Moore, 1977, Penecontemporaneous Deformation Zones Wolf Creek Generating Station; for Kansas Gas and Electric Co.
and Kansas City Power & Light Co., Dames & Moore, May 20, 1977.
1.2-27                        Rev. 14
 
WOLF CREEK TABLE 1.2-1 DESIGN ENVELOPE (Sheet 1) (1)
Parameter                    SNUPPS                        USAR Reference Hazards                  There are no hazards which have an        Section 2.2        Explosions from acci-adverse effect on structures                                  dents involving ex-plosives or gases were postulated in accordance with Regulatory Guide 1.91. The maximum overpressure and ground shock on the plant structures from such explosions are well below the design pressures for tornado protection and the design OBE ground accelerations.
Temperatures(2)                                                  Sections 2.3,        These temperatures 3.11(B).2.5, and 9.4 envelop the histori-
: 1. Design min. and    -60 F to 120 F                                                cally recorded mini-max. for exposed                                                                mum and maximum re-outside structures                                                              gional temperatures or components                                                                    and are within a range of 100-year
: 2. Design for ulti-    Cooling pond                                                  recurrence temperatures.
mate heat sink
: 3. Design winter      -25 F and 15 mph wind air conditions for ventilation
: 4. Design for summer  97 F dry bulb, 79 F wet bulb ventilation Flood level              Flooding is precluded by the            Section 2.4 and      No special flood elevation of the plant and by            Table 3.4-2          protection measures the site drainage system.                                    (such as external flood doors) are incorporated.
Maximum rainfall        7.4 in/hr, excess allowed to run        Section 2.4          Site drainage designed off roofs                                                    to preclude local flooding.
Rev. 6
 
WOLF CREEK TABLE 1.2-1 (Sheet 2)
Parameter                  SNUPPS                        USAR Reference    Remarks Ice and snow loading                                            Section 2.4  Basic snow load (100-year recurrence snow-
: 1. Basic snow load,  91 psf                                                pack) adjusted for normal and severe                                                      geometry and drifting environmental                                                          for roof design using Section 7.2 of ANSI
: 2. Basic snow load,  153 psf                                              A58.1-1972. "Extreme extreme environ-  environmental" includes mental                                                                  the effects of PMWP.
Ground water elevation  Maximum hydrostatic level is at          Section 2.4  Conservative assumption Grade                                                for buoyancy calcula-tions and computation of uplift pressure on foundation base mats.
Seismology (OBE and    OBE - 0.12g, SSE - 0.20g                Section 2.5  The standard plant SSE)
OBE and SSE were used with each site's soil properties to generate seismic structural loads. These loads were enveloped for design.
Floor response spectra were generated in the same manner. All items were designed either to the envelope or all of the individual floor response spectra so that these items could be interchangeable at all sites, thus the Wolf Creek site design was limited by the other SNUPPS plants floor response spectra.
Foundation character-istics
: 1. Settlement      Design settlements are within the      Section 2.5 following criteria:
: a. Total - 3 in.
: b. Post construction - 1 in.
: c. Post construction differen-tial (between buildings and/or columns) - 1/2 in.
Rev. 0
 
WOLF CREEK TABLE 1.2-1 (Sheet 3)
Parameter                            SNUPPS                      USAR Reference          Remarks
: 2. Static and dy-        The equations for the lateral earth      Section 2.5      Used for design of namic lateral          pressures, shown in Figure 2.5-152,                        subsurface Category earth pressures        are used in conjunction with the                            I walls soil parameters and the enveloping earthquake parameters (i.e., enveloping SSE and OBE) to compute the lateral pressures on the foundation walls.
The maximum earth pressure thus computed is taken as the envelope pressure and is used in design.
: 3. Liquefaction          Subsurface materials at all sites        Section 2.5 not susceptible to liquefaction
: 4. Local subsidence      No evidence of any actual or poten-      Section 2.4      WCGS is free from tial subsidence                                            major surface or sub-surface subsidence or collapse resulting from tectonic depressions, cavernous conditions, solutioning, or extraction of subsurface fluids or mining resulting from man-made activities.
Windspeed                    100 mph. Tornado maximum wind speed      Sections 3.3.1 and BC-TOP-3-A, ANSI-is 360 mph with 3 psi pressure drop      3.3.2              A58.1-1972, and Reg-in 1-1/2 secs                                              ulatory Guide 1.76.
Relative humidity(2)        97 F dry bulb                            Section 9.4 and    These are temperature 79 F wet bulb 45% (summer)              Table 9.4-1        conditions based
                            -25 F dry bulb (winter)                                    on 1972 ASHRAE Handbook of Fundamentals and are used for design of the plant HVAC systems for safety-related structures.
(1)  The design envelope was conservatively developed using data from all SNUPPS sites.
(2)  The winter temperature conditions have been re-evaluated for Wolf Creek. The acceptable design minimum temperature for exposed outside structures or components is -30°F. The acceptable design winter temperature for HVAC design is 7°F.
Rev. 14
 
WOLF CREEK 1.3  COMPARISON TABLES 1.3.1 COMPARISON WITH SIMILAR FACILITY DESIGNS Table 1.3-1 presents a design comparison of the major NSSS parameters or features of WCGS with Comanche Peak, Units 1 and 2 (Docket Nos. 50-445 and 50-446), W. B. McGuire, Units 1 and 2 (Docket Nos. 50-369 and 50-370), and Trojan (Docket No. 50-344). Wolf Creek and Callaway were both designed using the SNUPPS powerblock design. These comparisons were made at the time of application for the Operating License and are considered historic data. Table 1.3-1 will not be updated to reflect changes made at these facilities.
For a general design comparison of the major BOP systems utilized by WCGS with similar systems on other Bechtel-designed plants, refer to the general system descriptions in Section 1.2.
Refer to Table 1.3-2 for a listing of major analyses that have been used on WCGS but are not included in topical reports. Most of these analyses have been previously reviewed by the NRC on other dockets. Note that approved topical reports issued by Bechtel and Westinghouse are listed in Section 1.6.
1.3.2 COMPARISON OF FINAL AND PRELIMINARY INFORMATION Table 1.3-3 identifies all the significant changes that were made to the power block since submittal of the SNUPPS PSAR but prior to receipt of the operating license. Only items not reported in the PSAR and its subsequent amendments are listed in Table 1.3-3.
1.3.3 COMPLIANCE WITH NRC REGULATIONS Table 1.3-4 presents a list of NRC regulations and a corresponding description regarding the degree of compliance to each regulation. Compliance with 10 CFR Parts 20, 26, 50, 51, 55, 70, 71, 72, 73, and 100 is considered.
1.3-1                        Rev. 35
 
WOLF CREEK TABLE 1.3-1 DESIGN COMPARISON USAR Parameter or Feature    Chapter/Section    WCGS(b)    Comanche Peak W. B. McGuire  Trojan Reactor core heat      4.0, 5.0, 15.0    3,411          3,411        3,411        3,411 output, MWt Minimum DNBR for        4.1, 4.4, 15.0    >1.30          >1.30        >1.30        >1.30 design transients Total thermal flow      4.1, 4.4, 5.1      142.1          142.1        140.3        132.7 rate, 106 lb/hr Reactor coolant tem-    4.1, 4.4 peratures, F Core outlet                          621.4            621.4        620.8          619.5 Vessel outlet                        618.2            618.2        618.2          616.8 Core average                        591.8            591.8        589.4          585.9 Vessel average                      588.5            588.5        588.2          584.7 Core inlet                          558.8            558.8        558.1          552.5 Vessel inlet                        558.8            558.8        558.1          552.5 Average linear power,  4.1, 4.4          5.44            5.44        5.44            5.44 kW/ft Peak linear power for  4.1, 4.4          12.6            12.6        12.6            13.6 normal operation, kW/ft Heat flux hot channel  4.1, 4.4, 15.0    2.32            2.32        2.32            2.50 factor, FQ Fuel assembly array    4.1, 4.3          17 x 17          17 x 17      17 x 17        17 x 17 Rev. 7
 
WOLF CREEK TABLE 1.3-1 (Sheet 2)
DESIGN COMPARISON USAR Parameter or Feature    Chapter/Section      WCGS(b)    Comanche Peak W. B. McGuire    Trojan Number of fuel          4.1, 4.3            193            193          193            193 assemblies Uranium dioxide rods    4.1, 4.3            264            264          264            264 per assembly Fuel weight as uranium  4.1, 4.3            222,739        222,739      222,739        222,739 dioxide, lb Number of grids per      4.1, 4.3            8-Type R        8-Type R    8-Type R      8-Type R assembly Rod cluster control      4.1, 4.3 assemblies Number of full/part                      53/-            53/-          53/8            53/8 length assemblies Absorber material                        Ag-In-Cd        Ag-In-Cd    Ag-In-Cd      Ag-In-Cd Hafnium Clad material                            Stainless      Stainless    Stainless    Stainless Steel          Steel        Steel        Steel Clad thickness                          0.0185          0.0185      0.0185      0.0185 Equivalent core          4.1, 4.3            132.7          132.7        132.7        132.7 diameter, in.
Active fuel length, in. 4.1, 4.3            143.7          143.7        143.7        143.7 Fuel enrichment (weight` 4.1, 4.3            Core A          Unit 1 percent)
Region 1                                2.10            1.60        2.10        2.10 Region 2                                2.60            2.40        2.60        2.60 Region 3                                3.10            3.10        3.10        3.10 Rev. 7
 
WOLF CREEK TABLE 1.3-1 (Sheet 3)
DESIGN COMPARISON USAR Parameter or Feature          Chapter/Section  WCGS(b)    Comanche Peak W. B. McGuire      Trojan Number of coolant loops      5.0                  4              4              4              4 Total steam flow,            5.1              15.14          15.14          15.14          5.07 106 lb/hr Reactor vessel                5.3 Inside diameter, in.                      173              173            173            173 Inlet nozzle inside                      27-1/2          27-1/2        27-1/2            27-1/2 diameter, in.
Outlet nozzle inside                        29              29              29                29 diameter, in.
Number of reactor                          54              54              54                54 closure head studs Reactor coolant pumps        5.4.1 Horsepower                                7,000          7,000          7,000            6,000 Capacity, gpm                            100,600        99,000        99,000            88,500 Steam generators              5.4.2 Model                                        F              D              D                D Heat transfer areas, ft.2                55,000          48,000        48,000          51,500 Number of U-tubes                          5,626          4,578        4,674              3,388 Residual heat removal        5.4.7 Initiation pressure,                        ~425            ~425          ~425              ~400 psig Rev. 12
 
WOLF CREEK TABLE 1.3-1 (Sheet 4)
DESIGN COMPARISON USAR Parameter or Feature          Chapter/Section      WCGS(b)    Comanche Peak W. B. McGuire      Trojan Initiation/completion                        ~350/140        ~350/140      ~350/140      ~350/140 temperature, F Component cooling water                      105            105            95            95 design temperature, F Cooldown time after                          ~16            ~16            ~16            ~16 initiation, hr Heater exchanger removal                      39.0            39.1          34.15          34.2 capacity, 106 Btu/hr Pressurizer                      5.4.10 Heatup rate using                            55              55            55            55 heaters, F/hr Internal volume, ft3                          1,800          1,800          1,800          1,800 Pressurizer safety              5.4.13 valves Number                                          3                3                3              3 Maximum relieving                              420,000(c)      420,000          420,000      420,000 capacity, lb/hr Accumulators                  6.3 Number                                          4              4                4              4 Operating pressure,                            600            600              600            600 minimum, psig Minimum operating water                        819            950              950            870 volume, each, ft3 Rev. 16
 
WOLF CREEK TABLE 1.3-1 (Sheet 5)
DESIGN COMPARISON USAR Parameter or Feature    Chapter/Section      WCGS(b)    Comanche Peak W. B. McGuire  Trojan Centrifugal charging    6.3 pumps Number                                    2                2            2            2 Design flow, gpm                          150              150          150          150 Design head, ft                          5,800          5,800        5,800        5,800 Safety injection        6.3 pumps Number                                    2                2            2            2 Design flow, gpm                          440              425          425          425 Design head, ft                          2,780          2,680        2,500        2,500 Residual heat removal    5.4.7, 6.3 pumps Number                                    2                2            2            2 Design flow, gpm                          3,800            3,800        3,000        3,000 Design head, ft                          350              350          375          375 Instrumentation and      7.0                  (a)              (a)          (a)          (a) controls New fuel storage racks  9.1.1                21              21            21          21 center-to-center spacing, in.
Chemical and volume      9.3.4 control Total seal water                          32              32          32          32 supply flow rate, nominal, gpm Rev. 7
 
WOLF CREEK TABLE 1.3-1 (Sheet 6)
DESIGN COMPARISON USAR Parameter or Feature          Chapter/Section      WCGS(b)    Comanche Peak      W. B. McGuire      Trojan Total seal water                                12            12                  12              12 return flow rate, nominal, gpm Letdown flow, normal/maximum,                            75/120          75/120              75/120          75/120 gpm Charging flow,                                55/100          55/100              55/100          55/100 normal/maximum, gpm NOTES:
(a)    The instrumentation and control systems discussed in Chapter 7.0 are functionally similar to those systems implemented in Comanche Peak, W. B. McGuire, and Trojan.
(b)    The design conditions for WCGS listed in this Table have been changed by the Wolf Creek Power Rerate Program. However, since the comparisons were made at the time of application for the Operating License and are considered historic data, the Table will not be updated to reflect the new information.
(c)    The capacity is reduced to 415,764 lb/hr due to the setpoint change from 2485 psig to 2460 psig.
Rev. 16
 
WOLF CREEK TABLE 1.3-2 MAJOR ANALYSES NOT INCLUDED IN TOPICAL REPORTS Analysis                                            Previously Description and                    Applicable        Reviewed on Other Name                        USAR Section            Projects Control Room Habitability
: a. Control room air intake            2.3          Partial use in X/Q due to accidental                            Calvert Cliffs and releases of radiological                          Grand Gulf materials
: b. All other accidents, e.g.,          2.2          Grand Gulf explosions, toxic chemical spills, fire, etc.
Reactor Building
: a. Tendon Gallery                      3.8          Grand Gulf (1)
[CE 901 (STRUDL)]
[CE 800 (BSAP)]
: b. Base Slab Bending                  3.8          Grand Gulf (1)
[CE 779 (SAP 1.9)]
: c. Containment                        3.8          San Onofre Units Wall-Flexure                                      2 and 3 (1)
[CE 779 (SAP 1.9)]
Reactor Building Internals
: a. Secondary Shield Walls              3.8          San Onofre Units
[CE 779 (SAP 1.9)]                                2 and 3
: b. Refueling Pool                      3.8          San Onofre Units (CE 779 (SAP 1.9)]                                2 and 3
: c. Compartment Analysis                3.8          Grand Gulf (1)
[CE 901 (STRUDL)]
Other Category I Structures
: a. Structural Steel Framing            3.8          Grand Gulf Units
[CE 901 (STRUDL)]                                1 and 2
: b. Reinforced Concrete Analysis        3.8          Grand Gulf Units
[CE 901 (STRUDL)]                                1 and 2
[CE 800 (BSAP)]
Rev.0
 
WOLF CREEK TABLE 1.3-2 (Sheet 2)
Analysis                                      Previously Description and                Applicable  Reviewed on Other Name                    USAR Section      Projects Seismic
: a. Impedance Functions for        3.7(B)        Palo Verde Layered Soils
[CE 970 (LUCON)]
: b. Floor Response Spectra        3.7(B)        None (FLUSH). Although not specifically named, a description of this program is included in BC-TOP-4-A
: c. Seismic Displacement          3.7(B)        None Analysis
[CE 933 (FASS)] (DISCOM)
Piping Analysis
: a. ME-101                        3.9(B)      Grand Gulf, Farley ME-632 Used to calculate the stresses and loads in piping systems due to restrained thermal expansion; deadweight and seismic anchor movements, and earthquake
: b. ANSYS                          3.9(B)      Grand Gulf, Farley General static, thermal, and dynamic analysis for linear elastic and plastic analysis
: c. ME-210                        3.9(B)      Grand Gulf, Farley Computes local stresses in piping due to external loads Rev. 0
 
WOLF CREEK TABLE 1.3-2 (Sheet 3)
Analysis                                      Previously Description and                  Applicable    Reviewed on Other Name                      USAR Section      Projects
: d. ICES/STRUDL                      3.9(B)    Grand Gulf, Farley (CE 901)
Analysis of indeterminate frame structures, both spatial and plane.
Used to evaluate reactions, deflections, stresses, and code check Essential Service Water Vertical Loop Chase
: a. Foundation and Substructure      3.8        N/A Walls Analysis (020544.14.01-C-002)
: b. Vertical Loop Chase              3.8        N/A Structural Analysis (020544.14.01-C-005)
: c. ESW Waterhammer                  9.2.1.2    N/A Mitigation Analysis (EF-M-076)
(1)  Although this program was not necessarily used for analysis of the same structure on another plant, it was used for similar applications.
Rev. 29
 
WOLF CREEK TABLE 1.3-4 COMPLIANCE WITH NRC REGULATIONS, 10 CFR Regulation (10 CFR)              Compliance 20.1001(a) This regulation states the general purpose for which the Part 20 regulations are established and does not impose any independent obligations on licensees.
20.1001(b) This regulation describes the overall purpose of the Part 20 regulations to control the possession, use, and transfer of licensed material by any licensee, so that the total dose to an individual will not exceed the standards prescribed therein. It does not impose any independent obligations on licensees.
20.1101(b) Conformance with the ALARA principle stated in this regulation is ensured by the implementation of Company policies and appropriate Technical Specifications and health physics procedures.
Chapters 11.0 and 12.0 of the USAR describe the specific equipment and design features utilized in this effort.
20.1002    This regulation states the general purpose for the Part 20 regulations and imposes no independent obligations on those licensees to which they apply.
20.1003    The definitions contained in this regulation are adhered to in appropriate Technical Specifications and procedures and in applicable sections of the USAR.
20.1004    The units of radiation dose specified in this regulation are accepted and conformed to in all applicable station procedures.
20.1005    The units of radioactivity specified in this regulation are accepted and conformed to in all applicable station procedures.
20.1006    This regulation governs the interpretation of regulations by the NRC and does not impose independent obligations on licensees.
20.1007    This regulation gives the address of the NRC and does not impose independent obligations on licensees.
Rev. 9
 
WOLF CREEK TABLE 1.3-4 (Sheet 2)
Regulation (10 CFR)                Compliance 20.1201    The radiation dose limits specified in this regulation are complied with through the implementation of and adherence to administrative policies and controls and appropriate health physics procedures developed for this purpose.
Conformance is documented by the use of appropriate personnel monitoring devices and the maintenance of all required records.
20.2104(d) When required by this regulation, the accumulated dose for any individual permitted to exceed the exposure limits specified in 20.1201 is determined by the use of Form NRC-4. Appropriate health physics procedures and administrative policies control this process.
20.1204    Compliance with this regulation is ensured through the implementation of appropriate health physics procedures relating to air sampling for radioactive materials and bioassay of individuals for internal contamination. Administrative policies and controls provide adequate margins of safety for the protection of individuals against intake of radioactive materials. The systems and equipment described in Chapters 11.0 and 12.0 of the USAR provide the capability to minimize these hazards.
20.1701    Appropriate process and engineering controls and equipment, as described in Chapters 11.0 and 12.0 of the USAR, are installed and operated to maintain levels of airborne radioactivity as low as reasonably achievable.
20.1703    This regulation allows credit in estimating individual exposures for operators who are wearing respiratory protective equipment. Operating manuals contain procedures that ensure that approved respiratory protection equipment is being properly used and that plant practices are in compliance with Regulatory Guide 8.15, "Acceptable Programs for Respiratory Protection."
20.1703(c) This regulation prohibits the licensee from assigning protection factors higher than those specified in Appendix A and allows the Commission to authorize higher protection factors under certain conditions.
Rev. 9
 
WOLF CREEK TABLE 1.3-4 (Sheet 3)
Regulation (10 CFR)                      Compliance 20.1703(a)(2) This regulation requires the licensee to obtain specific authorization from the NRC in order to make allowance for certain respiratory protection equipment and provides the requirements for the application for such authorization.
20.1703(c)    This regulation states the requirements for respiratory equipment which can be used as emergency devices.
20.1703(d)    The proper notification specified by this regulation will be made to the appropriate authority within the appropriate time limit.
20.1207      Conformance with this regulation is ensured by appropriate company policies regarding employment of individuals under the age of 18 and the station procedures restricting these individuals' access to the station restricted areas.
20.1301(c)    Chapter 11.0 of the USAR provides the information and related radiation dose assessments specified by this regulation.
20.1301(c)    The radiation dose rate limits specified in this regulation are complied with through the implementation of station procedures, Technical Specifications, and administrative policies which control the use and transfer of radioactive materials. Appropriate surveys and monitoring devices document this compliance.
20.1301(a)    Conformance with the limits specified in this regulation is ensured through the implementation of station procedures and applicable Technical Specifications which provide adequate sampling and analyses and monitoring of radioactive materials in effluents prior to and during their release. The level of radioactivity in station effluents is minimized to the extent practicable by the use of appropriate equipment designed for this purpose, as described in Chapter 11.0 of the USAR.
20.1301(c)    The Operating Agent has not and does not currently 20.106(c) intend to include in any license or amendment application proposed limits higher than those specified in 20.1301(a), as provided for in these regulations.
Rev. 9
 
WOLF CREEK TABLE 1.3-4 (Sheet 4)
Regulation (10 CFR)                      Compliance 20.1206      This regulation allows for Planned Special Exposures that are authorized by the licensee. These exposures are tracked separately with their own exposure limits. The lifetime whole body limit for Planned Special Exposure is 25 rem.
20.1302(b)(2) Appropriate allowances for dilution and disper-sion of radioactive effluents are made in conformance with this regulation, and are described in detail in Chapter 11.0 of the USAR and in appropriate reports required by the Technical Specifications.
20.1301      This regulation provides criteria by which the Commission may impose further limitations on releases of radioactive materials made by a licensee. It imposes no independent obligations on licensees.
20.1301(a)    This regulation states that the provisions of 20.2003 do not apply to disposal of radioactive material into sanitary sewage systems. It imposes no independent obligations on licensees.
20.1301(d)    This regulation pertains to licensees engaged in Uranium fuel cycle operations and does not apply to WCGS.
20.1002      This regulation clarifies that the Part 20 regulations are not intended to apply to the intentional exposure of patients to radiation for the purpose of medical diagnosis or therapy. It does not impose any independent obligations on licensees.
20.1204      Necessary bioassay equipment and procedures, including Whole Body Counting, are utilized at the station to determine exposure of individuals to concentrations of radioactive materials.
Appropriate health physics procedures and administrative policies implement this require-ment.
20.1501(a)    The surveys required by this regulation are performed at adequate frequencies and contain such detail as to be consistent with the radiation hazard being evaluated. Applicable health physics procedures require these surveys and provide for their documentation in such a manner as to ensure compliance with the regulations of 10 CFR 20.
Rev. 9
 
WOLF CREEK TABLE 1.3-4 (Sheet 5)
Regulation (10 CFR)                    Compliance 20.1502    Applicable health physics procedures set forth policies and practices which ensure that all individuals are supplied with and required to use appropriate personnel monitoring equipment. Work procedures are established to provide additional control of personnel working in radiation areas and to ensure that the level of protection afforded to these individuals is consistent with the radiological hazards in the work place.
20.1501    The terminology set forth in this regulation is accepted and conformed to in all applicable station procedures, Technical Specifications, and those portions of the station procedures in which its use is made.
20.1501(c) This regulation pertains to personnel dosimeter processing and evaluation and is conformed to through appropriate Health Physics procedures.
20.1901(a) All materials used for labeling, posting, or otherwise designating radiation hazards or radioactive materials, and using the radiation symbol, conform to the conventional design pre-scribed in this regulation.
20.1902(a) This regulation is conformed to through the implementation of appropriate health physics procedures relating to posting of radiation areas, as defined in 10 CFR Part 20.1501.
20.1902(b) The requirements of this regulation for "High Radiation Areas" are conformed to by the imple-mentation of the Technical Specifications and appropriate health physics procedures. The controls and other protective measures set forth in the regulation are maintained under the surveillance of the station Health Physics group.
20.1902(d) Each Airborne Radioactivity Area, as defined in this regulation, is required to be posted in accordance with appropriate health physics procedures. These procedures also provide for the surveillance requirements necessary to determine airborne radioactivity levels.
Rev. 9
 
WOLF CREEK TABLE 1.3-4 (Sheet 6)
Regulation (10 CFR)                Compliance 20.1902    The area and room posting requirements set forth in this regulation pertaining to radioactive materials are complied with through the implementation of appropriate health physics procedures.
20.1902(e) The container labeling requirements set forth in this regulation are complied with through the implementation of appropriate health physics procedures.
20.1904(a) The posting requirement exceptions described in this regulation are used where appropriate and necessary at the station. Adequate controls are provided within the station health physics procedures to ensure safe and proper application of these exceptions.
20.1906    All of the requirements of this regulation per-taining to procedures for picking up, receiving, and opening packages of radioactive materials are implemented by the station procedures and appropriate health physics procedures. These procedures also provide for the necessary documentation to ensure an auditable record of compliance.
20.1801    The storage and control requirements for licensed 20.1802    materials in unrestricted areas are conformed to and documented through the implementation of station health physics procedures.
20.2001    The general requirements for waste disposal set forth in this regulation are complied with through station health physics procedures, the Technical Specifications, and the provisions of the station license. Chapter 11.0 of the USAR describes the solid waste disposal system installed at the station.
Rev. 9
 
WOLF CREEK TABLE 1.3-4 (Sheet 7)
Regulation (10 CFR)                  Compliance 20.2002    No such application for proposed disposal procedures as described in this regulation is contemplated.
20.2003    No such plans for disposal of licensed material by release into sanitary sewage systems as provided for in this regulation are contemplated.
20.2004    No such incineration of licensed material as provided for in this regulation is contemplated.
20.2005    This regulation pertains to disposal of specific wastes and is not applicable to WCGS.
20.2006    This regulation provides requirements which are designed to control transfers of radioactive waste intended for a land disposal facility and establishes a tracking system, in addition to supplementing existing requirements concerning transfer and record keeping. These requirements are met via implementation of appropriate health physics procedures.
20.2101    All of the requirements of this regulation are 20.2103    complied with through the implementation of appropriate Technical Specifications and health physics procedures pertaining to records of-surveys, radiation monitoring, and waste dis-posal. The retention periods specified for such records are also provided for in these specifications and procedures.
20.2201    The station has established an appropriate inventory and control program to ensure strict accountability for all licensed radioactive materials. Reports of theft or loss of licensed material are required by reference to the regulations of 10 CFR in the Technical Specifi-cations.
20.2202    Notifications of accidents, as described in this regulation, are assured by the requirements of the Technical Specifications and appropriate health physics procedures, which also provide for the necessary assessments to determine the occurrence of such incidents.
Rev. 9
 
WOLF CREEK TABLE 1.3-4 (Sheet 8)
Regulation (10 CFR)                  Compliance 20.2203    Reports of overexposures to radiation and the occurrence of excessive levels and concentra-tions, as required by this regulation, are provided for by reference in the Technical Specifications and in appropriate health physics procedures.
20.2206    The personnel monitoring required by this regulation is provided for by the Technical Specifications. Appropriate health physics procedures establish the data base from which this report is generated.
20.2206    The report of radiation exposure required by this regulation upon termination of an individual's employment or work assignment is generated through the provisions of a station health physics procedure.
20.2301    This regulation provides for the granting of exemptions from 10 CFR Part 20 regulations, provided that such exemptions are authorized by law and will not result in undue hazard to life or property. It does not impose independent obligations on licensees.
20.2302    This regulation describes the means by which the Commission may impose upon any licensee requirements which are in addition to the regulations of Part 20. It does not impose independent obligations on licensees.
20.2401    This regulation describes the remedies which the Commission may obtain in order to enforce its regulations, and sets forth those penalties or punishments which may be imposed for violations of its rules. It does not impose any independent obligations on licensees.
26.1-26.11  Subpart A, This subpart prescibes requirements and standards for the establishment, implementation, and maintenance of fitness-for-duty (FFD) programs.
26.21-26.41 Subpart B, This subpart requires the establishment, implementation, and maintenance of FFD progrzms.
26.51-26.71 Subpart C, This subpart specifies the requirements to grant initial authorization, authorization update, authorization reinstatement, or authorization with potentially disqualifying FFD information.
26.73-26.77 Subpart D, This section defines the minimum sanctions that licensees and other entities shall impose when an individual has violated the drug and alcohol provisions of an FFD policy.
Rev. 23
 
WOLF CREEK TABLE 1.3-4 (Sheet 9)
Regulation (10 CFR)                      Compliance 26.81-26.119  Subpart E, This subpart contains requirements for collecting specimens for drug testing and conducting alcohol tests by or on behalf of the licensees and other entities.
26.151-26.169 Subpart G, This subpart contains requirements for the HHS-certified laboratories that licensees and other entities who are subject to this part use for testing urine specimens for validty and the presence of drugs and drug metabolites.
26.181-26.189 Subpart H, This subpart contains requirements for determining whether a donor has violated the FFD policy and for making a determination of fitness.
26.201-26.211 Subpart I, This subpart contains requirements for work hour controls and rest-break periods for select categories of workers.
26.709-26.719 Subpart N, This subpart contains the requirements for maintaining records and submitting certain reports to the NRC.
26.821-26.825 Subpart O, This subpart requires the allowance of duly authorized NRC inspectors.
50.1          This regulation states the purpose of the Part 50 regulations and does not impose any independent obligations on licensees.
50.2          This regulation defines various terms and does not impose independent obligations on licensees.
50.3          This regulation governs the interpretation of the regulations by the NRC and does not impose independent obligations on licensees.
50.4          This regulation gives the address of the NRC and does not impose independent obligations on licensees.
50.7          This regulation provides the requirements for employee protection and provides for remedy on the part of the employee who is discriminated against for engaging in certain protected activities as well as the penalty for violation.
Rev. 23
 
WOLF CREEK TABLE 1.3-4 (Sheet 10)
Regulation (10 CFR)                  Compliance 50.8      This regulation provides the NRC information collection requirements and specifies associated OMB approval.
50.9      This regulation provides the requirements for completeness and accuracy of information provided by the licensee to the NRC.
50.10      These regulations specify the types of activities 50.11 that may not be undertaken without a license from the NRC. The Operating Agent does not propose to conduct any such activities at Wolf Creek without an NRC license.
50.12      This regulation provides for the granting of exemptions from 10 CFR Part 50 regulations, provided that such exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. It does not impose independent obligations on licensees.
50.13      This regulation says that a license applicant need not design against acts of war. It imposes no independent obligations on licensees.
50.20      These regulations described the types of licenses 50.21      that the NRC issues. They do not address the sub-50.22      stantive requirements that an applicant must sat-50.23      isfy to qualify for such licenses.
50.30      This regulation sets forth procedural requirements for the filing of license applications concerning items such as place of filing, oath or affirmation, number of copies of application, application for operating license, filing fees, and an environmental report. The procedural requirements of this regulation have been met in the license application and will continue to be met for subsequent amendments to the license application.
50.31      These regulations permit more efficient organiza 50.32      tion of the license application and impose no independent obligations on licensees.
50.33      This regulation requires the licensee's application to contain certain general information, such as identification of the applicant, information about the applicant's financial qualifications, and a list of regulatory agencies with jurisdiction over the applicant's rates and services. This information is provided in the operating license application.
Rev. 23
 
WOLF CREEK TABLE 1.3-4 (Sheet 11)
Regulation (10 CFR)                  Compliance 50.33a    This regulation requires applicants for construction permits to submit information required for the antitrust review. The requirements set forth by this regulation were satisfied at the time the application for a construction permit was submitted.
50.34(a)  This regulation sets forth requirements which govern the content of technical information in the Preliminary Safety Analysis Report and is relevant to the construction permit stage. The requirements of this regulation were satisfied as part of the construction permit application.
50.34(b)  An Updated Safety Analysis Report (USAR) has been prepared and submitted which addresses in the chapters indicated the information required:
: 1. Site evaluation factors - Chapter 2.0
: 2. Structures, systems, and components - Chapters 3.0, 4.0, 5.0, 6.0, 7.0, 8.0, 9.0, 10.0, 11.0, 12.0, and 15.0
: 3. Radioactive effluents and radiation protection
              - Chapters 11.0 and 12.0
: 4. Design and performance evaluation - ECCS performance is discussed and shown to meet the requirements of 10 CFR 50.46 in Chapters 6.0 and 15.0
: 5. Results of research program - Section 1.5
: 6. i. Organizational structure - Chapter 13.0 ii. Managerial and administrative controls -
Chapters 13.0 and 17.0. Chapter 17.0 discusses compliance with the quality assurance requirements of Appendix B.
iii. Preoperational testing and initial operations - Chapter 14.0 iv. Plans for conduct of normal operations -
Chapters 13.0 and 17.0. Surveillance and periodic testing is specified in the Technical Specifications.
: v. Plans for coping with emergencies -
Emergency Plan.
vi. Technical Specifications vii. Potential hazards analysis (Appendix 3B)
: 7. Technical qualifications - Chapter 13.0
: 8. Operator requalification program - Chapter 13.0 Rev. 23
 
WOLF CREEK TABLE 1.3-4 (Sheet 12)
Regulation (10 CFR)                  Compliance 50.34(c)  The information required in these sections was 50.34(d)  submitted for Wolf Creek pursuant to Paragraph 2.790(d) 10 CFR 2, "Rules of Practice." This information includes both the Physical Security Plan and the Safeguards Contingency Plan.
50.34(e)  This regulation requires that the licensee who prepares a physical security plan, a safeguards contingency plan, or a guard qualification and training plan protect the plans and other Safeguards Information against unauthorized disclosure in accordance with 10 CFR 73.21.
50.34(f)  This regulation provides additional TMI-related requirements for applicants for a construction permit whose application was pending as of February 16, 1982. Wolf Creek is not impacted by this regulation.
50.34a    This regulation sets forth the requirements for including in the construction permit application a description of the design objectives and the preliminary design of equipment to control the release of radioactive material in nuclear power reactor effluents. The requirements of this regulation were satisfied as part of the construction permit application.
50.35      This regulation is relevant to the construction permit stage rather than the operating license stage.
50.36      Technical Specifications are prepared for implemen-tation and include 1) safety limits and limiting safety settings, 2) limiting conditions for operations, 3) surveillance requirements, 4) design features, and 5) administrative controls.
Technical Specifications will take the form prescribed by NUREG 0452, Revision 3, dated November 1980 which are the "Standard Technical Specifications for Westinghouse Pressurized Water Reactors."
50.36(a)  Radiological Effluent Technical Specifications (RETS) were prepared for implementation as required by this regulation. The RETS have taken the form prescribed by NUREG 0472, Revision 2, dated July 1979.
Rev. 3
 
WOLF CREEK TABLE 1.3-4 (Sheet 13)
Regulation (10 CFR)                  Compliance 50.36b    This regulation allows the NRC to attach to and incorporate in the license additional conditions to protect the environment.
50.37      This regulation requires the applicant to agree to limit access to restricted data. This requirement was satisfied at the time of application for the construction permit.
50.38      This regulation prohibits the NRC from issuing a license to any person who is a citizen, national, or agent of a foreign country or any corporation or other entity which is owned, controlled, or dominated by an alien, a foreign corporation, or a foreign government. The licensees were eligible to apply for and obtain a license as stated in their applications for operating licenses. Therefore, the requirements of this regulation are not applicable.
50.39      This regulation provides that applications and related documents may be made available for public inspection. This imposes no direct obligations on applicants and licensees.
50.40      This regulation provides considerations to "guide" the Commission in granting licenses, as follows:
50.40(a)  The design and operation of the facility is to provide reasonable assurance that the health and safety of the public will not be endangered. The basis for the assurance that the regulations will be met and the public protected is contained in this document and in the license application and the related correspondence over the years.
Moreover, the lengthy process by which the plant was designed, constructed, and reviewed, including reviews by the architect-engineer, the NSSS vendor, the licensees individual staffs, and the NRC Staff, provides a great deal of assurance that the public health and safety will not be endangered.
50.40(b)  This regulation requires that the applicant be both technically and financially qualified to engage in the proposed activities as specified in the license application. Technical and financial adequacy of the applicants was determined to be satisfactory during the hearing process at the construction permit stage. Additional information was provided in the operating license application.
Rev. 3
 
WOLF CREEK TABLE 1.3-4 (Sheet 14)
Regulation (10 CFR)                  Compliance 50.40(c)  The issuance of a license to the applicants was not inimical to the common defense and security or to the health and safety of the public. The individual showings of compliance with particular regulations contained in this section as well as the contents of the USAR and related correspondence on the record, plus the lengthy process of design, construction, and review by the applicants, the architect-engineer, the NSSS vendor, and the government ensure that the license will not be inimical to the health and safety of the public.
Compliance with the requirements in 10 CFR 50.40(a) demonstrated that a license was not inimical to the common defense and security.
50.40(d)  The requirements set forth in this regulation were satisfied in that Environmental Reports were submitted in accordance with 10 CFR 51 as part of the operating license application.
50.41      This regulation applies to class 104 licensees, such as those for devices used in medical therapy.
The Operating Agent has not applied for a class 104 license, and therefore 50.41 is not applicable.
50.42      This regulation requires the Commission to consider additional standards in determining whether or not a license should be issued, i.e., 1) that the proposed activities will serve a useful purpose proportionate to the quantities of special nuclear material or source material to be utilized and 2) that due account will be taken of the antitrust advice provided by the Attorney General.
Information pertinent to these standards was made known to the Commission at the construction permit stage 1) by the licensing board verification of the need for power and 2) by the Attorney General's satisfactory review of the antitrust information.
An update of this information was provided with the operating license application, in accordance with Regulatory Guide 9.3.
Rev. 0
 
WOLF CREEK TABLE 1.3-4 (Sheet 15)
Regulation (10 CFR)                  Compliance 50.43      This regulation imposes certain duties on the NRC and addresses the applicability of the Federal Power Act and the right of government agencies to obtain NRC licenses. It imposes no direct obligations on licensees.
50.44      10 CFR 50.44 was revised in 2003. The revised 10 CFR 50.44 no longer defines a design-basis LOCA hydrogen release, and eliminates the requirements for hydrogen control systems to mitigate such a release. The installation of hydrogen recombiners and/or vent and purge systems required by 10 CFR 50.44(b)(3) was intended to address the limited quantity and rate of hydrogen generation that was postulated from a design-basis LOCA. The Commission has found that the hydrogen release is not risk-significant because the design-basis LOCA hydrogen release does not contribute to the conditional probability of a large relase up to approximately 24 hours after the onset of core damage. In addition, these systems were ineffective at mitigating hydrogen releases from risk-significant beyond design-basis accidents. License Amendment No. 157 was issued by the NRC on January 31, 2005 and deleted the Technical Specification requirements for the hydrogen recombiners and relocated the requirements for the hydrogen monitors.
The WCGS power block combustible gas control system is described in USAR Section 6.2.5.2. The system is designed to maintain the hydrogen concentration in containment at a safe level following a LOCA, without purging the containment atmosphere, as specified in 10 CFR 50.44(c).
The system consists of a hydrogen monitoring subsystem, and hydrogen recombiners.
Sections 6.2.5.2 and 6.2.5.3 of the USAR describe the hydrogen mixing provisions and indicate that adequate mixing occurs following a LOCA without reliance on the hydrogen mixing fans.
Section 6.2.5.2 of the USAR indicates that the recombiners or the hydrogen purge subsystem can be utilized in sufficient time to limit hydrogen concentration following a LOCA to less than 4 volume percent. In accordance with 50.44(d), the hydrogen contribution of the core metal-water reaction is assumed to be that resulting from reaction of 5 percent of the fuel cladding.
50.45      This regulation provides standards for construction permits rather than operating licenses and is therefore not pertinent to this operating license proceeding.
50.46      USAR Section 6.3 describes the emergency core cooling system and the methods used to analyze ECCS performance following the course of an accident. The results of the loss-of-coolant accident analyses presented in USAR Section 15.6.5 demonstrate conformance with 50.46.
Rev. 19
 
WOLF CREEK TABLE 1.3-4 (Sheet 16)
Regulation (10 CFR)                  Compliance 50.47      This regulation states that the NRC will not issue an operating license until adequate emergency plans have been assured based upon their evaluation of FEMA's assessments of state and local emergency plans and the NRC's assessment of the onsite emergency plans.
50.48      This regulation governs the fire protection plans required for operating nuclear power plants. USAR Section 9.5 describes the fire protection system designed to provide fire protection in accordance with 10 CFR 50 Appendix A, GDC3. USAR Appendix 9.5A provides a summary of the compliance with NRC Branch Technical Position APCSB 9.5-1, Appendix A. USAR Appendix 9.5B provides a summary of analyses performed to demonstrate that WCGS could meet the requirements of 10 CFR 50, Appendix R.
Table 9.5E-1 of USAR Appendix 9.5E provides a design comparison to 10 CFR 50 Appendix R.
50.49      This regulation provides the requirements for environmental qualification of electrical equipment important to safety for nuclear power plants.
50.50      This regulation provides that the NRC will issue a license upon determining that the application meets the standards and requirements of the Atomic Energy Act and the regulations and that the necessary notifications to other agencies or bodies have been duly made. It imposes no direct obligations on the licensees.
50.51      This regulation specifies the maximum duration of licenses. Compliance will be affected by the Commission's issuing the license in order to comply.
50.52      This regulation provides for the combining in a single license of a number of activities. It imposes no independent obligation on the licensee.
50.53      This regulation provides that licenses are not to be issued for activities that are not under or within the jurisdiction of the United States. The operation of WCGS will be within the United States and subject to the jurisdiction of the United States, as is evident from the description of the facility in Part A of the operating license application.
50.54      This regulation  specifies certain conditions that are incorporated in  every license issued. Compliance was effected by the  inclusion of these conditions in the license when it  was issued.
50.54(jj)  The regulation changes published in the Federal Register Vol. 79, No. 214, pages 65776 through 65814 on Nov. 5, 2014, relocated previous regulation 50.55a(a)(1) to 50.54(jj). As originally stated with regard to 50.55a(a)(1) and now with regard to 50.54(jj), Section 3.2 of the USAR describes compliance with this regulation.
Rev. 29
 
WOLF CREEK TABLE 1.3-4 (Sheet 17)
Regulation (10 CFR)                    Compliance 50.55        This regulation addresses conditions of con-struction permits, not operating licenses, and therefore it is not applicable to this applica-tion.
50.55a(a)    The regulation changes published in the Federal Register Vol. 79, No. 214, pages 65776 through 65814 on Nov. 5, 2014, relocated standards and other documents incorporated by reference from 50.55a(b) to a new 50.55a(a). Therefore, 50.55a(a) provides guidance concerning the approved edition and addenda of ASME Codes and IEEE Standards that are incorporated by reference in the regulations.
Note: The previous 50.55a introductory text and 50.55a(a)(2), which specified the requirements for systems and components and protection systems for nuclear power reactors, were moved into 50.55a(b),
50.55a(c), 50.55a(d), 50.55a(e), 50.55a(f), 50.55a(g) and 50.55a(h).
50.55a(b)(1) This regulation provides conditions on use of ASME BPV Code Section III.
50.55a(b)(2) This regulation provides conditions on use of ASME BPV Code Section XI 50.55a(b)(3) This regulation provides conditions on use of ASME OM Code.
50.55a(b)(4) This regulation provides conditions on use of ASME BPV Code Section III Code Cases for design, fabrication and materials.
50.55a(b)(5) This regulation provides conditions on use of ASME BPV Code Section XI Code Cases for inservce inspection and repair/replacement activities.
50.55a(b)(6) This regulation provides conditions on use of ASME OM Code Code Cases.
50.55a(c)    This regulation provides the code requirements for components which are part of the reactor coolant pressure boundary and for components which are connected to the reactor coolant system, including inservice inspection requirements.
Rev. 29
 
WOLF CREEK TABLE 1.3-4 (Sheet 18)
Regulation (10 CFR)                  Compliance Design and fabrication of the reactor vessel, reactor coolant system piping, reactor coolant pumps, and reactor coolant system valves were carried out in accordance with ASME Section III as described in Section 5 of the USAR.
50.55a(d)  These regulations apply to nuclear power plants 50.55a(e)  whose application for a construction permit was docketed after May 14, 1984.
50.55a(f)  Inservice testing (IST) requirements delineated in this part are specified in the Technical Specifications.
50.55a(g)  Inservice inspection (ISI) requirements delineated in this part are specified in the Technical Requirements Manual and Inservice Inspection Program.
50.55a(z)  The regulation changes published in the Federal Register Vol. 79, No. 214, pages 65776 through 65814 on Nov. 5, 2014, relocated te allowance for proposed alternatives contained in the previous 50.55a(a)(3) to a new paragraph 50.55a(z). 50.55a(z) allows for proposed alternatives to 50.55a paragraphs (b), (c), (d), (e),
(f), (g), and (h).
50.55(h)  As discussed in Chapter 7.0, Section 7.1, the protection systems meet IEEE 279-1971.
50.55b    This regulation has been revoked. 43 Fed. Reg.
49775.
50.56      This regulation provides that the Commission will, in the absence of good cause shown to the contrary, issue an operating license upon completion of the construction of a facility in compliance with the terms and conditions of the construction permit.
This imposes no independent obligations on the applicant.
50.57(a)  This regulation required the Commission to make certain findings prior to the issuance of the operating license.
50.57(b)  The license, as issued, contains appropriate conditions to ensure that items of construction or modification were completed on a schedule acceptable to the Commission.
50.57(c)  This regulation provided for a low-power testing license.
50.58      This regulation provided for the review and report of the Advisory Committee on Reactor Safeguards.
50.59      This regulation provides for the licensing of certain changes, tests, and experiments at a licensed facility. Technical Specifications and procedures provide implementation of this regulation.
Rev. 29
 
WOLF CREEK TABLE 1.3-4 (Sheet 19)
Regulation (10 CFR)                  Compliance 50.60      This regulation provides the acceptance criteria for fracture prevention measures for nuclear power reactors during normal operation. Section 5.3 of the USAR details vessel material parameters in terms of the fracture toughness requirements set forth in Appendices G and H of 10 CFR 50.
50.61      This regulation provides the fracture toughness requirements for protection against pressurized thermal shock events. Fracture toughness for the reactor pressure vessel is addressed in Section 5.3 of the USAR.
Compliance with Regulatory Guide 1.99 is addressed in Appendix 3A of the USAR.
50.62      This regulation specifies the requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water nuclear power plants.
50.63      This regulation pertains to the Station Blackout rule.
50.64      This regulation pertains to non-power reactors only and is not applicable to WCGS.
50.65      This regulation requires the implementation of a program to monitor the effectiveness of maintenance programs by monitoring performance of plant SSCs. Plant procedures implement and control this program.
50.68      This regulation provides the licensee with eight requirements that may be complied with in lieu of compliance with 10CFR70.24 for criticality monitoring.
WCGS complies with this regulation.
50.70      The Commission has assigned resident inspectors to WCGS and space was provided in conformance with 50.70(b)(1) through (3).
50.71      Records are and will be maintained and reports will be made in accordance with the requirements of sections (a) through (e) of this regulation and the license.
50.72      This regulation provides the immediate notification requirements for operating nuclear power reactors.
50.73      This regulation requires the licensee to submit Licensee Event Reports for certain specific events.
50.74      This regulation requires the licensee to notify the NRC pertaining to a change in Reactor Operator or Senior Reactor Operator status.
50.78      This regulation pertains to holders of construction permits and does not apply to WCGS.
50.80      This regulation provides that licenses may not be transferred without NRC consent. No application for transfer has been made by the WCGS Licensees.
Rev. 29
 
WOLF CREEK TABLE 1.3-4 (Sheet 20)
Regulation (10 CFR)                  Compliance 50.81      This regulation permits the creation of mort-gages, pledges, and liens on licensed facilities, subject to certain provisions. The regulation prohibits secured creditors from violating the Atomic Energy Act and the Commission's regulations.
50.82      This regulation provides for the termination of licenses. It does not apply to WCGS because no termination of licenses has been requested.
50.90      This regulation governs applications for amend-ments to licenses. Future request for license amendments will be made in accordance with these requirements.
50.91      This regulation provides guidance to the NRC regarding no significant hazards considerations, notices for public comment and state consultation.
50.92      This regulation provides guidance to the NRC in issuing license amendments including no significant hazards consideration determinations.
50.100    These regulations govern the revocation, suspen-50.101    sion, and modification of licenses by the Com-50.102    mission under unusual circumstances. No such 50.103    circumstances are present and these regulations are not applicable.
50.109    This regulation specifies the conditions under which the NRC may require the backfitting of a facility. This regulation imposes no independent obligations on a licensee unless the NRC proposes a backfitting requirement and, therefore, this regulation is not applicable.
50.110    This regulation governs enforcement of the Atomic Energy Act, the Energy Reorganization Act of 1974, and the NRC's regulations and orders. No enforcement action is at issue and, therefore, this regulation is not applicable.
50.120    This regulation provides guidance for the training and qualifications of Nuclear Power Plant Personnel.
This regulation establishes the requirements for a training program. Appropriate procedures control this program.
Appendix A USAR Section 3.1 discusses the extent to which the design criteria for the WCGS's plant structures, systems, and components important to safety comply with Title 10, Code of Federal Regulations, Part 50 (10 CFR 50), Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC). As presented in Section 3.1, each criterion is first quoted and then discussed in enough detail to demonstrate SNUPPS' compliance with each criterion. In some cases, detailed evaluations of compliance with the various general design criteria are incorporated in more appropriate USAR sections, and are located by reference.
Rev. 29
 
WOLF CREEK TABLE 1.3-4 (Sheet 21)
Regulation (10 CFR)                  Compliance Appendix B Chapter 17.0 describes in detail the provisions of the operating Quality program which have been implemented to meet all applicable requirements of Appendix B.
Appendix C This appendix provides a guide for establishing an applicant's financial qualifications. Financial qualifications were established at the construction permit stage, and it was found that there is reasonable assurance that the funds needed to operate the facility in compliance with the Commission's regulations are available.
Updated information addressing financial quali-fication was submitted with the operating license application.
Appendix D This appendix has been superseded by 10 CFR Part
: 51. As noted in the discussion for 10 CFR 50.40(d), the requirements of Part 51 have been satisfied.
Appendix E This appendix specifies requirements for emer-gency plans. Emergency plans were developed to provide reasonable assurance that appropriate measures can and will be taken in the event of an emergency to protect the public's health and safety and prevent damage to property. The new criteria for emergency planning developed subsequent to the event at Three Mile Island, Unit 2 were factored into the emergency plans for the WCGS utilities. The Emergency Plan and associated facilities (EOF and Alternate TSC) were updated in 2011 to comply with revised regulations pertaining to Enhancements to Emergency Preparedness Regulations (Reference Section 18).
Appendix F This appendix applies to fuel reprocessing plants and related waste management facilities, not to power reactors such as those found in WCGS plants and is, therefore, not applicable.
Appendix G Fracture toughness compliance can be found in USAR Section 5.3.1.5. Assurance of adequate fracture toughness of ferritic materials in the reactor coolant pressure boundary (ASME Code, Section III, Class 1 components) is provided by compliance with the requirements for fracture toughness testing included in NB-2300 to Section III of the ASME Code and Appendix G of 10 CFR 50.
Appendix H Reactor vessel material surveillance program requirements are delineated in this part. Technical Specifications and operating procedures have been established to implement their requirements.
Further information is provided in USAR Chapter 5.0.
Rev. 28
 
WOLF CREEK TABLE 1.3-4 (Sheet 22)
Regulation (10 CFR)                  Compliance Appendix I This appendix provides numerical guides for design objectives and limiting conditions for operation to meet the criteria "as low as is reasonably achievable" for radioactive material in light water-cooled nuclear power reactor effluents. USAR Chapters 2.0, 11.0, and 12.0 discuss the extent to which the criteria for Appendix I are met.
Appendix J Reactor containment leakage testing for water-cooled power reactors is delineated in this appendix. These requirements are given in the Technical Specifications. Additional information concerning compliance can be found in USAR Chapter 6.0, Sections 6.2.3 and 6.2.6.
Appendix K This appendix specifies features of acceptable ECCS evaluation models. As stated in USAR Section 6.3, the ECCS subsystem functional parameters are integrated so that the Appendix K requirements are met over the range of anticipated accidents and single failure assumptions.
In addition, the ECCS evaluation model used to demonstrate conformance with 10 CFR 50.46 (see USAR Section 15.6.5) is in conformance with Appendix K requirements.
Appendix L This appendix identifies the information required to be submitted by the applicant to the Attorney General to satisfy the requirements when applying for a facility license. The requirements of this appendix were satisfied prior to the time of application for the operating license.
Appendix M This appendix lists guidelines for the licensing of plants whose site requirements are not considered in the design of the plant structures. Since all WCGS sites are considered in the plant design, this appendix is not applicable.
Appendix N This appendix dictates the requirements applicable to duplicate plant designs on multiple sites. As allowed in this regulation, WCGS used a common Safety Analysis Report prior to issuance of the first update after receipt of the Operating License; however, where site specific needs were addressed Addenda were included for reference. The two reports have been merged into one Wolf Creek Specific Safety Analysis Report.
Rev. 9
 
WOLF CREEK TABLE 1.3-4 (Sheet 23)
Regulation (10 CFR)                  Compliance Appendix 0 Appendix 0 dictates guidelines for the Staff in reviewing standardization of design. No inde-pendent obligation on the licensee is required.
Appendix P Reserved.
Appendix Q Appendix Q dictates guidelines for the staff in early review of the site and does not deal with operating license review.
Appendix R Appendix R delineates the fire protection program for nuclear power facilities operating prior to January 1, 1979. Appendix 9B of the USAR provides a summary of analyses performed to demonstrate that WCGS could meet the requirements of Appendix R.
Table 9.5E-1 of the USAR appendix 9.5E provides a design comparison of WCGS to Appendix R.
51.1      This regulation states the general purpose and scope for which the Part 51 regulations are established and does not impose any independent obligations on licensees.
51.2      This regulation specifies that Subpart A of Part 51 implements section 102(2) of the NEPA act of 1969, as amended.
51.3      This regulation states that in any conflict between the general rule and Subpart A of Part 51 (or other applicable part of this chapter) the special rule governs.
51.4      The definitions contained in this regulation are adhered to in all appropriate documents.
51.5      This regulation governs the interpretation of regulations and does not impose independent obligations on licensees. This regulation specifies that interpretations of the regulations in Part 51 are not authorized other than a written interpretation by the General Counsel.
51.6      This regulation specifies the authority of the NRC in granting exemptions and does not impose independent obligations on licensees.
Subpart A 51.10      This regulation provides the purpose and scope of Subpart A which is to implement Section 102(2) of the National Environmental Policy Act (NEPA) in a manner consistent with NRC's domestic licensing and regulatory authority.
Rev. 9
 
WOLF CREEK TABLE 1.3-4 (Sheet 24)
Regulation (10 CFR)                  Compliance 51.11      [Reserved]
51.12      This regulation states that subpart A applies to the NRC's ongoing environmental work and does not require retroactive measures for environmental reports or supplements filed prior to June 7, 1984. This regulation does not impose independent obligations or licensees.
51.13      This regulation permits the NRC to take immediate action in emergencies where the health and safety of the public may be adversely affected without observing the NEPA regulations. This regulation does not impose independent obligations on licensees.
51.14      This regulation provides a pertinent definitions related to NEPA which are adhered to in all appropriate documents.
51.15      This regulation provides the requirements for establishing time schedules for NRC-NEPA processes.
51.16      This regulation provides the requirements for the submittal of proprietary information.
51.17      This regulation indicates that the NRC has submitted the information requirements to OMB related to this part of the Code of Federal Regulations.
51.20      This regulation sets forth the requirements for an applicant for filing an Environmental Impact Report. These requirements were satisfied during the review of the Environmental Reports that were submitted with the application for a construction permit and the application for an operating license.
51.21      This regulation specifies that all licensing and regulatory actions subject to Part 51 require an environmental assessment except those identified in 51.20(b) as requiring an environmental impact statement and those identified in 51.22(c) as categorical exclusions.
51.22      This regulation sets forth the criterion for an identification of licensing and regulatory actions eligible for categorical exclusion.
Rev. 3
 
WOLF CREEK TABLE 1.3-4 (Sheet 25)
Regulation (10 CFR)                  Compliance 51.23      This regulation indicates that the NRC will require no environmental reports, impact statements, assessments or other analyses in connection with issuance of license amendment for storage of spent fuel up to 30 years beyond the expiration of reactor operating license.51.25 This regulation states than an appropriate NRC staff director will determine when a categorical exclusion environmental impact statement or environment assessment should be prepared.
51.26      This regulation states that when an NRC staff director determines that an environmental impact statement will be prepared a notice of intent will be published in the Federal Register and a scoping process will be conducted.
51.27      This regulation describes the requirements for the Notice of Intent as required by 10 CFR 51.26.
51.28      This regulation specifies who the NRC staff director shall invite to participate in the scoping process for an environmental impact statement.
51.29      This regulation provides the requirements for the scoping process for an environmental impact statement.
51.30      This regulation sets fourth the requirements for an environmental assessment by the NRC.
51.31      This regulation states that the NRC staff director will make the determination (based upon environmental assessments) whether to prepare an environmental impact statement or a finding of no significant impact.
51.32-51.35 These regulations provide the requirements for a finding of no significant impact by the NRC.
51.40      This regulation provides guidance to prospective applicants or petitioners for rulemaking for consultation with the NRC staff.5 51.41      This regulation gives the NRC authority to require permit or license applicants amendment applicants or petitioners to submit information useful in aiding NRC compliance with Section 102(2) of NEPA.
Rev. 23
 
WOLF CREEK TABLE 1.3-4 (Sheet 26)
Regulation (10 CFR)                      Compliance 51.45-51.69    These regulations set fourth the requirements for environmental reports.
51.70-51.125  These regulations set fourth the requirements for environmental impact statements.
Appendix A to Subpart A This Appendix to Part 51 Subpart A provides the format for presentation of material in environmental impact statements.
55.1-55.71    These regulations set forth the requirements for nuclear power plant operator's licenses.
70.1          This regulation states the general purpose for which Part 70 regulations are established and does not impose any independent obligations on licensees.
70.2          This regulation states the general scope of Part 70 and does not impose any independent obligations on licensees.
70.3          This regulation gives the Commission the power to authorize licenses for the shipment and possession of special nuclear material.
70.4          The definitions contained in this regulation are adhered to in all appropriate documents.
70.5          This regulation sets forth the requirements for communications with the NRC regarding special nuclear materials and includes with addresses for the Director, Office of Nuclear material Safety and Safeguards.
70.6          This regulation governs the interpretation of regulations and does not impose any independent obligations on licensees.
70.7          This regulation prohibits discrimination against and otherwise protects employees of all licensee engaged in certain protected activities.
Rev. 23
 
WOLF CREEK TABLE 1.3-4 (Sheet 27)
Regulation (10 CFR)                  Compliance 70.8      This regulation sets forth the information collection requirements and specifies OMB approval.
70.9      This regulation addresses the completeness and accuracy of information provided to the NRC.
70.11 -    These regulations specify those persons exempted 70.14      from complying with Part 70. The licensees are not exempt from complying with the applicable requirements of Part 70.
70.15      Reserved.
70.18 -    These regulations list types of licenses issued 70.20b    for special nuclear material. WCGS adhered to all applicable requirements.
70.21      This regulation sets forth the requirements concerning the filing of special nuclear material license applications. The requirements of this regulation were satisfied.
70.22      This regulation sets forth the requirements concerning the contents of special nuclear material license applications. The requirements of this regulation were satisfied.
70.23      This regulation defines the requirements for the approval of an application for a license to possess special nuclear material. It does not impose independent obligations on licensees.
70.24      This regulation requires licensees to install monitors which have the capability of initiating audible alarms in the event of accidental criticality. On June 24, 1997, the NRC issued to WCNOC an exemption from the requirements of 10CFR70.24.
On November 12, 1998 the NRC issued 10CFR50.68, which provides eight criteria that may be followed in lieu of criticality monitoring per 10CFR70.24 and revised 10CFR70.24 to make any exemption ineffective so long as the licensee elects to comply to 10CFR50.68.
70.31      This regulation lists guidelines for the Com-mission to follow in issuing a license.
70.32      This regulation defines the conditions by which the licensee must abide in order to keep the special nuclear materials license. The health physics program found in Chapter 12.0, Section 12.5 provides information relating to the compliance of this regulation.
Rev. 13
 
WOLF CREEK TABLE 1.3-4 (Sheet 28)
Regulation (10 CFR)                  Compliance 70.33 -    These regulations dictate    procedural  require-70.35      ments for renewing or amending a license. The Operating Agent shall follow these guidelines when the need to renew or amend arises.
70.36      This regulation prohibits the transfer of the license. No such transfer is planned by WCGS.
70.37      This is a disclaimer of warranty and does not affect the Licensees.
70.38      This regulation sets forth the requirements for expiration and termination of licenses.
70.39      This regulation sets guidelines for the manu-facture of source material and does not apply to power units such as WCGS.
70.41      This regulation provides the requirements for authorized use of special nuclear material.
70.42      This regulation provides guidance on the transfer of special nuclear material. WCGS follows these guidelines as appropriate.
70.44      This regulation sets forth the requirements in regard to creditors concerning special nuclear material. Information concerning creditors has been included, as applicable, in the information submitted with the operating license applica-tions. The primary financial constituents have been identified and their relationships described.
70.51      This regulation sets forth the requirements in regard to licensees of special nuclear material that require them to maintain records and establish procedures for inventory of special nuclear material. At such a time when this regulation applies, records will be established and kept and procedures established to satisfy this regulation.
70.52      This regulation sets forth the requirements concerning reporting procedures in the event of an accidental criticality, loss or theft or attempted theft of special nuclear material. When applicable, the requirements of this regulation will be satisfied.
Rev. 3
 
WOLF CREEK TABLE 1.3-4 (Sheet 29)
Regulation (10 CFR)                  Compliance 70.53      This regulation sets forth the requirements for submitting Material Status Reports. Where applicable, proper procedures were developed and submitted to properly account for quantities of special nuclear material and to describe appropriate actions that should be taken in the event that material is unaccounted for.
70.54      This regulation sets forth the requirements for the reporting of special nuclear material transfers in the Nuclear Material Transaction Report. When applicable, the proper transfer documentation will be completed.
70.55      This regulation sets forth the requirements regarding the responsibilities of the licensees with respect to affording support and access to NRC inspection personnel. Provisions have been made to satisfy the requirements of this regulation in conjunction with granting approval on an application for license for special nuclear material.
70.56      This regulation sets forth the requirements for testing the administration of the regulations in 10 CFR 71. The Operating Agent will support such testing to the extent practicable under the regulation.
70.57      This regulation sets forth the requirements for operations other than those involved in the operation of a nuclear reactor licensed to Part 50, waste disposal operations or sealed sources. No such operations are contemplated; therefore, the requirements of this regulation are not applicable.
70.58      This regulation sets forth the requirements concerning use of special nuclear material other than licensed by Part 50 and in a waste disposal operation and as sealed sources.
No such use is contemplated; therefore, the re-quirements of this regulation are not applicable.
70.59      This regulation sets forth the requirements for effluent monitoring reporting for special nuclear material. This regulation pertains to fuel pro-cessing and fabrication and is not applicable to a utilization facility.
Rev. 23
 
WOLF CREEK TABLE 1.3-4 (Sheet 30)
Regulation (10 CFR)                  Compliance 70.61      These regulations allow the Commission to revoke 70.62      any license for special nuclear material. It does not impose independent obligations on licensees.
70.71      This regulation governs enforcement of the Atomic Energy Act, the Energy Reorganization Act of 1974, and the NRC's regulations and orders. No enforcement action is at issue and, therefore, this regulation is not applicable.
Subpart A 71.0      This regulation establishes the purpose, scope and applicability for the Part 71 regulations and does not impose any independent obligations on licensees .
71.1      This regulation provides the address for communications with the NRC.
71.2      This regulation states that only written interpretations of Part 71 by the NRC's General Counsel are binding.
71.3      This regulation prohibits delivery or transport of licensed material, except as authorized by the Commission.
71.4      The definitions contained in this regulation are adhered to in all appropriate documents.
71.5      This regulation specifies that transportation of licensed materials be done per the requirements of the Department of Transportation and Postal Service. This regulation shall be complied with per the revision of March 25, 1980.
71.6      This regulation states the information collection requirements submitted for OMB approval.
71.6a      This regulation governs the completeness and accuracy of information provided to the NRC.
Subpart B 71.7-71.10 These regulations delineate the exemptions from Part 71.
Rev. 3
 
WOLF CREEK TABLE 1.3-4 (Sheet 31)
Regulation (10 CFR)                    Compliance Subpart C 71.12        This regulation issues a general license for shipment in certain NRC approved containers and packages provided the licensee has an approved QA program. QA programs for WCGS are filed with the NRC during the operations phase as part of the license application.
71.13-71.24  These regulations set forth the general license requirements for shipments in specific packages or contains under a general license.
Subpart D 71.31-71.39  These regulations provide the requirements for an application for a proposed packaging design including the contents of the application, package description, package evaluation, and quality assurance requirements.
Subpart E 71.41-71.65  These regulations provide the requirements for packaging radioactive material for transport.
Compliance with each of the individual parts and paragraphs was demonstrated in the license application proceedings. These requirements of these regulations are referenced as the standards.
Subpart F 71.71-71.77  These regulations address the testing requirements for packages, containers and special form radioactive materials.
Subpart G 71.81-71.99  These regulations provide various operating controls and procedures pertinent to the transport and packaging of radioactive materials.
Subpart H 71.101-71.137 These regulations set forth the quality assurance requirements applying to design, purchases, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair and modification of components of packaging (for radioactive materials) which are important to safety.
Appendix A This regulation establishes the procedure for obtaining activity values A and A2 to be used in packaging and shipping processes.
Rev. 3
 
WOLF CREEK TABLE 1.3-4 (Sheet 32)
Regulation (10 CFR)                  Compliance 72        These regulations provide the requirements for onsite storage of spent fuel. WCNOC complies with the general license provisions of 10 CFR 72 Subpart K for the NUHOMS EOS System. For details refer to the NUHOMS EOS System UFSAR and Technical Specifications, Docket 72-1042, and the WCNOC plant specific 10 CFR 72.212 Evaluation Report.
73.1      This regulation states the general purpose and scope of Part 73 and does not impose independent obligations on the licensee.
73.2      The definitions contained in this regulation are adhered to in all appropriate documents.
73.3      This regulation governs the interpretation of regulations by the NRC and does not impose independent obligations on licensees.
73.4      This regulation gives the address of the NRC and does not impose any independent obligations on licensees.
73.5      This regulation allows the Commission to grant exemptions as long as they will not endanger life or property or the common defense and security. It does not impose independent obligations on licensees.
73.6      This regulation enumerates specific exemptions, including an exemption for the following: U-235 contained in uranium enriched to less than 20 percent in the U-235 isotope. Since this is the only special nuclear material for which WCGS is currently licensed, it is exempt from the requirements of 73.20, 73.25, 73.26, 73.27, 73.45, 73.46, 73.70, and 73.72. This regulation sets forth the information requirements established by the Commission and specifies OMB approval.
73.8      This regulation specifies the information collection requirements and submitted for OMB approval.
73.20      The licensee is exempt from the requirements of this regulation. See 73.6.
73.21      This regulation sets forth the requirements for the protection of safeguards information. These requirements are addressed by the WCGS Site Security Plan.
Rev. 35
 
WOLF CREEK TABLE 1.3-4 (Sheet 33)
Regulation (10 CFR)                  Compliance 73.24      This regulation sets forth the requirements concerning transport of special nuclear material in passenger aircraft and in quantities in excess of formula quantities. Shipments of special nuclear material use the requirements of this regulation for reference when such requirements are applicable.
73.25 -    The licensee is exempt from the requirements of 73.27      these regulations. See 73.6.
73.30 -    These requirements have been deleted.
73.36 73.37      This regulation sets forth the requirements regarding physical protection during the transport of irradiated reactor fuel.
73.40      This regulation sets forth the requirements regarding the establishment of and maintenance of physical security systems that provide physical protection against radiological sabotage and against theft of special nuclear material at fixed sites. Physical security systems are provided and maintained to provide adequate physical protection against sabotage and theft of special nuclear material. In addition, a safeguards contingency plan was prepared in accordance with the criteria in Appendix C of this part, submitted for Commission approval and implemented.
73.45      The licensee is exempt from the requirements of 73.46      these regulations. See 73.6.
73.50      This regulation sets forth the requirements for physical protection of licensed activities at other than nuclear power reactors.
73.55      This regulation sets forth the requirements for physical protection of licensed activities in nuclear power reactors against radiological sabotage. Features were implemented to provide for physical barriers, access control, detection aids, and communications along with a physical security organization that ensures physical protection. The requirements, as prescribed by this regulation, has been satisfied to the extent practicable.
73.57      This regulation establishes the requirements for Criminal History Checks of individuals granted unescorted access to a nuclear power facility or access to Safeguards Information by power reactor licensees. These requirements are addressed in the WCGS Site Security Plan.
Rev. 35
 
WOLF CREEK TABLE 1.3-4 (Sheet 34)
Regulation (10 CFR)                  Compliance 73.60      This regulation applies to non-power reactors and thus does not apply to WCGS.
73.67      This regulation sets forth licensee fixed site and in-transient requirements for the physical protection of special nuclear material of moderate and low strategic significance. The requirements of this regulation will be met in a manner similar to that described in the response to Paragraph 73.55.
73.70      This regulation sets forth the requirement for records for licensees subject to various Para-graphs in part 73. At this time the licensee is exempt from the requirements of this regulation (See 73.6).
73.71      This regulation sets forth requirements for reporting unaccounted for shipments, suspected theft, unlawful diversion, or radiological sabotage. The requirements of this regulation will be followed at such time as they become applicable.
73.72      This regulation sets forth the requirements for making advanced notice of shipment of special nuclear material. At this time the Licensee is exempt from the requirements of this regulation (See 73.6).
73.73      This regulation establishes the requirements for advance notice and protection of export shipments of special nuclear material of low strategy significance and does not apply to WCNOC since it is not licensed to export special nuclear material.
73.74      This regulation establishes the requirements for advance notice and protection of import shipments of nuclear materials from countries that are not party to the Convention of Physical Protection of Nuclear Material. It does not apply to WCNOC since it is not licensed to import special nuclear material.
73.80      This regulation governs enforcement of the Atomic Energy Act, the Energy Reorganization Act of 1974, and the NRC's regulations and orders. No enforcement action is at issue and so this regulation is not applicable.
Appendix A This appendix groups each state into regions to be supervised by the USNRC Inspection and Enforcement and requires no obligations by the licensee.
Rev. 35
 
WOLF CREEK TABLE 1.3-4 (Sheet 35)
Regulation (10 CFR)                  Compliance Appendix B The general criteria for security personnel are outlined in this appendix. The principles in this regulation were factored into the WCGS security plans.
Appendix C This regulation sets forth the requirements for licensee safeguards contingency plans. This plan has been developed and implemented.
Appendix D This appendix requires that licensees who transport or deliver to a carrier for transport irradiated reactor fuel assure that shipment escorts have completed a training program. These requirements were satisfied at the time of submittal of the operating license application.
Appendix E This regulation specifies the levels of physical protection to be applied in international transport of Nuclear material and does not apply to WCGS, since WCNOC is not involved in such transport.
Appendix F This regulation merely lists the nations that are parties to the convention on the physical protection of nuclear material.
Appendix G This regulation sets forth the requirements for 3 reportable safeguards events and is addressed by the WCGS Site Emergency Plan.
100.1      This regulation is explanatory and does not impose independent obligations on licensees.
100.2      This regulation is explanatory. WCGS is not novel in design and is not unproven as a prototype or pilot plant.
100.3      The definitions contained in this regulation are adhered to in all appropriate documents.
100.8      This regulation sets forth the information requirements established by the NRC and specifies OMB approval.
Rev. 3
 
WOLF CREEK TABLE 1.3-4 (Sheet 36)
Regulation (10 CFR)                  Compliance 100.10    The factors listed related to both the unit design and the site have been provided in the applica-tion. Site specifics, including seismology, meteorology, geology, and hydrology, are presented in Chapter 2.0 of the USAR. The exclusion area, low population zone, and population center dis-tance are provided and described. The USAR also describes the characteristics of reactor design and operation.
100.11    Exclusion areas have been established, as described in Section 2.1. The low population zone has been established in accordance with this requirement.
50.67      The USAR accident analyses, particularly those in Chapters 6.0 and 15.0, demonstrate that offsite doses resulting from postulated accidents would not exceed the criteria in this section of the regulation.
Appendix A Appendix A to 10 CFR Part 100 provides seismic and geologic siting criteria for nuclear power plants.
Site suitability was determined at the construction permit stage.
Rev. 34
 
WOLF CREEK 1.4  IDENTIFICATION OF AGENTS AND CONTRACTORS 1.4.1  APPLICANTS Kansas City Power & Light Company (KCPL), Kansas Electric Power Cooperative, Incorporated (KEPCo), and Kansas Gas and Electric Company (KG&E) are co-owners of WCGS, having 47, 6 and 47 percent participation, respectively. For the purposes of the operating license application KCPL and KG&E were considered co-applicants for Wolf Creek. KG&E was the lead applicant and was initially responsible for the design, construction and operation of WCGS.In Amendment No.
4 to the Operating License, Wolf Creek Nuclear Operating Corporation (WCNOC) was authorized to act as agent for KCPL, KG&E, and KEPCo and has exclusive responsibility and control over the physical construction, operation, and maintenance of the facility. WCNOC will be referred to as the Operating Agent in matters where the corporation acts in the interests of all three licensees.
KCPL is an independent investor-owned utility with headquarters in Kansas City, Missouri that provides electric service in a 5,700 square mile area of western Missouri and eastern Kansas. KCPL serves 331,000 customers and also provides electricity at wholesale to eight communities, three electric cooperatives and two utilities.
KEPCo is a rural electric cooperative association of 25 member cooperatives which provide electric service to the rural areas of Kansas. KEPCo is headquartered in Topeka, Kansas and was incorporated in 1975. KEPCo serves approximately 90,000 meters to provide electricity to nearly 325,000 consumers located throughout Kansas.
KG&E is an independent investor-owned utility that provides electric service in an 8,100 square mile area of south central and southeast Kansas. General offices of the company are in Wichita, Kansas. KG&E serves 212,000 retail customers and also provides, at wholesale, part or all of the electricity sold by 24 municipal electric systems and by 8 rural electric cooperatives. KG&E merged with Kansas Power & Light in 1992 to become Western Resources, and eventually Westar.
The owners have over 97 years of experience in the operation of electric generating plants. The owners do not maintain engineering and construction staffs for power plants but do engage reputable engineering and construction firms for these purposes. As of January 1, 1980, the owners had in operation 11 power stations in which they are full or partial owners with a total system generator nameplate capacity of 4420 MW.
Great Plains and Westar merged to form a new company named Evergy. The merger was finalized on June 4, 2018. As a result of this merger, Evergy owns a combined 94% of WCNOC and WCGS. The remaining 6% ownership interest is held by Kansas Electric Power Cooperative, Inc. (KEPCO).
1.4.2  SNUPPS KCPL, KEPCO, KG&E and Union Electric Company joined together to share costs and manage a project to design, purchase, and license two nuclear power plants of standardized design, the Standardized Nuclear Unit Power Plant System (SNUPPS) units (Wolf Creek and Callaway).
1.4-1                        Rev. 33
 
WOLF CREEK The SNUPPS utilities were signatories of an agreement for standardization for nuclear generating facilities known as the SNUPPS Agreement. The agreement specified objectives of the undertaking, defined the responsibilities of the utility participants, and established a method of sharing the costs.
Participation in the agreement was open to any entity proposing to construct or participate with others in constructing nuclear generating facilities of approximately 1,100 electrical megawatts on a site without active earthquake potential. The SNUPPS Agreement provided for cost sharing of the duplicate portions of the plant, established an organizational structure for management of the project, and defined a mechanism for reaching decisions on joint actions on the basis of one share and one vote per unit.
The basic shared activities were: (1) design of the standardized portion of the plants, known as the power block; (2) procurement of the NSSS; (3) procurement of the turbine generators; (4) procurement of essentially all other equipment and materials for the power block; and (5) design and fabrication of the first fuel loading. Activities which were the responsibility of each individual applicant utility were: (1) design and procurement of equipment and materials for nonstandardized facilities outside of the power block; (2) construction of both standardized and nonstandardized facilities; and (3) procurement of certain power block materials to standard specifications.
The SNUPPS utilities controlled the project through a management committee composed of one officer of each company. The members elected a chairman annually and held regular meetings. This senior executive group had overall responsibility for administration of the project and for resolution of technical, contractual, and schedular problems during evolution of the project.
The SNUPPS utilities entered into individual, basically identical contracts with four contractors to purchase the materials and services for the shared activities: 1) Bechtel Power Corporation to provide architect-engineering services for the power block; 2) Westinghouse to supply two identical NSSSs and, under a separate contract, and to supply the first fuel loading; 3)
General Electric to supply two identical turbine generators and directly related auxiliaries; and 4) Nuclear Projects, Inc., for project management and for furnishing the technical and administrative staff to represent the utility owners and to engage consulting services and contractors, as required. These contracts were administered as one project.
1.4-2                        Rev. 0
 
WOLF CREEK Nuclear Projects, Inc., established in May 1974, furnished services to the SNUPPS utilities for management of the SNUPPS project, as authorized and directed by the management committee. The SNUPPS Executive Director, appointed by the management committee, and the SNUPPS technical and administrative staff were employees of and consultants to Nuclear Projects, Inc.
They had the responsibility to act for the management committee and the utilities in the day-to-day administration of work under the lead architect-engineer contract. The lead architect-engineer, in turn, was delegated responsibility for administration of the turbine generator and NSSS procurement. The lead architect-engineer had responsibility for the power block and authority to procure equipment and materials for the utilities. The Executive Director also had authority to administer the contracts for design and fabrication of the first core.
Various utility committees augmented the SNUPPS staff and provided communication links between SNUPPS activities and each individual utility.
Committees included a technical committee, quality assurance committee, operations committee, legal committee, construction review group, licensing coordination group, and committees for records management, spare parts, finance and accounting, public relations, and numerous ad hoc groups and task forces for special problems.
Outside of the shared activities, each utility managed site unique activities and construction. Each utility retained a site architect-engineer (Sargent &
Lundy for WCGS) to design non-standardized facilities. Construction management at the Wolf Creek site was by Daniel International Corporation. Bechtel staff was located at each construction site to interpret plans and specifications and expedite procurements. A SNUPPS staff member was located at each active site to ensure that construction experience was made available to later plants.
1.4.3  NUCLEAR STEAM SUPPLY SYSTEM MANUFACTURER Westinghouse Electric Corporation (Westinghouse) was responsible for supplying the NSSS and first fuel load for WCGS.
Westinghouse has designed, developed, and manufactured nuclear facilities since the 1950s, beginning with the world's first large central station nuclear power plant (Shippingport), which has produced power since 1957. Completed or presently contracted commercial nuclear capacity totals in excess of 97,000 MW.
Westinghouse pioneered new nuclear design concepts, such as chemical shim control of reactivity and the rod cluster control 1.4-3                          Rev. 0
 
WOLF CREEK concept, throughout the last two decades. Among the company's own related manufacturing facilities are the Columbia Plant, Nuclear Fuel Division, the largest commercial nuclear fuel fabrication facility in the world, and the Tampa Division Plant, the world's most modern heat transfer equipment production facility.
1.4.4  STANDARD PLANT (LEAD) ARCHITECT/ENGINEER The Gaithersburg Power Division of Bechtel Power Corporation (Bechtel) was retained by the SNUPPS utilities to provide architect/engineer services, including procurement, for the standardized portions of the nuclear electric generating facilities.
The Bechtel Corporation, the parent of Bechtel Power Corporation, has been continuously engaged in construction and engineering activities since 1898.
Since the close of World War II, Bechtel has placed strong emphasis on electrical power generation projects. During this period, Bechtel has been responsible for the design of over 204 thermal generating units, representing more than 126,860 MW of new generating capacity. Of this number, a nuclear capacity of more than 65,800 MW has been or is being engineered by the company itself.
The ratings of thermal generating plants designed by Bechtel range up to 1,470 MW per unit and include most types of station designs and arrangements, such as reheat and nonreheat, indoor and outdoor stations, single and multiple units, and wide ranges of steam conditions up to 3,500 psig, 1,050/1,000 F. Also, some of the larger units are fully automated and computer controlled. The majority of contracts for these facilities provided Bechtel with complete responsibility for both engineering and construction, although several contracts have been engineering design assignments only.
For over 25 years, Bechtel has been actively working on nuclear projects involving power plants, as well as such facilities as nuclear accelerators, research laboratories, hot cells, experimental reactors, and nuclear fuel processing plants. Its responsibilities have covered design, construction, site surveys, license applications, feasibility studies, and equipment procurement.
1.4.5  TURBINE-GENERATOR MANUFACTURER The General Electric Company was responsible for the design, fabrication, and delivery of the turbine generators, and provided technical assistance for installation, startup, and operation of this equipment.
1.4-4                        Rev. 1
 
WOLF CREEK General Electric has a long history in the application of turbine generators for nuclear power plants.
1.4.6  SITE ARCHITECT/ENGINEER For the site-related work covered by the application, except for certain environmental studies, Sargent & Lundy Engineers (S&L) was retained as the architect-engineer and design consultant. In general, the responsibilities included the site layout, the location of the power block, the design of yard and construction facilities, and the location and design of the circulating water systems. They were responsible for the design of site-related systems and facilities which are nonseismic Category I and for seismic Category I dams, canals, ponds and earthwork.
Sargent & Lundy is an independent consulting engineering organization founded in Chicago in 1891. The firm has specialized in the design of generation, transmission and distribution systems for steam utilization, electric power, and related facilities. The firm has provided the complete engineering services for more than 600 turbine-generator units with a total installed capacity of 53,000 MW. Of this, some 9,800 MW is nuclear generating capacity.
Table 1.4-1 lists the nuclear plants S&L itself has completed or is currently designing. Table 1.4-2 lists other nuclear plants Sargent & Lundy has had partial responsibility for.
1.4.7  CONSULTANT FIRMS 1.4.7.1  SNUPPS Consultants Principal consultants for the SNUPPS portions (powerblock) of the WCGS and their related responsibilities are:
: a. Quadrex Corporation (formerly Nuclear Services Corporation This consultant assisted the SNUPPS staff to coordinate the owners' preparation of power block operating procedures and review and approval action by the owners of Bechtel-prepared flush, hydrostatic, preoperation, and special test procedures. The compilation of specific data lists, useful for operating procedure preparation, power plant operation, and maintenance, is assigned to this consultant on an as-needed basis. This consultant also performed third-level design reviews of selected systems for compliance with codes and regulations.
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: b. Southwest Research Institute This consultant reviewed portions of the SNUPPS unit design to assure that adequate provision was made for preservice and inservice inspection, including access engineering, and to verify the performance of mechanical equipment. In the latter category, this consultant has performed analog simulation of the reactor charging system and recommended the design of pulsation suppressors chosen for use in the SNUPPS plants.
: c. NUS Corporation The nuclear engineering, plant design, and nuclear power plant licensing skills and experience of this consultant were drawn upon on an as-needed basis to perform a number of activities. Examples included drafting specifications for a loose parts monitor, carrying out an independent review of Bechtel's calculations for shielding the reactor cavity, and reviewing the SNUPPS units' cold shutdown capability.
: d. Nuclear Water & Waste Technology, Inc.
This consultant, a specialist in water chemistry, re-viewed the design and assisted in the selection of fluid systems and equipment, particularly the condensate polisher, liquid radwaste systems, and process control instrumentation.
: e. Pickard, Lowe and Garrick, Inc.
This consultant was utilized early in the SNUPPS project to assist in bid evaluations and selection of the Stan-dard Plant A/E. This consultant remained available on an as-needed basis, and provided occasional assistance in matters related to nuclear design and performance, such as reviewing the performance of nuclear fuel designs.
: f. Professional Loss Control, Inc.
This consultant reviewed the WCGS fire protection system and assisted in making related design decisions.
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WOLF CREEK
: g. Energy Research & Consultants Corporation This consultant reviewed design and operation of pumps and other rotating equipment, including advising WCGS during the bid evaluation for several pumps, and per-forming tests necessary to evaluate the auxiliary feed-water pumps.
: h. Fauske and Associates, LLC This consultant developed calculations of atmospheric dispersion factors for the control room/TSC fresh air intake for use in control room/TSC accident dose calcula-tions.
: i. Energy Incorporated This consultant was engaged to assist the SNUPPS utilities to develop an independent plant transient and analysis capability using the RETRAN computer code.
: j. Essex Corporation This consultant was engaged to perform an independent design evaluation of the SNUPPS control room, emphasizing human factors considerations.
1.4.7.2    WCGS Specific Consultants
: a. Dames & Moore The independent consulting firm of Dames & Moore was retained to perform site investigations relating to demography, geography and land use, meteorology, hydrology, geology and seismology. Having performed such safety-realted and environmental impact related investigations for over 75 nuclear power plant sites, Dames & Moore is an acknowledged leader in the field of site investigations related to nuclear plant construction.
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WOLF CREEK Listed below are some of the nuclear power plants for which Dames & Moore has performed geotechnical and/or environmental investigations:
U.S. NORTHEAST Atlantic                James A. Fitzpatrick    Salem Burlington              Limerick                Seabrook Calvert Cliffs          Newbold Island          Shoreham Douglas Point            Nine-Mile Point          Somerset Forked River            Oyster Creek            Sterling Hope Creek              Peach Bottom            Summit Indian Point            Perryman                Susquehanna Jamesport                Robert E. Ginna          Yankee U.S. MIDWEST Bailly N 1              Dresden                  Midland Braidwood                Duane Arnold            Monticello Byron                    Fermi                    Palisades Carroll County          Fort Calhoun            Point Beach Central Iowa            Greenwood                Prairie Island Clinton                  Haven                    Quad-Cities Cooper                  Kewaunee                Zimmer Davis-Besse              LaSalle                  Zion Donald C. Cook          Marble Hill U.S. SOUTH Brunswick                Joseph M. Farley        Shearon Harris Catawba                  McGuire                  South Dade Cherokee                North Anna              St. Lucie Crystal River            Nuclear One              Surry DeSoto                  Oconee                  Turkey Point Edwin I. Hatch          Perkins                  Virgil C. Summer Isolte                  Robinson U.S. SOUTHWEST Allens Creek            Comanche Peak            South Texas Project U.S. WEST AND NORTHWEST Humboldt Bay            Skagit                  Trojan San Onofre              Sundesert 1.4-8                        Rev. 0
 
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: b. Ecological Analysts, Inc.
The private research and service organization of Ecological Analysts, Inc. was retained to collect, analyze and report environmental data related to the environmental impact of the WCGS. These studies included biological, chemical, and radiological investigations. The organization was founded in 1978 and was formerly part of Industrial BIO-TEST Labora-tories, Inc. (1968-75), NALCO Chemical Company (1975-78) and Hazleton Environmental Sciences Corporation (1978-80). The present full-time staff includes more than 160 scientists and technicians. Listed below are some of the nuclear power plants for which Ecological Analysts, Inc. has performed environmental studies and investigations:
Bailly                            Fort Calhoun Catawba                            Kewaunee Clinton                            LaSalle Cooper                            Quad-Cities Dresden                            Wm. H. Zimmer Duane Arnold                      Zion
: c. Hoad Engineers, Incorporated Hoad Engineers, Incorporated (HEI), a wholly-owned subsidiary of Blount, Incorporated, was retained to provide the design, plans and specifications along with equipment procurement and construction management services for the WCGS security system. The Operating Agent took over responsibility for these activities in late 1982. HEI prepared and provided documents for the Security Plan, Safeguard Contingency Plan and Security Training and Qualifications Plan. Since its inception in 1953, HEI has primarily served the investor-owned utilities in all of the usual engineering and architectural disciplines.
HEI has been engaged in providing the design plans, specifications and equipment procurement services for security systems at two nuclear power plants for the Consumers Power Company in Jackson, Michigan.
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: d. Phoenix Power Services, Incorporated Phoenix Power Services, Incorporated was retained to assist in preparing the early drafts of the Emergency Plan for WCGS. Phoenix has performed extensive work in the area of Emergency Planning.
: e. Professional Loss Control, Inc.
Professional Loss Control, Inc., (PLC), was retained to review the Fire Plan and implementing procedures and advise The Operating Agent of their adequacy. Founded in 1976, PLC provides services in the fields of fire protection, safety and environmental engineering.
Listed below are some of the utilities for which PLC has performed loss control services Carolina Power and Light Company Florida Power Corporation Jersey Central Power and Light Company Maine Yankee Atomic Power Company Niagara Mohawk Power Corporation Portland General Electric Company Power Authority of the State of New York Rochester Gas and Electric Company Tennessee Valley Authority Washington Public Power Supply System Wisconsin Electric Power Company Yankee Atomic Electric Company 1.4.8  CONSTRUCTOR Daniel International Corporation, herein referred to as Daniel, was assigned construction and construction management responsibilities for WCGS.
Daniel's scope of work consisted of receiving design information as prepared by Bechtel, Westinghouse and Sargent & Lundy; receiving manufactured items and materials as procured by Bechtel and Westinghouse; procuring additional bulk materials and consumable items; procuring the services of various subcontractors; planning and scheduling the activities of the construction forces and directly supervising the construction forces to assemble the power plant in accordance with the design.
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WOLF CREEK Daniel Construction Company was awarded the ASME Certificate of Authorization to perform nuclear code construction (N stamp) on September 11, 1973, following an ASME implementation and enforcement audit of Daniel's Quality Assurance Program.
Daniel's experience, past and present, includes construction of nuclear and fossil fueled power plants. The first project of this nature was construction of the nuclear power Carolinas-Virginia Tube Reactor at Parr, South Carolina.
This facility operated several years as a prototype plant. Other nuclear power plants under construction, in operation or on which Daniel is performing maintenance are specified below:
Callaway                Oconee H. B. Robinson          Shearon Harris Joseph M. Farley        Virgil Summer 1.4.9  DIVISION OF RESPONSIBILITIES 1.4.9.1  Utility Company The ultimate responsibility for the proper design, construction, and operation for the entire spectrum of safety of WCGS rests with The Operating Agent.
1.4.9.2  Standard Plant Architect/Engineer Bechtel Power Corporation was responsible for the design, engineering, and procurement of the standard power block, which included the following:
: a. Turbine building
: b. Reactor building
: c. Auxiliary building
: d. Fuel building
: e. Radwaste building
: f. Diesel generator building
: g. Control building 1.4-11                        Rev. 0
 
WOLF CREEK Bechtel was also responsible for the design of the standard plant storage tanks and transformer vaults. However, the individual utilities arranged to procure this equipment.
The NSSS portion of the facility was procurred by individual contract between The Operating Agent and the NSSS supplier. Similarly, the turbine generator is obtained by direct contract between the turbine generator supplier and The Operating Agent. However, Bechtel Power Corporation (acting as agent) retained responsibility for monitoring the design and integrating the system into the power block to ensure that the NSSS and turbine generator components being supplied were consistent with the needs of the facility. Other equipment and material for areas within their scope were procured by Bechtel Power Corporation.
Bechtel Power Corporation was also responsible for the design, engineering and procurement of the portions of the WCGS seismic Category I essential service water systems (ESWS) which lie outside the power block.
The design and engineering of all SSCs associated with the ESW vertical loops and chase were not part of the original SNUPPS standard plant design. Loops and chase were added to both trains of the ESW. Design and specification of these SSCs were obtained by direct contract between WCNOC and several contributing contractors. Interfaces were established and monitored by WCNOC to ensure compatibility in design between power block SSCs and the ESW vertical loops and chase. WCNOC is ultimately responsible for the the design, engineering, and procurement of the SSCs associated with the ESW vertical loops and chase.
1.4.9.3  SNUPPS Staff The SNUPPS Staff functioned as an extension of the management, engineering, and operations organizations of the SNUPPS Utilities. During design and construction phases, the SNUPPS Staff performed day-to-day administration of all of the shared activities, primarily by interfacing with and providing written direction to the Standard Plant Architect/Engineer. This required a close relationship between the SNUPPS Staff and SNUPPS Utilities, which was achieved by frequent communications and regularly scheduled meetings of the various committees and groups.
1 4.9.4  Site Architect/Engineer All systems, equipment and structures outside the power block except for the ESW components and station security-related systems were designed or specified by Sargent & Lundy. The ultimate heat sink was the only seismic Category I structure designed by Sargent & Lundy.
Interfaces were established and monitored by Bechtel Power Corporation to ensure compatibility in design between power block and site-related systems and equipment.
1.4.9.5  Security Consultant The station security-related systems were initially designed and specified by Hoad Engineers, Incorporated, who was retained to design a security system for WCGS to meet the requirements of 10 CFR 73.55.
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WOLF CREEK TABLE 1.4-1 NUCLEAR POWER PLANTS COMPLETED OR CURRENTLY UNDER DESIGN BY SARGENT & LUNDY NOMINAL GROSS*
UNIT                                            RATING (MWe)        POWER OPERATION EBWR                                              5                      1956 Elk River                                        22                      1962 La Crosse                                        60                      1969 SEFOR                                            20 (MWt)                1969 Dresden 2                                        850                    1969 Dresden 3                                        850                    1971 Quad-Cities 1                                    850                    1971 Quad-Cities 2                                    850                    1972 Zion 1                                            1085                    1973 Zion 2                                            1085                    1973 Fort St. Vrain, Unit 1                            330                    1973 La Salle County Station, Unit 1                  1122                    1982 La Salle County Station, Unit 2                  1122                    1983 Byron Station, Unit 1                            1175                    1984 Byron Station, Unit 2                            1175                    1985 Braidwood Station, Unit 1                        1175                    1985 Clinton Power Station, Unit 1                    992                      1985 Braidwood Station, Unit 2                        1175                    1986 Clinton Power Station, Unit 2                    992                      Cancelled Bailly Nuclear 1                                  685                    Cancelled Marble Hill, Unit 1                              1190                    Cancelled Marble Hill, Unit 2                              1190                    Cancelled Wm. H Zimmer, Unit 1                              840                    Cancelled Carroll County Station, Unit 1                    1175                    Future Carroll County Station, Unit 2                                            Future
*Note that this is a gross rating, not a net rating.
Rev. 0
 
WOLF CREEK TABLE 1.4-2 OTHER NUCLEAR POWER PLANTS WITH PARTIAL SARGENT & LUNDY DESIGN RESPONSIBILITY NOMINAL GROSS*                    SARGENT & LUNDY UNIT                            RATING (MWe)                      RESPONSIBILITY Bellefonte, Unit 1                  1235                      Consulting Services on Containment Bellefonte, Unit 2                  1235                      Design (Sructural Analysis of post-tension containment structures).
D. C. Cook, Unit 1                  1083                      Design of water intake structures, crib D. C. Cook, Unit 2                  1118                      house, turbine room foundations and miscellaneous electrical consulting services.
Enrico Fermi, Unit 2                1223                      Design of Residual Heat Removal complex, shielding design calculations and class 1 piping analysis.
Kaiseraugst                          992                      All design inside the containment including the containment itself.
Point Beach, Unit 1                  519                      Consulting services on water intake Point Beach, Unit 2                  519                      structures.
*Note that this is a gross rating, not a net rating.
Rev. 0
 
WOLF CREEK 1.5  REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION One of the design bases for WCGS has been to utilize well-developed and proven design concepts, systems, and equipment, in order to minimize the potential for cost and schedule overruns and to enhance the reliability of operation. As a consequence, there have been few requirements, as delineated by 10 CFR 50.34(a)(8), for research and development programs to confirm the adequacy of the design. Two such programs were identified at the construction permit stage. Those programs have been satisfactorily completed, as described in Sections 1.5.1 and 1.5.2. Other programs were identified at the construction permit stage as not required but as valuable to define margins of conservatism or possible design improvements. Relevant programs in this latter category are described in Section 1.5.3.
1.5.1  17 x 17 FUEL ASSEMBLY A comprehensive test program for the 17 x 17 assembly has been successfully completed by Westinghouse. Reference 1 contains a summary discussion of the program. The following sections present specific references documenting individual portions of the program.
1.5.1.1  Rod Cluster Control Spider Tests Rod cluster control spider tests have been completed. For a further discussion of these tests, refer to Section 4.2.4.3.
1.5.1.2  Grid Tests Verification tests of the structural adequacy of the grid design have been completed. Refer to Section 4.2.3.4 and Reference 2 for a discussion of these tests.
1.5.1.3  Fuel Assembly Structural Tests Fuel assembly structural tests have been completed. Refer to References 2 and 3 for a discussion of these tests.
1.5.1.4  Guide Tube Tests Verification tests of the structural adequacy of the guide tubes have been completed. Refer to References 3 and 4 for a discussion of these tests.
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WOLF CREEK 1.5.1.5  Prototype Assembly Tests Verification tests of the integrated fuel assembly and rod cluster control performance have been completed. Refer to References 3 and 4 for a discussion of these tests.
1.5.1.6  Departure from Nucleate Boiling Tests The test program for experimentally determining the effect of the fuel assembly geometry on the departure from nucleate boiling (DNB) heat flux has been completed. Refer to Reference 5 for a discussion of these tests.
1.5.1.7  Incore Flow Mixing The experimental test program to determine the effects of the fuel assembly geometry on mixing has been completed. Refer to Reference 6 for a discussion of these tests.
1.5.2  FIRE STOPS A test program to determine the adequacy of various fire stop designs has been completed. Penetration seals compatible with the WCGS design were successfully tested, using silicone foam sealant. Details of the tests are provided in Section 9.5.1.
1.5.3  OTHER PROGRAMS 1.5.3.1  Generic Programs of Westinghouse Reference 7 summarizes ongoing safety-related research and development programs that are being carried out for, or by, or in conjunction with the Westinghouse Nuclear Energy System Division and that are applicable to Westinghouse pressurized water reactors. These programs are applicable to WCGS and may lead to changes in safety analyses or modes of operation. Further progress on these programs is not required for safe operation of WCGS.
Experimental test programs to determine the thermal-hydraulic characteristics of 17 x 17 fuel assemblies and to obtain experimental reflooding heat transfer data under simulated LOCA conditions have been completed. Refer to Reference 8 for a discussion of these tests. A single rod burst test program to quantify the maximum assembly flow blockage which is assumed in the LOCA analyses has been completed. Refer to Reference 9 for a discussion of these tests. The results of these two test programs have been used in the ECCS analyses in Section 15.6.5.
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WOLF CREEK Two general types of model boiler tests were conducted by Westinghouse (1) to confirm the thermal-hydraulic analyses used for the Model-F steam generator and (2) to explore the potential for corrosion and other water-chemistry induced effects in the Model-F steam generator. The initial series of each of these tests were completed prior to startup of WCGS. Further information on the steam generator test programs of Westinghouse is given in Section 5.4.2.
1.5.3.2  Generic Programs of Bechtel Wolf Creek through SNUPPS has contributed, with other utilities, to tests of prototypical cable trays under seismically induced loads. A primary objective of the tests has been evaluation of damping coefficients under SSE conditions.
Mechanical bracing of cable trays at WCGS is verified by the results of this test program.
1.5.3.3  Test of a Wolf Creek Steam Generator One of the steam generators in the Wolf Creek plant was equipped with special pressure and temperature instrumentation that enabled thermal-hydraulic performance characteristics to be measured during the early stages of power operation. The objective of these tests was primarily to confirm Westinghouse's design analyses. The test results are proprietary but are available for NRC review.
1.
 
==5.4  REFERENCES==
: 1. Eggleston, F. T., "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries - Spring 1976," WCAP-8768, June, 1976.
: 2. Gesinski, L. and Chiang, D., "Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident," WCAP-8236 (Proprietary) and WCAP-8288 (Non-Proprietary), December, 1973.
: 3. DeMario, E. E., "Hydraulic Flow Test of the 17 x 17 Fuel Assembly," WCAP-8278 (Proprietary) and WCAP-8279 (Non-Proprietary), February, 1974.
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: 4. Cooper, F. W., Jr., "17 x 17 Driveline Component Tests -
Phase IB, II, III, D-Loop Drop and Deflection," WCAP-8446 (Proprietary) and WCAP-8449 (Non-Proprietary), December, 1974.
: 5. Hill, K. W., et al., "Effect of 17 x 17 Fuel Assembly Geometry on DNB," WCAP-8296-P-A (Proprietary) and WCAP-8297-A (Non-Proprietary), February, 1975.
: 6. Cadek, F. F., Motley, F. E. and Dominicis, D. P., "Effect of Axial Spacing on Interchannel Thermal Mixing with the R Mixing Vane Grid," WCAP-7941-P-A (Proprietary) and WCAP-7959-A (Non-Proprietary), January, 1975.
: 7. Eggleston, F. T., "Safety-Related Research and Development for Westinghouse Pressurized Water Reactors, Program Summaries - Winter 1977 - Summer 1978," WCAP-8768, Revision 2, October, 1978.
: 8. "Westinghouse ECCS Evaluation Model - October 1975 Version,"
WCAP-8622 (Proprietary) and WCAP-8623 (Non-Proprietary),
November, 1975.
: 9. Kuchirka, P. J., "17 x 17 Design Fuel Rod Behavior During Simulated Loss-of-Coolant Accident Conditions," WCAP-8289 (Proprietary) and WCAP-8290 (Non-Proprietary), November, 1974.
1.5-4                        Rev. 0
 
WOLF CREEK 1.6  MATERIAL INCORPORATED BY REFERENCE The Wolf Creek USAR incorporates, by reference, various topical reports as part of the application. Bechtel topical reports are listed in Table 1.6-1, and Westinghouse topical reports are listed in Table 1.6-2. The Bechtel and Westinghouse topical reports have been filed separately in support of this and similar applications.
Amendment No. 89 relocated various Technical Specifications to the USAR. The relocated Technical Specifications have subsequently been incorporated into the Technical Requirements Manual (TRM) with the same format and content they possessed in the Technical Specifications. The TRM is a physically separate document from the USAR, but by this specific reference, it is considered part of the USAR and is thereby incorporated by reference. Implementation of, and revision to, the TRM is controlled through administrative procedures.
Controlled drawings were removed from the USAR at Revision 17. The drawings are considered incorporated by reference. Table 1.6-3 identifies the controlled drawings that are incorporated by reference and also provides a cross-reference of the controlled drawings to the respective USAR figure number. The contents of the drawings are controlled by WCGS procedures.
Appendix 9.5B, Fire Hazard Analyses, was removed from the USAR at Revision 19, and is considered incorporated by reference. Table 1.6-4 identifies the section and provides a cross-reference of the controlled document, E-1F9905, Fire Hazard Analysis, which supersedes the information originally provided in Appendix 9.5B of the USAR. The contents of the Fire Hazard Analysis is controlled by WCGS procedures.
Chapter 17.2 Quality Assurance, was removed from the USAR at Revision 21.
Chapter 17.2 is considered incorporated by reference. The Quality Program Manual supercedes the information originally provided in Chapter 17.2 of the USAR. The contents of the Quality Program Manual are controlled by WCGS procedures.
Table 3.11(B)-1, Plant Environmental Normal Conditions; Table 3.11(B)-2, Environmental Qualification Parameters for SNUPPS NUREG-0588 (LOCA, MSLB and HELB); Table 3.11(B)-3, Identification of Safety-Related Equipment and Components: Equipment Qualification; Table 3.11(B)-4, Containment Worst Case Radiation Levels (MRADs); Table3.11(B)-5, Containment Spray Requirements; Table 3.11(B)-8, Exemptions from NUREG-0588 Qualification; Table 3.11(B)-10, Equipment Added for NUREG-0737; Figures 3.11(B)-1 through 3.11(B)-49, were removed from the USAR at Revision 28. The listed Tables and Figures are considered incorporated by reference. EQSD-I, EQ Summary Document Section I Program Description, and EQSD-II, EQ Master List Section II, supercedes the information provided by listed Tables and Figures. The contents of EQSD-I and EQSD-II are controlled by WCGS procedures.
1.6-1                        Rev. 28
 
WOLF CREEK TABLE 1.6-1 BECHTEL TOPICAL REPORTS INCORPORATED BY REFERENCE Bechtel                                                      USAR      Report Topical                                        Revision      Section    Submitted    Review(1)
Report No. Title                                Number        Reference  to the NRC    Status BC-TOP-1  Containment Building                  Rev. 1      3.7(B)-3      1/73          A Liner Plate Design                                  3.8 Report BC-TOP-3-A Tornado and Extreme                  Rev. 3      3.3          8/74          A Wind Design Criteria                              3.8 for Nuclear Power Plants BC-TOP-4-A Seismic Analyses of                  Rev. 3      3.7(B).2      11/74        A Structures and                                    3.7(B).3 Equipment for Nuclear                              3.8 Power Plants BC-TOP-5-A Prestressed Concrete                  Rev. 3      3.8          2/75          A Nuclear Reactor                                    3A Containment Structures BC-TOP-7  Full Scale Buttress                  Rev. 0      3.8          9/72          A Test for Prestressed                              3A Nuclear Containment Structures BC-TOP-8  Tendon End Anchor                    Rev. 0      3.8          9/72          A Reinforcement Test                    3A BC-TOP-9-A Design of Structures                  Rev. 2      3.8          9/74          A for Missile Impact                                  3.5.3.2 Rev. 0
 
WOLF CREEK TABLE 1.6-1 (Sheet 2)
BECHTEL TOPICAL REPORTS INCORPORATED BY REFERENCE Bechtel                                                              USAR      Report Topical                                                Revision      Section    Submitted    Review(1)
Report No.        Title                                Number        Reference  to the NRC    Status BN-TOP-1          Test Criteria for                    Rev. 1        3.8          11/72        A Integrated Leak Rate                              6.2 Testing of Primary Containment Structures for Nuclear Power Plants BN-TOP-2          Design for Pipe Break                Rev. 2        3.6          5/74          A Effects                              3.8 BN-TOP-3          Performance and Siz-                Rev. 3                      8/75          P ing of Dry Pressure                                6.2.1 Containments BN-TOP-4          Subcompartment                      Rev. 1        6.2.1        10/77        A Pressure and                                      3.6 Temperature Transient Analysis BP-TOP-1          Seismic Analysis of                  Rev. 3        3.7.(B).2    1/76          A Piping Systems                                    3.7.(B).3 3.9.(B).7 (1) See Notes on Table 1.6-2 Rev. 0
 
WOLF CREEK TABLE 1.6-2 WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                      USAR        Report Topical                                            Revision      Section      Submitted      Review(1)
Report No. Title                                Number        Reference    to the NRC    Status WCAP-2048      "The Doppler Effect for a            Rev. 0        4.3            7/62          0 Non-Uniform Temperature Distribution in Reactor Fuel Elements" WCAP-2850-L(P) "Single Phase Local Boiling          Rev. 0        4.4            5/66          0 WCAP-7916      And Bulk Boiling Pressure Drop Correlations" WCAP-2923      "In-Pile Measurement of              Rev. 0        4.4            3/66          0 UO2 Thermal Conductivity" WCAP-3269-8    "Hydraulic Tests of the              Rev. 0        4.4            6/64          0 San Onofre Reactor Model" WCAP-3269-26  "LEOPARD - A Spectrum                Rev. 0        4.3, 15.0,      9/63          0 Dependent Non-Spatial                              15.4 Depletion Code for the IBM - 7094" WCAP-3385-56  "Saxton Core II Fuel                Rev. 0        4.3, 4.4        7/70          0 Performance Evaluation,"
WCAP-3385-56, Part II "Evaluation of Mass Spectrometric and Radiochemical Materials Analyses of Irradiated Saxton Plutonium Fuel" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 2)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-3680-20 "Xenon-Induced Spatial              Rev. 0        4.3            3/68          0 Instabilities in Large Pressurized Water Reactors" (EURAEC-1974)
WCAP-3680-21 "Control Procedures for            Rev. 0        4.3            2/69          0 Xenon-Induced X-Y Instabilities in Large Pressurized Water Reactors" (EURAEC-2111)
WCAP-3680-22 "Xenon-Induced Spatial              Rev. 0        4.3            9/69          0 Instabilities in Three-Dimensions" (EURAEC-2116)
WCAP-3696-8  "Pressurized Water                  Rev. 0        4.3            10/68        0 Reactor pH-Reactivity Effect Final Report" (EURAEC-2074)
WCAP-3726-1  "Pu02 -U02 Fueled Critical          Rev. 0        4.3            7/67          0 Experiments" WCAP-6065    "Melting Point of                  Rev. 0        4.4            2/65          0 Irradiated U02" WCAP-6069    "Burnup Physics of                  Rev. 0        4.4            6/65          0 Heterogeneous Reactor Lattices" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 3)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                      USAR        Report Topical                                            Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-6073      "LASER - A Depletion                Rev. 0        4.3            4/66          0 Program for Lattice Calculations Based on MUFT and THERMOS" WCAP 6086      "Supplementary Report              Rev. 0        4.3            8/69          0 on Evaluation of Mass Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel, Including Isotopes of Elements Thorium Through Curium" WCAP-7015      "Subchannel Thermal                Rev. 1        4.4            2/14/69        0 Analysis of Rod Bundle Cores WCAP-7048      "The PANDA Code"                    Rev. 0        4.3            1/9/75        A P-A(P)
WCAP-7757-A WCAP-7198-L(P) "Evaluation of Protective          Rev. 0        6.1            4/23/69        0 WCAP-7825      Coatings for use in                                              12/16/71 Reactor Containment" WCAP-7213-    "The TURTLE 24.0 Diffusion          Rev. 0        4.3, 15.0,    1/9/75        A P-A(P)      Depletion Code"                                    15.4 WCAP-7758-A WCAP-7240(P)  "An Experimental Investi-          Rev. 0                        7/7/72        B gation of the Effect of Open Channel Flow on Thermal-Hydrodynamic Flow Stability" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 4)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                      USAR        Report Topical                                            Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-7308-L(P) "Evaluation of Nuclear              Rev. 0        4.3          7/9/70        U WCAP-7810      Hot Channel Factor                                                12/16/71 Uncertainties" WCAP-7359-L(P) "Application of the THINC          Rev. 0        4.4          9/8/69        0 WCAP-7838      Program to PWR Design"              1/17/72 WCAP-7397-L(P) "Seismic Testing of                Rev. 0        3.10(N)      2/6/70        U WCAP-7817      Electrical and Control                                          12/16/71 Equipment" WCAP-7397-L(P) "Seismic Testing of                Supple-        3.10(N)      1/27/71      U WCAP-7817      Electrical and Control              ment 1                      12/16/71 Equipment (WCID Process Control Equipment)"
WCAP-7477-L(P) "Sensitized Stainless              Rev. 0        5.2          3/26/70      A WCAP-7735      Steel in Westinghouse                                            8/12/71 PWR Nuclear Steam Supply Systems" WCAP-7488-L(P) "Solid State Logic                  Rev. 0        7.2, 7.3      3/24/71      B WCAP-7672      Protection System                                                5/27/71 Description" WCAP-7558      "Seismic Vibration Testing          Rev. 0        3.10(N)      9/25/72      U with Sine Beats" WCAP-7588      "An Evaluation of the              Rev. 1A        15.4          1/7/75        A Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 5)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                      USAR        Report Topical                                            Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-7595-A    (See WCAP-7941-P-A(P))
WCAP-7667-P-  "Interchannel Thermal Mixing        Rev. 0        4.4            1/27/75        A A(P)          with Mixing Vane Grids" WCAP-7755-A WCAP-7695-    "DNB Tests Results for              Rev. 0        4.4            1/21/75        A P-A(P)        New Mixing Vane Grids (R)"
WCAP-7958-A WCAP-7695,    "DNB Test Results for R            Rev. 0        4.4            1/21/75        A Addendum      Grid Thimble Cold Wall 1-P-A(P)      Cells" WCAP-7985, Addendum 1-A WCAP-7672      (See WCAP 7488-L(P))
WCAP-7705      "Testing of Engineered              Rev. 2                        5/5/76        B Safety Features Actuation System" WCAP-7706-L(P) "An Evaluation of Solid            Rev. 0        4.6, 7.1        9/2/71        U WCAP-7706      State Logic Reactor                                7.2, 7.3 Protection in Anticipated Transients" WCAP-7709-L(P) "Electrical Hydrogen                Rev. 0                        7/14/71        A WCAP-7820      Recombiner for Water                              6.2.5          12/16/71 Reactor Containments" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 6)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                      USAR        Report Topical                                            Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-7709-L(P) "Electric Hydrogen Recombiner      Supple-                      5/23/72        A WCAP-7820      for PWR Containments - Final        ment 1        6.2.5        5/31/72 Development Report" WCAP-7709-L(P) "Electric Hydrogen Recombiner      Supple-                      9/24/73        A WCAP-7820      for PWR Containments - Equip-      ment 2        6.2.5        11/2/73 ment Qualification Report" WCAP-7709-L(P) "Electric Hydrogen Recombiner      Supple-                      1/23/74        A WCAP-7820      for PWR Containments - Long-        ment 3        6.2.5        3/22/74 Term Tests" WCAP-7709-L(P  "Electric Hydrogen Recombiner      Supple-                      4/21/74        A WCAP-7820      for PWR Containments"              ment 4        6.2.5 WCAP-7709-L(P) "Electric Hydrogen Recombiner      Supple-                      1/7/76          A WCAP-7820      Special Tests"                      ment 5        6.2.5 WCAP-7709-L(P) "Electric Hydrogen Recombiner      Supple-                      11/5/76        A WCAP-7820      IEEE 323-1974 Qualification"        ment 6        6.2.5 WCAP-7709-L(P) "Electric Hydrogen Recombiner      Supple-                      9/21/77        A WCAP-7820      LWR Containments - Supple-          ment 7        6.2.5 mental Test Number 2" WCAP-7735      (See WCAP-7477-L(P))
Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 7)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                      USAR Report Topical                                    Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-7750      "A Comprehensive Space Time        Rev. 0        3.6.3        8/31/71        0 Dependent Analysis of Loss-of-Coolant (SATAN-IV Digital (Code)"
WCAP-7755-A    (See WCAP-7667-P-A(P))
WCAP-7757-A    (See WCAP-7048-P-A(P))
WCAP-7758-A    (See WCAP-7213-P-A(P))
WCAP-7769      "Overpressure Protection for        Rev. 1        5.2, 15.2      7/5/72        U Westinghouse Pressurized Water Reactors" WCAP-7798-L(P) "Behavior of Austenitic            Rev. 0        6.1            12/6/71        0 WCAP-7803      Stainless Steel in Post                                          1/4/72 Hypothetical Loss-of-Coolant Environment" WCAP-7800      "Nuclear Fuel Division              Rev. 4A        3A            4/28/75        A Quality Assurance Program                          4.2 Plan" WCAP-7803      (See WCAP-7798-L(P))
WCAP-7806      "Nuclear Design of Westing-        Rev. 0        4.3          21/16/71      B house Pressurized Water Reactors with Burnable Poison Rods" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 8)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-7810    (See WCAP-7308-L(P))
WCAP-7811    "Power Distribution Control        Rev. 0        4.3          12/16/71          0 of Westinghouse Pressurized Water Reactors" WCAP-7817    (See WCAP-7397-L(P))                Rev. 0 WCAP-7817    (See WCAP-7477-L(P))                Supple-ment 1 WCAP-7817    "Seismic Testing of Electrical      Supple-        3.10(N)      1/17/72            U and Control Equipment (Low          ment 2 Seismic Plants)"
WCAP-7817    "Seismic Testing of Electrical      Supple-        3.10(N)      1/17/72            U and Control Equipment (Westing      ment 3 house Solid State Protection System) (Low Seismic Plants)"
WCAP-7817    "Seismic Testing of Electrical      Supple-        3.10(N)      12/14/72          U and Control Equipment (WCID        ment 4 NUCANA 7300 Series) (Low Seismic Plants)"
WCAP-7817    "Seismic Testing of Electrical      Supple-        3.10(N)      12/14/72          U and Control Equipment (Instru-      ment 5 ment Bus Distribution Panel)
(Low Seismic Plants)"
WCAP-7817    "Seismic Testing of Electrical      Supple-        3.10(N)      8/00/74            U and control Equipment (Type DB      ment 6 Reactor Trip Switchgear)
Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 9)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-7820    (See WCAP 7709-L[P])
WCAP-7825    (See WCAP 7198-L[P])
WCAP-7832    "Evaluation of Steam Generator      Rev. 0        5.4          21/26/73          A Tube, Tube Sheet and Divider Plate Under Combined LOCA Plus SSE Conditions" WCAP-7836    "Inlet Orificing of Open PWR        Rev. 0        4.4          1/17/72          B Cores" WCAP-7838    (See WCAP 7359-L[P])
WCAP-7870    "Neutron Shielding Pads"            Rev. 0        3.9 (N)      7/17/72          A WCAp-7907    "LOFTRAN Code Description"          Rev. 0        5.2,        10/11/72          U 15.0, 15.1 15.2, 15.3, 15.4, 15.5 15.6 WCAP-7908    "FACTRAN - A FORTRAN-IV code        Rev. 0        15.0,        9/20/72          U for Thermal Transients in a                        15.3, U02 Fuel Rod"                                      15.4 WCAP-7909    "MARVEL - A Digital Computer        Rev. 0                    10/11/72          U Code for Transient Analysis of a Multiploop PWR System" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 10)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-7912-  "Power Peaking Factors"            Rev. 0        4.3, 4.4      1/16/75        A P-A (P)
WCAP-7912-A WCAP-7913    "Process Instrumentation            Rev. 0        7.2, 7.3      3/9/73          B for Westinghouse Nuclear Steam Supply System (4-Loop Plant Using WCID-7300 Series Process Instrumentation)"
WCAP-7916    (See WCAP 2850-L[P])
WCAP-7921-AR "Damping Values of Nuclear          Rev. 0        3.7(N), 3A    7/11/74          A Power Plant Components" WCAP-7924-A  "Basis for Heatup and              Rev. 0                      4/28/75          A Cooldown Limit Curves" WCAP-7941-  "Effect of Axial Spacing on        Rev. 0        1.5, 4.4      1/27/75        A P-A (P)    Interchannel Thermal Mixing WCAP-7595-A  with the R Mixing Van Grid" WCAP-7956    "THINC-IV - An Improved Program    Rev. 0        4.4          10/22/73        A for Thermal-Hydraulic Analysis of Rod Bundle Cores" WCAP-7958    (See WCAP-7695-P-A(P))
WCAP-7964    "Axial Zenon Transient Tests        Rev. 0        4.3            6/15/71        O at the Rochester Gas and Electric Reactor" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 11)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-7979-    "TWINKLE - A Multi-Dimensional      Rev. 0        15.0,        1/7/75        A P-A (P)      Neutron Kinetics Computer                          15.4 WCAP-8028-A  Code" WCAP-7985    (See WCAP-7695, Addendum 1-P-A[P]) Addendum 1A WCAP-8028-A  (See WCAP-7979-P-A(P))
WCAP-8054 (P) "Application of the THINC-IV        Rev. 0        4.4          12/7/73      A WCAP-8195    Program to PWR Design"                                          1/11/74 WCAP-8082-    "Pipe Breaks for the LOCA          Rev. 0        3.6.3        1/16/75      A P-A (P)      Analysis of the Westinghouse WCAP-8172-A  Primary Coolant Loop" WCAP-8099    "A Summary Analysis of the          Rev. 0                      4/20/73      B April 30 Incident at the San Onofre Nuclear Generation Station, Unit 1" WCAP-8163    "Reactor Coolant Pump              Rev. 0        3A, 5.4      9/20/73      U Integrity in LOCA" WCAP-8172-A  (See WCAP 8082-P-A[P])
Rev. 31
 
WOLF CREEK TABLE 1.6-2 (Sheet 12)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC      Status WCAP-8183    "Operational Experience with        Rev. 7        4.2            4/20/78        B Westinghouse Cores (up to December 31, 1977)"
WCAP-8195    (See WCAP-8054(P))
WCAP-8200 (P) "WFLASH - A FORTRAN-IV              Rev. 2        15.6          7/3/74          AE WCAP-8261    Computer Program for Simula-        Rev. 1 tion of Transients in a Multi-Loop PWR" WCAP-8218    "Improved Fuel Performance          Rev. 0        3A, 4.2, 4.4  June 1985      A P-A (P)      Models for Westinghouse Fuel WCAP-8219-A  Rod Design and Safety Evaluations."
WCAP-8236 (P) "Safety Analysis of the 17 x 17    Rev. 0        1.5, 4.2      2/28/74        U WCAP-8288    Fuel Assembly for Combined                                      3/1/74 Seismic and Loss-of-Coolant Accident" WCAP-8236 (P) "Safety Analysis of the 8-Grid      Addendum      3.7(N)        4/15/74          A WCAP-8288    17 x 17 Fuel Assembly for          1 Combined Seismic and Loss-of-Coolant Accident" Rev. 14
 
WOLF CREEK TABLE 1.6-2 (Sheet 13)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR          Report Topical                                          Revision      Section      Submitted    Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-8252    "Documentation of Selected          Rev. 1        3.6.3, 3.9(N)  7/19/77      U Westinghouse Structural Analysis Computer Codes" WCAP-8253    "Source Term Data for Westing-      Amendment                    2/13/76      B house Pressurized Water            1 Reactors" WCAP-8255    "Nuclear Instrumentation            Rev. 0        7.2, 7.7      4/9/74        B System" WCAP-8261    (See WCAP-8200(P))
WCAP-8278 (P) "Hydraulic Flow Test of the        Rev. 0        1.5, 4.2, 4.4  2/28/74      U WCAP-8279    17 x 17 Fuel Assembly"                                            3/1/74 WCAP-8288    (See WCAP-8236(P))
WCAP-8289 (P) "17 x 17 Design Fuel Rod            Rev. 0        1.5            11/18/74      A WCAP-8290    Behavior During Simulated Loss-of-Coolant Accident Conditions" WCAP-8296-    "Effect of 17 x 17 Fuel            Rev. 0        1.5            2/6/75        A P-A (P)      Assembly Geometry on DNB" WCAP-8297-A WCAP-8298-    "The Effect of 17 x 17              Rev. 0        4.4            1/28/75        0A P-A (P)      Fuel Assembly Geometry on WCAP-8299-A  Interchannel Thermal Mixing" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 14)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-8301 (P) "LOCTA-IV Program: Loss-of          Rev. 0        15.0, 15.6    7/12/74        AE WCAP-8305    Coolant Transient Analysis" WCAP-8303-    "Prediction of the Flow-Induced    Rev. 0        3.9(N)        7/18/75        A P-A (P)      Vibration of Reactor Internals WCAP-8317-A  by Scale Model Test" WCAP-8305    (See WCAP-8301(P))
WCAP-8306    (See WCAP-8302(P))
WCAP-8317-A  (See WCAP-8303-P-A(P))
WCAP-8324-A  "Control of Delta Ferrite in        Rev. 0        5.2          6/23/75        A Austenitic Stainless Steel Weldments" WCAP-8327 (P) "Containment Pressure Analysis      Rev. 0        15.6          7/3/74          AE WCAP-8326    Code (COCO)"
WCAP-8330    "Westinghouse Anticipated          Rev. 0        4.3, 4.6      9/25/74        U Transients Without Trip Analysis" Rev. 31
 
WOLF CREEK TABLE 1.6-2 (Sheet 15)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-8339    "Westinghouse Emergency Core        Rev. 0        15.6            7/3/74      AE Cooling System Evaluation Model - Summary" WCAP-8340 (P) "Westinghouse Emergency Core        Rev. 0        15.6            8/1/74      AE WCAP-8356    Cooling System - Plant Sensitivity Studies" WCAP-8341 (P) "Westinghouse Emergency Core        Rev. 0        15.6            7/3/74      AE WCAP-8342    Cooling System Evaluation Model-Sensitivity Studies" WCAP-8359    "Effects of Fuel Densification      Rev. 0        4.3            8/2/74      A Power Spikes on Clad Thermal Transients" WCAP-8370    "Quality Assurance Plan            Rev. 7A        3A              2/5/75        A Westinghouse Nuclear Energy Systems Divisions" WCAP-8370    "Westinghouse Water Reactor        Rev. 8A        3A              11/14/77      A Divisions Quality Assurance Plan" WCAP-8370    "Westinghouse Water Reactor        Rev. 9A        3A                            U Divisions Quality Assurance Plan WCAP-8373    "Qualification of Westinghouse      Rev. 0        3.10(N)        8/23/74      U Seismic Testing Procedure for Electrical Equipment Tested Prior to May 1974" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 16)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-8377 (P) "Revised Clad Flattening            Rev. 0        4.2          8/7/74          A WCAP-8381    Model"                                                          8/6/74 WCAP-8385 (P) "Power Distribution Control and    Rev. 0        4.3, 4.4      10/9/74        A WCAP-8403    Load Following Procedures" WCAP-8424    "An Evaluation of Loss of          Rev. 1        15.3          5/30/75        U Flow Accidents Caused by Power System Frequency Transients in Westinghouse PWRs" WCAP-8446 (P) "17 x 17 Driveline Com-            Rev. 0        1.5, 3.9(N)  12/31/74      A WCAP-8449    ponents Tests Phase IB, II, III D-Loop Drop and Deflection" WCAP-8453-A  "Analysis of Data from the          Rev. 0        4.4          5/10/76        A Zion (Unit 1), THINC Veri-fication Test" WCAP-8471 (P) "Westinghouse ECCS Evaluation      Rev. 0        15.6          2/10/75        AE WCAP-8472    Model - Supplementary Informa-                                  2/11/75 tion" WCAP-8485    "Safety-Related Research and        Rev. 0                      4/2/75          B Development for Westinghouse Pressurized Water Reactors, Program Summaries - Fall 1974" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 17)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                      USAR        Report Topical                                            Revision      Section      Submitted    Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-8498      "Incore Power Distribution          Rev. 0        4.3          7/22/75      U Determination in Westinghouse Pressurized Water Reactors, Program Summaries - Fall 1974" WCAP-8510      Method for Fracture Mechanics      Rev. 0        5.3          12/00/75      U Analysis of Nuclear Reactor Vessels Under Severe Thermal Transients WCAP-8516-P(P) UHI Plant Internals Vibration      Rev. 0        3.9(N)        4/11/75      A WCAP-8517      Measurement Program and Pre and Post Hot Functional Examinations" WCAP-8536(P)  "Critical Heat Flux Testing of      Rev. 0        4.4          5/30/75      A WCAP-8537      17 x 17 Fuel Assembly Geometry with 22-Inch Grid Spacing" WCAP-8565-    "Westinghouse ECCS-Four Loop        Rev. 0        15.6          7/17/75      A P-A (P)      Plant (17 x 17) Sensitivity WCAP-8566-A    Studies" WCAP-8577      "The Application of Preheat        Rev. 0        6.1          2/3/76        A Temperatures after Welding Pressure Vessel Steels" WCAP-8584 (P)  "Failure Mode and Effects          Rev. 1        4.6, 7.3      3/20/80      U WCAP-8760      Analysis (FMEA) of the Engineered Safety Features Actuation System" WCAP-8587      "Methodology for Qualifying        Rev. 6A        3.10(N),      11/00/83      U Westinghouse WRD Supplied                          3.11(N),
NSSS Safety-Related Electrical                    3A Equipment" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 18)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                        USAR        Report Topical                                              Revision      Section      Submitted      Review(1)
Report No.      Title                              Number        Reference    to the NRC    Status WCAP-8587        "Equipment Qualification            Rev. 1        3.11(N)        4/17/78        U Data Packages"                      Supple-        3.10(N) ment 1 WCAP-8622 (P)    "Westinghouse ECCS Evaluation      Rev. 0        1.5, 15.6    11/20/75        AE WCAP-8623        Model - October 1975 Version" WCAP-8624(P)    "General Method of Developing      Rev. 0        3.10(N)        0/00/00        U Multi-Frequency Biaxial Test Inputs for Bistables" WCAP-8682 (P)    "Experimental Verification of      Rev. 0        4.3            3/18/76        B WCAP-8683        Wet Fuel Storage Criticality Analyses" WCAP-8691 (P)    "Fuel Rod Bow Evaluation"          Rev. 0        4.2, 4.4      11/83            U WCAP-8692 WCAP-8693        "Delta Ferrite in Production        Rev. 0        5.2            3/16/76        B Austenitic Stainless Steel Weldments" WCAP-8708-P-A    "MULTIFLEX - A FORTRAN-IV          Rev. 0        3.6.3        9/16/77          A (P), Vol. I & II Computer Program for Analyzing                    3.9(N)
WCAP-8709-A,    Thermal-Hydraulic-Structure Volumes I & II  System Dynamics" WCAP-8720 (P)    "Improved Analytical Models        Rev. 0        4.2          11/2/76          A WCAP-8785        Used in Westinghouse Fuel Rod Design Computations" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 19)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-8768    "Safety-Related Research and        Rev. 2        1.5, 4.2,    9/28/78          B Development for Westinghouse                      4.3, 5.4 Pressurized Water Reactors, Program Summaries - Winter 1977 - Summer 1978" WCAP-8766 (P) "Verification of Neutron Pad        Rev. 0        3.9(N)        5/21/76          A WCAP-8780    and 17 x 17 Guide Tube Designs by Preoperational Tests on the Trojan 1 Power Plant" WCAP-8865-A  "Westinghouse ECCS - Four Loop      Rev. 0                      5/6/77          A Plant (17 x 17) Sensitivity Studies with Upper Head Fluid Temperature at THOT" WCAP-8872    "Design, Inspection, Operation      Rev. 0        12.1          4/27/77          B and Maintenance Aspects of the Westinghouse NSSS to Maintain Occupational Radiation Exposures as Low as Reasonably Achievable" WCAP-8892-A  "Westinghouse 7300 Series          Rev. 0        7.1            6/15/77        A Process Control System Noise Tests" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 20)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                    USAR        Report Topical                                          Revision      Section      Submitted      Review(1)
Report No. Title                              Number        Reference    to the NRC    Status WCAP-8929    "Benchmark Problem Solutions        Rev. 0        3.9(N)      5/26/77          U Employed for Verification of the WECAN Computer Program" WCAP-8963 (P) "Safety Analysis for the            Rev. 0        4.2          3/31/71          A WCAP-8964    Revised Fuel Rod Internal                                      8/11/77 Pressure Design Basis" WCAP-8970(P)  "Westinghouse emergency Core        Rev. 0        15.6        4/77              U WCAP-8971    Cooling System Small Break -
October 1975 Model" WCAP-8976    "Failure Mode and Effects          Rev. 0        4.6,        10/26/77          U Analysis (FMEA) of the                            7.7 Solid State Full Length Rod Control System" WCAP-9166    "Westinghouse Emergency Core        Rev. 0        15.6        2/00/78          U Cooling System Evaluation Model for Analyzing Large LOCA's During Operation With One Loop Out of Service for Plants Without Loop Isolation Values" WCAP-9168 (P) "Westinghouse Emergency Core        Rev. 0        15.6        9/27/77          U WCAP-9169    Cooling System Evaluation Model -
Modified October 1975 Version" WCAP-9179 (P) "Properties of Fuel and Core        Rev. 1        4.2          8/2/78            U Component Materials" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 21)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                        USAR        Report Topical                                              Revision      Section      Submitted      Review(1)
Report No.      Title                              Number        Reference    to the NRC    Status WCAP-9207 (P)    "Evaluation of Mispositioned        Rev. 0          6.3          3/21/78        U WCAP-8966        ECCS Valves" WCAP-9220-P-A(P) Westinghouse ECCS Evaluation        Rev. 0          15.6        2/00/78        U WCAP-9221-P-A    Model, February 1978 Version WCAP-9224, Ap-  "Hafnium"                          Rev. 0          10/00/80                    U pendix A WCAP-9226(P)    Reactor Core Response to                            15.1        7/00/78        U WCAP-9227        Excessive Secondary Steam Releases WCAP-9230 (P)    "Report on the Consequences        Rev. 0          15.2        1/27/78        U WCAP-9231        of a Postulated Main Feedline Rupture" WCAP-9279        "Combination of Safe Shutdown      Rev. 0          3.9(N)      3/21/78        U Earthquake and Loss-of-Coolant Accident Responses for Faulted Condition Evaluation of Nuclear Power Plants" WCAP-9283        "Integrity of the Primary          Rev. 0          3.9(N)      3/21/78        U Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events" Rev. 0
 
WOLF CREEK TABLE 1.6-2 (Sheet 22)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE Westinghouse                                                      USAR        Report Topical                                              Revision      Section      Submitted      Review(1)
Report No.      Title                                Number        Reference    to the NRC      Status WCAP-9292      "Dynamic Fracture Toughness          Rev. 0        5.2            3/17/78          U of ASME SA508 Class 2a and ASME SA533 Grade A Class 2 Base and Heat Affected Zone Material and Applicable Weld Metals" WCAP-9346      "Electric Hydrogen Recombiner        Rev. 0        6.2.5          7/00/78          U Qualification Testing for Model B WCAP-9714-PA(P) Methodology for the Seismic                        3.10.(N)      00/00/00        A WCAP-9750-A    Qualification of Westinghouse WRD Supplied Equip.
WCAP-9944(P)    "Verification of Upper Head In-      Rev. 0        3.9(N)        7/00/81          U WCAP-9945      jection Reactor Vessel Internals by Preoperational Tests on Sequoyah 1 Power Plant WCAP-10297-P-A  Dropped Rod Methodology for          Rev. 0        15.4          6/00/83          A Negative Flux Rate Trip Plants WCAP-10043      "Steam Generator Tube Plugging      Rev. 0        5.4.2.5        12/3/82          U Analysis for the Westinghouse Standardized Nuclear Power Plant (P) System" WCAP-10858P-A  "AMSAC Generic Design Package"      Rev. 1        7.7.1.11      07/25/85        A
& Addendum 1                                                                      02/26/87 WCAP-13589-A    Assessment of Clad Flattening        Rev. 0      4.3          01/18/93          A and Densificaiton Power Spike Factor Elimination in Westinghouse Nuclear Fuel WCAP-16009-P-A  Realistic Large-Break LOCA          Rev. 0        15.0, 15.6    06/02/03          A Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)
Rev. 31
 
WOLF CREEK TABLE 1.6-2 (Sheet 23)
WESTINGHOUSE TOPICAL REPORTS INCORPORATED BY REFERENCE (P) - Proprietary (1)  A legend to the review status code letters follows:
A - NRC review complete; NRC acceptance letter issued.
AE- NRC accepted as part of the Westinghouse emergency core cooling system (ECCS) evaluation model only; does not constitute acceptance for any purpose other than for ECCS analyses.
B - Submitted to NRC as background information; not undergoing formal NRC review.
O - On file with NRC; older generations report with current validity; not actively under formal NRC review.
U - Actively under formal NRC review.
P - Pending approval by the NRC.
Rev. 0
 
WOLF CREEK Table 1.6-3 USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                          Title                        Drawing #
1.1-1    1      Symbols and Legend for System Flow and Piping      M-120101 and Instrumentation Diagrams 1.1-1    2      Symbols and Legend for System Flow and Piping      M-120102 and Instrumentation Diagrams 1.1-1  3      Symbols and Legend for System Flow and Piping      M-020103 and Instrumentation Diagrams 1.1-1    4      Symbols and Legend for System Flow and Piping      M-020104 and Instrumentation Diagrams 1.2-1    0      Peninsular Plant Arrangement Standard Power        M-1G001 Systems & Structure Interface 1.2-2    0      Equipment Location Radwaste Building Plan El.      M-1G010 1976'-0" 1.2-3    0      Equipment Location Radwaste Building Plan El.      M-1G011 2000'-0" 1.2-4    0      Equipment Location Radwaste Building Plan El.      M-0G012 2022'-0" 1.2-5    0      Equipment Location Radwaste Building El. 2031'-    M-1G013 6" & Roof Plan 1.2-6    0      Equipment Location Radwaste Building Sections A    M-1G014
                &B 1.2-7    0      Equipment Location Radwaste Building Sections C    M-1G015
                &E 1.2-8    0      Equipment Location Radwaste Building Sections D    M-1G016
                &F 1.2-9    0      Equipment Location Reactor and Auxiliary Bldgs    M-1G020 Plan - Basement El. 1974'-0" 1.2-10    0      Equipment Location Auxiliary Building Partial Plan M-1G021 El. 1988'-0" & El. 2013'-6" 1.2-11    0      Equipment Location Reactor and Auxiliary Building  M-1G022 Plan Ground Floor Elevation 2000'-0" 1.2-12    0      Equipment Location Reactor and Auxiliary Building  M-1G023 Plan El. 2026'-0" 1.2-13    0      Equipment Location Reactor and Auxiliary          M-1G024 Buildings Plan Operating Floor El. 2047'-6" 1.2-14    0      Equipment Locations Reactor and Auxiliary          M-1G025 Buildings Plan El. 2068'-8" 1.2-15    0      Equipment Location Reactor and Auxiliary Building  M-1G026 Section A 1.2-16    0      Equipment Locations Reactor and Auxiliary          M-1G027 Buildings Section B Rev. 17
 
WOLF CREEK Table 1.6-3 (Sheet 2)
USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                        Title                        Drawing #
1.2-17    0    Equipment Location Reactor and Auxiliary Building M-1G028 Section C 1.2-18    0    Equipment Location Reactor and Auxiliary Building M-1G029 Section D 1.2-19    0    Equipment Location Auxiliary Building Sections E, M-1G030 F, & G 1.2-20    0    Equipment Location Fuel Building Plan Elevation  M-1G040 2000'- 0", 2026'-0" and 2047'-6" 1.2-21    0    Equipment Location Fuel Building Sections A, B, & M-1G041 C
1.2-22    0    Equipment Location Fuel Building Sections D, E, & M-1G042 F
1.2-23    0    Equipment Location Control Building &            M-1G050 Communication Corridor Plan Elevation 1974'- 0" &
1984'-0" 1.2-24    0    Equipment Location Control & Diesel Generator    M-1G051 Buildings & Communication Corridor Plan Elevation 2000'-0" & 2016'-0" 1.2-25    0    Equipment Location Control & Diesel Generator    M-1G052 Buildings & Communication Corridor Plan Elevation 2032'-0" & 2047'-6" 1.2-26    0    Equipment Location Control & Diesel Generator    M-1G053 Buildings & Corridor Plan Elevation 2061'- 6",
2066'-0" & 2073'-6" & Section D.
1.2-27    0    Equipment Location Control & Diesel Generator    M-1G054 Buildings & Communication Corridor Section A 1.2-28    0    Equipment Location Control & Diesel Generator    M-1G055 Buildings Sections B & C 1.2-29    0    Equipment Location Turbine Building Condenser    M-1G060 Pit Plan Elevation 1983'-0" 1.2-30    0    Equipment Location Turbine Building Ground Floor  M-1G061 Plan Elevation 2000'-0" 1.2-31    0    Equipment Location Turbine Building Partial Plan  M-1G062 Elevation 2015'-4" 1.2-32    0    Equipment Location Turbine Building Mezzanine    M-1G063 Floor Plan Elevation 2033'-0" 1.2-33    0    Equipment Location Turbine Building Operating    M-1G064 Floor Plan Elevation 2065'-0" Rev. 17
 
WOLF CREEK Table 1.6-3 (Sheet 3)
USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                        Title                    Drawing #
1.2-34    0    Equipment Location Turbine Building Section A M-1G065 1.2-35    0    Equipment Location Turbine Building Section B M-1G066 1.2-36    0    Equipment Location Turbine Building Section C M-1G067 1.2-37    0    Equipment Location Turbine Building Section D M-1G068 1.2-38    0    Equipment Location Turbine Building Section E M-1G069 1.2-39    0    Equipment Location Turbine Building Section F M-1G070 1.2-40    0    Equipment Location Turbine Building Section G M-0G071 1.2-41    0    Equipment Location Turbine Building Section H M-1G072 1.2-42    0    Turbine Component Laydown Area, Elevation    M-1G073 2065'-0" 1.2-44    0    Site Plan                                    8025-C-KG1202 2.4-3    2    Grading Plan Switchyard Area                  S-0172 2.4-3    3    Drainage Plan Plant Area                      S-0186 2.4-3    4A    Manhole, Pipe & Culvert Schedule              S-0189 Sheet 1 2.4-3    4B    Manhole, Pipe & Culvert Schedule              S-0189 Sheet 2 2.4-3    4C    Manhole, Pipe & Culvert Schedule              S-0189 Sheet 3 2.4-3    4D    Manhole, Pipe & Culvert Schedule              S-0189 Sheet 4 2.4-3    5    Manhole & Pipe Details                        S-0191 2.4-3    6A    Manhole & Pipe Details                        S-0296 Sheet 1 2.4-3    6B    Manhole & Pipe Details                        S-0296 Sheet 2 2.4-3    7    Plant Area Roadway Grading & Drainage        S-0297 5.1-1    1    Reactor Coolant System                        M-12BB01 5.1-1    2    Reactor Coolant System                        M-12BB02 5.1-1    3    Reactor Coolant System                        M-12BB03 5.1-1    4    Reactor Coolant System                        M-12BB04 5.4-7    0    Residual Heat Removal System                  M-12EJ01 5.4-21    0    Hot and Cold Leg Lateral Restraints          C-03BB53 6.2.2-1  0    Containment Spray System                      M-12EN01 6.2.2-2  1    Containment Spray System Reactor Building    M-13EN03 A & B Trains 6.2.2-2  2    Containment Spray System Reactor Building    M-13EN04 A & B Trains 6.2.2-2  3    Containment Spray System Reactor Building    M-13EN05 A & B Trains 6.2.5-1  0    Containment Hydrogen Control System          M-12GS01 6.2.6-1  0    Containment Integrated Leak Rate Test        M-12GP01 6.3-1    1    Borated Refueling Water Storage System        M-12BN01 6.3-1    2    High Pressure Coolant Injection System        M-12EM01 6.3-1    3    High Pressure Coolant Injection System        M-12EM02 6.3-1    4    Accumulator Safety Injection                  M-12EP01 Rev. 30
 
WOLF CREEK Table 1.6-3 (Sheet 4)
USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                        Title                        Drawing #
7.2-1    1    Functional Diagrams (Index and Symbols)          M-744-00018 7.2-1    2    Functional Diagrams (Reactor Trip Signals)      M-744-00019 7.2-1    3    Functional Diagrams (Nuclear Instrumentation and M-744-00020 Manual Trip Signals) 7.2-1    4    Functional Diagrams (Nuclear Instrumentation    M-744-00021 Permissives and Blocks) 7.2-1    5    Functional Diagrams (Primary Coolant System Trip M-744-00022 Signals 7.2-1    6    Functional Diagrams (Pressurizer Trip Signals)  M-744-00023 7.2-1    7    Functional Diagrams (Steam Generator Trip        M-744-00024 Signals) 7.2-1    8    Functional Diagrams (Safeguards Actuation        M-744-00025 Signals) 7.2-1    9    Functional Diagrams (Rod Controls and Rod        M-744-00026 Blocks) 7.2-1    10    Functional Diagrams (Steam Dump Control)        M-744-00027 7.2-1    11    Functional Diagrams (Pressurizer Pressure and    M-744-00028 Level Control) 7.2-1    12    Functional Diagrams (Pressurizer Heater Control) M-744-00029 7.2-1    13    Functional Diagrams (Feedwater Control and      M-744-00030 Isolation) 7.2-1    14    Functional Diagrams (Feedwater Control and      M-744-00031 Isolation) 7.2-1    15    Functional Diagrams (Auxiliary Feedwater Pumps  M-744-00032 Start-up) 7.2-1    16    Functional Diagrams (Turbine Trips, Runbacks and M-744-00033 Other Signals) 7.2-1    17    Functional Diagram (Pressurizer Pressure Relief  M-744-00039 System Train A) 7.2-1    18    Functional Diagram (Pressurizer Pressure Relief  M-744-00040 System Train B) 7.3-1    2    Logic Diagram Engineered Safety Features        J-104-00390 Actuation System (BOP) 7.6-4    1    Train B Functional Diagram Showing Logic        M-744-00039 Requirements for Pressurizer Pressure Relief System 7.6-4    2    Train A Functional Diagram Showing Logic        M-744-00040 Requirements for Pressurizer Pressure Relief System 8.2-3    0    Wolf Creek Substation General Plan              KD-7750 8.2-4    0    One-Line Diagram                                KD-7496 Rev. 17
 
WOLF CREEK Table 1.6-3 (Sheet 5)
USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                        Title                      Drawing #
8.3-1    1    Main Single Line Diagram                        E-11001 8.3-1    2    Single Line Diagram, Essential Service Water    E-K1001 System 8.3-1    3    Single Line Diagram Site Area Loads            E-1001 8.3-2    0    List of Loads Supplied by the Emergency Diesel  E-11005 Generator 8.3-3    0    Logic Diagram Standby Generation Excitation    E-12NE01 Control 8.3-4    0    Logic Diagram Standby Generator System          E-12NE02 Protection 8.3-5    0    Logic Diagram Standby Generator Engine and      E-12KJ01 Governor Control 8.3-6    1    DC Main Single Line Diagram                    E-11010 8.3-7    0    DC Main Single Line Diagram (PK03 and PK04      E-11010A Bus) 9.1-3    1    Fuel Pool Cooling and Cleanup System            M-12EC01 9.1-3    2    Fuel Pool Cooling and Cleanup System            M-12EC02 9.2-1    1    Service Water System                            M-12EA01 9.2-1    2    Service Water System                            M-12EA02 9.2-1    3    Service Water System                            M-0022 Sheet 1 9.2-2    1    Essential Service Water System                  M-12EF01 9.2-2    2    Essential Service Water System                  M-12EF02 9.2-2    3    Essential Service Water System                  M-K2EF01 9.2-2    4    Essential Service Water System                  M-K2EF03 9.2-3    0    ESW Pumphouse Equipment Location - Plan        M-KG080 9.2-4    0    ESWS Pumphouse Equipment Location - Sections    M-KG081 9.2-5    1    Makeup Demineralizer System                    M-0025 Sheet 1 9.2-5    2    Makeup Demineralizer System                    M-0025 Sheet 2 9.2-5    3    Makeup Demineralizer System                    M-0025 Sheet 3 9.2-5    4    Makeup Demineralizer System                    M-0025 Sheet 4 9.2-5    4A    Makeup Demineralizer System                    M-0025 Sheet 4A 9.2-5a    0    Potable Water System                            A-0503 Sheet 1 9.2-13    0    Reactor Make-up Water System                    M-12BL01 9.2-14    0    Closed Cooling Water System                    M-12EB01 9.2-15    1    Component Cooling Water System                  M-12EG01 9.2-15    2    Component Cooling Water System                  M-12EG02 9.2-15    3    Component Cooling Water System                  M-12EG03 9.2-16    0    Demineralized Water Storage and Transfer System M-12AN01 9.2-17    1    Domestic Water System                          M-12KD01 9.2-17    2    Domestic Water System                          M-12KD02 9.2-23    0    Condensate Storage and Transfer System          M-12AP01 Rev. 17
 
WOLF CREEK Table 1.6-3 (Sheet 6)
USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                          Title                      Drawing #
9.2-24    1    Waste Water Treatment Facility                  M-12WT01 9.2-25    1    Waste Water Treatment Facility                  M-12WT03 9.3-1    1    Compressed Air System                            M-12KA01 9.3-1    2    Compressed Air System (Service Air)              M-12KA02 9.3-1    3    Instrument Air System                            M-12KA03 9.3-1    4    Instrument Air System                            M-12KA04 9.3-1    5    Compressed Air System                            M-12KA05 9.3-1    6    Compressed Air System                            M-12KA06 9.3-1    7    Compressed Air System                            M-12KA07 9.3-2    1    Nuclear Sampling System                          M-12SJ01 9.3-2    2    Nuclear Sampling System                          M-12SJ03 9.3-3    0    Nuclear Sampling System                          M-12SJ02 9.3-4    1    Process Sampling System                          M-12RM01 9.3-4    2    Process Sampling System                          M-12RM02 9.3-4    3    Process Sampling System                          M-12RM03 9.3-5    1    Sanitary Lift Station & Turb. Bldg. Sanitary    M-12LA01 Drainage System 9.3-5    2    Comm. Corridor & Control Bldg. Sanitary Drainage M-12LA02 System 9.3-5    3    Chemical and Detergent Waste                    M-12LD01 9.3-5    4    Turbine Bldg. and Aux. Feedwater Pump Rooms      M-12LE01 Oily Waste System 9.3-5    5    Control and Diesel Generator Bldg. Oily Waste    M-12LE02 System 9.3-5    6    Turbine Bldg. and Aux. Boiler Room Oily Waste    M-12LE03 System 9.3-5    7    Tendon Access Gallery and Turbine Bldg. Oily    M-12LE04 Waste System 9.3-5    8    Auxiliary Building Floor and Equipment Drain    M-12LF01 (FED) System 9.3-5    9    Auxiliary Building Floor and Equipment Drain    M-12LF02 System 9.3-5    10    Auxiliary Building Floor and Equipment Drain    M-12LF03 System 9.3-5    11    Auxiliary Building Floor and Equipment Drain    M-12LF04 System 9.3-5    12    Auxiliary Building Floor and Equipment Drain    M-12LF05 System 9.3-5    13    Radwaste and Fuel Bldgs. FED System              M-12LF06 9.3-5    14    Radwaste Bldg. FED System                        M-12LF07 9.3-5    15    Control and Fuel Bldgs. FED System              M-12LF08 9.3-5    16    Reactor Bldg. and Hot Machine Shop FED System    M-12LF09 9.3-5    17    Radwaste Bldg. and Tunnel FED System            M-12LF10 Rev. 17
 
WOLF CREEK Table 1.6-3 (Sheet 7)
USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                        Title                        Drawing #
9.3-7    1    Reactor Building, Stainless Steel Liner Plate,    C-0L2931 Reactor Refueling Canal 9.3-7    2    Fuel Building-Area 1, Stainless Steel Liner Plate C-1L6111 Plan, Spent Fuel Pool 9.3-8    1    Chemical and Volume Control System                M-12BG01 9.3-8    2    Chemical and Volume Control System                M-12BG02 9.3-8    3    Chemical and Volume Control System                M-12BG03 9.3-8    4    Chemical and Volume Control System                M-12BG04 9.3-8    5    Chemical and Volume Control System                M-12BG05 9.3-9    1    Service Gas System                                M-12KH01 9.3-9    2    Service Gas System                                M-12KH02 9.3-11    1    Boron Recycle System                              M-12HE01 9.3-11    2    Boron Recycle System                              M-12HE02 9.3-11    3    Boron Recycle System                              M-12HE03 9.4-1    1    Control Building HVAC                            M-12GK01 9.4-1    2    Control Building HVAC                            M-12GK02 9.4-1    3    Control Building HVAC                            M-12GK03 9.4-1    4    Control Building HVAC                            M-12GK04 9.4-2    1    Fuel Building HVAC                                M-12GG01 9.4-2    2    Fuel Building HVAC                                M-12GG02 9.4-3    1    Miscellaneous Buildings HVAC                      M-12GF01 9.4-3    2    Miscellaneous Buildings HVAC                      M-12GF02 9.4-3    3    Auxiliary Building HVAC                          M-12GL03 9.4-3    4    Auxiliary Building HVAC                          M-12GL02 9.4-3    5    Auxiliary Building HVAC                          M-12GL01 9.4-4    1    Turbine Building HVAC                            M-12GE01 9.4-4    2    Turbine Building HVAC                            M-12GE02 9.4-4    3    Turbine Building HVAC                            M-12GE03 9.4-4    4    Turbine Building HVAC                            M-12GE04 9.4-5    1    Radwaste Building HVAC                            M-12GH01 9.4-5    2    Radwaste Building HVAC                            M-12GH02 9.4-6    1    Containment Cooling System                        M-12GN01 9.4-6    2    Containment Cooling System                        M-12GN02 9.4-6    3    Containment Atmospheric Control System            M-12GR01 9.4-6    4    Containment Purge Systems HVAC                    M-12GT01 9.4-7    0    Diesel Generators Building HVAC                  M-12GM01 9.4-8    0    Essential Service Water Pump House HVAC          M-K2GD01 9.4-9    1    Plant Heating System                              M-12GA01 9.4-9    2    Plant Heating System                              M-12GA02 9.4-10    0    Central Chilled Water System                      M-12GB01 9.4-11    0    Waste Water Treatment Facility HVAC              M-12VW01 Rev. 17
 
WOLF CREEK Table 1.6-3 (Sheet 8)
USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                          Title                    Drawing #
9.5-1    1    Fire Protection System (site)                  M-0023 Sheet 1 9.5-1    2    Fire Protection System (site)                  M-0023 Sheet 2 9.5-1    3    Fire Protection System (site)                  M-0023 Sheet 3 9.5-1    4    Fire Protection System (site)                  M-0023 Sheet 4 9.5-2    0    Outdoor Piping, Key Plan and General Notes      M-0051 9.5.1-1  1    Fire Protection Turbine Building                M-12KC01 9.5.1-1  2    Fire Protection System (power block)            M-12KC02 9.5.1-1  3    Fire Protection System (power block)            M-12KC03 9.5.1-1  4    Fire Protection (Halon) System                  M-12KC04 9.5.1-1  5    Fire Protection System (power block)            M-12KC05 9.5.1-1  6    Fire Protection (Halon) System                  M-12KC06 9.5.1-1  7    Fire Protection (Halon) System                  M-12KC07 9.5.1-2  1    Fire Area Delineation el. 1974                10466-A-1801 9.5.1-2  2    Fire Area Delineation el. 2000                10466-A-1802 9.5.1-2  3    Fire Area Delineation el. 2026                10466-A-1803 9.5.1-2  4    Fire Area Delineation el. 2047-6              10466-A-1804 9.5.2-1  0    Telephone System Riser Diagram                  E-14QE01 9.5.2-2  0    Public Address System Riser Diagram            E-1L9903 9.5.3-1  0    Lighting Distribution Riser Diagram            E-1L9901 9.5.4-1  0    Emergency Fuel Oil System                      M-12JE01 9.5.5-1  1    Standby Diesel Generator "A" Cooling Water      M-12KJ01 System 9.5.5-1  2    Standby Diesel Generator "B" Cooling Water      M-12KJ04 System 9.5.6-1  1    Standby Diesel Generator "A" Intake, Exh., F.0. M-12KJ02 and Starting Air System 9.5.6-1  2    Standby Diesel Generator "B" Intake, Exh., F.0. M-12KJ05 and Starting Air System 9.5.7-1  1    Standby Diesel Generator "A" Lube Oil System    M-12KJ03 9.5.7-1  2    Standby Diesel Generator "B" Lube Oil System    M-12KJ06 9.5.9-1  1    Auxiliary Boiler System                        M-12FA01 9.5.9-1  2    Auxiliary Steam System                          M-12FB01 9.5.9-1  3    Auxiliary Steam System                          M-12FB02 9.5.9-1  4    Auxiliary Steam Chemical Addition System        M-12FE01 9.5.10-1  1    Breathing Air System                            M-12KB01 9.5.10-1  2    Breathing Air System                            M-12KB02 9.5.10-1  3    Breathing Air System                            M-12KB03 Rev. 17
 
WOLF CREEK Table 1.6-3 (Sheet 9)
USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                        Title                          Drawing #
10.2-1    1    Main Turbine                                      M-12AC01 10.2-1    2    Main Turbine                                      M-12AC02 10.2-1    3    Main Turbine                                      M-12AC03 10.2-1    4    Main Turbine                                      M-12AC04 10.2-1    5    Lube Oil Storage, Transfer and Purification System M-12CF01 10.2-1    6    Lube Oil Storage, Transfer and Purification System M-12CF02 10.2-1    7    Main Turbine Control Oil System                    M-12CH01 10.2-1    8    Main Turbine Control Oil System                    M-12CH02 10.3-1    1    Main Steam System                                  M-12AB01 10.3-1    2    Main Steam System                                  M-12AB02 10.3-1    3    Main Steam System                                  M-12AB03 10.4-1    1    Circulating Water & Waterbox Drains System        M-12DA01 10.4-1    2    Circulating Water System                          M-0021 10.4-1    3    Circulating Water Waterbox Venting System          M-12DA02 10.4-1    4    Circulating Water Screenhouse Plans                M-0004 10.4-1    5    Circulating Water Screenhouse - Sections          M-0005 10.4-2    1    Condensate System                                  M-12AD01 10.4-2    2    Condensate System                                  M-12AD02 10.4-2    3    Condensate System                                  M-12AD03 10.4-2    4    Condensate System                                  M-12AD04 10.4-2    5    Condensate System                                  M-12AD05 10.4-2    6    Condensate System                                  M-12AD06 10.4-3    0    Condenser Air Removal                              M-12CG01 10.4-4    0    Steam Seal System                                  M-12CA01 10.4-5    1    Condensate Demineralizer System                    M-12AK01 10.4-5    2    Condensate Demineralizer System                    M-12AK02 10.4-5    3    Condensate Demineralizer System                    M-12AK03 10.4-6    1    Feedwater System                                  M-12AE01 10.4-6    2    Feedwater System                                  M-12AE02 10.4-6    3    Feedwater Heater Extraction Drains & Vents        M-12AF01 10.4-6    4    Feedwater Heater Extraction Drains & Vents        M-12AF02 10.4-6    5    Feedwater Heater Extraction Drains & Vents        M-12AF03 10.4-6    6    Feedwater Heater Extraction Drains & Vents        M-12AF04 10.4-6    7    Auxiliary Turbines S.G.F.P. Turbine A            M-12FC03 10.4-6    8    Auxiliary Turbines S.G.F.P. Turbine "B"            M-12FC04 10.4-7    1    Condensate Chemical Addition System                M-12AQ01 10.4-7    2    Feedwater Chemical Addition System                M-12AQ02 10.4-8    1    Steam Generator Blowdown System                    M-12BM01 10.4-8    2    Steam Generator Blowdown System                    M-12BM02 10.4-8    3    Steam Generator Blowdown System                    M-12BM03 10.4-8    4    Steam Generator Blowdown System                    M-12BM04 10.4-8    5    Steam Generator Blowdown System                    M-12BM05 Rev. 17
 
WOLF CREEK Table 1.6-3 (Sheet 10)
USAR Figure/Controlled Drawing Cross-Reference Figure # Sheet                        Title                        Drawing #
10.4-9    0    Auxiliary Feedwater System                        M-12AL01 10.4-10  0    Auxiliary Turbines Auxiliary Feedwater Pump      M-12FC02 Turbine 10.4-12  1    Secondary Liquid Waste System                    M-12HF01 10.4-12  2    Secondary Liquid Waste System                    M-12HF02 10.4-12  3    Secondary Liquid Waste System                    M-12HF03 10.4-12  4    Secondary Liquid Waste System                    M-12HF04 11.2-1    1    Liquid Radwaste System                            M-12HB01 11.2-1    2    Liquid Radwaste System                            M-12HB02 11.2-1    3    Liquid Radwaste System                            M-12HB03 11.2-1    4    Liquid Radwaste System                            M-12HB04 11.3-1    1    Gaseous Radwaste System                          M-12HA01 11.3-1    2    Gaseous Radwaste System                          M-12HA02 11.3-1    3    Gaseous Radwaste System                          M-12HA03 11.4-1    1    Solid Radwaste System                            M-12HC01 11.4-1    2    Solid Radwaste System                            M-12HC02 11.4-1    3    Solid Radwaste System                            M-12HC03 11.4-1    4    Solid Radwaste System                            M-12HC04 12.3-2    1    Radiation Zones for Normal Operation El. 1974    10466-A-1701 12.3-2    2    Radiation Zones for Normal Operation El. 2000    10466-A-1702 12.3-2    3    Radiation Zones for Normal Operation El. 2026    10466-A-1703 12.3-2    4    Radiation Zones for Normal Operation El. 2047-6 10466-A-1704 12.3-2    5    Radiation Zones for Normal Operation Turbine      10466-A-1705 Bldg El. 1983 & 2000 12.3-2    6    Radiation Zones for Normal Operation Turbine      10466-A-1706 Bldg El. 2033 & 2065 12.3-4    0    Decontamination System                            M-12HD01 18.2-15  0    Nuclear Sampling System                          M-12SJ04 Rev. 17
 
WOLF CREEK Table 1.6-4 (sheet 1)
Incorporated by Reference USAR Section/Controlled Document Cross-Reference Section                                  Title                        Document #
Table 3.11(B)-1                                                            EQSD-I, Plant Environmental Normal Conditions          Attachment A and B Table 3.11(B)-2                                                            EQSD-I, Environmental Qualification Parameters for SNUPPS Attachment A NUREG-0588 (LOCA, MSLB and HELB) and B Table 3.11(B)-3                                                            EQSD-I, Identification of Safety-Related Equipment and      Attachment A Components: Equipment Qualification            and B; EQSD-II, Tables 1 and 2 Table 3.11(B)-4                                                            EQSD-I, Containment Worst Case Radiation Levels (MRADs)
Attachment A Table3.11(B)-5                                                            EQSD-I, Containment Spray Requirements Attachment A Table 3.11(B)-8                                                          EQSD-I, Exemptions from NUREG-0588 Qualification Attachment C Table 3.11(B)-10                                                          EQSD-II, Equipment Added for NUREG-0737 Tables 1 and 2 Figures 3.11(B)-
EQSD-I, 1 through                                    Figures Attachment A 3.11(B)-49 Section                                  Fire Barriers
                  *Note: Only portions of this section have been relocated M-663-00017A 9.5.1.2.2.3*                    and Incorporated by Reference Appendix 9.5B.1              Fire Hazard Analyses - Introduction          E-1F9905 Table 9.5B-1          Minimum Equipment Required for Safe Shutdown        XX-E-013 Appendix 9.5B.2  Fire Hazard Analyses - Assumption on Plant Conditions    E-1F9905 Table 9.5B-2        Equipment Required for Shutdown following a Fire      XX-E-013 Fire Hazard Analyses - Fire Effects on Appendix 9.5B.3                                                            E-1F9905 Electrical Equipment and Safe Shutdown Information Rev. 28
 
WOLF CREEK Table 1.6-4 (sheet 2)
Incorporated by Reference USAR Section/Controlled Document Cross-Reference Section                                Title                      Document #
Table 9.5B-3          Safety-Related Fire Areas Containing Rooms    E-1F9905, Without Detection Provisions          Attachment A Fire Hazard Analyses - General Information on  E-1F9905, or Appendix 9.5B.4 Design Features                XX-E-013, or E-1F9900 Table 9.5B-4      Safety-Related Fire Areas Outside Containment With E-1F9905 Area Suppression Coverage            Attachment A Fire Hazard Analyses - Combustible Loadings Appendix 9.5B.5                                                      E-1F9905 and Flame Spread E-1F9905, Table 9.5B-5              Non-Safety Related Site Structures Attachment C Fire Hazard Analyses - Fire Hazard Review      E-1F9905, or Appendix 9.5B.6 Methodology                  XX-E-013 Appendix 9.5B.7        Fire Hazard Analyses - Power Block Fire Hazards Analysis                E-1F9905 Fire Hazard Analyses - Site Specific Fire Appendix 9.5B.8                    Hazards Analysis                E-1F9905 Design Basis Document for OFN AP-017, Control Room Appendix 9.5B                                                        E-1F9915 Evacuation Technical Chapter 16.0                Technical Requirements Manual          Requirements Manual Quality Chapter 17.2        Quality Assurance During the Operation Phase    Program Manual Rev. 29
 
WOLF CREEK 1.7  DRAWINGS AND OTHER DETAILED INFORMATION The engineering drawings listed in Tables 1.7-1, 1.7-2 and 1.7-3 reflect the detailed design configuration as described in USAR text and tables. The controlled drawings that were removed from the USAR in Revision 17 and incorporated by reference are listed in Table 1.6-3 1.7.1  Electrical, Instrumentation and Control Drawings Table 1.7-1 contains a list of electrical, instrumentation, and control drawings that are considered to be necessary to evaluate the safety-related features pertaining to the power block.
1.7.2  Piping and Instrumentation Diagrams Table 1.7-2 contains a list of each piping and instrumentation diagram and the corresponding USAR figure number as it appears at the end of the respective text section. The P&ID legend, Figure 1.1-1, provides an explanation of symbols and characters used in these USAR figures.
1.7.3  Miscellaneous Controlled Drawings Table 1.7-3 contains a list of other controlled drawings utilized as figures in the USAR and the corresponding USAR figure number as it appears at the end of the respective text section.
1.7-1                  Rev. 30
 
WOLF CREEK TABLE 1.7-1 ELECTRICAL, INSTRUMENTATION, AND CONTROL DRAWINGS DRAWING NUMBER                            TITLE E-K1001          SINGLE LINE DIAGRAM ESSENTIAL SERVICE WATER SYSTEM E-11NB01          LOWER MED VOLTAGE SYS CLASS 1E 4.16 KV SINGLE LN E-11NB02          LOWER MED VOLTAGE SYS CLASS 1E 4.16 KV SINGLE LN E-11NE01          STBY GENERATOR CLASS 1E 4.16 KV S/L E-11NG01          CLASS 1E LOW VOLTAGE 480V SYS S/L M & R E-K1NG01          LOW VOLTAGE SYSTEM CLASS 1E 480V SINGLE LINE METER & RELAY DIAGRAM E-11NG02          CLASS 1E LOW VOLTAGE 480V SYS S/L M & R E-11NG20          LOW VOLTAGE SYSTEM CLASS 1E MOTOR CONTROL CENTERS
 
==SUMMARY==
 
E-11NK01          CLASS 1E 125 DC SYS S/L M & R E-11NK02          CLASS 1E 125 DC SYS S/L M & R E-11PA01          HIGH MED VOLT SYS 13.8 KV M & R E-11PA02          HIGH MED VOLT SYS 13.8 KV M & K E-11PG06          NON-CLASS 1E LOW VOLTAGE SYS 480V S/L M & R E-11PN01          NON-CLASS 1E INSTRUMENT AC POWER E-12KJ01          STANDBY GENERATOR SYSTEM E-K2NG01          LOGIC DIAGRAM - 4.16 KV MOTOR CONTROL CENTER TRANSFORMER FEEDER BREAKERS (VOID)
E-K2NG02          LOGIC DIAGRAM - 480V MOTOR CONTROL CENTER TRANSFORMER FEEDER BREAKERS (VOID)
E-K2NG03          LOGIC DIAGRAM - 480V SYSTEM NOTES
                    & REFERENCES E-13AB01          MAINSTEAM SUPPLY VLV TO TURB DR AUX FEEDW PUMP E-13AB03A        MAIN STEAM LINE DRAIN VLV E-13AB03B        MAIN STEAM LINE DRAIN VLV E-13AB06A        MAIN STM ATMOS VENT VLV POS INT LIGHT Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 2)
DRAWING NUMBER                    TITLE E-13AB06B MAIN STM ATMOS VENT VLV POS INT LIGHT E-13AB08  MN STM COOLDOWN VLV E-13AB11A SCHEMAT DIAG MAIN DUMP VLVS E-13AB11B SCHEMAT DIAG MAIN STM DUMP VLV E-13AB11C SCHEMAT DIAG MAIN STM DUMP VLV E-13AB17  MN STM BYPASS VAL TO AUX FEEDWTR PUMP TURB E-13AB23A SCHEMAT DIAG MAIN STM BYPASS VLVS E-13AB23B SCHEMAT DIAG MAIN STM BYPASS VLVS E-13AB26  SCHEMATIC DIAGRAM MAIN STEAM ISO VALVES ALL CLOSE - SEPARATION GROUP 1 E-13AB27  SCHEMATIC DIAGRAM MAIN STEAM ISO VALVES ALL CLOSE - SEPARATION GROUP 4 E-13AB28  SCHEMATIC DIAGRAM MAIN STEAM ISO VALVES CONTROL PART 1 E-13AB29  SCHEMATIC DIAGRAM MAIN STEAM ISO VALVES CONTROL PART 2 E-13AB30  SCHEMATIC DIAGRAM MAIN STEAM AND FEEDWATER ISO VLV MISCELLANEOUS CIRCUITS E-13AB31  STEAM DUMP CONTROL & BYPASS INDICATION E-13AC38  MAIN TURB SYS WITH NSSS INTERFACE E-13AE05  SCH DIAG STEAM GENER CHEMICAL INJECT E-13AE06  SCH DIAG MAIN FEEDWTR CONTR VALVES E-13AE07  SCH DIAG MAIN FEEDWTR CONTR VALVES E-13AE14  SCHEMATIC DIAGRAM FEEDWATER ISOLATION VALVES ALL CLOSE - SEPARATION GROUP 1 E-13AE15  SCHEMATIC DIAGRAM FEEDWATER ISOLATION VALVES ALL CLOSE - SEPARATION GROUP 4 E-13AE16  SCHEMATIC DIAGRAM FEEDWATER ISO VALVES CONTROL PART 1 E-13AE17  SCHEMATIC DIAGRAM FEEDWATER ISO VALVES CONTROL PART 2 E-13AE18  BYPASS FEEDWATER CONTROL VALVES E-13AL00  AUX FEEDWATER SCHEMATIC INDEX SHEET E-13AL01A AUX FEEDWATER SYSTEM MOTOR DRIVEN PUMP E-13AL01B AUX FEEDWATER SYSTEM MOTOR DRIVEN PUMP E-13AL02A AUX FEEDWATER SYSTEM MOTOR OPERATED VALVES Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 3)
DRAWING NUMBER                    TITLE E-13AL02B AUX FEEDWATER SYSTEM MOTOR OPERATED VALVES E-13AL02C SCHEMATIC DIAGRAM MOTOR OPERATED VALVE ALHV0036 E-13AL03A AUX FEEDWATER PUMPS DISCH CONTR MOVS E-13AL03B AUX FEEDWATER PUMPS DISCH CONTROL MOVS E-13AL04A SUPPLY FROM ESS SERV WTR SYS E-13AL04B SUPPLY FROM ESS SERV WTR SYS E-13AL05A AUX FDWR PUMPS DISCH CONTR AIR OP VALVES E-13AL05B AUX FDWR PUMPS DISCH CONTR AIR OP VALVES E-13AL06  SCHEMATIC DIAGRAM INSTRUMENTATION & ALARMS E-13AL07A SCHEMATIC DIAGRAM INSTRUMENTATION & ALARMS E-13AL07B SCHEMATIC DIAGRAM INSTRUMENTATION & ALARMS E-13AL08  SCHEMATIC DIAGRAM INSTRUMENTATION & ALARMS E-13AL09  SCHEMATIC DIAGRAM MISCELLANEOUS CIRCUITS E-13AL10  SCHEMATIC DIAGRAM AUX FDWTR SYS E-13AP04  SCHEMATIC DIAGRAM CONDENSATE SYSTEM E-13BB03  S.D. RCP THERM BARRIER CCW ISO VLVS E-13BB04  S.D. SEAL WTR INJECT ISO VALVE E-13BB11  S.D. PRZR RELIEF TANK VENT TO WPS ISO VLV E-13BB12A S.D. RHR LOOP INLET ISO VALVE E-13BB12B S.D. RHR LOOP INLET ISO VALVE E-13BB13  S.D. PRZR RELIEF TK VENT TO WPRS ISO VLV E-13BB27  SCH DIAG REACTOR COOLANT PUMP MOTORS E-13BB28  SCHEMATIC DIAGRAM REACTOR COOLANT PUMP E-13BB30  SCHEMATIC DIAGRAM REACTOR COOLANT PUMP E-13BB31  SCHEMATIC DIAGRAM REACTOR COOLANT PUMP E-13BB33  SCHEMATIC DIAGRAM E-13BB34  SCHEMATIC DIAGRAM E-13BB35  SCHEMATIC DIAGRAM REACTOR COOLANT PUMP E-13BB36  SCHEMATIC DIAGRAM REACTOR COOLANT PUMP Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 4)
DRAWING NUMBER                    TITLE E-13BB37  SCHEMATIC DIAGRAM REACTOR COOLANT PUMP E-13BB38  SCHEMATIC DIAGRAM REACTOR COOLANT PUMP E-13BB39  SCHEMATIC DIAGRAM REACTOR COOLANT PUMP E-13BB40  SCHEMATIC DIAGRAM REACTOR COOLANT PUMP E-13BG01  CENTRIFUGAL CHARGING PUMPS SCHEMATIC DIAGRAM E-13BG11A CHARGING PUMP TO REACTOR COOLANT & MINIFLOW ISO E-13BG11B CHARGING PUMP TO REACTOR COOLANT & MINIFLOW ISO E-13BG11C SCHEMATIC DIAGRAM CHARGING PUMP TO REACTOR COOLANT AND MINIFLOW ISOLATION VALVE BGHV8111 E-13BG12  VOLUME CONT TANK OUTLET ISO VLVS SCH DIAG E-13BG12A VOLUME CONTROL TANK OUTLET ISO VALVE E-13BG13  BORIC ACID FILTER TO CHG PUMP VLV E-13BG17  LETDOWN LINE ISO VLV SCH DIAG E-13BG24  REACTOR COOLANT PUMP SEAL WATER ISO VLV SCH DIAG E-13BG27  BORIC ACID TRANSFER PUMPS E-13BG36  LETDOWN CONTAINMENT ISO VLV SCH DIAG E-13BG38  REACTOR COOLANT PUMP SEAL WATER ISO VLV SCH DIAG E-13BG48  EXCESS LETDOWN LINE ISO VLV SCH DIAG E-13BG52  RCP SEAL INJECTION FLOW THROTTLING VALVES E-13BL01  PRZR RELIEF TANK REACTOR TO MAKEUP WTR SUPPLY E-13BM01  STEAM GENERATOR BLOWDOWN SCHEMATIC E-13BM02  STEAM GENERATOR BLOWDOWN SCHEMATIC E-13BM03  STEAM GENERATOR BLOWDOWN SCHEMATIC E-13BM06A STEAM GENERATOR BLOWDOWN SCHEMATIC E-13BM06B STEAM GENERATOR BLOWDOWN SCHEMATIC E-13BM06C STEAM GENERATOR BLOWDOWN SCHEMATIC E-13BN01  REFUELG WTR STRG TK TO CHARGE PUMP MOV E-13BN01A SCHEMATIC DIAGRAM REFUELING WATER STORAGE TANK TO CHARGING PUMP MOV BNLCV0112E Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 5)
DRAWING NUMBER                    TITLE E-13BN02  REFUELG WTR STRG TK TO SFPCS PUMP MOV E-13BN03  REFUELG WTR STRG TK RHR PUMP MOV E-13BN04  RWST TO CONTMET SPRAY PUMP MOV E-13BN06  RWST TO SAFETY INJECT PUMP VLV E-13BN08  SIS PUMP MINIFLW ISO VLV E-13BN10  SCHEMATIC DIAGRAM E-13EC01  FUEL POOL COOLING PUMPS SCHEMATIC DIAGRAM E-13EC02  SCHEM DIAG CCW DISCHARGE VLVS FROM FUEL POOL CLG E-13EF01  ESW TO AIR COMP ISOL VALVES SCHEMATIC DIAGRAM E-K3EF01  SCHEMATIC DIAGRAM - ESW PUMPS E-13EF02  SCHEMATIC DIAG ESW TO SW SYS ISOL VLVS E-K3EF02  SCHEMATIC DIAGRAM - TRAVELING WATER SCREENS E-13EF03  SCHEMATIC DIAG ESW TO SW SYS ISOL VLVS E-K3EF03  SCHEMATIC DIAGRAM - SCREEN WASH WATER VALVE E-13EF04  SCHEM DIAG ESW FROM COMPONENT COOLG WTR HEAT EXC E-K3EF04  SCHEMATIC DIAGRAM - ESW SELF-CLEANING STRAINER E-13EF05  SCHEMATIC DIAGRAM ESW TO COMPONENT COOLG WTR HEA E-13EF05A SCHEMATIC DIAGRAM ESW TO COMPONENT COOLING WATER HEAT EXCHANGER ISOLATION VALVE EFHV0052 E-K3EF05  SCHEMATIC DIAGRAM - SELF-CLEANING STRAINER TRASH VALVE E-13EF06  SCHEMATIC DIAG ESW TO ULTIM HEAT SINK ISOL VALVE E-K3EF06  SCHEMATIC DIAGRAM - ESW PUMP DISCHARGE AIR RELEASE VALVE E-13EF07  ESWS CONTAIN COOLER VALVE E-13EF07A SCHEMATIC DIAGRAM ESW TO CONTAINMENT AIR COOLERS ISOLATION VALVE EFHV0032 E-K3EF07  SCHEMATIC DIAGRAM - SYSTEM BACKPRESSURE CONTROL VALVE E-13EF08  SCHEM DIAG ESW FROM CONTAIN-MENT AIR COOLERS ISOL E-K3EF08  SCHEMATIC DIAGRAM - MISCELLANEOUS CIRCUITS E-13EF08A SCHEMATIC DIAGRAM ESW TO CONTAINMENT AIR COOLERS ISOLATION VALVE EFHV0050 E-13EF09  SCHEM DIAG ESW TO/FROM CONTAIN AIR COOL ISOL VLV E-13EF09A ESW TO/FROM CONTAIN AIR COOLERS ISOLATION VALVES E-K3EF09  SCHEMATIC DIAGRAM - STATUS PANEL CIRCUITS E-13EF10  SCHEM DIAG ESW FROM CONTAIN AIR COOLERS ISOL VLV Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 6)
DRAWING NUMBER                    TITLE E-K3EF10  SCHEMATIC DIAGRAM - MISCELLANEOUS CIRCUITS E-13EG01A CCW PUMPS A & B E-13EG01B CCW PUMPS C & D E-13EG01C SCHEMATIC DIAGRAM COMPONENT COOLING WATER E-13EG01D SCHEMATIC DIAGRAM E-13EG03  CCW SURGE TANK VENT E-13EG04  ESW MAKEUP TO CCW SYS E-13EG05A CCWS SUPPLY RETURN FROM NUCLEAR AUX COMPONENT E-13EG05B CCWS SUPPLY TO NUCLEAR AUX COMPONENT E-13EG05C SCHEMATIC DIAGRAM COMPONENT COOLING WATER RETURN FROM NUCLEAR AUX.
COMPONENT EGHV0015 E-13EG05D SCHEMATIC DIAGRAM COMPONENT COOLING WATER SUPPLY FROM NUCLEAR AUX.
COMPONENT EGHV0054 E-13EG06  CCW TO CONTAINMENT ISO VALVES E-13EG07  CCW SUPPLY TO RHR HEAT EXCHANGER E-13EG07A COMPONENT COOLING WATER SUPPLY TO RHR HEAT EXCHANGER EGHV0102 E-13EG08  CCW SUPPLY RETURN FROM RADWASTE BLDG E-13EG09  CCW CONTAINMENT ISO VALVES E-13EG10  CONTAINMENT ISO VALVE RETURN FROM THERM BAR COOL E-13EG16  CCW HEAT EXCHANGER OUTLET TEMP CONT VALVES E-13EG17  SCHEMATIC DIAGRAM ISOMETRIC VALVES E-13EG18  SCHEMATIC DIAGRAM ISOMETRIC VALVES E-13EG19  SCHEMATIC DIAGRAM ISOMETRIC VALVES E-13EG20  SCHEMATIC DIAGRAM ISOMETRIC VALVES E-13EJ01  RESIDUAL HEAT REMOVAL PUMPS 1 & 2 SCHEM DIAGRAM E-13EJ02  NUCLEAR SAMPLE LINE VALVES SCHEMATIC DIAGRAM E-13EJ03  RHR CROSS CONNECT VALVES SCHEMATIC DIAGRAM E-13EJ04A RHR PUMP TO CHG PUMP VLV SCH DIA E-13EJ04B RHR PUMP TO CHARGING PUMP VALVE E-13EJ05A RHR PUMP TO CHG PUMP VLV SCH DIAG E-13EJ05B RHR LOOP INLET ISO VALVE E-13EJ06A RHR PUMP TO CHG PUMP VLV SCH DIAG E-13EJ06B SUMP TO RESIDUAL HEAT REMOVAL PUMP Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 7)
DRAWING NUMBER                    TITLE E-13EJ07  RESIDUAL HEAT REMOV HOT & COLD LEG TEST LINE E-13EJ08  RESIDUAL HEAT REMOV MINI FLOW VALVES E-13EJ08A SCHEMATIC DIAGRAM RESIDUAL HEAT REMOVAL PUMP MINIFLOW VALVE EJFCV0611 E-13EJ09A RHR COLD & HOT LEG VALVES E-13EJ09B RHR COLD & HOT LEG VALVES E-13EJ09C SCHEMATIC DIAGRAM RHR TO COLD AND HOT LEG VALVES EJHV8809B AND EJHV8840 E-13EJ13  SCHEMATIC DIAGRAM E-13EJ14  SCHEMATIC DIAGRAM E-13EM01  HIGH HEAD SAFETY INJECTION SCHEMATIC E-13EM02  BORON INJ TANK DISCH & INLET ISOLA VALVES E-13EM02A SCHEMATIC DIAGRAM BORON INJECTION TANK DISCHARGE & INLET ISOLATION VALVE EMHV8801B E-13EM02B SCHEMATIC DIAGRAM BORON INJECTION TANK DISCHARGE ISOLATION VALVE EMHV8803B E-13EM02C SCHEMATIC DIAGRAM BORON INJECTION TANK DISCHARGE AND INLET ISOLATION VALVE EMHV8801A E-13EM03  SIS MINI FLOW ISOLATION VALVES E-13EM04  SAFETY INJ & CHARGING PUMPS HOT
            & COLD LEG TEST E-13EM04A SCHEMATIC DIAGRAM CHARGING PUMPS COLD LEG E-13EM08  SUCTION HEADER & SAFETY INJ CROSS CONNECTION E-13EM09  SAFETY INJECTION PUMP SUCTION VALVES E-13EM11  SAFETY INJECTION PUMP TO HOT LEG MOV E-13EM12  ACCUMULATOR FILL LINE & TEST HEADER LINE E-13EM13A SAFETY INJ PUMP TO HOT & COLD LEGS E-13EM13B SAFETY INJ PUMP TO HOT & COLD LEGS E-13EM17  SCHEMATIC DIAGRAM E-13EN01  CONTAINMENT SPRAY SYS SCHEMAT DIAG E-13EN02  CONTAINMENT SPRAY PUMP SUCT SCHEMATIC DIAGRAM E-13EN03  CONT SPRAY P.P ISOLATION SCH DIAG E-13EN04  CONT SPRAY PP ISO DIAG E-13EP01  ACCUMULATOR N2 SUPP ISO VLV E-13EP02A ACCUMULATOR ISO VLV E-13EP02B ACCUMULATOR ISO VLV E-13EP07  SD ACCUMULATOR ISO VLV E-13EP09  SAFETY DRY ACCUMULATOR VENT VLVS Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 8)
DRAWING NUMBER                    TITLE E-13FC21 SCH DIAG SGFP E-13FC23 SCH DIAG AUX WATER TURB TRIP
          & THROTTLE VALVE E-13FC24 SCH DIAG AUX WATER TURB TRIP
          & THROTTLE VALVE E-13FC25 SCH DIAG AUX WATER TURB TRIP
          & THROTTLE VALVE E-13FC26 SCH DIAG AUX WATER TURB TRIP
          & THROTTLE VALVE E-13FC27 SGFPT ISOLATION INPUT TO ESFAS E-K3GD01 SCHEMATIC DIAGRAM - ESW PUMP ROOM SUPPLY FANS E-K3GD02 SCHEMATIC DIAGRAM - ESW PUMP ROOM UNIT HEATERS E-K3GD03 SCHEMATIC DIAGRAM - ESW PUMP ROOM EXHAUST VALVE E-K3GD04 SCHEMATIC DIAGRAM - ESW PUMP ROOM MISCELLANEOUS CIRCUITS E-K3GD05 SCHEMATIC DIAGRAM - ESW VALVE PIT UNIT HEATER E-13GE18 COND AIR REMOVAL FILTER SYS DAMPERS E-13GF01 AUX FDWTR PUMP RM COOLS FANS E-13GF07 MAIN STREAM ENCL BLDG EXHAUST FANS & DAMPERS E-13GF08 TENDON GALLERY SUPPLY/RETURN ISOL DAMPERS E-13GF13 MAIN STEAM ENCLOSURE BLDG MISC DAMPERS E-13GF14 MISC MOTOR SPACE HEATERS E-13GG01 EMERGING EXHAUST FAN & DIS-CHARGE DAMPERS E-13GG02 SPENT FUEL PUMP ROOM COOLERS E-13GG03 EMERGENCY EXHAUST HTG COILS E-13GG08 FUEL BLDG ISOL DAMPERS OUTSIDE AIR SUPPLY E-13GG09 FUEL BLDG EXHAUST TO EMERG FILT ADSORB UNITS ISO E-13GG10 EMERG FILTER ADSORB UNITS AUX BLDG INTAKE ISOL E-13GG11 SPENT FUEL POOL DISCHARGE TO AUX BLDG DAMPER E-13GG12 FUEL BLDG INSTRUMENTATION E-13GG15 EMERG EXHAUST CROSS CONNECTION DAMPER E-13GG17 SPENT FUEL POOL NORMAL/EMRGNCY EXHAUST RADIOACTIVITY SAMPLE Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 9)
DRAWING NUMBER                    TITLE E-13GG18  MANUAL INITIATION FUEL BLDG ISOL SIGNAL E-13GK01A SCHEMATIC DIAG CONTROL ROOM FAN DAMPER E-13GK01B CONTROL ROOM HVAC SYSTEM E-13GK02A CONTROL ROOM A/C UNIT FAN E-13GK02B CONTROL ROOM A/C UNIT FAN E-13GK02C CONTROL ROOM A/C UNIT SUP E-13GK03A FIRE ISOLATION DAMPERS E-13GK03B MISCELLANEOUS DAMPERS E-13GK07  MISCELLANEOUS DAMPERS E-13GK09A ISO DAMPERS E-13GK09B ISO DAMPERS E-13GK10A CONTROL ROOM PRESSURIZATION FAN & SUP DAMPERS E-13GK10B CONTROL ROOM PRESSURIZATION FAN
            & SUPPLY DAMPERS E-13GK11  SCH DIAG CONTROL ROOM E-13GK13  CLASS IE ELE A/C UNIT E-13GK13A SCHEMATIC DIAGRAM E-13GK17  SCH DIAGRAM ISO DAMPERS E-13GK19  SCH DIAGRAM AIR RETURN CONT RM ASB UNITS E-13GK23  SCH DIAG FIRE ISO DAMPERS E-13GK25  SCH DIAGRAM MISCELLANEOUS ALARMS E-13GK28  CONTROL ROOM HVAC SYSTEM E-13GK30A ISOLATION DAMPERS E-13GK30B ISOLATION DAMPERS E-13GK31  FIRE SIGN ISOMETRIC E-13GL02  AUX BLDG SUPPLY AIR UNIT SUPPLY DAMPERS E-13GL05  PUMP ROOM COOLERS E-13GL06  PUMP ROOM COOLERS E-13GL12  PENETRATION ROOM COOLERS E-13GL12A PENETRATION ROOM COOLER E-13GL14  SCH DIAGRAM ISOL DAMPERS E-13GL14A SCHEMATIC DIAGRAM E-13GL16  FUEL BLDG NORMAL EXHAUST ISOL DAMPERS E-13GL27  CCW PUMP ROOM EXHAUST DAMP E-13GL30  ISOLATION DAMPERS E-13GM01  DIESEL GENERATOR VENTILATION SUPPLY FANS SCHEM E-13GM02  DIESEL GENERATOR SUPPLY DAMPER CONTROL & MISC AL E-13GM04  DIESEL GENERATOR BLDG EXHAUST DAMPERS SCHEMATIC E-13GM04A SCHEMATIC DIAGRAM Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 10)
DRAWING NUMBER                    TITLE E-13GN01  HYDROGEN MIXING FANS E-13GN02  CONTAINMENT COOLER FANS E-13GN02A SCHEMATIC DIAGRAM MISC FAN SPACE HEATER E-13GN03  CRDM COOLING FANS & DISCHARGE DAMPERS E-13GN09  MISC FAN SPACE HTRS E-13GS01A SCHEMATIC DIAGRAM E-13GS01B SCHEMATIC DIAGRAM E-13GS02A SCHEMATIC DIAGRAM E-13GS02B SCHEMATIC DIAGRAM E-13GS03  SCHEMATIC DIAG E-13GS04  SCHEMATIC DIAG E-13GS05  SCHEMATIC DIAG E-13GS06  SCHEMATIC DIAG E-13GS07  SCHEMATIC DIAG E-13GS10  SCHEMATIC DIAGRAM E-13GS11  SCHEMATIC DIAGRAM E-13GS12  SCHEMATIC DIAGRAM E-13GS13  SCHEMATIC DIAGRAM E-13GS14  SCHEMATIC DIAGRAM E-13GT03  CONTAINMENT PURGE SYSTEM SCHEMATIC E-13GT04  CONTAINMENT PURGE SYSTEM SCHEMATIC E-13GT05  MINI PURGE FAN SCHEMATIC E-13GT06  SCH DIAGRAM CON PUR SUP EX DM E-13GT13  CTMT PURGE EXHAUST SMPL VLVS E-13HB02  REACTOR COOLANT DRAIN TK PUMP DISC & VENT ISOL VALVE E-13HB03  REACTOR COOLANT DRAIN TK VENT ISOL VLV E-13HB19  REACTOR COOLANT DRAIN TK PUMP DISCH VLV E-13JE01  EMERGENCY FUEL OIL TRANSFER PUMPS E-13JE01A SCHEMATIC DIAGRAM EMERGENCY FUEL OIL TRANSFER PUMP PJE01B E-13JE02  MISC CIRCUITS E-13JE03  FIRE SIGNAL ISOL RELAY E-13KA01A SCH DIA AIR COMPR ISOL CIRCUIT BREAKER E-13KA02  SCH DIA COMPRESSED AIR CONTAIN-MENT ISO VLV E-13KA04  SCH DIA HYDROGEN CONTR SYST MAKE-UP AIR VALVE E-13KC08  FIRE PROTECTION SYSTEM ISO VLV E-13KJ01A DIESEL GEN KKJ01A ENG CON SCH E-13KJ01B DIESEL GEN KKJ01A ENG CON SCH E-13KJ02  S.D. DIESEL GEN CONTROLS E-13KJ03A DG KKJ01B ENG CON SCH Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 11)
DRAWING NUMBER                    TITLE E-13KJ03B DG KKJ01B ENG CON SCH E-13KJ04  DG KKJ01B ANNUN MISC CKTS SCH DIAG E-13KJ05  JKT COOL & LUBE OIL LEVEL CON E-13KJ06  DIESEL GEN KKJ01B GOV CON E-13KJ07  DIESEL GEN KKJ01B GOV CON E-13KJ08  STANDBY JKT COOL HTR E-13KJ09  STANDBY JKT COOL CIRC PMP E-13KJ10  STANDBY LUBE OIL HTR E-13KJ11  SCH DIAGRAM LUBE OIL KEEP WARM PUMP E-13KJ12  SCH DIAG GEN SPACE HEATER E-13KJ13  ROCKER ARM PRE LUBE PUMP E-13KJ16  DIESEL GEN RTD'S THERMOCOUPLES E-13KJ20  STANDBY D/G STARTING AIR SYSTEM E-13LF07  SUMP DISCHARGE VALVES E-13LF08  SUMP PUMP DISCHARGE ISOLATION VALVE E-13LF09A REACTOR BLDG SUMP PUMP DISCHARGE ISO VLVS E-13LF09B REACTOR BLDG SUMP PUMP DISCHARGE ISO VALVES E-13NB01  LOWER MED VOLTAGE SYS CLASS IE 4.16 KV 3/L METER E-13NB02  LOWER MED VOLTAGE SYS CLASS IE 4.16 KV 3/L METER E-13NB03  LOWER MED VOLTAGE SYS CLASS IE 4.16 KV 3/L METER E-13NB04  LOWER MED VOLTAGE SYS CLASS IE 4.16 KV 3/L METER E-13NB05  LOWER MED VOLTAGE SYS CLASS IE 4.16 KV 3/L METER E-13NB06  LOWER MED VOLTAGE SYS CLASS IE 4.16 KV 3/L METER E-13NB10  SCHEMAT DIAGRAM FEEDER BRKR 13.8 KV E-13NB11  SCHEMAT DIAGRAM FEEDER BRKR 13.8 KV E-13NB12  SCHEMAT DIAGRAM FEEDER BRKR 13.8 KV E-13NB13  SCHEMAT DIAGRAM FEEDER BRKR 13.8 KV E-13NB14  SCHEMAT DIAGRAM FEEDER BRKR 13.8 KV E-13NB15  SCHEMAT DIAGRAM FEEDER BRKR 13.8 KV E-13NE01  STANDBY GENER. SYS. 3/L M&R DIAG E-13NE02  STANDBY GENER. SYS. 3/L M&R DIAG Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 12)
DRAWING NUMBER                    TITLE E-13NE10  SCHEMATIC DIAGRAM E-13NE11  SCHEMATIC DIAGRAM E-13NE12  STANDBY GEN SYS E-13NE13  STANDBY GEN SYS E-13NF01  LOAD SHEDDER AND LOAD SEQUENCER E-03NG01  CLASS IE LOW VOLTAGE SYSTEM 3/L M&R E-K3NG01  LOW VOLTAGE SYSTEM CLASS IE 480 V THREE LINE METER & RELAY DIAGRAM E-03NG10  CLASS IE 4.16 KV LC XFMR FDR BKRS E-K3NG10  SCHEMATIC DIAGRAM 4.16 KV TRANSFORMER FEEDER BREAKERS E-03NG10A SCHEMATIC DIAGRAM E-03NG11  BREAKER CLASS IE LOW VO E-03NG11A SCHEMATIC DIAGRAM E-03NG11B SCHEMATIC DIAGRAM 480 V LC MAIN FEEDER E-03NG12  SCHEMATIC DIAG 480 V LC TIE BREAKER MISC COMP INPU E-03NK10  125 VDC & 250 VDC POW SYS SCHEMATICS E-03NK11  125 VDC CLASS IE POWER SYSTEM E-13NN01  CLASS IE INSTRUMENT AC SCHEMATIC E-13PA02  HIGH MED VOLT SYS 13.8 KV 3/L M&R E-13PA05  HIGH MED VOLT SYS MED 3/L M&R E-03PG05  NON-CLASS IE L.V. SYSTEM 3/L M&R E-03PG12  NON-CLASS IE L.V. SYSTEM SCHEMATIC E-03PG12A SCHEMATIC DIAGRAM E-13PG15B 480 V & LC TIE BKR SCHEMATIC E-03PJ11  250 VDC POW SYS SCHEMAT E-13PN01  NON CLASS IE INST AC LINE DIAG E-03QB01  STANDBY LIGHTING SYSTEM PWR FDRS E-K3QB01  SCHEMATIC DIAGRAM STANDBY LIGHTING SYSTEM POWER FEEDER E-03QB02  STANDBY LIGHTING SYSTEM PWR FDRS E-13SB12A REACTOR TRIP & SAFETY INJEC SWITCHES E-13SB12B REACTOR TRIP AND SAFETY INJECT SWITCHES E-13SB13  SCH DIAG SPRAY ACTUATION &
CTMT ISO SWITCHES E-03SB14  SCH DIAG SWITCH DEVEL E-13SB15  CONTROL BOARD SWITCHES Rev. 15
 
WOLF CREEK TABLE 1.7-1 (SHEET 13)
DRAWING NUMBER                    TITLE E-13SB17  SAF INJ (RWST) RESET SWITCH/
SWITCHOVER STATUS INDICATOR E-13SB19  SCHEMATIC DIAGRAM E-13SJ01  NUC SAMP SYS CTMT ISO VLVS E-13SJ02  NUC SAMP SYS SCHEMAT CTMT ISO VLVS E-13SJ05  NUCLEAR SAMPLE SYSTEM E-13SJ06  NUCLEAR SAMPLE SYSTEM E-13SJ07  NUCLEAR SAMPLE SYSTEM ISO VLVS E-13SJ09  SCHEMATIC DIAGRAM E-13SJ11  SCHEMATIC DIAGRAM E-1R0223  RACEWAY PLAN - STATION SERVICE AND STARTUP XFMR AREA E-0R0224  RACEWAY PLAN - ESF TRANSFORMER AREA E-1R3211  RACEWAY PLAN CONTROL BLDG.
AREA 1 E-1R3221  RACEWAY PLAN - COMMUNICATION CORRIDOR AREA - 2 EL. 1974'-0" AND EL. 1984'-0" E-1R3321  RACEWAY PLAN - COMMUNICATION CORRIDOR AREA - 2 EL. 2000'-0" E-1R4321  RACEWAY PLAN - TURBINE BUILDING AREA - 2 EL. 2000'-0" E-1R4331  RACEWAY PLAN - TURBINE BUILDING AREA - 3 EL. 2000'-0" J-020101  SYMBOLS & LEGEND FOR LOGIC DIAGRAMS J-02AB01  CONTROL LOGIC DIAGRAM J-02AB02  MN STM ATMOS VENT VLVS INDICATING LIGHTS J-02AB02A CONTROL LOGIC DIAGRAM J-02AB02B CONTROL LOGIC DIAGRAM J-02AB03  MAIN STEAM SUPPLY TD AUX FEED-WATER PUMP J-12AB04  MAIN STEAM LINE DRAIN VALVES J-02AB10  MAIN STEAM SYSTEM BYPASS VALVE AUX FDWTR PUMP J-02AB12  MAIN STEAM STOP VALVE BYPASS VALVE J-02AC06  MAIN TURBINE TRIP LOGIC LEGENDS
            & NOTES J-02AC07  MAIN TURBINE TRIP BLOCK DIAGRAM J-02AC08  MAIN TURBINE TRIP LOGIC DIAGRAM J-12AC09  CONTROL LOGIC DIAGRAM J-12AE08  STEAM GENERATOR CHEMICAL INJEC VALVE J-02AL01  MOTOR DRIVEN AUX FEEDWATER PUMPS Rev. 27
 
WOLF CREEK TABLE 1.7-1 (SHEET 14)
DRAWING NUMBER                    TITLE J-02AL01A CONTROL LOGIC DIAGRAM AUX FW SYSTEM J-02AL01B CONTROL LOGIC DIAGRAM AUX FW SYSTEM J-02AL02  AUX FDWTR SYS SUPPLY FROM COND STORAGE TANK J-02AL02A CONTROL LOGIC DIAGRAM J-02AL03  AUX FEEDWATER PUMPS DISC CONTROL VALVES J-12AL04  AUX FEEDWATER PUMPS SUPPLY ESWS SYSTEM J-12AL04A CONTROL LOGIC DIAGRAM J-12AL05  AUX FEEDWATER SYS PUMPS SUCTION DISC PRESS ALAR J-02AL06  CONTROL LOGIC DIAGRAM J-02AL07  CONTROL LOGIC DIAGRAM J-02BB01  REACTOR COOLANT SYSTEM RCP THERM BAR ISO VALVE J-02BM01  STEAM GENERATOR BLOWDOWN LOGIC DIAG CTMNT VLVS J-02BM04  STU BEN BLOWDOWN ISOLATION VALVES J-02BN01  BORON REFUELING WATER STORAGE RWST HEATER VALVE J-02BN02  STEAM CONTROL VALVE J-12EC01  FUEL POOL COOLING & CLEANUP SYSTEM J-02EC02  FUEL POOL COOL CLEANUP SYS DIXC VLV HEAT EXCHAN J-12EC05  SYSTEM ALARMS J-12EF01A ESWS MOTOR OPERATED ISOLATION VALVES J-K2EF01A ESWS LOGIC DIAGRAM J-02EF01B ESWS MOTOR OPERATED ISOLATION VALVES J-K2EF01B ESWS LOGIC DIAGRAM J-02EF02  ESWS AIR COMPRESSORS ISOLATION VALVES J-K2EF02A ESWS ESW PUMPS LOGIC DIAGRAM J-K2EF02B ESWS ESW PUMP LUBE VALVE LOGIC DIAGRAM J-12EF03  ESWS CTMT AIR COOLERS ISOLATION VALVES J-K2EF03A ESWS SELF-CLEANING STRAINER LOGIC DIAGRAM J-K2EF03B ESWS SELF-CLEANING STRAINER LOGIC DIAGRAM J-12EF04  ESWS CTMT AIR COOLERS ISOL VALVE BYPASS VALVE Rev. 15
 
WOLF CREEK TABLE 1.7-1 (SHEET 15)
DRAWING NUMBER                    TITLE J-12EF05  ESWS MOTOR OPERATED ISOLATION VALVES J-K2EF06  ESWS ESW PUMP DISCHARGE AIR RELEASE VALVE LOGIC DIAGRAM J-02EF07  CONTROL LOGIC DIAGRAM STR SYS ESWS J-K2EF07  ESWS BACKPRESSURE CONTROL VALVE LOGIC DIAGRAM (VOID)
J-02EG01A COMPONENT COOLING WATER SYSTEM PUMPS J-02EG01B COMPONENT COOLING WATER SYSTEM PUMPS J-02EG01C CONTROL LOGIC DIAG COOL WTR SYSTEM J-12EG02  CCWS DEMIN WATER MAKEUP CCW SURGE TANK J-02EG04  CCW ESWS MAKE-UP SURGE TANK VENT J-02EG05A CCWS SUPPLY RETURN NUCLEAR AUX CMPNT SHEET A J-02EG05B CCWS SUPPLY RETURN NUCLEAR AUX CMPNT J-02EG05C CCWS SUP RTRN NUC AUX CMPNT VLV POSITION ALARM J-02EG06  CCWS HX'S DISC TEMP ALARM RHR HX'S FLOW ALARM J-02EG07  CCWS SUPPLY RHR HX'S J-12EG08A CCWS SUPPLY RETURN RADWASTE BLDG J-12EG08B CCWS SUPPLY RETURN RADWASTE BLDG J-12EG09  CCWS CONTAINMENT ISOLATION VALVE J-12EG10  CCWS INSIDE CTMT ISO VLV RTRN THRM BAR COOL CO J-02EG11  CCWS HEAT EXCHANGER OUTLET TEMP CONT J-12EG13A CONTROL LOGIC DIAGRAM J-02EG13B CONTROL LOGIC DIAGRAM J-12EG13C CONTROL LOGIC DIAGRAM J-12EG14  CONTROL LOGIC DIAGRAM J-02EJ01  RESIDUAL HEAT RMVL SYS NUC SAMP SYS VALVE J-02EJ03  RESIDUAL HEAT REMOVAL CONTROL LOGIC DIAGRAM J-02EJ04  CONTROL LOGIC DIAGRAM J-02EN01  CONTAINMENT SPRAY SYS CONTAIN SPRAY PUMPS J-02EN02  CONT SPRAY SYSTEM CTMT RECIRC SUMP ISOL VALVE Rev. 15
 
WOLF CREEK TABLE 1.7-1 (SHEET 16)
DRAWING NUMBER                    TITLE J-12EN03  CONT SPRAY SYSTEM SPRAY ADD TANK ISOL VLVS J-12EN04  CONT SPRAY SYSTEM SPRAY ADD TANK ALARMS J-02EN05  CONT SPRAY SYSTEM CTMT SPRAY NOZZLES ISOL VLV J-02FC18  AUX TURB AFP STEAMLINE WATER TRAP DRAIN J-02FC19  CONTROL LOGIC DIAGRAM J-K2GD01  ESSENTIAL SERVICE WATER PUMPHOUSE HVAC SUPPLY FANS LOGIC DIAGRAM J-K2GD02  ESSENTIAL SERVICE WATER PUMPHOUSE HVAC UNIT HEATER LOGIC DIAGRAM J-02GE08  TURB BLDG HVAC LOGIC DIAG COND AIR REMVL FILTER J-02GF03  MISC BLDG HVAC MAIN STM BLDG SUP EXH DAMPER J-12GF06  MISC BLDG HVAC TENDON GALL SUP RETURN DAMPER J-12GF06A CONTROL LOGIC DIAGRAM MISC.
BUILDING HVAC J-12GF07  MISC BLDG HVAC FDWTR PUMP ROOM COOLER FANS J-12GF09A CONTROL LOGIC DIAGRAM MISC.
BUILDING HVAC J-12GF09B CONTROL LOGIC DIAGRAM MISC.
BUILDING HVAC J-02GG04  FB HVAC ISOLATION DAMPERS J-02GG05  FUEL BLDG SPENT FUEL POOL ROOM COOLERS J-12GG06  FB HVAC FILTER UNITS INTAKE ISOL DAMPERS J-02GG07  FB HVAC FILTER UNITS AUX BLDG ISOLATION DAMPERS J-02GG08  FB HVAC MANUAL INITIATION OF FBIS J-02GG09  FB HVAC SPENT FUEL POOL DISC J-02GG10A FB HVAC EMERGENCY EXHAUST FAN J-02GG13  FB HVAC EMERGENCY EXHAUST CROSS CONNENC DAMPERS J-02GG14A FUEL BLDG HVAC SPENT FUEL POOL J-02GG14B FUEL BLDG HVAC SPENT FUEL POOL J-12GK01A LOGIC DIAG CONTROL ROOM FILTRATION FAN J-12GK01B FILTER ABSORBER UNIT SUPPLY J-12GK01C FILTER ABSORBER UNIT DISK DAMPER J-12GK01D CONTROL ROOM RECIRC DAMPERS CONTROL BLDG HVAC Rev. 15
 
WOLF CREEK TABLE 1.7-1 (SHEET 17)
DRAWING NUMBER                    TITLE J-12GK01E CONTROL BLDG HVAC CONTROL A/C UNIT DISCHARG DMP J-02GK02A CONTROL ROOM A/C UNIT FAN & TEMP J-02GK02B CONTROL ROOM A/C UNIT SUPPLY DAMPER J-02GK02C CONTROL BLDG HVAC CONT ROOM A/C UNIT DISC DAMPE J-02GK03A CONTROL BLDG HVAC FIRE ISO-LATION DAMPERS J-02GK07  MISC DAMPERS J-02GK09  ISO DAMPERS J-12GK10A CONTROL ROOM PRESS FAN J-12GK10B CONTROL ROOM PRESS SUPPLY DAMPER J-12GK10C CONT BLDG HVAC CONT ROOM PRESS SYS UNIT SUP DAM J-12GK11  CONTROL ROOM PRESS SYSTEM FIL UNIT J-12GK13  CLASS IE ELEC EQUIP A/C UNIT J-12GK15  CHLORINE ALAR J-12GK17A CONTROL LOGIC DIAGRAM CONTROL BLDG HVAC J-12GK17B CONTROL LOGIC DIAGRAM CONTROL BLDG HVAC J-02GK19  CONT BLDG HVAC RETRN DAMPERS RM FLTR ABSORB UNI J-02GK23  CONTROL BLDG HVAC FIRE ISO DAMPERS J-02GK25  CONTROL LOGIC DIAGRAM FOR CONTROL BLDG HVAC J-02GK26  CONTROL LOGIC DIAGRAM FOR CONTROL BLDG HVAC J-02GK27  CONTROL LOGIC DIAGRAM FOR CONTROL BLDG HVAC J-02GL01A AUX BLDG HVAC AIR SUPPLY UNIT J-02GL01B AUX BLDG HVAC AIR SUPPLY J-12GL03A AUX BLDG HVAC MISC ROOM COOLERS J-12GL03B AUX BLDG HVAC MISC ROOM COOLER J-12GL03C AUX BLDG HVAC MISC ROOM COOLER J-12GL11  AUX BLDG HVAC PENE RM COOLER FANS J-02GL13  AUX BLDG HVAC ISO DAMPERS J-02GL15  AUX BLDG HVAC FUEL BLDG HVAC DISCH ISO DAMPERS J-02GL21  AUX BLDG HVAC CCW PUMP ROOM EXH DAMPERS J-12GL23  AUX BLDG HVAC AUX/FUEL BLDG EXH FANS DISC Rev. 15
 
WOLF CREEK TABLE 1.7-1 (SHEET 18)
DRAWING NUMBER                    TITLE J-12GM01A DIESEL GENERATOR BLDG HVAC FAN
            & DAMPER J-02GM01B DIESEL GENERATOR BLDG EXH DAMPERS J-12GN01  CONTAINMENT COOLING J-12GN02A CONTROL LOGIC DIAGRAM CONTAINMENT COOLING SYS J-12GN02B CONTROL LOGIC DIAGRAM CONTAINMENT COOLING SYS J-02GN03A CONTAINMENT COOLING FANS J-02GS02  CTMT HYDROGEN CONTROL THERMAL HYDROGEN RECOM J-02GS03  CTMT HYDROGEN CONTROL SOLENOID ISO VLV J-02GS06  CTMT HYDROGEN CONTROL PURGE SUBSYS ISO VLV J-02GS08  CONTROL LOGIC DIAGRAM J-02GS09  CONTROL LOGIC DIAGRAM J-02GS10  CONTROL LOGIC DIAGRAM J-02GT03  CONTAINMENT PURGE SYSTEM VALVES J-02GT06  CONTAINMENT PURGE SYS ISO DAMPERS J-02GT10  CONTROL LOGIC DIAGRAM J-12JE01  EMERGENCY FUEL OIL TRANSFER PUMPS J-12JE02  CONTROL LOGIC DIAGRAM J-02KA02  COMPRESSED AIR SYS COMP AIR CONTAIN ISO VALVE J-02KA03  COMPRESSED AIR SYS HYDROGEN CS M/U AIR VALVE J-02KA08  CONTROL LOGIC DIAGRAM COMPRESSED AIR SYSTEM J-02KC08  CONTROL LOGIC DIAG FIRE PROTECTION SYSTEM J-12KJ02  CONTROL LOGIC DIAGRAM J-02KJ03  CONTROL LOGIC DIAGRAM J-12LF03  FLOOR & EQUIP DRAINS REACT BLDG SUMP PUMP ISO VL J-02LF04  FLOOR & EQUIP DRAINS REACT BLDG SUMP PMP ISO VLV J-02LF08  FLOOR & EQUIP DRAINS DISCHARGE VALVES J-02RP01  CONTROL LOGIC DIAGRAM J-02RP01A CONTROL LOGIC DIAGRAM J-12SA03  CONTROL LOGIC DIAGRAM J-12SA04  CONTROL LOGIC DIAGRAM J-12SA05  CONTROL LOGIC DIAGRAM J-02SJ01  NUC SYSTEM CON ISOLATION VALVES J-02SJ03  CONTROL LOGIC DIAGRAM Rev. 15
 
WOLF CREEK TABLE 1.7-1 (SHEET 19)
DRAWING NUMBER                  TITLE J-14001 CONTROL ROOM EQUIPMENT ARRANGEMENT J-14002 REACT COOL SUPRT SYSTEM CONSOLE RL001 & RL002 J-14003 REACTOR OPERATORS CONSOLE RL003
          & RL004 J-14004 TURB GEN & FEEDWATER CONSOLE RL005 & RL006 J-14005 SITE RELATED MAIN CONT BOARD RL013 & RL014 J-14006 STATION ELEC DIST MAIN CONT BOARD RL015 & RL016 J-14007 ENG SAFETY FEATURES MAIN CONT BOARD RL017 & RL018 J-14008 ENG SAFETY FEATURES MAIN CONT BOARD RL019 & RL020 J-14009 REACT AUX MAIN CONTROL BOARD RL021 & RL022 J-14010 TURBOGENERATORS & FDWTR MAIN CONT BD RL023 & RL024 J-14011 TURBOGENERATORS & FDWTR MAIN CONT BD RL025 & RL026 J-14013 END SECTION MAIN CONT BRD RL011
          & RL012 J-04014 MAIN CONTROL BOARD DETAILS J-04015 OPERATORS CONSOLE DETAILS J-14016 BILL OF MAT MAIN CONTROL PANEL J-05001 AUX CONTROL PANEL DWGS J-05002 AUX SHUTDOWN PANEL DETAILS J-05003 AUX SHUTDOWN PANEL RP 118 7 SHTS J-05021 LOCAL CONT PANEL RPO 68 MISC BOP INST PANEL J-15023 MISC BOP INST PANEL RP068 BILL OF MAT J-K5041 ESSENTIAL SERVICE WATER PANEL (EF 155 & EF 156) DRAWING J-K5042 ESSENTIAL SERVICE WATER PANEL (EF 155) BILL OF MATERIAL J-K5043 ESSENTIAL SERVICE WATER PANEL (EF 156) BILL OF MATERIAL J-16002 ANN WINDOW ARRGT RK016.018.020 J-16003 ANN WINDOW ARRGT RK022.024.026 Rev. 15
 
WOLF CREEK TABLE 1.7-1 (SHEET 20)
DRAWING NUMBER                              TITLE M-761-0066 Through 0104 and 0459, 0460, 0494 (8756d37*)        PROCESS CONTROL BLOCK DIAGRAM (1 THROUGH 42)
M-762-0001        NIS SOURCE RANGE FUNCTIONAL (5655D49*)        BLOCK DIAGRAM M-762-0002        NIS INTERMEDIATE RANGE (5655D50*)        FUNCTIONAL BLOCK DIAGRAM M-762-0417        NIS POWER RANGE FUNCTIONAL (9552D32*)        BLOCK DIAGRAM M-762-0032        NIS AUXILIARY CHANNELS (5655D52*)        FUNCTIONAL BLOCK DIAGRAM M-767-221 THROUGH 240 (8761D17*)        SAFEGUARDS TEST CABINET (1 THROUGH 20)
J-104-00347        INSTRUCTION MANUAL ESFAS/LSELS J-104-00437        SIGNAL FLOW BLOCK DIAGRAM FOR THE ENGINEERED SAFETY FEATURES ACTUATION SYSTEMS (SHEET 1)
J-104-00438        ISOLATION AND DISTRIBUTION OF ANALOG SIGNALS TO COMPUTER (ESFAS SHEET lA)
J-104-00439        CHANNEL IV BLOCK DIAGRAM (ESFAS SHEET 2)
J-104-00440        CHANNEL II BLOCK DIAGRAM (ESFAS SHEET 3)
J-104-00441        ATI BLOCK DIAGRAM (ESFAS SHEET 4)
J-104-00442        BISTABLE RACK - CHANNEL I (ESFAS SHEET 5)
J-104-00443        ISOLATION RACK - CHANNEL I (ESFAS SHEET 6)
J-104-00444        ACTUATION INPUTS - CHANNEL I (ESFAS SHEET 7)
J-104-00445        ACTUATION OUTPUTS AND STATUS INDICATIONS - CHANNEL I (ESFAS SHEET 8)
J-104-00446        ANNUNCIATOR/COMPUTERS OUTPUTS -
CHANNEL I (ESFAS SHEET 9)
* Drawings suplied by NSSS Vendor.
Rev. 15
 
WOLF CREEK TABLE 1.7-1 (SHEET 21)
DRAWING NUMBER                      TITLE J-104-00447 ANALOG SIGNALS - CHANNEL II (ESFAS SHEET 10)
J-104-00448 ISOLATION RACK LOGIC SIGNALS -
CHANNEL II (ESFAS SHEET 11)
J-104-00449 RELAY OUTPUTS - CHANNEL II (ESFAS SHEET 12)
J-104-00450 BISTABLE RACK - CHANNEL IV (ESFAS SHEET 13)
J-104-00451 ISOLATION RACK - CHANNEL IV (ESFAS SHEET 14)
J-104-00452 ACTUATION INPUTS - CHANNEL IV (ESFAS SHEET 15)
J-104-00453 ACTUATION OUTPUTS AND STATUS INDICATIONS - CHANNEL IV (ESFAS SHEET 16)
J-104-00454 ANNUNCIATOR/COMPUTER OUTPUTS -
CHANNEL IV (ESFAS SHEET 17)
J-104-00455 ATI MODULE A (ESFAS SHEET 18)
J-104-00456 ATI MODULE B (ESFAS SHEET 19)
J-200-00029 TURB SUPERVISORY MAIN CONT BRD RL027 & RL028 Rev. 15
 
WOLF CREEK TABLE 1.7-2 PIPING AND INSTRUMENTATION DIAGRAMS Drawing  Figure Number  Number Sheet  Title M-120101 1.1-1    1    Symbols and Legend for System Flow and Piping
                        & Instrumentation Diagrams M-120102 1.1-1    2    Symbols and Legend for System Flow and Piping
                        & Instrumentation Diagrams M-120103 1.1-1    3    Symbols and Legend for System Flow and Piping
                        & Instrumentation Diagrams M-120104 1.1-1    4    Symbols and Legend for System Flow and Piping
                        & Instrumentation Diagrams M-12AB01 10.3-1  1    Main Steam System M-12AB02 10.3-1  2    Main Steam System M-12AB03 10.3-1  3    Main Steam System M-12AC01 10.2-1  1    Main Turbine M-12AC02 10.2-1  2    Main Turbine M-12AC03 10.2-1  3    Main Turbine M-12AC04 10.2-1  4    Main Turbine M-12AD01 10.4-2  1    Condensate System M-12AD02 10.4-2  2    Condensate System M-12AD03 10.4-2  3    Condensate System M-12AD04 10.4-2  4    Condensate System M-12AD05 10.4-2  5    Condensate System M-12AD06 10.4-2  6    Condensate System M-12AE01 10.4-6  1    Feedwater System M-12AE02 10.4-6  2    Feedwater System M-12AF01 10.4-6  3    Feedwater Heater Extraction Drains & Vents M-12AF02 10.4-6  4    Feedwater Heater Extraction Drains & Vents M-12AF03 10.4-6  5    Feedwater Heater Extraction Drains & Vents M-12AF04 10.4-6  6    Feedwater Heater Extraction Drains & Vents M-12AK01 10.4-5  1    Condensate Demineralizer System M-12AK02 10.4-5  2    Condensate Demineralizer System M-12AK03 10.4-5  3    Condensate Demineralizer System Rev. 27
 
WOLF CREEK TABLE 1.7-2 (SHEET 2)
Drawing  Figure Number  Number Sheet  Title M-12AL01 10.4-9        Auxiliary Feedwater System M-12AN01 9.2-16        Demineralized Water Storage and Transfer System M-12AP01 9.2-23        Condensate Storage and Transfer System M-12AQ01 10.4-7  1    Condensate Chemical Addition System M-12AQ02 10.4-7  2    Feedwater Chemical Addition System M-12BB01 5.1-1  1    Reactor Coolant System M-12BB02 5.1-1  2    Reactor Coolant System M-12BB03 5.1-1  3    Reactor Coolant System M-12BB04 5.1-1  4    Reactor Coolant System M-12BG01 9.3-8  1    Chemical and Volume Control System M-12BG02 9.3-8  2    Chemical and Volume Control System M-12BG03 9.3-8  3    Chemical and Volume Control System M-12BG04 9.3-8  4    Chemical and Volume Control System M-12BG05 9.3-8  5    Chemical and Volume Control System M-12BL01 9.2-13        Reactor Makeup Water System M-12BM01 10.4-8  1    Steam Generator Blowdown System M-12BM02 10.4-8  2    Steam Generator Blowdown System M-12BM03 10.4-8  3    Steam Generator Blowdown System M-12BM04 10.4-8  4    Steam Generator Blowdown System M-12BM05 10.4-8  5    Steam Generator Blowdown System M-12BN01 6.3-1  1    Borated Refueling Water Storage System M-12CA01 10.4-4        Steam Seal System M-12CF01 10.2-1  5    Lube Oil Storage, Transfer and Purification System M-12CF02 10.2-1  6    Lube Oil Storage, Transfer and Purification System M-12CG01 10.4-3        Condenser Air Removal M-12CH01 10.2-1  7    Main Turbine Control Oil System M-12CH02 10.2-1  8    Main Turbine Control Oil System Rev. 21
 
WOLF CREEK TABLE 1.7-2 (SHEET 3)
Drawing  Figure Number  Number  Sheet  Title M-12DA01 10.4-1  1    Circulating Water & Water Box Drains System M-12DA02 10.4-1  3    Circulating Water & Water Box Venting System M-12EA01 9.2-1    1    Service Water System M-12EA02 9.2-1    2    Service Water System M-12EB01 9.2-14        Closed Cooling Water System M-12EC01 9.1-3    1    Fuel Pool Cooling and Cleanup System M-12EC02 9.1-3    2    Fuel Pool Cooling and Cleanup System M-12EF01 9.2-2    1    Essential Service Water System M-12EF02 9.2-2    2    Essential Service Water System M-K2EF01 9.2-2    3    Essential Service Water System M-K2EF03 9.2-2    4    Essential Service Water System M-12EG01 9.2-15  1    Component Cooling Water System M-12EG02 9.2-15  2    Component Cooling Water System M-12EG03 9.2-15  3    Component Cooling Water System M-12EJ01 5.4-7          Residual Heat Removal System M-12EM01 6.3-1    2    High Pressure Coolant Injection System M-12EM02 6.3-1    3    High Pressure Coolant Injection System M-12EN01 6.2.2-1        Containment Spray System M-12EP01 6.3-1    4    Accumulator Safety Injection M-12FA01 9.5.9-1  1    Auxiliary Boiler System M-12FB01 9.5.9-1  2    Auxiliary Steam System M-12FB02 9.5.9-1  3    Auxiliary Steam System M-12FC02 10.4-10        Auxiliary Feedwater Pump Turbine M-12FC03 10.4-6  7    S.G.F.P. Turbine "A" M-12FC04 10.4-6  8    S.G.F.P. Turbine "B" M-12FE01 9.5.9-1  4    Auxiliary Steam Chemical Addition System M-12GA01 9.4-9    1    Plant Heating System M-12GA02 9.4-9    2    Plant Heating System M-12GB01 9.4-10        Central Chilled Water System M-K2GD01 9.4-8    1    Essential Service Water Pump House HVAC Rev. 17
 
WOLF CREEK TABLE 1.7-2 (SHEET 4)
Drawing  Figure Number  Number  Sheet  Title M-12GE01 9.4-4    1    Turbine Building HVAC M-12GE02 9.4-4    2    Turbine Building HVAC M-12GE03 9.4-4    3    Turbine Building HVAC M-12GE04 9.4-4    4    Turbine Building HVAC M-12GF01 9.4-3    1    Miscellaneous Buildings HVAC M-12GF02 9.4-3    2    Miscellaneous Buildings HVAC M-12GG01 9.4-2    1    Fuel Building HVAC M-12GG02 9.4-2    2    Fuel Building HVAC M-12GH01 9.4-5    1    Radwaste Building HVAC M-12GH02 9.4-5    2    Radwaste Building HVAC M-12GK01 9.4-1    1    Control Building HVAC M-12GK02 9.4-1    2    Control Building HVAC M-12GK03 9.4-1    3    Control Building HVAC M-12GK04 9.4-1    4    Control Building HVAC M-12GL01 9.4-3    5    Auxiliary Building HVAC M-12GL02 9.4-3    4    Auxiliary Building HVAC M-12GL03 9.4-3    3    Auxiliary Building HVAC M-12GM01 9.4-7          Diesel Generators Building HVAC M-12GN01 9.4-6    1    Containment Cooling System M-12GN02 9.4-6    2    Containment Cooling System M-12GP01 6.2.6-1        Containment Integrated Leak Rate Test M-12GR01 9.4-6    3    Containment Atmospheric Control System M-12GS01 6.2.5-1        Containment Hydrogen Control System M-12GT01 9.4-6    4    Containment Purge System HVAC M-12HA01 11.3-1  1    Gaseous Radwaste System M-12HA02 11.3-1  2    Gaseous Radwaste System M-12HA03 11.3-1  3    Gaseous Radwaste System M-12HB01 11.2-1  1    Liquid Radwaste System M-12HB02 11.2-1  2    Liquid Radwaste System M-12HB03 11.2-1  3    Liquid Radwaste System M-12HB04 11.2-1  4    Liquid Radwaste System M-12HC01 11.4-1  1    Solid Radwaste System M-12HC02 11.4-1  2    Solid Radwaste System M-12HC03 11.4-1  3    Solid Radwaste System M-12HC04 11.4-1  4    Solid Radwaste System M-12HD01 12.3-4        Decontamination System M-12HE01 9.3-11  1    Boron Recycle System M-12HE02 9.3-11  2    Boron Recycle System M-12HE03 9.3-11  3    Boron Recycle System Rev. 17
 
WOLF CREEK TABLE 1.7-2 (SHEET 5)
Drawing  Figure Number  Number  Sheet  Title M-12HF01 10.4-12  1    Secondary Liquid Waste System M-12HF02 10.4-12  2    Secondary Liquid Waste System M-12HF03 10.4-12  3    Secondary Liquid Waste System M-12HF04 10.4-12  4    Secondary Liquid Waste System M-12JE01 9.5.4-1        Emergency Fuel Oil System M-12KA01 9.3-1    1    Compressed Air System M-12KA02 9.3-1    2    Compressed Air System (Service Air)
M-12KA03 9.3-1    3    Instrument Air System M-12KA04 9.3-1    4    Instrument Air System M-12KA05 9.3-1    5    Compressed Air System M-12KA06 9.3-1    6    Compressed Air System M-12KA07 9.3-1    7    Compressed Air System M-12KB01 9.5.10-1  1    Breathing Air System M-12KB02 9.5.10-1  2    Breathing Air System M-12KB03 9.5.10-1  3    Breathing Air System M-12KC01 9.5.1-1  1    Fire Protection System M-12KC02 9.5.1-1  2    Fire Protection System M-12KC03 9.5.1-1  3    Fire Protection System M-12KC04 9.5.1-1  4    Fire Protection (Halon)
System M-12KC05 9.5.1-1  5    Fire Protection System M-12KC06 9.5.1-1  6    Fire Protection (Halon)
System M-12KC07 9.5.1-1  7    Fire Protection (Halon)
System M-12KD01 9.2-17    1    Domestic Water System M-12KD02 9.2-17    2    Domestic Water System M-12KH01 9.3-9    1    Service Gas System M-12KH02 9.3-9    2    Service Gas System M-12KJ01 9.5.5-1  1    Standby Diesel Generator "A" Cooling Water System M-12KJ04 9.5.5-1  2    Standby Diesel Generator "B" Cooling Water System M-12KJ02 9.5.6-1  1    SDG "A" Intake, Exh., F.O.
                        & Starting Air System M-12KJ05 9.5.6-1  2    SDG "B" Intake, Exh., F.O.
                        & Starting Air System M-12KJ03 9.5.7-1  1    Standby Diesel Generator "A" Lube Oil System M-12KJ06 9.5.7-1  2    Standby Diesel Generator "B" Lube Oil System Rev. 17
 
WOLF CREEK TABLE 1.7-2 (SHEET 6)
Drawing  Figure Number  Number Sheet Title M-12LA01 9.3-5    1  Sanitary Lift Station &
Turb. Bldg. Sanitary Drainage System M-12LA02 9.3-5    2  Comm. Corridor & Control Bldg. Sanitary Drainage System M-12LD01 9.3-5    3  Chemical and Detergent Waste M-12LE01 9.3-5    4  Turb. Bldg. & Aux. Feedwater Pump Rooms Oily Waste System M-12LE02 9.3-5    5  Control & Diesel Gen. Bldg.
Oily Waste System M-12LE03 9.3-5    6  Turb. Bldg. & Aux. Boiler Room Oily Waste System M-12LE04 9.3-5    7  Tendon Access Gallery &
Turb. Bldg. Oily Waste System M-12LF01 9.3-5    8  Aux. Bldg. Floor and Equip-ment Drain System M-12LF02 9.3-5    9  Aux. Bldg. Floor and Equip-ment Drain System M-12LF03 9.3-5    10  Aux. Bldg. Floor and Equip-ment Drain System M-12LF04 9.3-5    11  Aux. Bldg. Floor and Equip-ment Drain System M-12LF05 9.3-5    12  Aux. Bldg. Floor and Equip-ment Drain System M-12LF06 9.3-5    13  Radwaste & Fuel Bldgs.
FED System M-12LF07 9.3-5    14  Radwaste Bldg. FED System M-12LF08 9.3-5    15  Control and Fuel Bldgs.
FED System M-12LF09 9.3-5    16  Reactor Bldg. & Hot Machine Shop FED System M-12LF10 9.3-5    17  Radwaste Bldg. and Tunnel FED System M-12RM01 9.3-4    1  Process Sampling System M-12RM02 9.3-4    2  Process Sampling System M-12RM03 9.3-4    3  Process Sampling System M-12SJ01 9.3-2    1  Nuclear Sampling System Primary Sampling System M-12SJ02 9.3-3        Nuclear Sampling System Radwaste Sampling System M-12SJ03 9.3-2    2  Nuclear Sampling System Primary Sampling System Rev. 27
 
WOLF CREEK TABLE 1.7-2 (SHEET 7)
Drawing  Figure Number  Number  Sheet  Title M-12SJ04 18.2-15        Nuclear Sampling System M-12WT01 9.2-24  1    Wastewater Treatment Facility M-12WT03 9.2-25  1    Wastewater Treatment Facility M-12VW01 9.4-11        Wastewater Treatment Facility HVAC System M-0021  10.4-1  2    Circulating Water System M-0022  9.2-1    3    Service Water System M-0023  9.5-1    1    Fire Protection System (Site)
M-0023  9.5-1    2    Fire Protection System (Site)
M-0023  9.5-1    3    Fire Protection System (Site)
M-0023  9.5-1    4    Fire Protection System (Site)
M-0025  9.2-5    1    Demineralized Water Makeup System M-0025  9.2-5    2    Demineralized Water Makeup System M-0025  9.2-5    3    Demineralized Water Makeup System M-0025  9.2-5    4    Demineralized Water Makeup System M-0025  9.2-5    4A    Demineralized Water Makeup System M-0051  9.5-2          Outdoor Piping, Key Plans &
General Notes Rev. 21
 
WOLF CREEK TABLE 1.7-3 ADDITIONAL CONTROLLED DRAWINGS USED IN THE USAR Drawing      Figure Number        Number  Sheet  Title A-0503        9.2-5A          Potable Water System 10466-A-1701  12.3-2    1    Radiation Zones For Normal Operation 10466-A-1702  12.3-2    2    Radiation Zones For Normal Operation 10466-A-1703  12.3-2    3    Radiation Zones For Normal Operation 10466-A-1704  12.3-2    4    Radiation Zones For Normal Operation 10466-A-1705  12.3-2    5    Radiation Zones For Normal Operation 10466-A-1706  12.3-2    6    Radiation Zones for Normal Operation 10466-A-1801  9.5.1-2  1    Fire Delineation Floor Plan EL. 1974'-0" 10466-A-1802  9.5.1-2  2    Fire Delineation Floor Plan EL. 2000'-0" 10466-A-1803  9.5.1-2  3    Fire Delineation Floor Plan EL. 2026'-0" 10466-A-1804  9.5.1-2  4    Fire Delineation Floor Plan EL. 2047'-0" 8025-C-KG1202 1.2-44          Site Plan C-0L2931      9.3-7    1    Reactor Building Stainless Steal Liner Plate Reactor Refueling Canal C-1L6111      9.3-7    2    Reactor Building Stainless Steel Liner Plate Reactor Refueling Canal C-03BB53      5.4-21          Hot and Cold Leg Lateral Restraints E-1L9901      9.5.3-1        Lighting Distribution Riser Diagram E-1L9903      9.5.2-2        Public Address System Riser Diagram E-11005      8.3-2          List of Loads Supplied by the Emergency Diesel Generator E-1001        8.3-1    3    Single Line Diagram Site Area Loads E-K1001      8.3-1    2    Single Line Diagram Essential Service Water System E-11001      8.3-1    1    Main Single Line Diagram E-11010      8.3-6    1    DC Main Single Line Diagram E-11010A      8.3-7          DC Main Single Line Diagram (PK03 & PK04 Bus)
E-12KJ01      8.3-5          Standby Generator Engine and Governor Control Logic Diagram Rev. 21
 
WOLF CREEK TABLE 1.7-3 (Sheet 2)
Drawing    Figure Number      Number  Sheet  Title E-12NE01    8.3-3          Logic Diagram Standby Generator Excitation Control E-12NE02    8.3-4          Logic Diagram Standby Generator System Protection E-14QE01    9.5.2-1        Telephone System Riser Diagram J-104-00390 7.3-1    2      Logic Diagram Engineered Safety Features Actuation System (BOP)
KD-7496    8.2-4          WCGS Electrical One-Line Diagram KD-7750    8.2-3          Wolf Creek Substation General Plan M-0004      10.4-1  4      Circulating Water Screenhouse Planview M-0005      10.4-1  5      Circulating Water Screenhouse Section View M-1G001    1.2-1          Peninsular Plant Arrangement Standard Power System & Structure Interface M-1G010    1.2-2          Equipment Location Radwaste Building Plan EL 1976'-0" M-1G011    1.2-3          Equipment Location Radwaste Building Plan EL 2000'-0" M-0G012    1.2-4          Equipment Location Radwaste Building Plan EL 2022'-0" M-1G013    1.2-5          Equipment Location Radwaste Building Plan EL 2031'-6" and Roof Plan M-1G014    1.2-6          Equipment Location Radwaste Building Sections A & B M-1G015    1.2-7          Equipment Location Building Sections C & E M-1G016    1.2-8          Equipment Location Building Sections D & F M-1G020    1.2-9          Equipment Location Reactor and Auxiliary Building Plan Basement EL. 1974'-0" M-1G021    1.2-10          Equipment Location Auxiliary Building Partial Plan EL. 1988'-0" and 2013'-6" M-1G022    1.2-11          Equipment Location Reactor and Auxiliary Building Plan Ground Floor Elevation 2000'-0" M-1G023    1.2-12          Equipment Location Reactor and Auxiliary Building Plan EL. 2026'-0" M-1G024    1.2-13          Equipment Location Reactor and Auxiliary Building Plan Operating Floor Elevation 2047'-6" Rev. 21
 
WOLF CREEK TABLE 1.7-3 (Sheet 3)
Drawing Figure Number  Number Sheet    Title M-1G025 1.2-14          Equipment Location Reactor and Auxiliary Building Plan Elevation 2068'-8" M-1G026 1.2-15          Equipment Location Reactor and Auxiliary Building Section A M-1G027 1.2-16          Equipment Location Reactor and Auxiliary Building Section B M-1G028 1.2-17          Equipment Location Reactor and Auxiliary Building Section C M-1G029 1.2-18          Equipment Location Reactor and Auxiliary Building Section D M-1G030 1.2-19          Equipment Location Auxiliary Building Sections E, F & G M-1G040 1.2-20          Equipment Location Fuel Building Plan Elevation 2000'-0", 2026'-0" and 2047'-6" M-1G041 1.2-21          Equipment Location Fuel Building Sections A, B & C M-1G042 1.2-22          Equipment Location Fuel Building Sections D, E & F M-1G050 1.2-23          Equipment Location Control Building & Communication Corridor Plan Elevation 1974'-0" & 1984'-0" M-1G051 1.2-24          Equipment Location Control and Diesel Generator Buildings &
Communication Corridor Plan Elevation 2000'-0" and 2016'-0" M-1G052 1.2-25          Equipment Location Control and Diesel Generator Buildings &
Communication Corridor Plan Elevation 2032'-0" & 2047'-6" M-1G053 1.2-26          Equipment Location Control and Diesel Generator Buildings &
Communication Corridor Plan Elevation 2061'-6", 2066'-0" &
2073'-6" & Section D M-1G054 1.2-27          Equipment Location Control and Diesel Generator Building Communication Corridor Section A M-1G055 1.2-28          Equipment Location Control and Diesel Generator Building Sections B & C M-1G060 1.2-29          Equipment Location Turbine Building Condenser Pit Plan Elevation 1983'-0" M-1G061 1.2-30          Equipment Location Turbine Building Ground Floor Plan Elevation 2000'-0" Rev. 17
 
WOLF CREEK TABLE 1.7-3 (Sheet 4)
Drawing    Figure Number      Number  Sheet    Title M-1G062    1.2-31            Equipment Location Partial Plan Elevation 2015'-4" M-1G063    1.2-32            Equipment Location Turbine Building Mezzanine Floor Plan Elevation 2033'-0" M-1G064    1.2-33            Equipment Location Turbine Building Operating Floor Plan Elevation 2065'-0" M-1G065    1.2-34            Equipment Location Turbine Building Section "A" M-1G066    1.2-35            Equipment Location Turbine Building Section "B" M-1G067    1.2-36            Equipment Location Turbine Building Section "C" M-0G068    1.2-37            Equipment Location Turbine Building Section "D" M-1G069    1.2-38            Equipment Location Turbine Building Section "E" M-0G070    1.2-39            Equipment Location Turbine Building Section "F" M-0G071    1.2-40            Equipment Location Turbine Building Section "G" M-1G072    1.2-41            Equipment Location Turbine Building Section "H" M-0G073    1.2-42            Turbine Component Laydown Area Elevation 2065'-0" M-13EN03    6.2.2-2  1        Containment Spray System Reactor Building A & B Trains M-13EN04    6.2.2-2  2        Containment Spray System Reactor Building A & B Trains M-13EN05    6.2.2-2  3        Containment Spray System Reactor Building A & B Trains M-KG080    9.2-3            ESWS Pumphouse    Equipment Location - Plan M-KG081    9.2-4            ESWS Pumphouse    Equipment Location - Sections M-744-00018 7.2-1    1        Functional Diagrams Index and Symbols M-744-00019 7.2-1    2        Functional Diagrams (Reactor Trip Signals)
M-744-00020 7.2-1    3        Functional Diagrams (Nuclear Instrumentation and Manual Trip Signals)
M-744-00021 7.2-1    4        Functional Diagrams (Nuclear Instrumentation Permissives and Blocks)
M-744-00022 7.2-1    5        Functional Diagrams (Primary Coolant System Trip Signals)
M-744-00023 7.2-1    6        Functional Diagrams (Pressurizer Trip Signals)
M-744-00024 7.2-1    7        Functional Diagrams (Steam Generator Trip Signals)
Rev. 21
 
WOLF CREEK TABLE 1.7-3 (Sheet 5)
Drawing    Figure Number      Number Sheet  Title M-744-00025 7.2-1  8      Functional Diagrams (Safe-guards Activation Signals)
M-744-00026 7.2-1  9      Functional Diagrams (Rod Controls and Rod Blocks)
M-744-00027 7.2-1  10    Functional Diagrams (Steam Dump Control)
M-744-00028 7.2-1  11    Functional Diagrams (Pressur-izer Pressure & Level Control)
M-744-00029 7.2-1  12    Functional Diagrams (Pressur-izer Heater Control)
M-744-00030 7.2-1  13    Functional Diagrams (Feedwater Control and Isolation)
M-744-00031 7.2-1  14    Functional Diagrams (Feedwater Control and Isolation)
M-744-00032 7.2-1  15    Functional Diagrams (Auxiliary Feedwater Pumps Startup)
M-744-00033 7.2-1  16    Functional Diagram (Turbine Trip Runbacks and Other Signals)
M-744-00039 7.2-1  17    Functional Diagram (Pressurizer Pressure Relief)
M-744-00040 7.2-1  18    Functional Diagram (Pressurizer Pressure Relief)
SK-C-250    3B-2          Plan and Elevation View of Main Steam/Main Feedwater Isolation Valve Compartment S-0172      2.4-3  2      Grading Plan Switchyard Area S-0186      2.4-3  3      Drainage Plan Plant Area S-0189-1    2.4-3  4A    Manhole, Pipe & Culvert Schedule S-0189-2    2.4-3  4B    Manhole, Pipe & Culvert Schedule S-0189-3    2.4-3  4C    Manhole, Pipe & Culvert Schedule S-0189-4    2.4-3  4D    Manhole, Pipe & Culvert Schedule S-0191      2.4-3  5      Manhole & Pipe Details S-0296      2.4-3  6      Manhole & Pipe Details S-0297      2.4-3  7      Plant Area Roadway Grading &
Drainage Rev. 21
 
WOLF CREEK TABLE 1.7-4 ERECTED SCAFFOLD EXPECTED TO BE IN PLACE THROUGHOUT NEXT REQUIRED USAR UPDATE CYCLE This Table has been deleted Rev. 30
 
WOLF CREEK TABLE 1.7-4 (sheet 2)
ERECTED SCAFFOLD EXPECTED TO BE IN PLACE THROUGHOUT NEXT REQUIRED USAR UPDATE CYCLE This Table has been deleted Rev. 30
 
WOLF CREEK TABLE 1.7-4 (sheet 3)
ERECTED SCAFFOLD EXPECTED TO BE IN PLACE THROUGHOUT NEXT REQUIRED USAR UPDATE CYCLE This Table has been deleted Rev. 30
 
WOLF CREEK TABLE 1.7-4 (sheet 4)
ERECTED SCAFFOLD EXPECTED TO BE IN PLACE THROUGHOUT NEXT REQUIRED USAR UPDATE CYCLE This Table has been deleted Rev. 30
 
WOLF CREEK TABLE 1.7-4 (sheet 5)
ERECTED SCAFFOLD EXPECTED TO BE IN PLACE THROUGHOUT NEXT REQUIRED USAR UPDATE CYCLE This Table has been deleted Rev.30
 
WOLF CREEK 1.8  CONFORMANCE TO NRC REGULATORY GUIDES A discussion of the extent to which WCGS complies with each of the NRC Division 1 Regulatory Guides is provided in Appendix 3A. Appendix 3A gives a brief statement of WCGS compliance and refers to the most appropriate section of the USAR for the complete description of how the design complies with the regulatory recommendations.
1.8-1                  Rev. 0
 
WOLF CREEK 1.9  NRC REGULATORY REQUIREMENTS REVIEW COMMITTEE CATEGORY 2, 3, AND 4 MATTERS The Office of Nuclear Reactor Regulation (NRR) established a Regulatory Requirements Review Committee (RRRC) which reviewed proposed changes to the regulatory requirements issued by the staff and recommended a course of action to the Office of NRR. The course of action includes an implementation schedule. The Director's approval was then used by the NRR staff as review guidance on individual licensing matters.
The RRRC developed a categorization nomenclature to aid in the uniform implementation of new and revised regulatory staff concerns. The system included four categories (1 through 4) which correspond to the evaluation by the RRRC of the need for applying the regulatory concerns to new and ongoing license applications. The four categories are defined as follows:
Category 1:    Matters whose applicability is to be applied to applications in accordance with the implementation section of the published guide. The RRRC considers it necessary to forward fit (on new applications) the requirements of these matters.
Category 2:    A new position whose applicability is to be deter-mined on a case-by-case basis. The NRC staff will give further consideration to the need for backfit-ting certain identified items of the regulatory concerns.
Category 3:    Positions to which the NRC staff considers confor-mance necessary, either by direct implementation or by implementation of an acceptable alternative.
These positions could be the cause of backfitting if an acceptable alternative is not available.
Category 4:    Positions of concern to the NRC staff which have not been reviewed by the RRRC and subsequently categorized as Category 1, 2, or 3. Since these items are of concern to the NRC staff, for review purposes, they are to be considered on the same basis as Category 2, potential for backfitting certain identified regulatory concerns.
1.9-1                          Rev. 0
 
WOLF CREEK The Office of NRR, by {{letter dated|date=November 21, 1978|text=letter dated November 21, 1978}}, transmitted a list of Category 2, 3, and 4 matters for consideration in preparing the WCGS FSAR. A discussion of how Wolf Creek complies with each of the listed matters is contained in Tables 1.9-1 through 1.9-4. The RRRC Category Designation columns in the tables correspond to those contained in the {{letter dated|date=November 21, 1978|text=November 21, 1978 letter}}.
Table 1.9 Lists all Category 2, 3, and 4 regulatory guides and references the location in which the regulatory guides are addressed.
Table 1.9 Lists all Category 2, 3, and 4 branch technical positions (BTPs),
provides remarks on the extent to which the recommendations of the BTPs are met, and references the location of more complete discussions of the RRRC matter.
Tables 1.9-3 and 1.9 Address the Category 4 SRP criteria and other Category 4 positions, respectively.
1.9-2                        Rev. 0
 
WOLF CREEK TABLE 1.9-1 CATEGORY 2, 3, AND 4 REGULATORY GUIDES*
(Historical Information)
Regulatory Guide            RRRC Category Number      Revision            2  3  4 1.12          2                        X 1.13          1                        X 1.14          1                        X 1.27          2                X 1.52          2                X 1.56          1                    X 1.59          2                X 1.63          2                X 1.68.2        1                    X 1.75          2                        X 1.76          0                        X 1.79          1                        X 1.80          0                        X 1.82          0                        X 1.83          1                        X 1.89          1                        X 1.91          1                X 1.93          0                        X 1.97          2                    X 1.99          2                    X 1.101          1,2                  X 1.102          1                X 1.104          0                        X 1.105          1                X 1.108          1                X 1.114          1                    X 1.115          1                X 1.117          1                X 1.121          0                    X 1.124          1                X 1.127          1                    X 1.130          1                X 1.137          0                X 1.141          0                X 8.8            2                X
*All regulatory guides are addressed in Appendix 3A with the exception of Regulatory Guide 8.8, which is addressed in Section 12.1.
Rev. 32
 
WOLF CREEK TABLE 1.9-2 CATEGORY 2, 3, AND 4 BRANCH TECHNICAL POSITIONS Branch Technical                      RRRC Position          Title          Cat.      Remarks ASB 9.5-1,      Guidelines for Fire    2    The recommendations of Rev. 1          Protection for Nu-          this BTP are met to clear Power Plants          the extent described in Section 9.5.1.
MTEB 5-7        Material Selection    2    The recommendations of and Processing              this BTP are not ap-Guidelines for BWR          plicable to the WCGS Coolant Pressure            (PWR) design.
Boundary RSB 5-1,        Design Requirements    3    The recommendations of Rev. 1          of the Residual Heat        this BTP are met to the Removal System              extent described in Sec-tions 5.2.2 and 7.6.6.
RSB 5-2          Overpressurization    3    The recommendations of Protection of Pres-        this BTP are met to the surized Water Reactors      extent described in Sec-While Operating at Low      tions 5.2.2 and 7.6.6.
Temperatures MTEB 5-3        Monitoring of Sec-  4.B.1  The recommendations of ondary Side Water          this BTP are met. Refer Chemistry in PWR            to Sections 9.3.2 and Steam Generators            10.3.5.
CSB 6-1          Minimum Containment 4.B.2  The recommendations of Pressure Model for          this BTP are met. Refer PWR ECCS Performance        to Sections 6.2.1 and Evaluation                  15.6.5.
CSB 6-2          Control of Combus-  4.B.3  The recommendations of tible Gas Concentra-        this BTP are met. Refer tions in Containment        to Section 6.2.5.
Following a Loss-of-Coolant Accident Rev. 0
 
WOLF CREEK TABLE 1.9-2 (Sheet 2)
Branch Technical                      RRRC Position          Title          Cat.      Remarks CSB 6-3          Determination of    4.B.4  The recommendations of Bypass Leakage Paths        BTP are not applicable in Dual Containment        to the WCGS design, Plants                      since there is no dual containment.
CSB 6-4          Containment Purging  4.B.5  The recommendations of During Normal Plant        this BTP are met to the Operations                  extent described in Table 9.4-13.
ASB 9.1          Overhead Handling    4.B.6  The recommendations of Systems for Nu-            this BTP are met. No clear Power Plants          critical loads are handled. Refer to Section 9.1.4.
ASB 10.1        Design Guidelines    4.B.7  The recommendations of for Auxiliary Feed-        this BTP are met. Refer water System Pump          to Section 10.4.9.
Drive and Power Sup-ply Diversity for PWR Plants Rev. 0
 
WOLF CREEK TABLE 1.9-3 CATEGORY 4 SRP CRITERIA SRP      RRRC Section  Cateqory          Title 3.5.3    4.B.8    Procedures for Composite The recommendations of this (Par II.1.C)      Section Local Damage Pre- SRP are met to the extent diction                    described in Section 3.5.3.
3.7.1    4.B.9    Development of Design      The recommendations of (Par. II.2)        Time History for Soil-      this SRP are met to the Structure Interaction      extent described in Analysis                    Section 3.7(B).1.
3.7.2    4.B.10  Procedures for Seismic      The recommendations of (Par. II)          System Analysis            this SRP are met. Refer to Sections 3.7(B).2 and 3.7(N).2.
3.7.3    4.B.11  Procedures for Seismic      The recommendations of (Par. II)          Subsystem Analysis          this SRP are met. Refer to Sections 3.7(B).3 and 3.7(N).3 3.8.1    4.B.12  Design and Construction    The design of the con-(Par. II)          of Concrete Containments    tainment structure is described in Section 3.8.1. The load combi-nations used meet or exceed ACI 359/SRP criteria.
3.8.2    4.B.13  Design and Construction    The recommendations of (Par. II)          of Steel Containments      this SRP are not applic-able to WCGS.
3.8.3    4.B.14  Structural Design Cri-      The design meets or ex-(Par. II)          teria for Category I        ceeds the load combina-Structures Inside Con-      tions of ACI 359/SRP tainment                    criteria. Refer to Sections 3.8.3 and 5.4, respectively, for dis-cussion of the Bechtel and Westinghouse com-ponent supports.
3.8.4    4.B.15  Structural Design Cri-      The design meets or ex-(Par. II)          teria for Other Seismic    ceeds the load combina-Category I Structures      tions of ACI 359/SRP criteria. Refer to Section 3.8.4.
Rev. 0
 
WOLF CREEK TABLE 1.9-3 (Sheet 2)
SRP      RRRC Section  Category        Title                        Remarks 3.8.5    4.B.16  Structural Design Cri-        The design meets or exceeds (Par. II)          teria for Foundations          the load combinations of ACI 359/SRP criteria.
Refer to Section 3.8.5.
The safety factors for sliding are discussed in Section 3.8.5.5.
3.7      4.B.17  Seismic Design Require-        The recommendations of 11.2              ments for Radwaste Systems    this SRP are met as 11.3              and their Housing Struc-      described in Appendix 11.4              tures (SRP Section 11.2,      3A in the response BTP ETSB 11-1, Par. B.v.)      to Regulatory Guide 1.143. Refer to Chapter 11.0. Section 3.8.6 des-cribes the seismic deisgn capabilities of the rad-waste building.
3.3.2    4.B.18  Tornado Load Effect Com-      The recommendations of (Par. II.2.d)      binations                      this SRP are met.
Refer to Section 3.3.2.
3.4.2    4.B.19  Dynamic Effects of Wave        The recommendations of (Par II)          Action                        this SRP are met.
Refer to Section 3.4.2.
10.4.7    4.B.20  Water Hammer for Steam        The WCGS steam qenerators (Par. I.2.b)      Generators with Pre-          (Model F) have no pre-heaters                        heaters. Refer to Sections 5.4.2 and 10.4.7.
4.4      4.B.21  Thermal-Hydraulic Sta-        The recommendations of (Par. II.5)        bility                        this SRP are met as dis-cussed in Section 4.4.4.6.
5.2.5    4.B.22  Intersystem Leakage Detec-    Intersystem leakage de-(Par II.4)        tion (See RG 1.45)            tection requirements and capabilities are discussed in Section 5.2.5.
3.2.2    4.B.23  Main Steam Isolation Valve    The recommendations of Leakage Control System        this SRP are not appli-(SRP Section 10.3, Par.        cable to the WCGS III.3 and BTP RSB-3.2)        (PWR) design.
Rev. 0
 
WOLF CREEK TABLE 1.9-4 OTHER CATEGORY 4 POSITIONS SRP    RRRC Section Category          Title                    Remarks 3.5.3  4.C.1    Ductility of Reinforced    The recommendations of this Concrete and Steel Struc-  item are met to the ex-tural Elements Subjected    tent described in Section to Impactive or Impulsive  3.5.3.
Loads 3.7.1  4.C.2    Response Spectra in Verti-  The recommendations of this cal Direction              item are met. Refer to Section 3.7(B).1. West-inghouse utilizes the damping values of WCAP 7921-AR. See also the response to Regulatory Guide 1.60 in Appendix 3A.
3.8.1  4.C.3    BWR Mark III Containment    The recommendations of this 3.8.2            Pool Dynamics              item are not applicable to the WCGS (PWR) design.
3.8.4  4.C.4    Air Blast Loads            Air blast loads from trans-portation are less than the external pressure design capabilities described in Section 3.8.
3.5.3  4.C.5    Tornado Missile Impact      The recommendations of this item are met. Refer to Section 3.5.3.1.
6.3    4.C.6    Passive Failures During    The recommendations of this Long-Term Cooling Follow-  item are met to the ex-ing LOCA                    tent described in Sections 3.1 and 6.3.
6.3    4.C.7    Control Room Position      The recommendations of this Indication of Manual        item are met. Refer to (Handwheel) Valves in      Sections 7.5.2.2.1 and the ECCS                    7.5.2.2.2.
15.1.5  4.C.8    Long-Term Recovery from    The recommendations of this Steamline Break: Opera-    item are met to the extent tor Action to Prevent      described in Section Overpressurization          15.0.13. Operator action is not assumed for 10 minutes.
Rev. 0
 
WOLF CREEK TABLE 1.9-4 (Sheet 2)
SRP    RRRC Section Category          Title                      Remarks 5.4.6  4.C.9    Pump Operability Require-    The recommendations of this 5.4.7            ments                        item are met. Refer to 6.3                                          Section 6.2.2.1.2.2 and Section 6.3.2.5.
3.5.1  4.C.10  Gravity Missiles, Vessel    The recommendations of this Seal Ring Missiles Inside    item are met. Refer to Containment                  Appendix 3B. Section 9.1.4.2.2 discusses the reactor cavity seal ring.
4.4    4.C.11  Core Thermal-Hydraulic      The recommendations of this Analysis                    item are met. However, Westinghouse is generically reducing rod bow penalties through experience gained by test surveillance. Refer to Section 4.2.3.1.
8.3    4.C.12  Degraded Grid Voltage        The recommendations of this Conditions                  item are met to the ex-tent described in Section 8.3.1.1.3 and Technical Specifications.
6.2.1.2 4.C.13  Asymmetric Loads on Com-    The recommendations of this ponents Located Within      item are met. Refer to Containment Subcompart-      Section 6.2.1.2.
ments 6.2.6  4.C.14  Containment Leak Testing    The recommendations of this Program                      item are met. Refer to Section 6.2.6.
6.2.1.4 4.C.15  Containment Response Due    The recommendations of this to Main Steamline Break      item are met. Refer to and Failure of MSLIV to      Sections 6.2.1, 3.11(B),
Close                        and 3.11(N).
3.6.1  4.C.16  Main Steam and Feedwater    The recommendations of this 3.6.2            Pipe Failures                item are met. Refer to Sections 3.6.1 and 3.6.2.
Rev. 0
 
WOLF CREEK TABLE 1.9-4 (Sheet 3)
Category          Title                      Remarks 9.2.2  4.C.17  Design Requirements for      The recommendations of this Cooling Water to Reactor    item are met to the extent Coolant Pumps                described in Sections 5.4.1, 9.2.2, and 9.3.4 10.4.7 4.C.18  Design Guidelines for        The design meets the rec-Water Hammer in Steam        ommendations of this item; Generators with Top          however, no testing was Feedring Design (BTP        performed. Refer to ASB-10.2)                    Section 10.4.7.
3.11  4.C.19  Environmental Control        The recommendations of this Systems for Safety-Re-      item are met to the ex-lated Equipment              tent described in Section 3.11(B).
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Latest revision as of 23:47, 17 November 2024

5 to Updated Safety Analysis Report, Chapter 1, Introduction and General Description of the Plant
ML22151A138
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 05/18/2022
From:
Wolf Creek
To:
Office of Nuclear Reactor Regulation
Shared Package
ML22152A025 List: ... further results
References
WO 22-0006
Download: ML22151A138 (200)
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