ML17151A993
| ML17151A993 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/09/2017 |
| From: | Wolf Creek |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17151A982 | List:
|
| References | |
| WO 17-0045 | |
| Download: ML17151A993 (199) | |
Text
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 T itle Drawing #*1.1-1 1 Symbols and Legend for System Flow and Piping and Instrumentation Diagrams M-120101 1.1-1 2 Symbols and Legend for System Flow and Piping and Instrumentation Diagrams M-120102 1.1-1 3 Symbols and Legend for System Flow and Piping and Instrumentation Diagrams M-020103 1.1-1 4 Symbols and Legend for System Flow and Piping and Instrumentation Diagrams M-020104 1.2-1 0 Peninsular Plant Arrangement Standard Power Systems & Structure Interface M-1G001 1.2-2 0 Equipment Location Radwaste Building Plan El.
1976'-0" M-1G010 1.2-3 0 Equipment Location Radwaste Building Plan El.
2000'-0" M-1G011 1.2-4 0 Equipment Location Radwaste Building Plan El.
2022'-0" M-0G012 1.2-5 0 Equipment Location Radwaste Building El. 2031'-
6" & Roof Plan M-1G013 1.2-6 0 Equipment Location Radwaste Building Sections A
& B M-1G014 1.2-7 0 Equipment Location Radwaste Building Sections C
& E M-1G015 1.2-8 0 Equipment Location Radwaste Building Sections D
& F M-1G016 1.2-9 0 Equipment Location Reactor and Auxiliary Bldgs Plan - Basement El. 1974'-0" M-1G020 1.2-10 0 Equipment Location Auxiliary Building Partial Plan El. 1988'-0" & El. 2013'-6" M-1G021 1.2-11 0 Equipment Location Reactor and Auxiliary Building Plan Ground Floor Elevation 2000'-0" M-1G022 1.2-12 0 Equipment Location Reactor and Auxiliary Building Plan El. 2026'-0" M-1G023 1.2-13 0 Equipment Location Reactor and Auxiliary Buildings Plan Operating Floor El. 2047'-6" M-1G024 1.2-14 0 Equipment Locations Reactor and Auxiliary Buildings Plan El. 2068'-8" M-1G025 1.2-15 0 Equipment Location Reactor and Auxiliary Building Section A M-1G026 1.2-16 0 Equipment Locations Reactor and Auxiliary Buildings Section B M-1G027 1.2-17 0 Equipment Location Reactor and Auxiliary Building Section C M-1G028 1.2-18 0 Equipment Location Reactor and Auxiliary Building Section D M-1G029 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 T itle Drawing #* 1.2-19 0 Equipment Location Auxiliary Building Sections E, F, & G M-1G030 1.2-20 0 Equipment Location Fuel Building Plan Elevation 2000'- 0", 2026'-0" and 2047'-6" M-1G040 1.2-21 0 Equipment Location Fuel Building Sections A, B, &
C M-1G041 1.2-22 0 Equipment Location Fuel Building Sections D, E, &
F M-1G042 1.2-23 0 Equipment Location Control Building &
Communication Corridor Plan Elevation 1974'- 0" &
1984'-0" M-1G050 1.2-24 0 Equipment Location Control & Diesel Generator Buildings & Communication Corridor Plan Elevation 2000'-0" & 2016'-0" M-1G051 1.2-25 0 Equipment Location Control & Diesel Generator Buildings & Communication Corridor Plan Elevation
2032'-0" & 2047'-6" M-1G052 1.2-26 0 Equipment Location Control & Diesel Generator Buildings & Corridor Plan Elevation 2061'- 6",
2066'-0" & 2073'-6" & Section D.
M-1G053 1.2-27 0 Equipment Location Control & Diesel Generator Buildings & Communication Corridor Section A M-1G054 1.2-28 0 Equipment Location Control & Diesel Generator Buildings Sections B & C M-1G055 1.2-29 0 Equipment Location Turbine Building Condenser Pit Plan Elevation 1983'-0" M-1G060 1.2-30 0 Equipment Location Turbine Building Ground Floor Plan Elevation 2000'-0" M-1G061 1.2-31 0 Equipment Location Turbine Building Partial Plan Elevation 2015'-4" M-1G062 1.2-32 0 Equipment Location Turbine Building Mezzanine Floor Plan Elevation 2033'-0" M-1G063 1.2-33 0 Equipment Location Turbine Building Operating Floor Plan Elevation 2065'-0" M-1G064 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'-
0" M-1G073 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 are:
Kansas City Power & Light Company (47 percent), Kansas Electric Power Cooperative, Incorporated (6 percent), and Kansas Gas and Electric Company (47
percent). 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. 2 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. 30 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 (T hot) 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 100. 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. 14 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 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. 29 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 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 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. 0 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 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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
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
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 of 1097.5 ft. msl., resulting from the
probable maximum flood routed through the cooling lake with coincident wave
activity, is below the plant site grade of 1099.5 ft. msl.
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. 1 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);
1.2-6 Rev. 30 WOLF CREEK u. Prescreening and Main Security Buildings;
- 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
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.
1.2-7 Rev. 30 WOLF CREEK 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.
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.
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).
1.2-8 Rev. 18 WOLF CREEK 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. 5 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.
1.2-14 Rev. 0 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 voltage is 345 kV for Kansas Gas and Electric
Company (KG&E) and Kansas City Power & Light Company (KCP&L). 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. 27 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 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
transfer.
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. 14 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, 73, and 100 is considered.
1.3-1 Rev. 23 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
10 6 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
- 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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
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.
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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.
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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.
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.
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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
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.
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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.
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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.
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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.
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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.
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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.
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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.
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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.
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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.
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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.
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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.
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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.
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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.
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Regulation (10 CFR) Compliance 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.
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.
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Regulation (10 CFR) Compliance 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.
73.60 This regulation applies to non-power reactors and thus does not apply to WCGS.
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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.
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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.
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. 3 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.
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.
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. 2 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. 1.4-5 Rev. 0 WOLF CREEK 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. 1.4-6 Rev. 0 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. Dr. James Halitsky This consultant developed calculations of atmospheric dispersion parameters for the control room fresh air intake for use in control room 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 Consultantsa. 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. 1.4-7 Rev. 0 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 WOLF CREEK 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. 1.4-9 Rev. 0 WOLF CREEK 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. 1.4-10 Rev. 0 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.
1.4-12 Rev. 30 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. 1.5-1 Rev. 0 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. 1.5-2 Rev. 0 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. 1.5-3 Rev. 0 WOLF CREEK 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-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
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-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-8170 (P) "Calculational Model for Core Rev. 0 15.16 7/3/74 AE WCAP-8171 Reflooding After a Loss-of-Coolant Accident (WREFLOOD
Code)"
WCAP-8172-A (See WCAP 8082-P-A[P])
Rev. 0 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-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-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-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-8302 (P) "SATAN-VI Program: Compre- Rev. 0 15.0, 15.6 7/12/74 AE WCAP-8306 hensive Space-Time Dependent Analysis of Loss-of-Coolant"
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-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
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. 0 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
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-8693 "Delta Ferrite in Production Rev. 0 5.2 3/16/76 B Austenitic Stainless Steel
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"
Rev. 11 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 and Instrumentation Diagrams M-120101 1.1-1 2 Symbols and Legend for System Flow and Piping and Instrumentation Diagrams M-120102 1.1-1 3 Symbols and Legend for System Flow and Piping and Instrumentation Diagrams M-020103 1.1-1 4 Symbols and Legend for System Flow and Piping and Instrumentation Diagrams M-020104 1.2-1 0 Peninsular Plant Arrangement Standard Power Systems & Structure Interface M-1G001 1.2-2 0 Equipment Location Radwaste Building Plan El.
1976'-0" M-1G010 1.2-3 0 Equipment Location Radwaste Building Plan El.
2000'-0" M-1G011 1.2-4 0 Equipment Location Radwaste Building Plan El.
2022'-0" M-0G012 1.2-5 0 Equipment Location Radwaste Building El. 2031'-
6" & Roof Plan M-1G013 1.2-6 0 Equipment Location Radwaste Building Sections A
& B M-1G014 1.2-7 0 Equipment Location Radwaste Building Sections C
& E M-1G015 1.2-8 0 Equipment Location Radwaste Building Sections D
& F M-1G016 1.2-9 0 Equipment Location Reactor and Auxiliary Bldgs Plan - Basement El. 1974'-0" M-1G020 1.2-10 0 Equipment Location Auxiliary Building Partial Plan El. 1988'-0" & El. 2013'-6" M-1G021 1.2-11 0 Equipment Location Reactor and Auxiliary Building Plan Ground Floor Elevation 2000'-0" M-1G022 1.2-12 0 Equipment Location Reactor and Auxiliary Building Plan El. 2026'-0" M-1G023 1.2-13 0 Equipment Location Reactor and Auxiliary Buildings Plan Operating Floor El. 2047'-6" M-1G024 1.2-14 0 Equipment Locations Reactor and Auxiliary Buildings Plan El. 2068'-8" M-1G025 1.2-15 0 Equipment Location Reactor and Auxiliary Building Section A M-1G026 1.2-16 0 Equipment Locations Reactor and Auxiliary Buildings Section B M-1G027 Rev. 17 WOLF CREEK Table 1.6-3 (Sheet 2)
USAR Figure/Controlled Drawing Cross-Reference Figure #Sheet TitleDrawing #1.2-17 0 Equipment Location Reactor and Auxiliary Building Section C M-1G028 1.2-18 0 Equipment Location Reactor and Auxiliary Building Section D M-1G029 1.2-19 0Equipment Location Auxiliary Building Sections E, F, & G M-1G030 1.2-20 0Equipment Location Fuel Building Plan Elevation 2000'- 0", 2026'-0" and 2047'-6" M-1G040 1.2-21 0Equipment Location Fuel Building Sections A, B, &
C M-1G041 1.2-22 0Equipment Location Fuel Building Sections D, E, &
F M-1G042 1.2-23 0 Equipment Location Control Building &
Communication Corridor Plan Elevation 1974'- 0" &
1984'-0" M-1G050 1.2-24 0 Equipment Location Control & Diesel Generator Buildings & Communication Corridor Plan Elevation 2000'-0" & 2016'-0" M-1G051 1.2-25 0 Equipment Location Control & Diesel Generator Buildings & Communication Corridor Plan Elevation 2032'-0" & 2047'-6" M-1G052 1.2-26 0 Equipment Location Control & Diesel Generator Buildings & Corridor Plan Elevation 2061'- 6", 2066'-0" & 2073'-6" & Section D.
M-1G053 1.2-27 0 Equipment Location Control & Diesel Generator Buildings & Communication Corridor Section A M-1G054 1.2-28 0 Equipment Location Control & Diesel Generator Buildings Sections B & C M-1G055 1.2-29 0Equipment Location Turbine Building Condenser Pit Plan Elevation 1983'-0" M-1G060 1.2-30 0Equipment Location Turbine Building Ground Floor Plan Elevation 2000'-0" M-1G061 1.2-31 0Equipment Location Turbine Building Partial Plan Elevation 2015'-4" M-1G062 1.2-32 0Equipment Location Turbine Building Mezzanine Floor Plan Elevation 2033'-0" M-1G063 1.2-33 0Equipment Location Turbine Building Operating Floor Plan Elevation 2065'-0" M-1G064 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 2065'-0" M-1G073 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 A & B Trains M-13EN03 6.2.2-2 2 Containment Spray System Reactor Building A & B Trains M-13EN04 6.2.2-2 3 Containment Spray System Reactor Building A & B Trains M-13EN05 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 #SheetTitleDrawing #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 Manual Trip Signals) M-744-00020 7.2-1 4 Functional Diagrams (Nuclear Instrumentation Permissives and Blocks) M-744-00021 7.2-1 5 Functional Diagrams (Primary Coolant System Trip SignalsM-744-00022 7.2-1 6 Functional Diagrams (Pressurizer Trip Signals) M-744-00023 7.2-1 7 Functional Diagrams (Steam Generator Trip Signals)M-744-00024 7.2-1 8 Functional Diagrams (Safeguards Actuation Signals)M-744-00025 7.2-1 9 Functional Diagrams (Rod Controls and Rod Blocks)M-744-00026 7.2-1 10 Functional Diagrams (Steam Dump Control) M-744-00027 7.2-1 11 Functional Diagrams (Pressurizer Pressure and Level Control) M-744-00028 7.2-1 12 Functional Diagrams (Pressurizer Heater Control) M-744-00029 7.2-1 13 Functional Diagrams (Feedwater Control and Isolation)M-744-00030 7.2-1 14 Functional Diagrams (Feedwater Control and Isolation)M-744-00031 7.2-1 15 Functional Diagrams (Auxiliary Feedwater Pumps Start-up)M-744-00032 7.2-1 16 Functional Diagrams (Turbine Trips, Runbacks and Other Signals) M-744-00033 7.2-1 17 Functional Diagram (Pressurizer Pressure Relief System Train A) M-744-00039 7.2-1 18 Functional Diagram (Pressurizer Pressure Relief System Train B) M-744-00040 7.3-1 2 Logic Diagram Engineered Safety Features Actuation System (BOP)
J-104-00390 7.6-4 1 Train B Functional Diagram Showing Logic Requirements for Pressurizer Pressure Relief SystemM-744-00039 7.6-4 2 Train A Functional Diagram Showing Logic Requirements for Pressurizer Pressure Relief SystemM-744-00040 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 #SheetTitleDrawing #8.3-1 1 Main Single Line Diagram E-11001 8.3-1 2 Single Line Diagram, Essential Service Water System E-K10018.3-1 3 Single Line Diagram Site Area Loads E-1001 8.3-2 0 List of Loads Supplied by the Emergency Diesel Generator E-110058.3-3 0 Logic Diagram Standby Generation Excitation Control E-12NE018.3-4 0 Logic Diagram Standby Generator System Protection E-12NE028.3-5 0 Logic Diagram Standby Generator Engine and Governor Control E-12KJ018.3-6 1 DC Main Single Line Diagram E-11010 8.3-7 0 DC Main Single Line Diagram (PK03 and PK04 Bus)E-11010A 9.1-3 1Fuel Pool Cooling and Cleanup System M-12EC01 9.1-3 2Fuel Pool Cooling and Cleanup System M-12EC02 9.2-1 1 Service Water System M-12EA019.2-1 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-12KD019.2-17 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 #SheetTitleDrawing #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 Drainage System M-12LA01 9.3-5 2 Comm. Corridor & Control Bldg. Sanitary Drainage
System M-12LA02 9.3-5 3 Chemical and Detergent Waste M-12LD01 9.3-5 4 Turbine Bldg. and Aux. Feedwater Pump Rooms
Oily Waste System M-12LE01 9.3-5 5 Control and Diesel Generator Bldg. Oily Waste
System M-12LE02 9.3-5 6 Turbine Bldg. and Aux. Boiler Room Oily Waste
System M-12LE03 9.3-5 7 Tendon Access Gallery and Turbine Bldg. Oily
Waste System M-12LE04 9.3-5 8 Auxiliary Building Floor and Equipment Drain (FED) System M-12LF01 9.3-5 9 Auxiliary Building Floor and Equipment Drain
System M-12LF02 9.3-5 10 Auxiliary Building Floor and Equipment Drain
System M-12LF03 9.3-5 11 Auxiliary Building Floor and Equipment Drain
System M-12LF04 9.3-5 12 Auxiliary Building Floor and Equipment Drain
System M-12LF05 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 #SheetTitleDrawing #9.3-7 1 Reactor Building, Stainless Steel Liner Plate, Reactor Refueling Canal C-0L29319.3-7 2 Fuel Building-Area 1, Stainless Steel Liner Plate Plan, Spent Fuel Pool C-1L61119.3-8 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-12GF029.4-3 3 Auxiliary Building HVAC M-12GL039.4-3 4 Auxiliary Building HVAC M-12GL029.4-3 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 #SheetTitleDrawing #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 System M-12KJ01 9.5.5-1 2 Standby Diesel Generator "B" Cooling Water
System M-12KJ04 9.5.6-1 1 Standby Diesel Generator "A" Intake, Exh., F.0.
and Starting Air System M-12KJ02 9.5.6-1 2 Standby Diesel Generator "B" Intake, Exh., F.0.
and Starting Air System M-12KJ05 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 #SheetTitleDrawing #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 3Circulating Water Waterbox Venting System M-12DA02 10.4-1 4Circulating Water Screenhouse Plans M-0004 10.4-1 5Circulating 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 0Condenser 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 1Steam Generator Blowdown System M-12BM01 10.4-8 2Steam Generator Blowdown System M-12BM02 10.4-8 3Steam Generator Blowdown System M-12BM03 10.4-8 4Steam Generator Blowdown System M-12BM04 10.4-8 5Steam Generator Blowdown System M-12BM05 Rev. 17 WOLF CREEK Table 1.6-3 (Sheet 10)
USAR Figure/Controlled Drawing Cross-Reference Figure #Sheet TitleDrawing #10.4-9 0Auxiliary Feedwater System M-12AL01 10.4-10 0 Auxiliary Turbines Auxiliary Feedwater Pump Turbine M-12FC02 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 1Radiation Zones for Normal Operation El. 1974 10466-A-1701 12.3-2 2Radiation Zones for Normal Operation El. 2000 10466-A-1702 12.3-2 3Radiation Zones for Normal Operation El. 2026 10466-A-1703 12.3-2 4Radiation Zones for Normal Operation El. 2047-6 10466-A-1704 12.3-2 5Radiation Zones for Normal Operation Turbine Bldg El. 1983 & 2000 10466-A-1705 12.3-2 6Radiation Zones for Normal Operation Turbine Bldg El. 2033 & 2065 10466-A-1706 12.3-4 0 Decontamination System M-12HD01 18.2-15 0Nuclear Sampling System M-12SJ04 Rev. 17 WOLF CREEK Table 1.6-4 (sheet 1)
Incorporated by Reference USAR Section/Controlled Do cument Cross-Reference Section Title Document # Table 3.11(B)-1 Plant Environmental Normal Conditions EQSD-I, Attachment A and B Table 3.11(B)-2 Environmental Qualification Parameters for SNUPPS NUREG-0588 (LOCA, MSLB and HELB)
EQSD-I, Attachment A and B Table 3.11(B)-3 Identification of Safety-Related Equipment and Components: Equipment Qualification EQSD-I, Attachment A and B; EQSD-II, Tables 1 and 2 Table 3.11(B)-4 Containment Worst Case Radiation Levels (MRADs)
EQSD-I, Attachment A Table3.11(B)-5 Containment Spray Requirements EQSD-I, Attachment A Table 3.11(B)-8 Exemptions from NUREG-0588 Qualification EQSD-I, Attachment C Table 3.11(B)-10 Equipment Added for NUREG-0737 EQSD-II, Tables 1 and 2 Figures 3.11(B)-
1 through 3.11(B)-49 Figures EQSD-I, Attachment A Section 9.5.1.2.2.3*
Fire Barriers *Note: Only portions of this section have been relocated and Incorporated by Reference M-663-00017A 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 Appendix 9.5B.3 Fire Hazard Analyses - Fire Effects on Electrical Equipment and Safe Shutdown Information E-1F9905 Rev. 28 WOLF CREEK Table 1.6-4 (sheet 2)
Incorporated by Reference USAR Section/Controlled Do cument Cross-Reference Section Title Document #
Table 9.5B-3 Safety-Related Fire Areas Containing Rooms Without Detection Provisions E-1F9905, Attachment A Appendix 9.5B.4 Fire Hazard Analyses - General Information on Design Features E-1F9905, or XX-E-013, or
E-1F9900 Table 9.5B-4 Safety-Related Fire Areas Outside Containment With Area Suppression Coverage E-1F9905 Attachment A Appendix 9.5B.5 Fire Hazard Analyses - Combustible Loadings and Flame Spread E-1F9905 Table 9.5B-5 Non-Safety Related Site Structures E-1F9905, Attachment C Appendix 9.5B.6 Fire Hazard Analyses - Fire Hazard Review Methodology E-1F9905, or XX-E-013 Appendix 9.5B.7 Fire Hazard Analyses - Power Block Fire Hazards Analysis E-1F9905 Appendix 9.5B.8 Fire Hazard Analyses - Site Specific Fire Hazards Analysis E-1F9905 Appendix 9.5B Design Basis Document for OFN AP-017, Control Room Evacuation E-1F9915 Chapter 16.0 Technical Requirements Manual Technical Requirements
Manual Chapter 17.2 Quality Assurance During the Operation Phase Quality 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 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 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*
Regulatory Guide RRRC Category Number Revision 2
3 4 1.12 1 X 1.13 1 X
1.14 1 X 1.27 2 X 1.52 1 X
1.56 1 X
1.59 2 X
1.63 2 X 1.68.2 1 X 1.75 1 X
1.76 0 X
1.79 1 X
1.80 0 X
1.82 0 X
1.83 1 X
1.89 0 X 1.91 1 X
1.93 0 X
1.97 1 X
1.99 1 X
1.101 1 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 0 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. 0 WOLFCREEKTABLE1.9-2CATEGORY2,3,AND4BRANCHTECHNICALPOSITIONSBranchTechnicalRRRCPositionTitle Cat.RemarksASB9.5-1,GuidelinesforFire2TherecommendationsofRev.1ProtectionforNu-thisBTParemettoclearPowerPlantstheextentdescribedinSection9.5.1.MTEB5-7MaterialSelection2TherecommendationsofandProcessingthisBTParenotap-GuidelinesforBWRplicabletotheWCGSCoolantPressure(PWR)design.
BoundaryRSB5-1,DesignRequirements3TherecommendationsofRev.1oftheResidualHeatthisBTParemettotheRemovalSystemextentdescribedinSec-tions5.2.2and7.6.6.RSB5-2Overpressurization3TherecommendationsofProtectionofPres-thisBTParemettothesurizedWaterReactorsextentdescribedinSec-WhileOperatingatLowtions5.2.2and7.6.6.
TemperaturesMTEB5-3MonitoringofSec-4.B.1TherecommendationsofondarySideWaterthisBTParemet.ReferChemistryinPWRtoSections9.3.2and SteamGenerators10.3.5.CSB6-1MinimumContainment4.B.2TherecommendationsofPressureModelforthisBTParemet.ReferPWRECCSPerformancetoSections6.2.1andEvaluation15.6.5.CSB6-2ControlofCombus-4.B.3TherecommendationsoftibleGasConcentra-thisBTParemet.RefertionsinContainmenttoSection6.2.5.FollowingaLoss-of-CoolantAccidentRev.0 WOLFCREEKTABLE1.9-2(Sheet2)BranchTechnicalRRRCPositionTitle Cat.RemarksCSB6-3Determinationof4.B.4TherecommendationsofBypassLeakagePathsBTParenotapplicableinDualContainmenttotheWCGSdesign,Plantssincethereisnodual containment.CSB6-4ContainmentPurging4.B.5TherecommendationsofDuringNormalPlantthisBTParemettothe OperationsextentdescribedinTable 9.4-13.ASB9.1OverheadHandling4.B.6TherecommendationsofSystemsforNu-thisBTParemet.NoclearPowerPlantscriticalloadsarehandled.Referto Section9.1.4.ASB10.1DesignGuidelines4.B.7TherecommendationsofforAuxiliaryFeed-thisBTParemet.Refer waterSystemPumptoSection10.4.9.DriveandPowerSup-plyDiversityforPWRPlantsRev.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
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).
Rev. 0