ML18271A132
| ML18271A132 | |
| Person / Time | |
|---|---|
| Site: | 05200046 |
| Issue date: | 09/28/2018 |
| From: | William Ward NRC/NRO/DLSE/LB2 |
| To: | |
| Ward W /NRO/415-7038 | |
| Shared Package | |
| ML18087A364 | List:
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| References | |
| Download: ML18271A132 (55) | |
Text
APPENDIX B. REFERENCES American Concrete Institute (ACI)
ACI 117, Specification for Tolerances for Concrete Construction and Materials and Commentary, American Concrete Institute, 2010.
ACI 211.1, Standard Practice for Selecting Proportions for Normal, Heavy Weight, and Mass Concrete, American Concrete Institute, 1991 (R2007).
ACI 214, Recommended Practice for Evaluation of Strength Test Results of Concrete, American Concrete Institute, 1991 (R1997).
ACI 304 R, Guide for Measuring, Mixing, Transporting, and Placing Concrete, American Concrete Institute, 2000 (R2009).
ACI 305 R, Hot Weather Concreting, American Concrete Institute, 1999.
ACI 306 R, Cold Weather Concreting, American Concrete Institute, 1998 (R2002).
ACI 308, Standard Practice for Curing Concrete, American Concrete Institute, 1992 (R1997).
ACI 309 R, Guide for Consolidation of Concrete, American Concrete Institute, 2005.
ACI 311.1 R, ACI Manual of Concrete Inspection, American Concrete Institute, 1999.
ACI 315, Details and Detailing of Concrete Reinforcement, American Concrete Institute, 1999.
ACI 318, Building code requirements for reinforced concrete, American Concrete Institute, 2008.
ACI 340R, Design of Structural Reinforced Concrete Elements in accordance with the Strength Design Method of ACI 318-95, American Concrete Institute, 1997.
ACI 347, Guide to Formwork for Concrete, American Concrete Institute, 2004.
ACI 349, Code Requirements for Nuclear Safety-Related Concrete Structures, American Concrete Institute, 1997.
ACI-349-01, Appendix B, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, 2001.
ACI 349-06, Code Requirements for Nuclear Safety-Related Concrete Structures (ACI 349-97) and Commentary, American Concrete Institute, September 2007.
ACI 349-1997, Code Requirements for Nuclear Safety Related Concrete Structures, American Concrete Institute, 1997.
ACI SP-66, ACI Detailing Manual, American Concrete Institute, 2004.
B-1
ACI 301, Specifications for Structural Concrete for Building, American Concrete Institute, 2010.
American Institute of Steel Construction (AISC)
AISC 360, Specification for Structural Steel Buildings, American Institute of Steel Construction, 2005.
American Iron and Steel Institute (AISI)
AISI S100, Specification for the Design of Cold-Formed Steel Structural Members, American Iron and Steel Institute, 2012.
American National Standards Institute (ANSI)
ANSI N14.6, Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4,500 Kg) or More American National Standards Institute/American Institute of Steel Construction (ANSI/AISC)
ANSI/AISC 303, Code of Standard Practice for Steel Buildings and Bridges, American Institute of Steel Construction, 2010.
ANSI/AISC 360-05, Specification for Structural Steel Buildings, American Institute of Steel Construction, 2005 B-2
ANSI/AISC N690 including Supplement 2 (2004), Specification for the Design, Fabrication and Erection of Steel SafetyRelated Structures for Nuclear Facilities, American Institute of Steel Construction, 1994.
ANSI/AISC N690-1994, Specification for the Design, Fabrication and Erection of Steel Safety Related Structures for Nuclear Facilities, American Institute of Steel Construction, 1994.
American National Standards Institute/American Nuclear Society (ANSI/ANS)
ANSI/ANS 2.8-1992, Determining Design Basis Flooding at Power Reactor Sites, American Nuclear Society, 1992.
ANSI/ANS 3.2, Managerial, Administrative, and Quality Assurance Controls for Operational Phase of Nuclear Power Plants, American Nuclear Society, 2012.
ANSI/ANS 5.1, Decay Heat Power in Light Water Reactors ANSI/ANS 8.1, Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors, American Nuclear Society, 1998.
ANSI/ANS-18.1-1999, entitled Radioactive Source Terms for Normal Operation of Light Water Reactors.
ANSI/ANS 8.17, Criticality Safety Criteria for Handling, Storage, and Transportation of LWR Fuel Outside Reactors, American Nuclear Society, 1984.
ANSI/ANS-8.3-2003, Criticality Accident Alarm System ANSI/ANS-18.1-1999, Radioactive Source Term for Normal Operation for Light Water Reactors.
ANSI/ANS 51.1-1983, American National Standard Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plant, American Nuclear Society, 1983.
ANSI/ANS 56.10-1987, Subcompartment Pressure and Temperature Transient Analysis in Light Water Reactors, American Nuclear Society, 1987.
ANSI/ANS 56.11-1988, Design Criteria for Protection against the Effects of Compartment Flooding in Light Water Reactor Plants, American Nuclear Society, 1988.
ANSI/ANS-57.1-1992, Design Requirements for Light Water Reactor Fuel Handling Systems.
ANSI/ANS 57.2, Design Requirements for Light Water Reactor Spent Fuel Storage Facilities at Nuclear Power Plants, American Nuclear Society, 1983.
ANSI/ANS 57.3, Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants B-3
ANSI/ANS 58.2-1988, American National Standard Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture, American Nuclear Society, 1988.
ANSI/ANS-59.51-1997, Fuel Oil Systems for Safety-Related Emergency Diesel Generators, ANSI/ANS-HPSSC 6.8.1-1981, Location and Design Criteria for Area Radiation Monitoring Systems for Light Water Nuclear Reactors.
American National Standards Institute/Instrument Society of America (ANSI / ISA)
ANSI/ISA 7.0.01-1996, Quality Standard for Instrument Air ANSI/TIA-603-C-2004, Land Mobile FM or PM - Communications Equipment -Measurement and Performance Standards American Society of Civil Engineers (ASCE)
ASCE 4-98, Seismic Analysis of Safety Related Nuclear Structures and Commentary, American Society of Civil Engineers, January 2000.
ASCE 7-05, Minimum Design Loads for Buildings and Other Structures, American Society of Civil Engineering/Structural Engineering Institute, 2006.
M. A. Haroun and G. W. Housner, Seismic Design of Liquid Storage Tanks, Journal of the Technical Councils of ASCE, Vol. 107, No. TC1, pp. 191-207, American Society of Civil Engineers, April 1981.
American Society of Civil Engineers/Structural Engineering Institute (ASCE/SEI)
ASCE/SEI 7-05, Minimum Design Loads for Buildings and Other Structures, American Society of Civil Engineers/Structural Engineering Institute, 2006.
ASCE/SEI 37-02, Design Loads on Structures during Construction Standard, American Society of Civil Engineers, 2002.
B-4
ASCE/SEI 43-05, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, American Society of Civil Engineers, 2005.
American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)
Code American Society of Mechanical Engineers (ASME) Code, Sections III, V, and IX ASME B31.1, Code for Pressure Piping, Power Piping, The American Society of Mechanical Engineers, the 2010 Edition.
ASME B31.1, Code for Pressure Piping, Power Piping, The American Society of Mechanical Engineers, the 2010 Edition.
ASME B31.3, Code for Pressure Piping, Process Piping, The American Society of Mechanical Engineers, the 2010 Edition.
ASME Boiler and Pressure Vessel Code,Section III, Division 1, Code Cases: Nuclear Components, Boiler and Pressure Vessel Code, The American Society of Mechanical Engineers, the 2007 Edition.
ASME Boiler and Pressure Vessel Code,Section III, Division 1, Rules for Construction of Nuclear Facility Components, The American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.
ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendix O, Rules for Design of Safety Valve Installation, The American Society of Mechanical Engineers, the 2007 Edition.
ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NE, Class MC Components, The American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.
ASME Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NF, Supports, The American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.
ASME Boiler and Pressure Vessel Code,Section III, Division 2, Code for Concrete Containments, Subsection CC, The American Society of Mechanical Engineers, the 2001 Edition with the 2003 Addenda.
ASME Boiler and Pressure Vessel Code,Section III, SA240, Specification for Chromium and Chromium-Nickel Stainless Steel Plate, Sheet, and Strip for Pressure Vessels and for General Applications, The American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.
ASME Boiler and Pressure Vessel Code,Section IX, Welding and Brazing Qualifications, The American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.
B-5
ASME Boiler and Pressure Vessel Code,Section XI, Inservice Inspection of Nuclear Power Plants, The American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.
ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, The American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.
ASME Code Case N-597-2, Requirements for Analytical Evaluation of Pipe Wall Thinning ASME Code Case N-4-13, Special Type 403 Modified Forgings or Bars,Section III, Division 1, Class 1 and CS, dated February 12, 2008.
ASME Section III, Division 1, Subsection NE, Class MC Components, The American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.
ASME Section III, Division 2, Rules of Construction of Nuclear Facility Components - Code for Concrete Containments, American Society of Mechanical Engineers, the 2007 Edition with the 2008 Addenda.
Other ASME Documents ANS/ASME-58-22-201x draft, Low-Power and Shutdown PRA Methodology," American Nuclear Society, American Society of Mechanical Engineers, July 2013.
ASME B31.1, Power Piping, The American Society of Mechanical Engineers, 2010.
ASME Code Case N-249-14, Additional Materials for Subsection NF, Class 1, 2, 3, and MC Component Supports Fabricated without Welding,Section III, Division 1, dated February 20, 2004.
ASME Code Case N-60-5, Material for Core Support Structures,Section III, Division 1, dated February 20, 2004.
ASME Code Case N-71-18, Additional Materials for Subsection NF, Class 1, 2, 3, and MC Supports Fabricated by Welding,Section III, Division 1, dated May 9, 2003.
ASME Code Case N-759-2, Alternative Rules for Determining Allowable External Pressure and Comprehensive Stress for Cylinders, Cones, Spheres, and Formed Heads,Section III, Division 1, dated January 4, 2008.
ASME NQA-1, Quality Assurance Requirements for Nuclear Facility Applications, American Society of Mechanical Engineers, 2008 Edition with 2009 Addendum.
ASME NQA-1-2008, Quality Assurance Program Requirements for Nuclear Facilities, The American Society of Mechanical Engineers, 2008.
B-6
ASME NQA-1a-2009, Addenda to ASME NQA-1-2008 Quality Assurance Program Requirements for Nuclear Facilities, The American Society of Mechanical Engineers, 2009.
ASME OM Code Cases OMN-13, Requirements for Extending Snubber Inservice Visual Examination Interval at LWR Power Plants, American Society of Mechanical Engineers, the 2004 Edition.
ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, American Society of Mechanical Engineers, the 2004 Edition with the 2005 and 2006 Addenda.
ASME OM-S/G-1990, Standards and Guides For Operations of Nuclear Power Plants, Part 3, Requirements for Preoperational and Initial Start-Up Vibration Testing of Nuclear Power Plant Piping Systems, and Part 7, Requirements for Thermal Expansion Testing of Nuclear Power Plant Piping Systems.
ASME Paper No. 75-PVP-56.
ASME QME-1, Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,"
The American Society of Mechanical Engineers, the 2007 Edition.
ASME QME-1-2007 Qualification of Active Mechanical Equipment Used in Nuclear Power Plants, The American Society of Mechanical Engineers, 2007.
ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (Revision 1 RA-S-2002), American Society of Mechanical Engineers, April 2008.
ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, American Society of Mechanical Engineers, February 2009.
ASME/ANS RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk for Nuclear Power Plant Applications, The American Society of Mechanical Engineers, 2009.
American Society of Testing and Materials (ASTM)
ASTM A36, Standard Specification for Carbon Structural Steel, American Society for Testing and Materials, 2013.
ASTM A416, Standard Specification for Steel Strand, Uncoated Seven-Wire for Prestressed Concrete, American Society for Testing and Materials, 2002.
ASTM A513, Standard Specification for Electric-Resistance-Welded Carbon and Alloy Steel Mechanical Tubing, American Society for Testing and Materials, 2000.
ASTM A519, Standard Specification for Seamless Carbon and Alloy Steel Mechanical Tubing, American Society for Testing and Materials, 2003.
B-7
ASTM A576, Standard Specification for Steel Bars, Carbon, Hot-Wrought, Special Quality, American Society for Testing and Materials, 1990.
ASTM A615, Standard Specification for Deformed and Plain Carbon-Steel Bars for Concrete Reinforcement, American Society for Testing and Materials, 2006.
ASTM C109, Standard Test Method for Compressive Strength of Hydraulic Cement Mortars, American Society for Testing and Materials, 2013.
ASTM C150, Standard Specification for Portland Cement, American Society for Testing and Materials, 2012.
ASTM C191, Standard Test Method for Time of Setting of Hydraulic Cement by Vicat Needle, American Society for Testing and Materials, 2013.
ASTM C260, Standard Specification for Air-Entraining Admixtures for Concrete, American Society for Testing and Materials, 2010.
ASTM C33, Standard Specification for Concrete Aggregates, American Society for Testing and Materials, 2013.
ASTM C494, Standard Specification for Chemical Admixtures for Concrete, American Society for Testing and Materials, 2013.
ASTM C618, Standard Specification for Coal Fly Ash and Raw or Calcined Natural Pozzolan for Use in Concrete, American Society for Testing and Materials, 2012.
ASTM C750, Standard Specification for Nuclear Grade Boron Carbide Powder, American Society for Testing and Materials, 2009(R2014).
ASTM E190, Standard Test Method for Guided Bend Test for Ductility of Welds, American Society for Testing and Materials, 2014.
ASTM E3, Standard Guide for Preparation of Metallographic Specimens, American Society for Testing and Materials, 2011.
Combustion Engineering Documents (CE)
CENPD-107, CESEC - Digital Simulation of a Combustion Engineering Nuclear Steam Supply System, CE, April 1974.
LD-82-001 (dated 1/6/82), CESEC - Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," Enclosure 1-P to letter from A.E. Scherer to D.G. Eisenhut, CE, December 1981.
CENPD-161-P, TORC Code - A Computer Code for Determining the Thermal Margin of a Reactor Core, CE, July 1975.
CENPD-206-P-A, "TORC Code-Verification and Simplified Modeling Methods," CE, June 1981.
B-8
CEN-214(A)-P, "CETOP-D Code Structure and Modeling Methods for Arkansas Nuclear One-Unit 2, CE, July 1982.
CENPD-98-A COAST Code Description, CE, April 1973.
CE-CES-159, Revision 0-P, "HRISE Users Manual, December 1992.
CENPD-254-P-A, "Post-LOCA LTC Evaluation Model", June 1980 (ML15358A220).
CENPD-190-A, CE Method for Control Element Assembly Ejection Analysis, CE, January 1976, (ML15240A186 (Proprietary)).
CENPD-135P, Supplement 2, STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modifications), CE, February 1975.
CENPD-135, Supplement 4-P, STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program, CE, August 1976.
CENPD-135, Supplement 5-P, STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program, CE, April 1977.
CENPD-188-A, "HERMITE A Multi-Dimensional Space-Time Kinetics Code for PWR Transients," CE, Reprinted July 1976.
CE-CES-048, Users Manual for CEKLAPS, Revision 3-P, August 1994.
CENPD-266-P-A, The ROCS and DIT Computer Codes for Nuclear Design, April 1983.
CE-CES-091-P, Users Manual for HERMITE - Space-Time Neutronics and Thermal Hydraulics Code, Revision 4, Westinghouse Electric Co., September 2001.
CENPD-133P, Supplement 1, CEFLASH-4AS, A Computer Program for Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident, CE, August 1974 (ML15267A400 (Proprietary)).
CENPD-133, Supplement 3-P, CEFLASH-4AS, A Computer Program for the Reactor Blowdown Analysis of the Small Break Loss-of-Coolant Accident, CE, January 1977.
CENPD-134P, COMPERC-II, A Program for Emergency Refill-Reflood of the Core, CE, August 1974.
CENPD-134P, Supplement 1, COMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modifications), CE, February 1975.
CENPD-134P, Supplement 2, COMPERC-II, A Program for Emergency Refill-Reflood of the Core, CE, June 1985.
CENPD-138P, PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup, CE, August 1974.
CENPD-138P, Supplement 1, PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup (Modifications), CE, February 1975.
B-9
CENPD-138, Supplement 2-P, PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup, CE, January 1977.
CENPD-137, Supplement 1-P, Calculative Methods for the CE Small Break LOCA Evaluation Model, CE Power Systems, January 1977.
CENPD-135P, STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program, CE, August 1974.
CENPD-158, Revision01, Analysis of Anticipated Transients without Reactor Scram in Combustion Engineering NSSSs, May 1976.
CENPD-263-P, ATWS Early Verification, Combustion Engineering, Inc., November 1979.
CENPD-107, CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System, April 1974.
CENPD-107, Supplement 1, ATWS Model Modifications to CESEC, September 1974.
CENPD-107, Supplement 1, Amendment 1-P, ATWS Model Modifications to CESEC, November 1975.
CENPD-107, Supplement 2, ATWS Models for Reactivity Feedback and Effect of Pressure on Fuel, September 1974.
CENPD-107, Supplement 3, ATWS Model Modifications to CESEC, August 1975.
CENPD-107, Supplement 4-P, ATWS Model Modifications to CESEC, December 1975.
CENPD-252-P-A, Method for the Analysis of Blowdown Induced Forces in a Reactor Vessel, Combustion Engineering, Inc., July 1979.
CENPD-133P, CEFLASH-4A: A Fortran-IV-Digital Computer Program for Reactor Blowdown Analysis, Combustion Engineering, Inc., August 1974.
CENPD-133P, CEFLASH-4A: A Fortran-IV Digital Computer Program for Reactor Blowdown Analysis (Modifications), Combustion Engineering, Inc., Supplement 2, February 1975.
Scherer, A. E., Licensing Manager, (C-E), Letter to D. F. Ross, Assistant Director of Reactor Safety Division of Systems Safety, LD-76-026, March 1976 (Proprietary).
CEN-267-(V)-P, Final Report on the Performance Evaluation of the Palo Verde Control Element Assembly Shroud, Combustion Engineering, Inc., Rev. 1-P, 1984.
CEN-263(V)-P, A Comprehensive Vibration Assessment Program for Palo Verde Nuclear Generating Station Unit 1 (System 80 Prototype), Combustion Engineering, Inc., Rev. 1-P, January 1985.
CENPD-178, Structural Analysis of Fuel Assemblies for Seismic and Loss-of-Coolant Accident Loading, Combustion Engineering, Inc., Rev. 1, August 1981.
Combustion Engineering Topical Report, CENPD-137P, Calculative Methods for the CE Small Break LOCA Evaluation Model, August 1974.
B-10
Combustion Engineering Topical Report, CENPD-137, Supplement 1-P, Calculative Methods for the CE Small Break LOCA Evaluation Model, January 1977.
Combustion Engineering Topical Report CENPD-137P, Supplement 2-P-A, Calculative Methods for the CE Small Break LOCA Evaluation Model, April 1998.
CENPD-42, Topical Report on Dynamic Analysis of Reactor Vessel Internals Under Loss-of-Coolant Accident Conditions with Application of Analysis to C-E 800 MWe Class Reactors, Combustion Engineering, Inc., August 1972.
CEN-133(B), FIESTA: A One Dimensional, Two Group Space-Time Neutronics Code for Calculating PWR Scram Reactivities, CE, November 1970. Approval: Letter, R.A. Clark (NRC) to A.E. Lundvall, Jr. (BG&E), Docket Nos. 50-317 and 50-318, Approval of CEN-133(B),
March 13, 1981.
Electric Power Research Institute (EPRI)
NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Nuclear Energy Institute, July 2000.
J. A. Achenbach et al., Large-scale Hydrogen Burn Equipment Experiments, NP-4354, Electric Power Research Institute, Palo Alto, CA, December 1985.
FAI/1994-034, MAAP4-DOSE Users Manual, Modular Accident Analysis Program Users Manual, Volume 4, Electric Power Research Institute, Palo Alto, CA, May 1994.
NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Nuclear Energy Institute, July 2000.
J. A. Achenbach et al., Large-scale Hydrogen Burn Equipment Experiments, NP-4354, Electric Power Research Institute, Palo Alto, CA, December 1985.
FAI/1994-034, MAAP4-DOSE Users Manual, Modular Accident Analysis Program Users Manual, Volume 4, Electric Power Research Institute, Palo Alto, CA, May 1994.
Advanced Light Water Reactor Utility Requirements Document, Rev. 7, Electric Power Research Institute, 1995.
EPRI Interim Technical Report, Treatment of Loss of Offsite Power (LOOP) in Probabilistic Risk Assessments: Technical Basis and Guidelines, Electric Power Research Institute, September 2009.
EPRI TR-100259, An Approach to the Analysis of Operator Actions in Probabilistic Risk Assessment, Electric Power Research Institute, June 1992.
EPRI TR-101869, Severe Accident Management Guidance Technical Basis Report: Volumes 1 and 2, Electric Power Research Institute, April 1993.
B-11
EPRI NP-6041, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Electric Power Research Institute, August 1991.
EPRI 1016735, Fire PRA Methods Enhancements: Additions, Clarifications, and Refinements to EPRI 1019189, Electric Power Research Institute, December 2008.
EPRI 1021086, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments (PRAs), Electric Power Research Institute, October 2010.
EPRI 1002988, Seismic Fragility Application Guide, Electric Power Research Institute, December 2002.
EPRI Interim Technical Report, Treatment of Loss of Offsite Power (LOOP) in Probabilistic Risk Assessments: Technical Basis and Guidelines, Electric Power Research Institute, September 2009.
EPRI TR-100259, An Approach to the Analysis of Operator Actions in Probabilistic Risk Assessment, Electric Power Research Institute, June 1992.
EPRI TR-101869, Severe Accident Management Guidance Technical Basis Report: Volumes 1 and 2, Electric Power Research Institute, April 1993.
EPRI NP-6041, A Methodology for Assessment of Nuclear Power Plant Seismic Margin, Electric Power Research Institute, August 1991.
EPRI 1016735, Fire PRA Methods Enhancements: Additions, Clarifications, and Refinements to EPRI 1019189, Electric Power Research Institute, December 2008.
EPRI 1021086, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments (PRAs), Electric Power Research Institute, October 2010.
EPRI 1002988, Seismic Fragility Application Guide, Electric Power Research Institute, December 2002.
EPRI Interim Technical Report, Treatment of Loss of Offsite Power (LOOP) in Probabilistic Risk Assessments: Technical Basis and Guidelines, September 2009.
EPRI TR-102323-R3, Guidelines for Electromagnetic Interference Testing of Power Plant Equipment EPRI TR-102323, Guidelines for Electromagnetic Interference Testing of Power plant Equipment, Rev. 2, Electric Power research Institute, 2000.
EPRI NP-5639, Guidelines for Piping System Reconciliation, Electric Power Research Institute, May 1988.
EPRI NP-6766, Water Hammer Prevention, Mitigation, and Accumulation, Volume 1: Plant Water Hammer Experience, Final Report, Electric Power Research Institute, July 1992.
EPRI MRP-111, Materials Reliability Program (MRP), Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111),
Electric Power Research Institute, March 2004.
B-12
EPRI NP-1931, V. Kumar, M.D. German, and C.F. Shih, An Engineering Approach for Elastic-Plastic Fracture Analysis, Electric Power Research Institute, July 1981.
EPRI NP-3596-SR, PICEP: Pipe Crack Evaluation Program, Rev. 1, Electric Power Research Institute, December 1992.
EPRI Report NP-6695, Guidelines for Nuclear Plant Response to an Earthquake, Electric Power Research Institute, December 1989.
EPRI Report TR-100082, Standardization of the Cumulative Absolute Velocity, Electric Power Research Institute, December 1992.
EPRI Draft White Paper, Considerations for NPP Equipment and Structures Subjected to Response Levels Caused by High Frequency Ground Motions, Electric Power Research Institute, March 2007.
EPRI TR-1023389, Evaluation of Seismic Hazards at Central and Eastern U.S. Nuclear Power Sites, Electric Power Research Institute, June 2011.
EPRI TR-102631, Soil-Structure Interaction Analysis Incorporating Spatial Incoherence of Ground Motions, Electric Power Research Institute, October 1997.
EPRI TR-1015110, Effects of Spatial Incoherence on Seismic Ground Motions, Electric Power Research Institute, November 2007.
EPRI TR-1015108, Program on Technology Innovation: The Effects of High-Frequency Ground Motion on Structures, Components, and Equipment in Nuclear Power Plants, Electric Power Research Institute, June 2007.
EPRI TR-1015109, Program on Technology Innovation: Seismic Screening of Components Sensitive to High-Frequency Vibratory Motions, Electric Power Research Institute, October 2007.
EPRI NP-2230, ATWS: A Reappraisal, Part 3: Frequency of Anticipated Transients, Electric Power Research Institute, January 1982.
UNCERT Users Manual," Version 3.0, EPRI, Palo Alto, CA, 2009.
UNCERT Users Manual," Version 3.0, EPRI, Palo Alto, CA, 2009.
FAI/12-0005, MAAP 4.0.8 Transmittal Document, Electric Power Research Institute, February 2012.
Modular Accident Analysis Program (MAAP) Version 4.0.8, Electric Power Research Institute, Palo Alto, CA, August 2012.
Modular Accident Analysis Program (MAAP) Version 4.0.8, Electric Power Research Institute, Palo Alto, CA, August 2012.
FAI/12-0005, MAAP 4.0.8 Transmittal Document, Electric Power Research Institute, February 2012.
B-13
The EPRI HRA Calculator Version 4.21 (Software Product ID #1022814) Software Manual, Electric Power Research Institute, May 2011.
The EPRI HRA Calculator Version 4.21 (Software Product ID #1022814) Software Manual, Electric Power Research Institute, May 2011.
CAFTA 6.0b, Software Manual, EPRI, Palo Alto, CA, 2014.
Korea Hydro Nuclear Power Co. (KHNP)
KHNP Procedure DC-DG-03-01, Design Change Control.
KHNP Procedure DC-DG-03-05, Technical Audit at Suppliers Facility.
KHNP Procedure DC-DG-03-09, Implementation of the Reliability Assurance Program (RAP).
KHNP Procedure DC-DG-03-10, Expert Panel Roles and Responsibilities.
KHNP Procedure DC-DG-03-11, Risk Significance Determination of RAP SSCs.
KHNP Procedure DC-DG-03-23, Implementation of Severe Accident Mitigation Design Alternatives.
KHNP Procedure DC-DG-03-24, Risk Management Procedure.
KHNP Procedure DC-DG-16-01, Corrective Action Program.
KHNP Technical Reports APR1400-E-B-NR-16001-P, Evaluation of Main Steam and Feedwater Piping applied to the graded approach for the APR1400, KEPCO & KHNP, March 2016.
APR1400-E-B-NR-16002-P, Evaluation of Safety Injection and Shutdown Cooling Piping applied to the graded approach for the APR1400," Rev. 1, KEPCO & KHNP, May 2018.
APR1400-H-N-NR-14005-P, Summary Stress Report for Primary Piping, Rev. 1, KEPCO &
KHNP, December 2015.
APR1400-K-Q-TR-11005-NP-A, KHNP Quality Assurance Program Description (QAPD) for the APR1400 Design Certification, Rev. 2, KHNP, October 2016.
B-14
APR1400-K-X-FS-14002-P, Tier 2, Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, Rev. 3, KHNP, July 2018.
APR1400-K-X-FS-14002-NP, Tier 2, Chapter 14, Verification Programs, Rev. 1, KHNP, March 2017.
APR1400-K-Q-TR-11005-NP-A, KHNP Quality Assurance Program Description (QAPD) for the APR1400 Design Certification, Rev. 2, KHNP, October 2016.
APR1400-E-S-NR-14004-P, Evaluation of Effects of HRHF Response Spectra on SSCs, Rev.
3, KEPCO & KHNP, December 2017.
APR1400-E-X-NR-14001-P, Equipment Qualification Program, Rev. 4, KHNP, July 2018.
APR1400-E-X-NR-14001-P, Equipment Qualification Program, Rev. 4, KEPCO & KEPCO &
KHNP, July 2018.
APR1400-F-A-NR-14002-P, The Effect of Thermal Conductivity Degradation on APR1400 Design and Safety Analyses, KEPCO & KHNP, September 2014.
APR1400-F-M-TR-13001-NP, Revision 1, PLUS7 Fuel Design for the APR1400 (ML17237A023).
APR1400-Z-M-NR-14010-NP, Revision 2, Structural Analysis of Fuel Assemblies for Seismic and Loss of Coolant Accident Loading (ML17228A787).
APR1400-Z-A-NR-14015, Revision 1, Neutron Fluence Calculation Methodology for Reactor Vessel, (ML17094A121)
APR1400-F-C-NR-14002, Revision 1 Functional Design Requirements for a Core Operating Limit Supervisory System for APR1400, (ML17094A132)
APR1400-F-C-NR-12001-NP, Revision 2, Thermal Design Methodology (ML17181A052),
APR1400-Z-J-NR-14004-NP, Revision 1, Uncertainty Methodology and Application for Instrumentation, (ML17094A127)
APR1400-Z-M-NR-14017-NP, Revision 0, Evaluation of Irradiation Assisted Stress Corrosion Cracking and Void Swelling on Reactor Vessel Internals (ML16096A281)
APR1400-F-C-NR-14003-NP, Functional Design Requirements for a Core Protection Calculator System for APR1400 (ML15009A192)
APR1400-E-A-NR-14002-P-SGI, Physical Security Design Features, Rev. 1, KEPCO & KHNP, May 2016.
APR1400-E-A-NR-14002-P-SGI, Physical Security Design Features, Rev. 1, KEPCO & KHNP, May 2016.
APR1400-F-C-NR-14001-P, CPC Setpoint Analysis Methodology for APR1400, Rev. 3, KHNP, June 2018.
B-15
APR1400-F-C-NR-14001-P, CPC Setpoint Analysis Methodology for APR1400, Rev. 3, KHNP, June 2018.
APR1400-E-I-NR-14001, Human Factors Engineering Program Plan, (ML18212A336).
APR1400-E-I-NR-14002, Operating Experience Review Implementation Plan, (ML18081A107).
APR1400-E-I-NR-14003, Functional Requirements Analysis and Function Allocation Implementation Plan, (ML18081A107).
APR1400-E-I-NR-14004, Task Analysis Implementation Plan, (ML18178A202).
APR1400-E-I-NR-14006, Treatment of Important Human Actions Implementation Plan, (ML18178A202).
APR1400-E-I-NR-14007, Human-System Interface Design Implementation Plan, (ML18178A202).
APR1400-E-I-NR-14008, Human Factors Verification and Validation Implementation Plan, (ML18178A202).
APR1400-E-I-NR-14010, Human Factors Verification and Validation Scenarios, (ML18081A107).
APR1400-E-I-NR-14011, Basic Human-System Interface, (ML18178A202).
APR1400-E-I-NR-14012, Style Guide, (ML18081A107).
APR1400-K-I-NR-14005, Staffing and Qualifications Implementation Plan, (ML17094A129).
APR1400-K-I-NR-14009, Design Implementation Plan, (ML17094A153).
APR1400-Z-A-NR-14019, CCF Coping Analysis, (ML18086B746).
APR1400-Z-J-NR-14001, Safety I&C System, (ML18212A336).
APR1400-Z-J-NR-14012, "Control System CCF Analysis," (ML18212A336).
APR1400-E-S-NR-14002-P, Finite Element Seismic Models for SSI Analyses of the NI Buildings, Rev. 2, KEPCO & KHNP, December 2017.
APR1400-E-S-NR-14005-P, Evaluation of Structure-soil-structure Interaction (SSSI) Effects, Rev. 2, KEPCO & KHNP, December 2017.
APR1400-E-S-NR-14003-P, SSI Analysis Methodology and Results of NI Buildings, Rev. 2, KEPCO & KHNP, December 2017.
APR1400-E-S-NR-14004-P, Evaluation of Effects of HRHF Response Spectra on SSCs, Rev.
3, KEPCO & KHNP, December 2017.
APR1400-E-P-NR-14005-P, Revision 2, Evaluations and Design Enhancements to Incorporate Lessons Learned from the Fukushima Dai-Ichi Nuclear Accident, B-16
APR1400 Design Control Document Tier 2 Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, APR1400-K-X-FS-14002-P, Revision 0., December 2014 APR1400-E-N-NR-14004-P, Summary Report of High-Energy Piping Rupture Analysis, Rev.
2, KEPCO & KHNP, October 2017.
APR1400-E-S-NR-14001-P, Seismic Design Bases, Rev. 2, KEPCO & KHNP, October 2017.
APR1400-Z-M-NR-14009-P, Comprehensive Vibration Assessment Program for the Reactor Vessel Internals, Rev. 1, KEPCO & KHNP, February 2017.
APR1400-F-C-TR-12002-P, Revision 0, KCE-1 Critical Heat Flux Correlation for PLUS7' Thermal Design Topical Report, KHNP and KEPCO, November 2012 (ML130180119).
Technical Report, APR1400-F-A-NR-14001-P, Revision 0, Small Break LOCA Evaluation Model, September 2014.
Technical Report, APR1400-E-N-NR-14001-P/-NP, Revision 0, Design Features to Address GSI-191, December 2014 (ML15009A323 (Proprietary), ML15009A130 (Non-Proprietary).
Technical Report, APR1400-Z-A-NR-14007-P, Revision0, LOCA Mass and Energy Release Methodology.
Technical Report, APR1400-E-N-NR-14001-P/-NP, Revision 1 Design Features to Address GSI-191, March 2017, (ML17094A172 (Proprietary), ML17094A122 (Non-Proprietary).
Technical Report, APR1400-K-A-NR-14001-P/-NP, Revision1, In-vessel Downstream Effect Tests for the APR1400, July 2015, (ML15195A015 (Proprietary), ML15195A016 (Non-Proprietary).
Technical Report, APR1400-K-A-I(RA)-13001-P (R3), Test Plan for In-vessel Downstream Effect (IDE) of the APR1400, July, 2014.
Technical Report, APR1400-K-A-NR-14001-P, Revision 2, In-vessel Downstream Effect Tests for the APR1400, July 2015 (ML17094A173 (Proprietary), ML17094A123 (Non-Proprietary).
APR1400-Z-J-NR-14002, Diversity and Defense-in-Depth Technical Report (D3 TeR)
APR1400-Z-A-NR-14019, CCF Coping Analysis Technical Report (CCF Coping Analysis TeR)
APR1400-E-J-NR-14001, Component Interface Module Technical Report (CIM TeR)
APR1400-F-C-NR-14003, Functional Design Requirements for a Core Protection Calculator System for APR1400 Technical Report (CPCS TeR)
APR1400-Z-J-NR-14012, Control System CCF Analysis Technical Report (CSCCF TeR)
APR1400-Z-J-NR-14013, Response Time Analysis of Safety I&C System Technical Report (Response Time Analysis TeR)
APR1400-A-J-NR-14003-, APR1400 Disposition of Common Q Topical Report NRC Generic Open Items and Plant Specific Action Items Technical Report (Common Q Platform Disposition TeR)
B-17
APR1400-A-J-NR-14004-, Common Q Platform Supplemental Information in Support of APR1400 Design Certification Technical Report (Common Q Supplemental TeR)
APR1400-Z-J-NR-14002-P, Revision 0, Diversity and Defense-in-Depth, November 2014.
KHNP Topical Reports Topical Report APR1400-F-A-TR-12004-P/-NP, Revision 0, Realistic Evaluation Methodology for Large-Break LOCA of the APR1400, December 2012 (ML130230128).
Topical Report APR1400-Z-M-TR-12003-P/-NP, Revision 0, Fluidic Device Design for the APR1400, December 2012 (ML130180120).
Topical Report APR1400-F-M-TR-13001, Revision 1, PLUS7 Fuel Design for the APR1400, August 2017 (ML17223B416 (Proprietary), ML17237A023 (Non-Proprietary).
Topical Report APR1400-F-C-TR-12002-P/-NP, KCE-1 Critical Heat Flux Correlation for PLUS7 Thermal Design, Revision 0, November 2012 (ML130180119).
Korea Electric Power Corporation Documents (KEPCO)
E-P-NU-907-1.3/DC, SAREX 1.3 Software Registration, KEPCO E&C, July 2013.
E-P-NU-1341-1.6/DC, FTREX 1.6 Software Registration, KEPCO E&C, July 2013.
9-911-Z-301-001C (CN-RAM-07-009, Rev. 3), Unavailability Analysis for the Plant Protection System (and Engineered Safety Feature - Component Control System) for SKN Power Plant Units 3 & 4, KEPCO E&C.
Full Power Level 2 PRA - PDS Analysis, APR1400-K-P-NR-013601, Revision 0, July 2013 Full Power Level 2 PRA - CET/DET Analysis, APR1400-K-P-NR-013602, Revision 0, July 2013 Full Power Level 2 PRA - Source Term Category Analysis, APR1400-K-P-NR-013603, Revision 0, July 2013 LPSD Level 2 MAAP Analyses, APR1400-K-P-NR-013761-P, Revision 0, December 2014 MELCOR calculation notebook: APR1400 Pressurized Water Reactor, ERI/NRC 14-212, Revision 3, April 2015.
B-18
Simulation of Representative Design Basis and Beyond Design Basis Accidents in APR1400, ERI/NRC 15-202, Revision 1, June 2015.
MELCOR Confirmatory Analysis for APR1400, meeting between NRC and KHNP, December 16-17, 2015. (ML16019A047)
Independent MELCOR Confirmatory Analysis of Selected Scenarios for APR1400 PWR, SPRA-16-01, November 2016 (ML16330A143)
KEPCO E&C/ND/TR/12-022, Thermal-hydraulic Analyses for PSA Support of NRC DC 1400 Plant, KEPCO E&C, December 2012.
11E47-1-603-M370-002, Rev. 0, Room Heatup Calculation for Electrical and I&C Equipment Rooms, KEPCO E&C, July 2015.
9-911-Z-301-001C (CN-RAM-07-009, Rev. 3), Unavailability Analysis for the Plant Protection System (and Engineered Safety Feature - Component Control System) for SKN Power Plant Units 3 & 4, KEPCO E&C.
KEPCO E&C/ND/TR/12-022, Thermal-hydraulic Analyses for PSA Support of NRC DC 1400 Plant, KEPCO E&C, December 2012.
11E47-1-603-M370-002, Rev. 0, Room Heatup Calculation for Electrical and I&C Equipment Rooms, KEPCO E&C, July 2015.
Westinghouse Documents Westinghouse Electric Corp. Report, Review and Evaluation of MHI BACCHUS PWR Vessel Mixing Tests, WCAP-16317-P, November 2004.
WCAP-18067-P, Rev. 0/APR1400-A-M-NR-16001-P, PRA Model for RCP Seal Failure Given Loss of Seal Cooling for APR1400 KSB HDD-254 Type F RCP Seals, Rev. 0, Westinghouse, July 2016.
WSRC-TR-93-262, Rev. 1, Savannah River Site Generic Data Base Development, Westinghouse Safety Management Solution, May 1998.
LTR-RAM-17-81, WO 89 - Response to Digital I&C Questions, Westinghouse, Revision 0-A, 2017.
WCAP-10697-P-A, Revision 3, Common Q Platform Topical Report, (Common Q Platform Topical Report).
Westinghouse Topical Report, Post-LOCA Long Term Cooling Evaluation Model, CENPD-254-P-A, June 1980.
B-19
Other KHNP Documents 00000-SS-VV-030, Software Verification and Validation Report of CESEC-III 89300 MOD5CS, Revision 00, KHNP, June 14, 2013.
APR1400-E-P-NR-14001-P, PRA Summary Report, Revision 0, KHNP.
APR1400-E-P-NR-14003-P, Severe Accident Analysis Technical Report, KHNP, December 2014.
APR1400-E-P-NR-14006-P, Severe Accident Mitigation Design Alternatives for the APR1400, Revision 0, KHNP, December 2014.
APR1400-F-A-TR-12004-P, Realistic Evaluation Methodology for Large-Break LOCA of the APR1400, Revision 0, KHNP, December 2012 (ML130230128).
APR1400-F-C-NR-12001-P, Thermal Design Methodology Technical Report, KHNP and KEPCO, September 2012.
APR1400-F-C-NR-12001-P, Thermal Design Methodology, Revision 1, KHNP and KEPCO, November 2014.
APR1400-K-Q-TR-11005, KHNP Quality Assurance Program Description (QAPD) for the APR1400 Design Certification, (ML16123A404).
APR1400-K-X-ER-14001, APR1400 Design Certification: Applicant's Environmental Report - Standard Design Certification, KHNP, December 2014.
APR1400-Z-A-NR-14006-P, Non-LOCA Safety Analysis Methodology, Revision 0, KHNP, September 2014 (ML15012A027 (Proprietary), ML15009A204 (Non-Proprietary).
APR1400-Z-A-NR-14014-P, ATWS Evaluation, Revision 0, KHNP, November 2014.
KHNP Calculation Note 1-035-N389-101, Revision 3, Hydrogen Generation and Control during Severe Accidents KHNP Calculation Note 1-035-N389-102, Revision 1, Assessment of AICC Pressure Load Due to Hydrogen Combustion in Containment KHNP Calculation Note 1-035-N389-103, Revision 2, Analysis of Local DDT Potential in the APR1400 Containment Technical Report, APR1400-F-A-NR-14003-P, Post-LOCA Long Term Cooling Evaluation Model, KHNP, September 2014.
Design Control Document, Tier 1, APR1400-K-X-IT-14001-P, Revision 0, December 2014.
APR1400 Design Control Document Tier 2, Chapter 15, Transient And Accident Analyses, APR1400-K-X-FS-14002-NP, Revision 0, December 2014.WCAP-16793-NP, Revision 1, Evaluation of Long-Term Cooling Considering Particulate, Fibrous and Chemical Debris in the Recirculating Fluid, Westinghouse Electric Corporation, Revision 2, October 2011.
B-20
APR1400 Design Control Document Tier 2 Chapter 19, Probabilistic Risk Assessment and Severe Accident Evaluation, APR1400-K-X-FS-14002-P, Revision 0, December 2014 Full Power Level 2 PRA - PDS Analysis, APR1400-K-P-NR-013601, Revision 0, July 2013 Full Power Level 2 PRA - CET/DET Analysis, APR1400-K-P-NR-013602, Revision 0, July 2013 Full Power Level 2 PRA - Source Term Category Analysis, APR1400-K-P-NR-013603, Revision 0, July 2013 LPSD Level 2 MAAP Analyses, APR1400-K-P-NR-013761-P, Revision 0, December 2014 MELCOR calculation notebook: APR1400 Pressurized Water Reactor, ERI/NRC 14-212, Revision 3, April 2015.
Simulation of Representative Design Basis and Beyond Design Basis Accidents in APR1400, ERI/NRC 15-202, Revision 1, June 2015.
MELCOR Confirmatory Analysis for APR1400, meeting between NRC and KHNP, December 16-17, 2015. (ML16019A047)
Independent MELCOR Confirmatory Analysis of Selected Scenarios for APR1400 PWR, SPRA-16-01, November 2016 (ML16330A143)
APR1400-E-P-NR-14005-P, Revision 2, Evaluations and Design Enhancements to Incorporate Lessons Learned from the Fukushima Dai-Ichi Nuclear Accident, APR1400-E-X-NR-14001-P, Equipment Qualification Program, Rev. 4, KEPCO & KHNP, July 2018.
APR1400-E-S-NR-14006-P, Stability Check for NI Common Basemat, Rev. 5, KEPCO &
KHNP, May 2018.
Institute for Electrical and Electronics Engineers (IEEE)
IEEE Std 384-1981, IEEE Standard Criteria for Independence of Class 1E Equipment and Circuits.
IEEE Std 420-1982, IEEE Standard for the Design and Qualification of Class 1E Control Boards, Panels, and Racks Used in Nuclear Power Generating Stations."
IEEE Std 497-2002, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations.
IEEE Std 603-1991, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating
- Stations, B-21
IEEE Std. 1050-1996, IEEE Guide for Instrumentation and Control Equipment Grounding in Generating Stations IEEE Std. 112-2004, IEEE Standard Test Procedure for Polyphase Induction Motors and Generators, Institute of Electrical and Electronics Engineers, 2004.
IEEE Std. 1202-1991 (R1996), IEEE Standard for Flame Propagation Testing of Wire and Cable, Institute of Electrical and Electronics Engineers, 1991.
IEEE Std. 1205-2000, IEEE Guide for Assessing, Monitoring, and Mitigating Aging Effects on Class 1E Equipment Used in Nuclear Generating Stations, Institute of Electrical and Electronics Engineers, 2000.
IEEE Std. 1290-1996 (R2005), IEEE Guide for Motor Operated valve (MOV) Motor Application, Protection, Control, and Testing in Nuclear Power Generation Station, Institute of Electrical and Electronics Engineers, 1996.
IEEE Std. 1313.2, Guide for the Application of Insulation Coordination IEEE Std. 141-1993, IEEE Recommended Practice for Electric Power Distribution for Industrial Plants IEEE Std. 1613-2003, IEEE Standard Environmental and Testing Requirements for Communications Networking Devices in Electric Power Substations IEEE Std. 242-2001, IEEE Recommended Practice for Protection and Coordination of Industrial and Commercial Power Systems IEEE Std. 269-2002, IEEE Standard Methods for Measuring Transmission Performance of Analog and Digital Telephone Sets, Handsets, and Headsets IEEE Std. 308-2001, IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations IEEE Std. 308-2001, IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers , 2001.
IEEE Std. 317-1983 (R2003), IEEE Standard for Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 1983.
IEEE Std. 323, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations IEEE Std. 323-2003, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2004.
IEEE Std. 323-2003, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2003.
B-22
IEEE Std. 334-2006, IEEE Standard for Qualifying Continuous-Duty Class 1E Motors for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2006.
IEEE Std. 338-1987, Standard Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems IEEE Std. 344-2004 (R2009), Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2005.
IEEE Std. 344-2004 (Reaffirmed 2009), Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, June 2005.
IEEE Std. 344-2004, IEEE Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2005.
IEEE Std. 344-2004, Recommended Practice for Seismic Qualification of Class 1E Equipment for Nuclear Power Generating Stations IEEE Std. 379-2000, Application of the Single-Failure Criterion to Nuclear Power Generating Station Safety Systems IEEE Std. 382-2006, IEEE Standard for Qualification of Actuators for Power-Operated Valve Assemblies with Safety-Related Functions for Nuclear Power Plants, Institute of Electrical and Electronics Engineers, 2007.
IEEE Std. 382-2006, IEEE Standard for Qualification of Safety Related Actuators for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2006.
IEEE Std. 383-2003, IEEE Standard for Type Test of Class 1E Electric Cables and Field Splices for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2003.
IEEE Std. 384-1992, IEEE Standard Criteria for Independence of Class 1E Equipment and Circuits IEEE Std. 387-1995, Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations IEEE Std. 387-1995, IEEE Standard Criteria for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 1995.
IEEE Std. 387-1995, Standard for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations IEEE Std. 387-1998 (R2007), IEEE Standard Criteria for Diesel Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 1995.
B-23
IEEE Std. 484-2002, IEEE Recommended Practice for Installation Design and Installation of Vented Lead-Acid Batteries for Stationary Applications IEEE Std. 485-2010, IEEE Recommended Practice for Sizing Lead-Acid Batteries for Stationary Applications IEEE Std. 487-2000, IEEE Recommended Practice for the Protection of Wire-Line Communication Facilities Serving Electric Supply Locations IEEE Std. 497-2002, IEEE Standard Criteria for Accident Monitoring Instrumentation for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2002.
IEEE Std. 519-1992, IEEE Recommended Practices and Requirements for Harmonic Control in Electrical Power Systems IEEE Std. 535-2006, Standard for Qualification of Class 1E Vented Lead Acid Storage Batteries for Nuclear Power Generating Stations IEEE Std. 535-2013, IEEE Standard for Qualification of Class 1E Vented Lead Acid Storage batteries for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2013.
IEEE Std. 572-2006, IEEE Standard for Qualification of Class 1E Connection Assemblies for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2006.
IEEE Std. 603-1991, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating
- Stations, IEEE Std. 603-1991, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 1991.
IEEE Std. 603-1991, IEEE Standard Criteria for Safety Systems for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 1991.
IEEE Std. 603-1991, Standard Criteria for Safety Systems for Nuclear Power Generating Stations IEEE Std. 627-2010, IEEE Standard for Design Qualification of Safety Systems Equipment Used in Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2010.
IEEE Std. 628-2001, IEEE Standard Criteria for Design, Installation and Qualification of Raceway Systems, Institute of Electrical and Electronics Engineers, 2001.
IEEE Std. 638-1992, IEEE Standard for Qualification of Class 1E Transformers for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 1992.
IEEE Std. 649-2006, IEEE Standard for Qualifying Class 1E Motor Control Centers for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2006.
B-24
IEEE Std. 650-2006, IEEE Standard for Qualification of Class 1E Static battery Chargers and Inverters for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2006.
IEEE Std. 666-1991 (reaffirmed 1996), IEEE Design Guide for Electric Power Service Systems for Generating Stations, IEEE Std. 7-4.3.2-2003,IEEE Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2003.
IEEE Std. 741, Standard Criteria for the Protection of Class 1E Power Systems and Equipment in Nuclear Power Generating Stations IEEE Std. 765 IEEE Standard for Preferred Power Supply (PPS) for Nuclear Power Generating Stations (NPGS)
IEEE Std. 80-2000, IEEE Guide for Safety in AC Substation Grounding IEEE Std. C37.013, Standard for AC High-Voltage Generator Circuit Breakers Rated on a Symmetrical Current Basis IEEE Std. C37.105-2010, IEEE Standard for Qualifying Class 1E Protective Relays and Auxiliaries for Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers, 2010.
IEEE Std. C37.82-1987 (R2004), IEEE Standard for the Qualification of Switchgear Assemblies for Class 1E Applications in Nuclear Power Generating Stations, Institute of Electrical and Electronics Engineers,1987.
IEEE Std. C62.23-1995 (reaffirmed in 2001), and IEEE Application Guide for Surge Protection of Electric Generating Plants IEEE Std. C62.82.1-2010, Standard for Insulation Coordination - Definitions, Principles, and Rules IEEE Std. 317-1983, IEEE Standard for Electric Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations IEEE Std. 450-2010, Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications IEEE 338-1987, IEEE Standard Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems."
IEEE Std. 665-1995 (reaffirmed in 2001), IEEE Guide for Generating Station Grounding IEEE Std. 946-2004, IEEE Recommended Practice for the Design of DC Auxiliary Power Systems for Generating Stations IEEE/ANSI C63.12-1999, American National Standard Recommended Practice for Electronic Compatibility Limits B-25
Nuclear Energy Institute (NEI)
NEI 00-04, 10 CFR 50.69 SSC Categorization Guideline, Rev. 0, Nuclear Energy Institute, July 2005.
NEI 05-01, Severe Accident Mitigation Alternatives (SAMA) Analysis - Guidance Document, Revision A, Nuclear Energy Institute (NEI), November 2005.
NEI 06-13A, Template for an Industry Training Program Description, Rev. 2, Nuclear Energy Institute, March 2009.
NEI 06-14, Quality Assurance Program Description (QAPD), Rev. 9, Nuclear Energy Institute, May 2010.
NEI 08-01, Revision 4, Industry Guidelines for the ITAAC Closure Process Under 10 CFR 52, July 2013 (ML13224A027).
NEI 08-01, Revision 5, Industry Guidelines for the ITAAC Closure Process Under 10 CFR 52, July 2010 (ML102010051).
NEI 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-12-051, To NEI 12-02, Revision 1, Industry Guidance for Compliance with NRC Order EA-12-051, To Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation NEI 12-06, Diverse and Flexible Coping Strategies (FLEX) Implementation Guide (ML12242A378).
NEI 97-06, Steam Generator Program Guidelines.
NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Rev. 6, Nuclear Energy Institute, December 2012.
Nuclear Energy Institute (NEI) guidance, NEI 04-07, "Pressurized Water Reactor Sump Performance Evaluation Methodology, May 2004.
U.S. Nuclear Regulatory Commission (NRC)
Audit Summary Reports APR1400 PRA and SA Audit Summary Report, (completion in project Phase 4)
B-26
CCVR-TH-02-02, CETOP Version 1 Mod4_kce1a Computer Code Verification Report, Revision 00, KHNP, January 30, 2002 (Audit Report).
VV-FE-0416, Software Verification and Validation Report - HERMITE Rev 1.6 Mod 0, Revision 0, CE, February 2, 1998 (Audit Report).
Commission Papers - Secretary of the Commission (SECY)
SECY-05-0197, Review of Operational Programs in a Combined License Application and Generic Emergency Planning Inspections, Tests, Analyses, and Acceptance Criteria, U.S.
Nuclear Regulatory Commission, October 2005.
SECY-05-0197, Review of Operational Programs in a Combined License Application and Generic Emergency Planning Inspections, Tests, Analyses, and Acceptance Criteria, October 2005.
SECY-05-0197, Review of Operational Programs in a Combined License Application and Generic Emergency Planning Inspections, Tests, Analyses, and Acceptance Criteria, October 28, 2005 (ML052770225).
SECY-11-0014 - Enclosure 1, The Use Of Containment Accident Pressure In Reactor Safety Analysis, (ML102110167).
SECY-12-0025 (February 17, 2012), Proposed Orders and Requests for Information in Response to Lessons Learned from Japans March 11, 2011, Great Tohoku Earthquake and Tsunami.
SECY-17-0075, Planned Improvements in Design Certification Tiered Information Designations, July 2017 (ML16196A321).
SECY-90-016, Evolutionary Light-Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements, U.S. Nuclear Regulatory Commission, June 1990.
SECY-90-241, Level of Detail Required for Design Certification under Part 52, July 1990.
SECY-90-377, Requirements for Design Certification under 10 CFR Part 52, November 1990.
SECY-91-178, Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications and Combined Licenses, June 1991.
SECY-91-210, Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Requirements for Design Review and Issuance of a Final Design Approval (FDA), July 1991.
SECY-92-053, Use of Design Acceptance Criteria during 10 CFR Part 52 Design Certification Process, February 1992.
B-27
SECY-92-214, Development of Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications, June 1992.
SECY-93-087, Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor Designs, U.S. Nuclear Regulatory Commission, Letter issued April 2, 1993 and Staff Requirements Memoranda issued July 21, 1993.
Staff Requirements Memorandum on SECY-93-087, Item II.Q, Defense Against Common-Mode Failures in Digital Instrumentation and Control Systems, U.S. Nuclear Regulatory Commission, July 1993.
Staff Requirements Memorandum to SECY-05-0197, Staff Requirements - SECY 0197 -
Review of Operational Programs in a Combined License Application and Generic Emergency Planning Inspections, Tests, Analyses, and Acceptance Criteria, U.S. Nuclear Regulatory Commission, February 2006.
Staff Review Memorandum for SECY-93-087, SECY-93-087 - Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs, U.S.
Nuclear Regulatory Commission, July 1993.
Interim Staff Guidance DC/COL-ISG-020, Interim Staff Guidance on Implementation of a Probabilistic Risk Assessment Based Seismic Margin Analysis for New Reactors, U.S. Nuclear Regulatory Commission.
JLD-ISG-2012-03, Revision 0, Compliance with Order EA-12-051, Reliable Spent Fuel Pool Instrumentation, issued August 29, 2012 LD-ISG-2012-01, Revision 0, Compliance with Order EA-12-049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External.
NUREG Series Reports NUREG/BR-0184, Regulatory Analysis Technical Evaluation Handbook, U.S. Nuclear Regulatory Commission, 1997.
NUREG/CR-1161RD, Recommended Revisions to Nuclear Regulatory Commission Seismic Design Criteria, U.S. Nuclear Regulatory Commission, May 1980.
NUREG/CR-1278, Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plants, U.S. Nuclear Regulatory Commission, August 1983.
B-28
NUREG/CR-1363, Data Summaries of LERs of Valves at U.S. Commercial Nuclear Power Plants, U.S. Nuclear Regulatory Commission, 1982.
NUREG/CR-1677, Piping Benchmark Problems, Vol. 1 and Vol. 2, U.S. Nuclear Regulatory Commission, August 1980.
NUREG/CR-1842, Evaluation of Human Reliability Analysis Methods Against Good Practices, U.S. Nuclear Regulatory Commission, September 2006.
NUREG/CR-2059, Compilation of Data Concerning Known and Suspected Water Hammer Events in Nuclear Power Plants, CY 1969-May 1981, U.S. Nuclear Regulatory Commission, May 1982.
NUREG/CR-2300, PRA Procedures Guide, U.S. Nuclear Regulatory Commission, January 1983.
NUREG/CR-2781, Evaluation of Water Hammer Events in Light Water Reactor Plants, U.S.
Nuclear Regulatory Commission, July 1982.
NUREG/CR-2913, Two-Phase Jet load, U.S. Nuclear Regulatory Commission, January 1983.
NUREG/CR-3331, A Methodology for Allocation of Nuclear Power Plant Control Functions to Human and Automated Control, August 1983.
NUREG/CR-3862, Development of Transient Initiating Event Frequencies for Use in Probabilistic Risk Assessment, U.S. Nuclear Regulatory Commission, May 1985.
NUREG/CR-4146, Simulation of an EPRI-Nevada Test Site (NTS) Hydrogen Burn Test at the Central Receiver Test Facility, U.S. Nuclear Regulatory Commission, June 1985.
NUREG/CR-4324, Testing of Nuclear Qualified Cables and Pressure Transmitters in Simulated Hydrogen Deflagrations to Determine Survival Margins and Sensitivities, U.S. Nuclear Regulatory Commission, December 1985.
NUREG/CR-4334, An Approach to the Quantification of Seismic Margins in Nuclear Power Plants, U.S. Nuclear Regulatory Commission, August 1985.
NUREG/CR-4482, Recommendations to the Nuclear Regulatory Commission on Trial Guidelines for Seismic Margin Reviews of Nuclear Power Plants, U.S. Nuclear Regulatory Commission, 1986.
NUREG/CR-4527, An Experimental Investigation of Internally Ignited Fires in Nuclear Power Plant Control Cabinets, Part II: Room Effects Tests, U.S. Nuclear Regulatory Commission, April 1987.
NUREG/CR-4550, Analysis of Core Damage Frequency: Internal Events Methodology, Volume 1, Rev. 1, U.S. Nuclear Regulatory Commission, January 1990.
NUREG/CR-4551, Evaluation of Severe Accident Risks: Methodology for the Containment, Source Term, Consequence, and Risk Integration Analyses, U.S. Nuclear Regulatory Commission, December 1993 B-29
NUREG/CR-4639 (EGG-2458), Nuclear computerized Library for Assessing Reactor Reliability (NUCLARR), May 1990.
NUREG/CR-4763, Safety-Related Equipment Survival in Hydrogen Burns in Large Dry PWR Containment Buildings, U.S. Nuclear Regulatory Commission, March 1988.
NUREG/CR-4772, Accident Sequence Evaluation Program Human Reliability Analysis Procedure, U.S. Nuclear Regulatory Commission, February 1987.
NUREG/CR-4780, Procedures for Treating Common Cause Failures in Safety and Reliability Studies, U.S. Nuclear Regulatory Commission, January 1988.
NUREG/CR-5096, Evaluation of Seals for Mechanical Penetrations of Containment Buildings, U.S. Nuclear Regulatory Commission, August 1988.
NUREG/CR-5132, Severe Accident Insights Report, U.S. Nuclear Regulatory Commission, April 1988.
NUREG/CR-5163, Power Burst Facility (PBF) Severe Fuel Damage Test 1-4 Test Results Report, U.S. Nuclear Regulatory Commission, April 1989.
NUREG/CR-5249, Quantifying Reactor Safety Margins, Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large-Break, Loss-of-Coolant Accident, EGG-2552, Revision 4, Idaho National Engineering Laboratory and EG&G Idaho, Inc., December 1989 (ML030380503).
NUREG/CR-5334, Severe Accident Testing of Electrical Penetration Assemblies, U.S. Nuclear Regulatory Commission, November 1989.
NUREG/CR-5347, Recommendations for Resolutions of Public Comments on USIA-40, Seismic Design Criteria, U.S. Nuclear Regulatory Commission, June 1989.
NUREG/CR-5485, Guidelines on Modeling Common Cause Failures in Probabilistic Risk Assessment, U.S. Nuclear Regulatory Commission, November 1998.
NUREG/CR-5497, CCF Parameter Estimations, 2010 Update, http://nrcoe.inl.gov/results/CCF/ParamEst2010/ccfparamest.htm, U.S. Nuclear Regulatory Commission, January 2012.
NUREG/CR-5500, Vol. 10, Reliability Study: Combustion Engineering Reactor Protection System, 1984-1998, U.S. Nuclear Regulatory Commission, November 2001.
NUREG/CR-5535, RELAP5/MOD3 Code Manual, U.S. Nuclear Regulatory Commission, June 1995.
NUREG/CR-5535, RELAP5/MOD3.3 Code Manual, Rev. P3, U.S. Nuclear Regulatory Commission, March 2006.
NUREG/CR-5535, Volumes 1 through 7, RELAP5/MOD3 Code Manual.
NUREG/CR-5564, Core-Concrete interactions Using Molten Urania with Zirconium on a Basaltic Basemat: The SURC-2 Experiment, U.S. Nuclear Regulatory Commission, August 1992.
B-30
NUREG/CR-5597, In-Vessel Zircaloy Oxidation/Hydrogen Generation Behavior during Severe Accidents, U.S. Nuclear Regulatory Commission, September 1990.
NUREG/CR-5750 (INEEL/EXT-98-00401), Rates of Initiating Event at U.S. Nuclear Power Plants: 1987 - 1995, U.S. Nuclear Regulatory Commission, February 1999.
NUREG/CR-5750, Rates of Initiating Events at U.S. Nuclear Power Plants: 1987-1995, December 1998.
NUREG/CR-6075, The Probability of Containment Failure by Direct Containment Heating in Zion, NUREG/CR-6075, December 1994.
NUREG/CR-6144 (BNL-NUREG-52399), Evaluation of Potential Severe Accidents During Low Power and Shutdown Operations at Surry, Unit 1, U.S. Nuclear Regulatory Commission, June 1994.
NUREG/CR-6303, "Method for Performing Diversity and Defense-in-Depth Analyses of Reactor Protection Systems," December 1994.
NUREG/CR-6338, Resolution of the Direct Containment Heating Issue for All Westinghouse Plants with Large Dry Containments or Sub-atmospheric Containment, U.S. Nuclear Regulatory Commission, February 1996.
NUREG/CR-6365, "Steam Generator Tube Failures," U.S. Nuclear Regulatory Commission, April 1996.
NUREG/CR-6400, Human Factors Engineering (HFE) Insights for Advanced Reactors Based upon Operating Experience, January 1997 (ML063480112).
NUREG/CR-6728, Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk-consistent Ground Motion Spectra Guidelines, U.S. Nuclear Regulatory Commission, October 2001.
NUREG/CR-6728, Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk-consistent Ground Motion Spectra Guidelines, U.S. Nuclear Regulatory Commission, March 2007.
NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, U.S.
Nuclear Regulatory Commission, September 2005.
NUREG/CR-6850, Supplement 1, Fire Probabilistic Risk Assessment Methods Enhancement, U.S. Nuclear Regulatory Commission, September 2010.
NUREG/CR-6883, The SPAR-H Human Reliability Analysis Method, U.S. Nuclear Regulatory Commission, August 2005.
NUREG/CR-6890, Reevaluation of Station Blackout Risk at Nuclear Power Plants, U.S.
Nuclear Regulatory Commission, December 2005.
NUREG/CR-6906, Containment Integrity Research at Sandia National Laboratories - An Overview, U.S. Nuclear Regulatory Commission, July 2006.
B-31
NUREG/CR-6928 (INL/EXT-06-11119), Industry-Average Performance for Components and Initiating Events at U.S. Commercial Nuclear Power Plants, U.S. Nuclear Regulatory Commission, February 2007.
NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S.
Commercial Nuclear Power Plants, January 2007, U.S. Nuclear Regulatory Commission, Industry Average Parameter Estimates, 2010 Update, http://nrcoe.inl.gov/resultsdb/AvgPerf, September 2012.
NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S.
Commercial Nuclear Power Plants - Initiating Event Data Sheets - Update 2010, U.S. Nuclear Regulatory Commission, January 2012.
NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S.
Commercial Nuclear Power Plants, January 2007, U.S. Nuclear Regulatory Commission, Industry Average Parameter Estimates, 2010 Update, http://nrcoe.inl.gov/resultsdb/AvgPerf, September 2012.
NUREG/CR-6928, Industry-Average Performance for Components and Initiating Events at U.S.
Commercial Nuclear Power Plants - Initiating Event Data Sheets - Update 2010, U.S. Nuclear Regulatory Commission, January 2012.
NUREG/CR-7114, A Framework for Low Power/Shutdown Fire PRA, U.S. Nuclear Regulatory Commission, September 2013.
NUREG/CR-7150, Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE), May 2014.
NUREG-0003, Statistical Studies of Vertical and Horizontal Earthquake Spectra, U.S. Nuclear Regulatory Commission, January 1976.
NUREG-0017, Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors: PWR-GALE Code, 1985.
NUREG-0460, Anticipated Transients Without Scram, Staff Report, Division of Systems Safety, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, April 1978.
NUREG-0484, Methodology for Combining Dynamic Responses, Rev. 1, U.S. Nuclear Regulatory Commission, May 1980.
NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, Revision 1, U.S. Nuclear Regulatory Commission, July 1981.
NUREG-0611, Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Westinghouse - Designed Operating Plants, January 1980 (ML082590256)
NUREG-0612, Control of Heavy load at Nuclear Power Plant: Resolution of Generic Technical Activity A-36, U.S. Nuclear Regulatory Commission, July 2007.
NUREG-0635, Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in Combustion Engineering - Designed Operating Plants, January 1980 (ML083540575)
B-32
NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, November 1980 (supplemented by March 2002 Addenda).
NUREG-0696, Functional Criteria for Emergency Response Facilities, U.S. Nuclear Regulatory Commission, February 1981 (ML051390358).
NUREG-0700, Human-System Interface Design Review Guidelines, May 2002 (ML021700373).
NUREG-0711, Human Factors Engineering Program Review Model, Revision 3, November 2012 (ML12324A013).
NUREG-0737, Clarification of TMI Action Plan Requirements, U.S. Nuclear Regulatory Commission, November 1980 (ML051400209).
NUREG-0737, Supplement No. 1, Clarification of TMI Action Plan Requirements:
Requirements for Emergency Response Capability, U.S. Nuclear Regulatory Commission, January 1983 (ML102560009).
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, U.S. Nuclear Regulatory Commission, various dates and revisions.
NUREG-0835, Human Factors Acceptance Criteria for the Safety Parameter Display System, October 1981 (ML102520360).
NUREG-0899, Guidelines for the Preparation of Emergency Operating Procedures, U.S.
Nuclear Regulatory Commission, August 1982 (ML102560007).
NUREG-0927, Evaluation of Water Hammer Occurrence in Nuclear Power Plants, March 1984 (ML071030267)
NUREG-0927, Evaluation of Water-Hammer Occurrence in Nuclear Power Plants: Technical Findings Relevant to Unresolved Safety Issue A-1, Rev. 1, U.S. Nuclear Regulatory Commission, March 1984.
NUREG-0933, Resolution of Generic Safety Issues, Rev. 34, U.S. Nuclear Regulatory Commission, September 2011, (includes Supplements 1-34).
NUREG-0933, Appendix B, Applicability of NUREG-0933 Issues to Operating and Future Reactor Plants, Rev. 25, U.S. Nuclear Regulatory Commission, September 2011.
NUREG-1061, Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Potential for Pipe Breaks, Volume 3, U.S. Nuclear Regulatory Commission, November 1984 (ML093170485).
NUREG-1061, Report of the U.S. Nuclear Regulatory Commission Piping Review Committee, Evaluation of Other Loads and Load Combinations, Volume 4, U.S. Nuclear Regulatory Commission, December 1984.
B-33
NUREG-1116, A Review of the Current Understanding of the Potential for Containment Failure from In-Vessel Steam Explosions, U.S. Nuclear Regulatory Commission, June 1985.
NUREG-1122, Knowledge and Abilities Catalog for Nuclear Power Plant Operators:
Pressurized Water Reactors, Revision 2, Supplement 1, October 2007 (ML071580631).
NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, U.S. Nuclear Regulatory Commission, December 1990.
NUREG-1335, Individual Plant Examination: Submittal Guidance, U.S. Nuclear Regulatory Commission, August 1989.
NUREG-1342, A Status Report Regarding Industry Implementation of Safety Parameter Display Systems, April 1989 (ML090060858).
NUREG-1367, Functional Capability of Piping Systems, U.S. Nuclear Regulatory Commission, November 1992 (ML083510056).
NUREG-1407, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities Final Report, U.S. Nuclear Regulatory Commission, June 1991.
NUREG-1417, Safety Evaluation Report: Related to Hydrogen Control Owners Group Assessment of Mark III Containments, U.S. Nuclear Regulatory Commission, October 1990.
NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, U.S. Nuclear Regulatory Commission, September 1993.
NUREG-1462, Final Safety Evaluation Report Related to the Certification of the System 80+
Design, Volume 1 (Docket No.52-002,) U.S. Nuclear Regulatory Commission, August 1994 (ML100780157).
NUREG-1462, Final Safety Evaluation Report Related to the Certification of the System 80+
Design, Volume 2 (Docket No.52-002,) U.S. Nuclear Regulatory Commission, August 1994 (ML100430017).
NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, U.S. Nuclear Regulatory Commission, February 1995.
NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, August 2011.
NUREG-1524, A Reassessment of the Potential for an Alpha-Mode Containment Failure and a Review of the Current Understanding of Broader Fuel-Coolant Interaction Issues, U.S. Nuclear Regulatory Commission, August 1996.
NUREG-1570, Risk Assessment of Severe Accident-Induced Steam Generator Tube Rupture, U.S. Nuclear Regulatory Commission, March 1998.
NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report, April 2001 (ML011080726)
B-34
NUREG-1829, Estimating Loss-of-Coolant Accident (LOCA) Frequencies Through the Elicitation Process, Volume 1, April 2008.
NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines, U.S. Nuclear Regulatory Commission, November 2009.
NUREG-1935, State-of-the-Art Reactor Consequence Analyses (SOARCA) Report, November 2012.
NUREG-2169, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database, U.S. Nuclear Regulatory Commission, January 2015.
Regulatory Guides Regulatory Guide 1.8, Qualification and Training of Personnel for Nuclear Power Plants, Rev.
3, U.S. Nuclear Regulatory Commission, May 2000.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition), U.S. Nuclear Regulatory Commission, June 2007.
Regulatory Guide 1.157, Best Estimate Calculations of Emergency Core Cooling System Performance.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants.
NRC Regulatory Guide 1.54, Revision 2, Service Level I, II, and III Protective Coatings Applied to Nuclear Power Plants, U. S. Nuclear Regulatory Commission, October 2010.
Regulatory Guide 1.53, Application of the Single-Failure Criterion to Safety Systems, Revision 2, November 2003.
Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, U.S. Nuclear Regulatory Commission, March 2009.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission, June 2007.
Regulatory Guide 1.102, Flood Protection for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, September 1976.
Regulatory Guide 1.216, Containment Structural Integrity Evaluation for Internal Pressure Loadings above Design Basis Pressure, U.S. Nuclear Regulatory Commission, August 2010.
Regulatory Guide 1.7, Control of Combustible Gas Concentrations in Containment, August 2013 (ML18018B335).
B-35
Regulatory Guide 1.9, Application and Testing of Safety-Related Diesel Generators in Nuclear Power Plants, March 2007 (ML070380553.)
Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, October 2015 (ML15274A351).
Regulatory Guide 1.20, Comprehensive Vibration Assessment Programs for Internals during Preoperational and Initial Startup Testing, March 2007 (ML070260376).
Regulatory Guide 1.28, Quality Assurance Program Criteria (Design and Construction), June 2010, (ML100160003).
Regulatory Guide 1.45, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, May 2008 (ML073200271).
Regulatory Guide 1.52, Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Post-Accident Engineered-Safety-Feature Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, September 2012 (ML12159A013).
Regulatory Guide 1.68, Initial Test Programs for Water-Cooled Nuclear Power Plants, March 2007 (ML070260039).
Regulatory Guide 1.69, Concrete Radiation Shields and Generic Shield Testing for Nuclear Power Plants, May 2009 (ML090820425).
Regulatory Guide 1.75, Physical Independence of Electric Systems, February 2005, (ML043630448).
Regulatory Guide 1.79, Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors, September 1975 (ML003740351).
Regulatory Guide 1.118, Periodic Testing of Electric Power and Protection Systems, August 2012 (ML17082A552).
Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, June 2003 (ML031530505).
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants, June 2007 (ML070720184).
Regulatory Guide 1.215, Guidance for ITAAC Closure Under 10 CFR Part 52," July 2015 (ML15105A447).
Regulatory Guide 5.7, Entry/Exit Control for Protected Areas, Vital Areas, and Material Access Areas, April 2015 (ML15099A080).
Regulatory Guide 5.65, Vital Area Access Controls, Protection of Physical Security Equipment, and Key and Lock Controls, September 2015 (ML15271A155).
Regulatory Guide 5.79, Protection of Safeguards Information, April 2011 (ML103270219).
Regulatory Guide 8.38, Control of Access to High and Very High Radiation Areas in Nuclear Power Plants, May 2006 (ML061350096).
B-36
Regulatory Guide 1.47, Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems, February 2010 (ML092330064).
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants, June 2007 (ML070720184).
Risk Assessment Results for Risk-Informed Activities, Revision 2, U.S. Nuclear Regulatory Commission, March 2009.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission, June 2007.
Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Regulatory Guide 1.102, Flood Protection for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, September 1976.
Regulatory Guide 1.216, Containment Structural Integrity Evaluation for Internal Pressure Loadings above Design Basis Pressure, U.S. Nuclear Regulatory Commission, August 2010.
RG 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, USNRC, Rev. 1, December 2001.
RG 1.91, Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants, U.S. NRC, Rev. 2, April 2013.
RG 1.115, Protection against Low-Trajectory Turbine Missiles, U.S. NRC, Rev. 2, January 2012.
RG 1.78, Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release, USNRC, Rev. 1, December 2001.
RG 1.91, Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants, U.S. NRC, Rev. 2, April 2013.
RG 1.115, Protection against Low-Trajectory Turbine Missiles, U.S. NRC, Rev. 2, January 2012.
Regulatory Guide 1.8, Qualification and Training of Personnel for Nuclear Power Plants, Rev.
3, U.S. Nuclear Regulatory Commission, May 2000.
Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Rev. 3, U.S.
Nuclear Regulatory Commission, June 2013.
Regulatory Guide 1.101, Emergency Planning and Preparedness for Nuclear Power Reactors, Rev. 5, U.S. Nuclear Regulatory Commission, June 2005.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition), U.S. Nuclear Regulatory Commission, June 2007.
Regulatory Guide 1.101, Emergency Planning and Preparedness for Nuclear Power Reactors, Rev. 5, U.S. Nuclear Regulatory Commission, June 2005.
B-37
Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Rev. 3, U.S.
Nuclear Regulatory Commission, June 2013.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition), U.S. Nuclear Regulatory Commission, June 2007.
Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Rev. 4, U.S. Nuclear Regulatory Commission, June 2006.
Regulatory Guide 1.152, Criteria for Use of Computers in Safety Systems of Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, July 2011.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition), U.S. Nuclear Regulatory Commission, June 2007.
Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Rev. 4, U.S. Nuclear Regulatory Commission, June 2006.
Regulatory Guide 1.152, Criteria for Use of Computers in Safety Systems of Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, July 2011.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition), U.S. Nuclear Regulatory Commission, June 2007.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants (LWR Edition), U.S. Nuclear Regulatory Commission, June 2007.
RG 1.37, Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power Plants RG 1.54, Service level I, II, and III Protective Coatings Applied to Nuclear Power Plants Regulatory Guide 1.29, Seismic Design Classification, Rev. 4, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, November 2001.
Regulatory Guide 1.151, Instrument Sensing Lines, Rev. 1, U.S. Nuclear Regulatory Commission, July 2010.
Regulatory Guide 1.189, Fire Protection for Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, October 2009.
Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants, Rev. 4, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.117, Tornado Design Classification, Rev. 1, U.S. Nuclear Regulatory Commission, April 1978.
B-38
Regulatory Guide 1.76, Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.221, Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants, U.S Nuclear Regulatory Commission, October 2011.
Regulatory Guide 1.59, Design Basis Floods for Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, August 1977.
Regulatory Guide 1.102, Flood Protection for Nuclear Power Plants, Rev. 1, Nuclear Regulatory Commission, September 1976.
Regulatory Guide 1.76, Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, February 2007.
Regulatory Guide 1.221, Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, October 2011.
Regulatory Guide 1.206, Combined License Applications for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission, June 2007.
Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Rev. 2, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.27, Ultimate Heat Sink, Rev. 2, U.S. Nuclear Regulatory Commission, January 1976.
Regulatory Guide 1.115, Protection Against Low-Trajectory Turbine Missiles, Rev. 2, U.S.
Nuclear Regulatory Commission, January 2012.
Regulatory Guide 1.117, Tornado Design Classification, Rev. 1, U.S. Nuclear Regulatory Commission, April 1978.
Regulatory Guide 1.142, Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor Vessels and Containments), Rev. 2, U.S. Nuclear Regulatory Commission, November 2001.
Regulatory Guide 1.11, Instrument Lines Penetrating Primary Containment, Rev. 1, U.S.
Nuclear Regulatory Commission, March 2010.
Regulatory Guide 1.45, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, Rev. 1, U.S. Nuclear Regulatory Commission, May 2008.
Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, July 2014.
Regulatory Guide 1.208, A Performance-based Approach to Define the Site-specific Earthquake Ground Motion, Rev. 4, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, March 2007.
B-39
Regulatory Guide 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-supported Equipment or Components, Rev. 1, U.S. Nuclear Regulatory Commission, February 1978.
Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Rev. 3, U.S. Nuclear Regulatory Commission, October 2012.
Regulatory Guide 1.12, Nuclear Power Plant Instrumentation for Earthquakes, Rev. 2, U.S.
Nuclear Regulatory Commission, March 1997.
Regulatory Guide 1.166, Pre-earthquake Planning and Immediate Nuclear Power Plant Operator Postearthquake Actions, U.S. Nuclear Regulatory Commission, March 1997.
Regulatory Guide 1.167, Restart of a Nuclear Power Plant Shut Down by a Seismic Event, U.S. Nuclear Regulatory Commission, March 1997.
Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, July 2014.
Regulatory Guide 1.100, Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, September 2009.
Regulatory Guide 1.35.1, Determining Prestressing Forces for Inspection of Prestressed Concrete Containments, U.S. Nuclear Regulatory Commission, July 1990.
Regulatory Guide 1.136, Design Limits, Loading Combinations, Materials, Construction, and Testing of Concrete Containments, Rev. 3, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.7, Control of Combustible Gas Concentrations in Containment, Rev. 3, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.57, Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components, Rev. 2, U.S. Nuclear Regulatory Commission, May 2013.
Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, July 2014.
Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Rev. 3, U.S. Nuclear Regulatory Commission, October 2012.
Regulatory Guide 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components, Rev. 1, U.S. Nuclear Regulatory Commission, February 1978.
Regulatory Guide 1.142, Safety-Related Concrete Structures for Nuclear Power Plants, Rev.
2, U.S. Nuclear Regulatory Commission, November 2001.
Regulatory Guide 1.199, Anchoring Components and Structural Supports in Concrete, U.S.
Nuclear Regulatory Commission, November 2003.
B-40
Regulatory Guide 1.29, Seismic Design Classification, Rev. 4, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.69, Concrete Radiation Shields and Generic Shield Testing for Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, May 2009.
Regulatory Guide 1.91, Evaluations of Explosions Postulated to Occur on Transportation Routes Near Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, April 2013.
Regulatory Guide 1.115, Protection Against Low-Trajectory Turbine Missiles, Rev. 2, U.S.
Nuclear Regulatory Commission, January 2012.
Regulatory Guide 1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants, Rev. 2, U.S. Nuclear Regulatory Commission, November 2001.
Regulatory Guide 1.160, Monitoring The Effectiveness of Maintenance at Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, May 2012.
Regulatory Guide 1.127, Inspection of Water-Control Structures associated with Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, March 1978.
Regulatory Guide 1.136, Design Limits, Loading Combinations, Materials, Construction, and Testing of Concrete Containments, Rev. 3, U.S Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.221, Design-Basis Hurricane and Hurricane Missiles for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, October 2011.
Regulatory Guide 1.216, Containment Structural Integrity Evaluation for Internal Pressure Loadings Above Design-Basis Pressure, U.S. Nuclear Regulatory Commission, August 2010.
Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Rev. 3, U.S. Nuclear Regulatory Commission, October 2012.
Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.20, Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing, Rev. 3, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, June 1984.
Regulatory Guide 1.73, "Qualification Tests for Safety-Related Actuators in Nuclear Power Plants," U.S. Nuclear Regulatory Commission, October 2013.
Regulatory Guide 1.124, Service Limits and Loading Combinations for Class 1 Linear-Type Supports, Rev. 3, U.S. Nuclear Regulatory Commission, July 2013.
Regulatory Guide 1.130, Service Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Supports, Rev. 3, U.S. Nuclear Regulatory Commission, July 2013.
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Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code, Rev. 1, U.S. Nuclear Regulatory Commission, June 2014.
Regulatory Guide 1.68, Initial Test Programs for Water-Cooled Nuclear Power Plant, Rev. 4, U.S. Nuclear Regulatory Commission, June 2013.
Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Rev. 4, U.S. Nuclear Regulatory Commission, June 2006.
Regulatory Guide 1.100, Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, September 2009.
Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.9, Application and Testing of Safety-Related Diesel Generators in Nuclear Power Plants, Rev. 4, U.S. Nuclear Regulatory Commission, June 2007.
Regulatory Guide 1.89, Environmental Qualification of Certain Electrical Equipment Important to Safety for Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, June 1984.
Regulatory Guide 1.40, Qualification of Continuous Duty Safety-Related Motors for Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, February 2010.
Regulatory Guide 1.63, Electric Penetration Assemblies in Containment Structures for Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, February 1987.
Regulatory Guide 1.73, Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, October 2013.
Regulatory Guide 1.97, Criteria for Accident Monitoring Instrumentation for Nuclear Power Plants, Rev.4, U.S. Nuclear Regulatory Commission, June 2006.
Regulatory Guide 1.156, Qualification of Connection Assemblies for Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, July 2011.
Regulatory Guide 1.158, Qualification of Safety-Related Lead Storage Batteries for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, February 1989.
Regulatory Guide 1.180, Guidelines for Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related Instrumentation and Control Systems, Rev. 1, U.S. Nuclear Regulatory Commission, October 2003.
Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, U.S. Nuclear Regulatory Commission, July 2000.
Regulatory Guide 1.209, Guidelines for Environmental Qualification of Safety-Related Computer-Based Instrumentation and Control Systems in Nuclear Power Plants, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.210, Qualification of Safety-Related Battery Chargers and Inverters for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, June 2008.
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Regulatory Guide 1.211, Qualification of Safety-Related Cables and Field Splices for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, April 2009.
Regulatory Guide 1.213, Qualification of Safety-Related Motor Control Centers for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, May 2009.
Regulatory Guide 1.84, Design, Fabrication, and Materials Code Case Acceptability, Rev. 36, U.S. Nuclear Regulatory Commission, August 2014.
Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, Rev. 1, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.122, Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components, Rev. 1, U.S. Nuclear Regulatory Commission, February 1978.
Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Rev. 3, U.S. Nuclear Regulatory Commission, October 2012.
Regulatory Guide 1.207, Guidelines for Evaluating Fatigue Analyses incorporating the Life Reduction of Metal Components Due to the Effects of the Light Water Reactor Environment for New Reactors, U.S. Nuclear Regulatory Commission, March 2007.
Regulatory Guide 1.199, Anchoring Components and Structural Supports in Concrete, U.S.
nuclear Regulatory Commission, November 2003.
Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, May 2012.
Regulatory Guide 1.28, Quality Assurance Program Criteria (Design and Construction), Rev.
4, U.S. Nuclear Regulatory Commission, June 2010.
Regulatory Guide 1.100, Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants, Rev. 3, U.S. Nuclear Regulatory Commission, September 2009.
Regulatory Guide 1.151, Instrument Sensing Lines, Rev. 1, U.S. Nuclear Regulatory Commission, July 2010.
RG 1.29, Seismic Design Classification RG 1.45, Guidance on Monitoring and Responding to Reactor Coolant System Leakage, Safety Evaluation Reports TAC No. 01142, Safety Evaluation Report, "CESEC Digital Simulation of a Combustion Engineering Nuclear Steam Supply System," Enclosure to Letter from C.O. Thomas (NRC) to A.E. Scherer (CE), NRC, April 3, 1984.
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Safety Evaluation Report for APR1400-F-C-TR-12002-P, KCE-1 Critical Heat Flux Correlation for PLUS7' Thermal Design Topical Report, (ML17348A152).
Safety Evaluation for Topical Report APR1400-F-A-TR-12004-P/NP, Revision 1, Realistic Evaluation Methodology for Large-Break LOCA of the APR1400 for Safety Evaluation, June 5, 2018. (ML18156A042).
Safety Evaluation for Topical Report APR1400-F-A-TR-12004-P/NP, Revision 1, Realistic Evaluation Methodology for Large-Break LOCA of the APR1400 for Safety Evaluation, June 5, 2018 (ML18156A042).
U.S. Code of Federal Regulations (CFR) 10 CFR Part 50, Appendix A, General Design Criterion 60, Control of Release of Radioactive Materials to the Environment, U.S. Nuclear Regulatory Commission.
10 CFR Part 20, Appendix B, Annual Limits on Intake (ALIs) and Derived Air Concentrations (DACs) of Radionuclides for Occupational Exposure; Effluent Concentrations; Concentrations for Release to Sewerage, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion As Low As is Reasonably Achievable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents, U.S. Nuclear Regulatory Commission.
10 CFR 50.34, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix A, General Design Criterion 54, Systems Penetrating Containment, U.S. Nuclear Regulatory Commission.
10 CFR 50.62, Requirements for Reduction of Risk from Anticipated Transients without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants, U.S. Nuclear Regulatory Commission.
10 CFR 73.2, Definitions, U.S. Nuclear Regulatory Commission.
10 CFR 73.55, Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage, U.S. Nuclear Regulatory Commission.
10 CFR 73.21, Protection of Safeguards Information: Performance Requirements, U.S.
Nuclear Regulatory Commission.
10 CFR Part 26, Fitness for Duty Programs, U.S. Nuclear Regulatory Commission.
10 CFR Part 26, Fitness for Duty Programs, U.S. Nuclear Regulatory Commission.
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10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, U.S. Nuclear Regulatory Commission.
10 CFR 50.54, Conditions of Licenses, U.S. Nuclear Regulatory Commission.
10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, U.S.
Nuclear Commission.
10 CFR 52.79, Contents of Applications; Technical Information in Final Safety Analysis Report, U.S. Nuclear Regulatory Commission.
10 CFR Part 55, Operators Licenses, U.S. Nuclear Regulatory Commission.
10 CFR 73.2, Definitions, U.S. Nuclear Regulatory Commission.
10 CFR 73.21, Protection of Safeguards Information: Performance Requirements, U.S.
Nuclear Regulatory Commission.
10 CFR 73.55, Requirements for Physical Protection of Licensed Activities in Nuclear Power Reactors Against Radiological Sabotage, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Commission.
10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, U.S.
Nuclear Commission.
10 CFR 52.79, Contents of Applications; Technical Information in Final Safety Analysis Report, U.S. Nuclear Regulatory Commission.
10 CFR 50.54, Conditions of Licenses, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, U.S. Nuclear Regulatory Commission.
General Design Criteria for Nuclear Power Plants, Title 10, U.S. Code of Federal Regulations, 10 CFR Part 50, Appendix A.
10 CFR 52.47, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission, November 2012.
10 CFR 50.49, Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, November 2012.
10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission, November 2012.
10 CFR 52.47, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission, November 2012.
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10 CFR 50.49, Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, November 2012.
10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission, November 2012.
10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Regulatory Commission.
10 CFR Part 55, Operators Licenses, U.S. Nuclear Regulatory Commission.
10 CFR Part 55, Operators Licenses, U.S. Nuclear Regulatory Commission.
10 CFR 52.47, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission.
10 CFR 50.34(f),Additional TMI-related Requirements, U.S. Nuclear Regulatory Commission.
10 CFR Part 21, Reporting of Defects and Noncompliance, U.S. Nuclear Regulatory Commission.
10 CFR Part 21, Reporting of Defects and Noncompliance, U.S. Nuclear Regulatory Commission.
10 CFR 50.34, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission.
10 CFR 50.34(f),Additional TMI-related Requirements, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix A, General Design Criterion 54, Systems Penetrating Containment, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix A, General Design Criterion 60, Control of Release of Radioactive Materials to the Environment, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix I, Numerical Guides for Design Objectives and Limiting Conditions for Operation to Meet the Criterion As Low As is Reasonably Achievable for Radioactive Material in Light-Water-Cooled Nuclear Power Reactor Effluents, U.S. Nuclear Regulatory Commission.
10 CFR 50.62, Requirements for Reduction of Risk from Anticipated Transients without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants, U.S. Nuclear Regulatory Commission.
10 CFR 52.47, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission.
10 CFR 52.47, Contents of Applications; Technical Information, U.S. Nuclear Regulatory Commission.
10 CFR 50.55a, Codes and Standards, U.S. Nuclear Regulatory Commission.
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10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission.
10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, U.S. Nuclear Regulatory Commission.
10 CFR 100.11, Determination of Exclusion Area, Low Population Zone, and Population Center Distance, U.S. Nuclear Regulatory Commission.
10 CFR Part 100, Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission.
10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Section 50.55a, Codes and Standards, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Regulatory Commission.
10 CFR Part 52 License, Certifications, and Approvals for Nuclear Power Plant, U.S. Nuclear Regulatory Commission.
10 CFR Part 100 Reactor Site Criteria, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Regulatory Commission.
10 CFR Part 100, Reactor Site Criteria, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix A, General Design Criterion 2, Design Bases for Protection Against Natural Phenomena, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission.
10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants, U.S.
Nuclear Regulatory Commission.
10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, U.S. Nuclear Regulatory Commission.
10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, U.S. Nuclear Regulatory Commission.
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NRC Letters and Memorandums Letter from R.L. Baer (NRC) to A.E. Scherer (CE) dated September 6, 1978.
Letter from R.L. Tedesco (NRC) to A.E. Scherer (CE), Acceptance for Referencing of Topical Report CENPD-206(P), TORC Code Verification and Simplified Modeling Methods, NRC, December 11, 1980.
Letter from R.A. Clark (NRC) to W. Cavanaugh III (AP&L), "Operation of ANO-2 During Cycle 2," NRC, July 21, 1981 (Safety Evaluation and Amendment No. 26 to Facility Operating License No. NPF-6 for ANO-2).
Letter from Olan D. Parr (NRC) to F.M. Stern (CE), NRC, December 4, 1974.
LD-WO-3900, Letter from C.B. Brinkman (NRC) to A.E. Scherer (CE), "Macbeth CHF Correlation Approval," NRC, August 2, 1983.
Letter from Olan D. Parr (NRC) to F.M. Stern (CE), NRC, June 13, 1975.
Letter from K. Kniel (NRC) to A.E. Scherer (CE) dated November 12, 1976.
Letter from Olan D. Parr (NRC) to A.E. Scherer (CE), NRC, June 10, 1976.
Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 61 to Facility Operating License No. NPF-41. Arizona Public Service, et al., Palo Verde Nuclear Generating Station, Unit No. 1, Docket No. STN 50-528, NRC, April 3, 1992, (ML021700246).
Letter from K. Kniel (NRC) to A.E. Scherer (CE) dated September 27, 1977.
Letter from Dennis M. Crutchfield (NRC) to A.E. Scherer (CE) dated July 31, 1986.
Letter from Robert L. Baer (NRC) to A.E. Scherer (CE), NRC, July 30, 1979.
NRC letter from Robert A. Gramm to James A. Gresham, Westinghouse Electric Co.,
Suspension of NRC Approval for Use of Westinghouse Topical Report CENPD-254-P, "Post-LOCA Long-Term Cooling Model, Due To Discovery Of Non-Conservative Modeling Assumptions During Calculations Audit, August 1, 2005.
NRC letter from Daniel S. Collins to James A. Gresham, Westinghouse Electric Co.,
Clarification of NRC Letter Dated August 1, 2005, Suspension of NRC Approval for Use of Westinghouse Topical Report CENPD-254-P, "Post-LOCA Long-Term Cooling Model," Due to Discovery of Non-Conservative Modeling Assumptions During Calculations Audit, November 23, 2005.
Letter, O.D. Parr (NRC) to F.M. Stern (CE), June 13, 1975.
Letter, O.D. Parr (NRC) to A.E. Scherer (CE), December 9, 1975.
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Letter, Karl Kniel (NRC) to A.E. Scherer (CE), September 27, 1977.
Letter, D.M. Crutchfield (NRC) to A.E. Scherer (CE), July 31, 1986.
Nuclear Regulatory Commission Inspection of Korea Hydro & Nuclear Power, Ltd., Report No.
05200046/206-201, April 8, 2016 (ML16081A081).
U.S. Nuclear Regulatory Commission, CCF Parameter Estimations, 2010 Update, http://nrcoe.inl.gov/results/CCF/ParamEst2010/ccfparamest.htm, January 2012.
Nuclear Regulatory Commission (U.S.) (NRC). (ML061580448).
Memorandum L. Kopp to T. Collins, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, dated August 19, 1998 (ML11088A013)
BulletinsBulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant System, U.S.
Nuclear Regulatory Commission, June 22, 1988.
Bulletin 79-13, Cracking in Feedwater System Piping, U.S. Nuclear Regulatory Commission, June 25, 1979.
Bulletin 88-11, Pressurizer Surge Line Thermal Stratification, U.S. Nuclear Regulatory Commission, December 20, 1988.
Bulletin 79-02, Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts, Rev. 2, U.S. Nuclear Regulatory Commission, November 18, 1979.
Bulletin 88-11, Pressurizer Surge Line Stratification Test, December 1988.
Other NRC Documents Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, U.S. Nuclear Regulatory Commission, February 2005.
Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, U.S. Nuclear Regulatory Commission, February 2005.
Generic Letter GL 84-15, Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability, July 1984 B-49
U.S. NRC Notice of Violation (NOV), NOV 99901453/2014-201-01(a), NOV 99901453/2014-201-01(b), NOV 99901453/2014-201-03, and NOV 99901453/2014-201-04(b) [ML14302A743].
Appendix A.
Regulatory Audit Plan for Topical Report APR1400-F-A-TR-12004-P, Realistic Evaluation Methodology for Large Break LOCA of the APR1400, (ML15208A199).
U.S. Nuclear Regulatory Commission, CCF Parameter Estimations, 2010 Update, http://nrcoe.inl.gov/results/CCF/ParamEst2010/ccfparamest.htm, January 2012.
ERI/NRC 16-208, Revision 2, Assessment of Combustible Gas Control during Severe Accidents in APR1400, (ML16314E431)
Order EA-12-049, Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events, dated March 12, 2012.
DC/COL-ISG-01, Interim Staff Guidance on Seismic Issues Associated with High Frequency Ground Motion in Design Certification and Combined License Applications, U.S Nuclear Regulatory Commission, 2009.
Generic Letter 90-09, Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions, U.S. Nuclear Regulatory Commission, December 11, 1990.
Regulatory Issue Summary 2000-03, Resolution of Generic Safety Issue 158: Performance of Safety-Related Power-Operated Valves Under Design Basis Conditions, U.S. Nuclear Regulatory Commission, March 15, 2000.
Generic Letter 89-10, Safety-Related Motor-Operated Valves, Testing and Surveillance, U.S.
Nuclear Regulatory Commission, June 28, 1989.
Generic Letter 96-05, Periodic Verification of Design Basis Capability of Safety-Related Motor-Operated Valves, U.S. Nuclear Regulatory Commission September 18, 1996.
Other Documents NFPA 70-2011, National Electrical Code (NEC)
NFPA 72-2007, National Fire Alarm Code MIL-STD-810F, Environmental Engineering Considerations and Laboratory Tests NEMA 250-2004, Enclosures for Electrical Equipment IEC61000-4-2-2008, Electromagnetic Compatibility (EMC) Testing and Measurement Techniques- Electrostatic Discharge Immunity Test IEC61000-4-3-2010, Electromagnetic Compatibility (EMC) Testing and Measurement Techniques- Radiated, Radio Frequency, Electromagnetic Field immunity Test B-50
IEC61000-4-5-2005, Electromagnetic Compatibility (EMC) Testing and Measurement Techniques - Surge Immunity Test IEC 61226, International Standard - Nuclear Power Plants - Instrumentation and Control Important to Safety - Classification of Instrumentation and Control Functions, International Electrotechnical Commission (IEC), Edition 3.0, 2009.
BNWL-1411, Experimental High Enthalpy Water Blowdown from a Simple Vessel through a Bottom Outlet, R.T. Allemann, et al., June, 1970.
BNWL-1463, Coolant Blowdown Studies of a Reactor Simulator Vessel Containing a Perforated Sieve Plate Separator, R.T. Allemann, et al., February 1971.
Entergy Letter W3F1-2005-0012 from Timothy G. Mitchell to USNRC, Supplement to Amendment Request NPF-38-249 Extended Power Uprate, Waterford Steam Electric Station, Unit 3, February 16, 2005.
The Structural Engineer, The Effect of Elevated Temperatures on the Strength Properties of Reinforcing and Prestressing Steels, M. Holmes, R.D. Anchor, G.M.E. Cook & R.N. Cook, Vol.60B, No.1, March 1982.
INL/EXT-15-37873, Analysis of Loss-of-Offsite-Power Events 1997-2014, February 2016.
NEA/CSNI/R(2000)7, Flame Acceleration and Deflagration-to-Detonation Transition in Nuclear Safety, OECD/NEA, October 2000.
M. T. Farmer et al., The CORQUENCH Code for Modeling of Ex-Vessel Corium Coolability under Top Flooding Conditions - Code Manual-Version 3.03, Revision 2, OECD/MCCI-2010-TR03, OECD/NEA, January 2011.
DOE/ID-10271, Prevention of Early Containment Failure due to High Pressure Melt Ejection and Direct Containment Heating for Advanced Light Water Reactors, March 1, 1990.
NEA/CSNI/R(97)30, OECD/CSNI Specialist Meeting on Fuel-Coolant Interactions, OECD/NEA, May 1997.
M. L. Corradini et al., A Users Manual for TEXAS-V: A One Dimensional Transient Fluid Model for Fuel-Coolant Interaction Analysis, UW Nuclear Engineering and Engineering Physics, August 2000.
R. H. Cole, Underwater Explosions, Princeton University Press, 1948.
Bureau of Labor Statistics' Producer Price Index for the commodity of Electric Power (BLS 2011l Producer Price Index-Commodities: Series Id: WPU054 2012/1993) (retrieved March 21, 2013).
NUMARC 87-00, Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors, Regulatory Issue Summary 2008-05, Revision 1, Lessons Learned to Improve Inspections, Tests, Analyses, and Acceptance Criteria Submittal, September 2010 (ML102500244).
E-P-NU-907-1.3/DC, SAREX 1.3 Software Registration, KEPCO E&C, July 2013.
B-51
E-P-NU-1341-1.6/DC, FTREX 1.6 Software Registration, KEPCO E&C, July 2013.
The Structural Engineer, The Effect of Elevated Temperatures on the Strength Properties of Reinforcing and Prestressing Steels, M. Holmes, R.D. Anchor, G.M.E. Cook & R.N. Cook, Vol.60B, No.1, March 1982.
IEC 61226, International Standard - Nuclear Power Plants - Instrumentation and Control Important to Safety - Classification of Instrumentation and Control Functions, International Electrotechnical Commission (IEC), Edition 3.0, 2009.
INL/EXT-15-37873, Analysis of Loss-of-Offsite-Power Events 1997-2014, February 2016.
NEA/CSNI/R(2000)7, Flame Acceleration and Deflagration-to-Detonation Transition in Nuclear Safety, OECD/NEA, October 2000.
M. T. Farmer et al., The CORQUENCH Code for Modeling of Ex-Vessel Corium Coolability under Top Flooding Conditions - Code Manual-Version 3.03, Revision 2, OECD/MCCI-2010-TR03, OECD/NEA, January 2011.
DOE/ID-10271, Prevention of Early Containment Failure due to High Pressure Melt Ejection and Direct Containment Heating for Advanced Light Water Reactors, March 1, 1990.
NEA/CSNI/R(97)30, OECD/CSNI Specialist Meeting on Fuel-Coolant Interactions, OECD/NEA, May 1997.
M. L. Corradini et al., A Users Manual for TEXAS-V: A One Dimensional Transient Fluid Model for Fuel-Coolant Interaction Analysis, UW Nuclear Engineering and Engineering Physics, August 2000.
R. H. Cole, Underwater Explosions, Princeton University Press, 1948.
Bureau of Labor Statistics' Producer Price Index for the commodity of Electric Power (BLS 2011l Producer Price Index-Commodities: Series Id: WPU054 2012/1993) (retrieved March 21, 2013).
NUMARC 87-00, Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors, Computer code ABAQUS Version 6.10-1; Installed on DELL Workstation with Windows Server 2008; Verification Document No. 00000-SM-VV-038, January 2011.
Computer code, CLEVER Version 1.0; Computers with Windows XP O/S; Verification Document No. 00000-SM-VV-037, Rev. 01, Oct. 2012.
Computer code, HeadPR Version 1; Computers with Windows XP, Windows 2000, Windows 7 O/S; Verification Document No. ND-G-CV-033, Rev. 2, October 2014.
Doherty, P. K., Software Verification and Validation Report of CEFLASH-4B, Version f4b.1.1, VV-FF-0178, Rev. 1, January 1995.
Computer code, ANSYS Version 12.1; Installed on IBM P6 570 24Core; Verification Document No. DAVM121, December 2010.
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Computer code, AFP2D Version 3; Installed on IBM P6 570 24Core; Verification Document No.
ND-G-CV-019, Rev. 8, December 2014.
Computer code, TSPOST Version 0; Installed on IBM P6 570 24Core; Verification Document No. ND-G-CV-018, Rev. 4, October 2014.
Computer code, AFPOST Version 2; Installed on IBM P6 570 24Core; Verification Document No. ND-G-CV-027, Rev. 7, December 2014.
Computer code, ATHOS3 Mod-01; Installed on IBM P6 570 24Core; Verification Document No.
ND-G-CV-017, Rev. 3, October 2014.
Computer code, PTXIG Version 1.0; Computers with Microsoft.Net 2.0 O/S; Verification Document No. 00000-RM-VV-002, Rev. 02, Sep, 2012.
DST Computer Services SA, a nuclear and non-nuclear piping analysis program, PIPESTRESS Version 3.9.0, Geneva, Switzerland, 2016.
REFORC-DEC User Manual, REF 03.7.483-1.0, Rev. 1, D.J. Pichurski, S&L, 21 January 1994.
RELAP5/MOD3.1, Transient Hydraulic Analysis Program, Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho, USA.
Computer code, NOZPROG Version 1; Installed on IBM P6 570 24Core; Verification Document No. ND-G-CV-006, Rev. 10, October 2014.
Computer code ANSYS Release 10.0 (2005), 12.0 (2009), 13.0 (2010), 14.0 (2011), 14.5 (2012) and 15.0 (2013), ANSYS, Inc.
Dynamic Stress Analysis of Axisymmetric Structures under Arbitrary Loading, Ghosh, S. and Wilson, E., EERC 69-10, University of California, Berkeley, September 1969.
Gabrielson, V. K., SHOCK, A Computer Code to Solve the Dynamic Response of Lumped-Mass Systems, SCL-DR-69-98, November 1969.
E. Simiu and R. H. Scanlan, Wind Effects on Structures: Fundamentals and Applications to Design, John Wiley & Sons, Inc., New York, 3rd Edition, 1996.
R. P. Kennedy, A Review of Procedures for the Analysis and Design of Concrete Structures to Resist Missile Impact Effects, Nuclear Engineering and Design, Volume 37, Number 2, 183-203, 1976.
Russell, C.R., Reactor Safeguards, MacMillan, 1962.
R.T. Lahey, Jr., and F.J. Moody, Pipe Thrust and Jet Loads, The Thermal Hydraulics of a Boiling Water Nuclear Reactor, Subsection 9.2.3, pp. 375-409, Published by American Nuclear Society, Prepared by the Division of Technical Information, United States Energy Research and Development Administration, 1977.
ACS SASSI NQA Version 2.3.0, An Advanced Computational Software for 3D Dynamic Analysis Including Soil-Structure Interactim, Including Option A and NQA Option FS, User Manual, Rev. 5, Ghiocel Predictrue Technologies, Inc., January 2012.
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ANSYS, Advanced Analysis Techniques Guide, Release 14.0, ANSYS Inc., November 2011.
GTSTRUDL Version 31, Analysis GTSTRUDL User Guide, Rev. 7, Georgia Institute of Technology, August 2010.
ANSYS, Advanced Analysis Techniques Guide, Release 14.0, ANSYS Inc., November 2011.
GTSTRUDL Version 31, GTSTRUDL User Guide, Georgia Institute of Technology, August 2010.
Research Council on Structural Connections, Specification for Structural Joints Using ASTM A325 or A490 Bolts, 2004.
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