Handbook of Generation IV Nuclear Reactors

 
 
Woodhead Publishing
  • 1. Auflage
  • |
  • erschienen am 9. Juni 2016
  • |
  • 940 Seiten
 
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978-0-08-100162-2 (ISBN)
 

Handbook of Generation IV Nuclear Reactors presents information on the current fleet of Nuclear Power Plants (NPPs) with water-cooled reactors (Generation III and III+) (96% of 430 power reactors in the world) that have relatively low thermal efficiencies (within the range of 32 36%) compared to those of modern advanced thermal power plants (combined cycle gas-fired power plants - up to 62% and supercritical pressure coal-fired power plants - up to 55%).

Moreover, thermal efficiency of the current fleet of NPPs with water-cooled reactors cannot be increased significantly without completely different innovative designs, which are Generation IV reactors. Nuclear power is vital for generating electrical energy without carbon emissions.

Complete with the latest research, development, and design, and written by an international team of experts, this handbook is completely dedicated to Generation IV reactors.

  • Presents the first comprehensive handbook dedicated entirely to generation IV nuclear reactors
  • Reviews the latest trends and developments
  • Complete with the latest research, development, and design information in generation IV nuclear reactors
  • Written by an international team of experts in the field
  • Englisch
  • Kent
Elsevier Science
  • 28,72 MB
978-0-08-100162-2 (9780081001622)
0081001622 (0081001622)
weitere Ausgaben werden ermittelt
  • Front Cover
  • Handbook of Generation IV Nuclear Reactors
  • Related titles
  • Handbook of Generation IV Nuclear Reactors
  • Copyright
  • Contents
  • List of contributors
  • Woodhead Publishing Series in Energy
  • Foreword
  • Preface
  • Nomenclature
  • Symbols
  • Subscripts
  • Acronyms/Abbreviations
  • 1 - Introduction: a survey of the status of electricity generation in the world*
  • 1.1 Statistics on electricity generation in the world
  • 1.2 Thermal power plants3
  • 1.3 Modern nuclear power plants5
  • 1.4 Conclusions
  • Abbreviations
  • Acknowledgments
  • References
  • One - Generation IV nuclear-reactor concepts
  • 2 - Introduction: Generation IV International Forum
  • 2.1 Origins of the Generation IV International Forum
  • 2.2 Generation IV goals
  • 2.3 Selection of Generation IV systems
  • 2.4 Six Generation IV nuclear energy systems
  • 2.4.1 Very high temperature reactor
  • 2.4.2 Gas-cooled fast reactor
  • 2.4.3 Sodium-cooled fast reactor
  • 2.4.4 Lead-cooled fast reactor
  • 2.4.5 Molten salt reactor
  • 2.4.6 Supercritical water-cooled reactors
  • 2.5 Summary
  • Acknowledgments
  • References
  • 3 - Very high-temperature reactor
  • 3.1 Development history and current status
  • 3.2 Technology overview
  • 3.2.1 Reactor design types
  • 3.2.2 Design features
  • 3.2.2.1 Safety
  • 3.2.2.2 Fuel cycle
  • 3.2.2.3 Multipurpose
  • 3.3 Detailed technical description
  • 3.3.1 Fuel design
  • 3.3.2 Fuel burnup
  • 3.3.2.1 Uranium fuel
  • 3.3.2.2 Plutonium fuel
  • 3.3.3 Reactor design
  • 3.3.3.1 Prismatic core reactor design
  • 3.3.3.2 Pebble bed core reactor design
  • 3.3.4 Reactor safety
  • 3.3.5 Plant design
  • 3.3.6 Plant operations
  • 3.3.6.1 Startup, rated operation, and shutdown
  • 3.3.6.2 Dynamic operation
  • 3.4 Applications and economics
  • 3.4.1 Power generation
  • 3.4.2 Cogeneration
  • 3.4.2.1 Hydrogen cogeneration
  • 3.4.2.2 Desalination cogeneration
  • 3.4.3 Industrial application
  • 3.4.4 Economics
  • 3.4.4.1 Cost of electricity generation
  • Capital cost
  • Operating cost
  • Fuel cost
  • Power generation cost
  • 3.4.4.2 Cost of hydrogen production
  • 3.4.4.3 Cost of desalination cogeneration
  • 3.5 Summary
  • Acronyms
  • References
  • 4 - Gas-cooled fast reactors
  • 4.1 Rationale and generational research and development bridge
  • 4.2 Gas-cooled fast reactor technology
  • 4.3 Conclusions
  • References
  • 5 - Sodium-cooled fast reactor
  • 5.1 Introduction
  • 5.2 Development history
  • 5.3 System characteristics
  • 5.3.1 Design features with sodium properties
  • 5.3.2 Core configurations
  • 5.3.3 Plant system
  • 5.3.4 Loop type and pool type
  • 5.3.5 Consistency with fuel cycle system (fuel cycle technology)
  • 5.4 Safety issues
  • 5.4.1 Safety design criteria and safety design guidelines
  • 5.4.2 Safety characteristics and safety design
  • 5.4.2.1 Reactor shutdown
  • 5.4.2.2 Decay heat removal
  • 5.4.2.3 Design measure against sodium chemical reactions
  • 5.4.2.4 Containment measures
  • 5.5 Future trends and key challenges
  • References
  • 6 - Lead-cooled fast reactor
  • 6.1 Overview and motivation for lead-cooled fast reactor systems
  • 6.2 Basic design choices
  • 6.2.1 Lead versus LBE
  • 6.2.2 Design choices for reactors with lead as the coolant
  • 6.2.3 Primary system concept: evolution and challenges
  • 6.2.3.1 Early conceptual designs derived from sodium-cooled fast reactor concepts
  • 6.2.3.2 Primary system development and current conceptual designs
  • 6.3 Safety principles
  • 6.4 Fuel technology and fuel cycles for the lead-cooled fast reactor
  • 6.4.1 Fuel assembly characteristics
  • 6.4.2 Fuel cycle for the lead-cooled fast reactor
  • 6.5 Summary of advantages and key challenges of the lead-cooled fast reactor
  • 6.5.1 Advantages of the lead-cooled fast reactor
  • 6.5.2 Key challenges of the lead-cooled fast reactor
  • 6.6 Overview of Generation IV lead-cooled fast reactor designs
  • 6.6.1 Reference Generation IV systems
  • 6.6.1.1 The European lead-cooled fast reactor
  • 6.6.1.2 The BREST-OD-300 reactor
  • 6.6.1.3 The Small Secure Transportable Autonomous Reactor
  • 6.6.2 Additional Generation IV systems under study or development
  • 6.6.2.1 The South Korea URANUS-40 system
  • 6.6.2.2 The Chinese CLEAR-I reactor
  • 6.6.2.3 The Pb-Bi-Cooled Direct Contact Boiling Water Fast Reactor
  • 6.6.2.4 The SVBR-100 (Zrodnikov et al., 2009
  • Toshinsky et al., 2013)
  • 6.7 Future trends
  • 6.7.1 The LFR-AS-200
  • 6.7.2 Swedish Advanced Lead Reactor
  • 6.7.3 Protection from lead corrosion by means of coatings
  • Sources of further information
  • Nomenclature
  • References
  • 7 - Molten salt fast reactors
  • 7.1 Introduction
  • 7.2 The molten salt fast reactor concept
  • 7.2.1 Core and system description
  • 7.2.2 Transient calculations
  • 7.3 Fuel salt chemistry and material issues
  • 7.3.1 Overview of the processing schemes
  • 7.3.2 Impact of the salt composition on the corrosion of the structural materials
  • 7.4 Molten salt fast reactor fuel cycle scenarios
  • 7.5 Safety issues
  • 7.5.1 Safety approach and risk analysis
  • 7.5.2 Liquid-fueled reactor specificities
  • 7.5.3 Decay heat removal
  • 7.5.4 Preliminary accidental transient identification
  • 7.6 Concept viability: issues and demonstration steps
  • 7.6.1 Identified limits
  • 7.6.2 Progression in safety demonstration and design optimization
  • 7.6.3 Presently ongoing laboratory-scale experiments
  • 7.6.4 Other research and development activities on molten salt systems
  • 7.7 Conclusion and perspectives
  • Nomenclature
  • Abbreviations and acronyms
  • Symbols
  • Acknowledgments
  • References
  • Bibliography web sources
  • 8 - Super-critical water-cooled reactors
  • 8.1 Introduction
  • 8.2 Types of supercritical water-cooled reactor concepts and main system parameters
  • 8.3 Example of a pressure vessel concept
  • 8.4 Example of a pressure tube concept
  • 8.5 Fuel cycle technology
  • 8.6 Fuel assembly concept
  • 8.6.1 High-performance light water reactor fuel assembly concept
  • 8.6.2 Fast reactor fuel assembly concept
  • 8.6.3 Canadian SCWR fuel assembly concept
  • 8.7 Safety system concept
  • 8.7.1 Safety system in a pressure vessel-type supercritical water-cooled reactor concept
  • 8.7.2 Safety system in the Canadian SCWR concept
  • 8.7.2.1 Containment pool
  • 8.7.2.2 Automatic depressurization system
  • 8.7.2.3 Gravity-driven core flooding system
  • 8.7.2.4 Isolation condensers
  • 8.7.2.5 Reserve water pool
  • 8.7.2.6 Atmospheric air heat exchangers
  • 8.7.2.7 Passive moderator cooling system
  • 8.8 Dynamics and control
  • 8.9 Start-up
  • 8.9.1 Start-up system in a pressure tube-type supercritical water-cooled reactor concept
  • 8.10 Stability
  • 8.11 Advantages and disadvantages of supercritical water-cooled reactor concepts
  • 8.12 Key challenges
  • 8.13 Future trends
  • Acronyms
  • Nomenclature
  • References
  • Two - Current status of Generation IV activities in selected countries
  • 9 - Generation IV: USA
  • 9.1 Generation IV program evolution in the United States
  • 9.2 Energy market in the United States and the potential role of Generation IV systems: electricity, process heat, and waste ma ...
  • 9.3 Electrical grid integration of Generation IV nuclear energy systems in the United States
  • 9.4 Industry and utilities interests in Generation IV nuclear energy systems in the United States
  • 9.5 Deployment perspectives for Generation IV systems in the United States and deployment schedule
  • 9.6 Conclusions
  • Abbreviations
  • References
  • 10 - Euratom research and training program in Generation-IV systems: breakthrough technologies to improve sustainability, s ...
  • 10.1 Background: Euratom (nuclear fission research and training) within the Energy Union (European Union energy mix policy)
  • 10.2 Generation-I, -II, -III, and -IV of nuclear fission reactors: research, development, and continuous improvement for more th ...
  • Short history of Generation-IV (GIF and IAEA/INPRO/approaches)
  • 10.3 "Goals for Generation-IV nuclear energy systems" and "technology roadmap" for the six GIF systems (2002 and 2013)
  • 10.4 "European sustainable nuclear industrial initiative" and Euratom research and training program in fast neutron reactor systems
  • NB: Historical reminder about the "European fast reactor" project (1984-93)
  • 10.5 Sustainability (efficient resource utilization and minimization of radioactive waste)
  • 10.6 Safety and reliability (maximum safety performance through design, technology, regulation, and culture)
  • 10.7 Socioeconomics (economic advantage over other energy sources and better governance structure in energy decision-making process)
  • 10.8 Proliferation resistance and physical protection (protection against all kinds of terrorism)
  • 10.9 Conclusion: a new way of "developing/teaching science," closer to the end-user needs of the 21-st century (society and indu ...
  • Nomenclature
  • Appendix: Tentative training scheme for preconceptual Generation-IV design engineers (knowledge, skills, attitudes)
  • Outline placeholder
  • Abstract
  • Bibliographic references:
  • A1. Introduction
  • A2. Trainees prerequisites
  • A3. Knowledge, skills, and attitudes required for a Generation-IV engineer
  • A4. Learning outcomes related to knowledge, skills, and attitudes for a Generation-IV engineer
  • A4.1 Learning outcomes in the knowledge area (learning to know)
  • A4.1.1 General knowledge on Generation-IV systems and technology
  • A4.1.2 Design specific knowledge for the lead fast reactor
  • A4.1.3 Design specific knowledge for the sodium fast reactor
  • A4.1.4 Design specific knowledge for the gas fast reactor
  • A4.1.5 Design specific knowledge for the very high temperature reactor
  • A4.1.6 Design specific knowledge for the super critical water reactor
  • A4.1.7 Design specific knowledge for the molten salt reactor
  • A4.2 Learning outcomes in the skills area (learning to do)
  • A4.3 Learning outcomes in the attitude area (learning to live together and/or learning to be)
  • 11 - Generation IV concepts: Japan
  • 11.1 Introduction
  • 11.2 JSFR design and its key innovative technologies
  • 11.2.1 General design features of JSFR
  • 11.2.2 Key innovative technologies in the Japan sodium-cooled fast reactor design
  • 11.2.2.1 High burn-up core
  • 11.2.2.2 Safety enhancement
  • 11.2.2.3 Compact reactor system
  • 11.2.2.4 Two-loop cooling system
  • 11.2.2.5 Integrated intermediate heat exchanger/pump component
  • 11.2.2.6 Reliable steam generator
  • 11.2.2.7 Natural-circulation decay heat removal system
  • 11.2.2.8 Simplified fuel handling system
  • 11.2.2.9 Steel plate-reinforced concrete containment vessel
  • 11.2.2.10 Advanced seismic isolation system
  • 11.3 Update of the Japan sodium-cooled fast reactor design with lessons learned from the Fukushima-Daiichi accident
  • 11.4 Concluding remarks
  • References
  • 12 - Generation IV concepts: USSR and Russia
  • 12.1 Introduction
  • 12.2 History of the Soviet fast reactor program
  • 12.3 Sodium fast reactors
  • 12.3.1 BN-800
  • 12.3.2 Multipurpose fast neutron research reactor
  • 12.3.3 BN-1200
  • 12.4 Heavy liquid metal reactors
  • 12.4.1 SVBR-100
  • 12.4.2 BREST-OD-300
  • 12.4.3 BREST-1200
  • 12.5 Supercritical water reactor
  • 12.6 Conclusion
  • Nomenclature
  • References
  • 13 - Generation IV concepts in Korea
  • 13.1 Current status of nuclear power in Korea
  • 13.2 Plans for advanced nuclear reactors in Korea
  • 13.2.1 Sodium-cooled fast reactor
  • 13.2.2 Very high temperature gas-cooled reactor
  • 13.3 Current research and development on Generation IV reactor in Korea
  • 13.3.1 Sodium-cooled fast reactor
  • 13.3.1.1 Development of a 150MWel prototype sodium-cooled fast reactor
  • Top-tier design requirements
  • Core design
  • Fuel design
  • Fluid system design
  • Mechanical structure design
  • 13.3.1.2 Research and development activities
  • Large-scale sodium thermal-hydraulic test program
  • Metal fuel development
  • Reactor physics experiment
  • 13.3.2 Very high temperature reactor
  • 13.3.2.1 Design and analysis codes
  • 13.3.2.2 TRISO fuel technology
  • 13.3.2.3 High-temperature materials
  • 13.3.2.4 Hydrogen production
  • 13.3.3 Lead fast reactor
  • 13.3.4 Molten salt reactor
  • Acronyms
  • References
  • 13. Appendix: Paper list related to PEACER (including P-demo and Pyroprocess), PASCAR, URANUS, and other SNU-NUTRECK activities
  • 14 - Generation IV concepts: China
  • 14.1 Current status of nuclear power in China
  • 14.2 Plans for advanced nuclear reactors in China
  • 14.3 Current research and development on Generation IV reactors in China
  • 14.3.1 Sodium-cooled fast reactor research and development
  • 14.3.1.1 Research before SFR construction
  • 14.3.1.2 SFR development strategy
  • CEFR
  • CDFR
  • Post-CDFR
  • 14.3.2 Very-high-temperature reactor research and development
  • 14.3.2.1 Early development of the high-temperature gas-cooled reactor program in China
  • 14.3.2.2 HTR-10 test module project
  • Concept design and objectives of HTR-10
  • HTR-10 engineering experiments
  • Experiences learned by constructing HTR-10
  • 14.3.2.3 High-temperature reactor pebble-bed modular project
  • The overall high-temperature reactor pebble-bed modular project
  • Design of HTR-PM
  • 14.3.3 Supercritical water-cooled reactor research and development
  • 14.3.3.1 Mixed spectrum supercritical water-cooled reactor conceptual design
  • 14.3.3.2 The 1000-MWel Chinese supercritical water-cooled reactor concept design
  • 14.3.4 Molten salt reactor research and development
  • 14.3.4.1 Thermal-hydraulic modeling and safety analysis
  • 14.3.4.2 Neutronic modeling
  • 14.3.4.3 Thermo-hydraulics and neutronics coupling analysis
  • 14.3.4.4 Molten salt test loops
  • 14.3.4.5 Material and salts research
  • 14.3.5 Lead-cooled fast reactor research and development
  • 14.3.5.1 China LEad-Alloy-cooled Reactor-0
  • 14.3.5.2 China LEad-Alloy-cooled Reactor-I
  • 14.3.5.3 China LEad-Alloy-cooled Reactor-II
  • 14.3.5.4 China LEad-Alloy-cooled Reactor-III
  • Nomenclature
  • Abbreviations and acronyms
  • References
  • 15 - Generation IV concepts: India
  • 15.1 Introduction
  • 15.2 Advanced heavy water reactors
  • 15.2.1 Design features of AHWR-300
  • 15.2.2 Enhanced safety features
  • 15.2.2.1 Inherent safety features
  • 15.2.2.2 Passive safety systems
  • 15.2.2.3 Features to deal with severe accidents and Fukushima types of scenarios
  • 15.2.3 Safety goals
  • 15.2.4 Proliferation resistance
  • 15.2.5 Physical protection
  • 15.2.6 Improved economics
  • 15.2.7 Research and development activities
  • 15.3 High-temperature reactors
  • 15.3.1 General description of compact high-temperature reactors
  • 15.3.2 Reactor physics design
  • 15.3.3 Thermal hydraulics design
  • 15.3.4 Fuel development
  • 15.3.5 Materials development
  • 15.3.6 Inherent safety features and passive heat removal systems
  • 15.3.7 Research and development activities
  • 15.3.8 Innovative high-temperature reactor
  • 15.4 Fast breeder reactor
  • 15.4.1 Fast reactor program in India
  • 15.4.2 Fast breeder test reactors and their current status
  • 15.4.3 The prototype fast breeder reactor and its current status
  • 15.4.4 Motivation for improvements for future fast breeder reactors beyond the prototype fast breeder reactor (FBR-600)
  • 15.4.5 Conceptual design features of FBR-600
  • 15.4.6 Enhanced safety features
  • 15.4.7 Research and development status
  • 15.5 Molten salt reactors
  • 15.5.1 Conceptual designs of IMSBR
  • 15.5.2 Design challenges
  • 15.5.3 Research and development activities
  • 15.6 Conclusions
  • Nomenclature
  • Reference
  • Bibliography
  • Three - Related topics to Generation IV nuclear reactor concepts
  • 16 - The safety of advanced reactors
  • 16.1 Basic safety principles
  • 16.2 Safety and reliability goals
  • 16.2.1 Subsidiary safety requirements and licensing review
  • 16.2.2 The safety focus for advanced concepts
  • 16.2.3 Emerging and new safety design criteria
  • 16.2.4 The safety goal and objective of "practical elimination"
  • 16.3 Safety objectives and the classification of advanced reactor types
  • 16.4 Generic safety objectives and safety barriers
  • 16.4.1 Reduce the likelihood of initiating events
  • 16.4.2 Ensure long-term cooling
  • 16.4.3 Ensure effective elimination of emergency response
  • 16.4.4 Manage rare and extreme events
  • 16.4.5 Ensure Rickover safeguards for public well-being
  • 16.5 Risk informing safety requirements by learning from prior events
  • 16.6 Major technical safety issues
  • 16.7 Multiple modules and plant risk
  • 16.8 The role of safety research and development for advanced reactors
  • 16.9 Natural-circulation loop and parallel channel thermal-hydraulics
  • 16.9.1 Introduction
  • 16.9.2 Natural-circulation flows
  • 16.10 Literature review
  • 16.10.1 The early investigations
  • 16.10.2 Three Mile Island issues
  • 16.10.3 Boiling water reactor stability in the time and frequency domains
  • 16.10.4 Numerical methods and artifacts
  • 16.10.5 Generation IV passive residual heat removal systems
  • 16.10.6 Coupled natural-circulation loops
  • 16.10.7 Supercritical fluid states and natural-circulation loops
  • 16.10.8 Computational fluid dynamics
  • 16.10.9 Nanofluids
  • 16.10.10 Sodium fast reactors
  • 16.10.11 Parallel channels
  • 16.11 Modeling natural-circulation loops
  • 16.11.1 Single channels and parallel channels
  • 16.11.2 Single natural-circulation loop
  • 16.11.3 Coupled natural-circulation loops
  • 16.12 Conclusions
  • Nomenclature
  • Greek symbols
  • Nondimensional numbers
  • Subscripts
  • Acronyms and abbreviations
  • References
  • 17 - Nonproliferation for advanced reactors: political and social aspects
  • 17.1 Introduction
  • 17.1.1 Nonproliferation: past influence and future directions
  • 17.1.2 Past dreams and present realities of the politics of power
  • 17.1.3 The genesis of the NPT and its bargain
  • 17.1.4 Effects of the treaty
  • 17.1.5 Shortcomings of the treaty
  • 17.1.6 Attempts to improve the treaty system
  • 17.2 Nuclear history and basic science
  • 17.2.1 Commercial nuclear power
  • 17.2.2 Present situation and issues on research and sustainability
  • 17.2.2.1 Research for advanced reactors
  • 17.2.2.2 Commercial fuel supply
  • 17.2.2.3 Alternate fuel cycles for advanced reactors
  • 17.2.2.4 Enrichment
  • 17.2.2.5 Reprocessing and recycling
  • 17.2.2.6 Future policy implications of nuclear fuel cycles
  • 17.3 A look at the future
  • 17.3.1 Alternate fuel cycles
  • 17.3.2 Advanced reactors and the Nuclear Non-Proliferation Treaty
  • 17.4 The wider context
  • 17.5 Fuel cycles: sustainable recycling of used fuel compared to retrievable storage
  • 17.5.1 Introduction: the cost of not burying the past
  • 17.5.2 Economic and social aspects of recycling
  • 17.5.3 The cost savings of the future
  • 17.5.4 Waste to energy: burning the benefits
  • 17.5.4 Overcoming the ostrich syndrome
  • Appendix 1: Euratom
  • Appendix 2: The 1997 IAEA additional protocol at a glance
  • The additional protocol
  • Acronym
  • References
  • 18 - Thermal aspects of conventional and alternative fuels
  • 18.1 Introduction
  • 18.2 Metallic fuels
  • 18.3 Ceramic fuels
  • 18.3.1 Oxide fuels
  • 18.3.1.1 Uranium dioxide
  • 18.3.1.2 Mixed oxide
  • 18.3.1.3 Thorium dioxide
  • 18.3.2 Carbide fuels
  • 18.3.2.1 Uranium carbide
  • 18.3.2.2 Uranium dicarbide
  • 18.3.3 Nitride fuels
  • 18.3.3.1 Uranium mononitride
  • 18.3.4 Discussion
  • 18.4 Hydride fuels
  • 18.4.1 Uranium-zirconium hydride fuel
  • 18.4.2 Uranium-thorium-zirconium fuels
  • 18.5 Composite fuels
  • 18.5.1 Uranium dioxide-silicon carbide
  • 18.5.2 Uranium dioxide-graphite
  • 18.5.3 Uranium dioxide-beryllium oxide
  • References
  • 19 - Hydrogen cogeneration with Generation IV nuclear power plants
  • 19.1 Introduction
  • 19.2 Hydrogen review
  • 19.2.1 Uses of hydrogen
  • 19.2.1.1 Transportation
  • 19.2.1.2 Industrial
  • 19.2.1.3 Electrical
  • 19.2.1.4 Commercial
  • 19.2.2 Benefits of hydrogen
  • 19.3 Hydrogen production methods
  • 19.3.1 Hydrogen production from fossil fuels
  • 19.3.2 Hydrogen production by electrolysis
  • 19.3.3 Hydrogen production by photoelectrolysis
  • 19.3.4 Hydrogen production from solar energy
  • 19.3.5 Thermochemical cycles
  • 19.4 Thermochemical cycles for hydrogen production
  • 19.4.1 Sulfur-iodine cycle
  • 19.4.2 Hybrid sulfur cycle
  • 19.4.3 Copper-chlorine cycle
  • 19.4.4 Iron oxide cycle
  • 19.4.5 Cerium-cerium oxide cycle
  • 19.4.6 Zinc-zinc oxide cycle
  • 19.5 Hydrogen cogeneration with Generation IV reactors
  • 19.5.1 Heat requirements for hydrogen production
  • 19.5.2 Pressure considerations
  • 19.5.3 Isolation
  • 19.6 Conclusions
  • References
  • 20 - Advanced small modular reactors
  • 20.1 Introduction
  • 20.2 Early designs of small modular reactors
  • 20.3 Nuclear reactors
  • 20.4 Reactor coolant system components
  • 20.5 Fuels
  • 20.6 Containment
  • 20.7 Emergency core cooling system
  • 20.8 Economic and financing evaluation
  • 20.9 Security of small modular reactors
  • 20.10 Flexibility of small modular reactors
  • 20.11 Conclusions and future trends
  • Abbreviations
  • Acknowledgment
  • References
  • A1: Additional materials (schematics, layouts, Tes diagrams, basic parameters, and photos) on thermal and nuclear power plants1
  • A1.1 Fossil fuel thermal power plants (listed here just for reference purposes)
  • A1.1.1 Combined cycle power plants
  • A1.1.2 Coal-fired power plants
  • A1.2 Current nuclear power reactors and nuclear power plants (listed here just for reference purposes)
  • A1.2.1 Pressurized water reactors
  • A1.2.2 Boiling water reactors
  • A1.2.3 Pressurized heavy water reactors
  • A1.2.4 Advanced gas-cooled reactors and gas-cooled reactors
  • A1.2.5 Light water-cooled graphite-moderated reactors: RBMK and EGP
  • A1.2.6 Sodium-cooled fast reactor: BN-600 and BN-800
  • Nomenclature
  • Acknowledgments
  • References
  • A2: Comparison of thermophysical properties of reactor coolants1
  • A2.1 Introduction
  • A2.1.1 Generations II, III, and III+ reactor coolants
  • A2.1.2 Generation IV reactor coolants
  • A2.2 Reactor coolants by type
  • A2.2.1 Fluid coolants
  • A2.2.2 Gas coolants
  • A2.2.3 Liquid metal coolants
  • A2.2.4 Molten salt coolants
  • A2.3 Thermophysical properties of proposed Generation II, III, III+, and IV reactor coolants
  • A2.4 Heat transfer coefficients in nuclear power reactors
  • A2.5 Conclusions
  • Nomenclature
  • Greek symbols
  • Nondimensional Numbers
  • Subscripts
  • Acronyms
  • References
  • A3: Thermophysical properties of fluids at subcritical and critical/ supercritical conditions1
  • A3.1 Introduction
  • A3.1.1 Historical note on using supercritical pressure fluids
  • A3.1.2 Definitions of terms and expressions related to critical and supercritical regions
  • A3.2 Thermophysical properties at critical and supercritical pressures
  • A3.3 Conclusions
  • Nomenclature
  • Greek letters
  • References
  • A4: Heat transfer and pressure drop in forced convection to fluids at supercritical pressures1
  • A4.1 Introduction
  • A4.1.1 Historical note on using supercritical pressure fluids
  • A4.1.2 Definitions of terms and expressions related to supercritical pressure heat transfer
  • A4.2 Specifics of forced convection heat transfer at supercritical pressures
  • A4.2.1 Basics of supercritical heat transfer
  • A4.2.2 Pseudo-boiling and pseudo-film boiling phenomena
  • A4.2.3 Horizontal flows
  • A4.2.4 Heat transfer enhancement
  • A4.2.5 Practical prediction methods for forced-convection heat transfer at supercritical pressures
  • A4.3 Hydraulic resistance
  • A4.4 Conclusions
  • Nomenclature
  • Greek letters
  • Non-dimensional numbers
  • Subscripts or superscripts
  • Abbreviations and acronyms
  • References
  • A5: World experience in nuclear steam reheat1
  • A5.1 Introduction
  • A5.2 US experience in nuclear steam reheat2
  • A5.3 Russian experience in nuclear steam reheat
  • A5.3.1 General information
  • A5.3.2 Thermodynamic cycle development
  • A5.3.2.1 Layout (a)
  • A5.3.2.2 Layout (b)
  • A5.3.2.3 Layout (c)
  • A5.3.2.4 Layout (d)
  • A5.3.2.5 Layout (e)
  • A5.3.2.6 Layout (f)
  • A5.3.3 Beloyarsk NPP reactor design
  • A5.3.4 Physical parameters of Beloyarsk NPP reactors
  • A5.3.5 Boiling water channels
  • A5.3.6 Superheated steam channels
  • A5.3.7 Hydrodynamic stability of the Beloyarsk NPP channels during reactor startup
  • A5.3.8 Startup of Beloyarsk NPP reactors
  • A5.3.9 Pumps
  • A5.3.10 Water chemistry
  • A5.3.11 Modular reactor with steam reheat
  • A5.4 Conclusions
  • Appendix A5 Nomenclature
  • Appendix A5 Greek letters
  • Appendix A5 Subscripts
  • Appendix A5 Abbreviations and acronyms
  • References
  • A6: Comparison of thermophysical properties of selected gases at atmospheric pressure
  • A7: Supplementary tables
  • A8: Unit conversion
  • A8.1 Primary or fundamental dimensions and their units in SI (International System)
  • A8.2 Standard prefixes in SI units
  • A8.3 Unit conversion
  • A8.3.1 Area
  • A8.3.2 Density
  • A8.3.3 Electrical resistivity specific
  • A8.3.4 Energy, work, and heat amount
  • A8.3.5 Specific Enthalpy
  • A8.3.6 Flow rate volumetric (or volume flow rate)
  • A8.3.7 Force
  • A8.3.8 Heat flux
  • A8.3.9 Heat flux volumetric
  • A8.3.10 Heat transfer coefficient
  • A8.3.11 Heat transfer rate and power
  • A8.3.12 Length
  • A8.3.13 Mass
  • A8.3.14 Pressure
  • A8.3.15 Specific heat
  • A8.3.16 Temperature scales
  • A8.3.17 Temperature difference
  • A8.3.18 Thermal conductivity
  • A8.3.19 Viscosity dynamic
  • A8.3.20 Viscosity kinematic
  • A8.3.21 Volume
  • A8.4 Some physical constants and definitions
  • A8.5 Thermophysical property software for gases and liquids
  • Index
  • A
  • B
  • C
  • D
  • E
  • F
  • G
  • H
  • I
  • J
  • K
  • L
  • M
  • N
  • O
  • P
  • Q
  • R
  • S
  • T
  • U
  • V
  • X
  • Y
  • Z
  • Back Cover

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