Handbook of Generation IV Nuclear Reactors

1st Edition - June 9, 2016
Editor: Igor Pioro
Language: English
Hardback ISBN:
9 7 8 - 0 - 0 8 - 1 0 0 1 4 9 - 3
eBook ISBN:
9 7 8 - 0 - 0 8 - 1 0 0 1 6 2 - 2

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 4… Read more

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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.

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List of contributors
Woodhead Publishing Series in Energy
Foreword
Preface
Nomenclature
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 plants
- 1.3. Modern nuclear power plants
- 1.4. Conclusions
- Abbreviations
Part One. Generation IV nuclear-reactor concepts
- Preface to Part One
- 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
  - 2.5. Summary
- 3. Very high-temperature reactor
  - 3.1. Development history and current status
  - 3.2. Technology overview
  - 3.3. Detailed technical description
  - 3.4. Applications and economics
  - 3.5. Summary
  - Acronyms
- 4. Gas-cooled fast reactors
  - 4.1. Rationale and generational research and development bridge
  - 4.2. Gas-cooled fast reactor technology
  - 4.3. Conclusions
- 5. Sodium-cooled fast reactor
  - 5.1. Introduction
  - 5.2. Development history
  - 5.3. System characteristics
  - 5.4. Safety issues
  - 5.5. Future trends and key challenges
- 6. Lead-cooled fast reactor
  - 6.1. Overview and motivation for lead-cooled fast reactor systems
  - 6.2. Basic design choices
  - 6.3. Safety principles
  - 6.4. Fuel technology and fuel cycles for the lead-cooled fast reactor
  - 6.5. Summary of advantages and key challenges of the lead-cooled fast reactor
  - 6.6. Overview of Generation IV lead-cooled fast reactor designs
  - 6.7. Future trends
  - Sources of further information
  - Nomenclature
- 7. Molten salt fast reactors
  - 7.1. Introduction
  - 7.2. The molten salt fast reactor concept
  - 7.3. Fuel salt chemistry and material issues
  - 7.4. Molten salt fast reactor fuel cycle scenarios
  - 7.5. Safety issues
  - 7.6. Concept viability: issues and demonstration steps
  - 7.7. Conclusion and perspectives
  - Nomenclature
- 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
  - 8.7. Safety system concept
  - 8.8. Dynamics and control
  - 8.9. Start-up
  - 8.10. Stability
  - 8.11. Advantages and disadvantages of supercritical water-cooled reactor concepts
  - 8.12. Key challenges
  - 8.13. Future trends
  - Acronyms
  - Nomenclature
Part Two. Current status of Generation IV activities in selected countries
- Preface to Part Two
- 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 management
  - 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
- 10. Euratom research and training program in Generation-IV systems: Breakthrough technologies to improve sustainability, safety and reliability, socioeconomics, and proliferation resistance
  - 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 than five decades
  - 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 industry)
  - Nomenclature
  - Appendix: Tentative training scheme for preconceptual Generation-IV design engineers (knowledge, skills, attitudes)
- 11. Generation IV concepts: Japan
  - 11.1. Introduction
  - 11.2. JSFR design and its key innovative technologies
  - 11.3. Update of the Japan sodium-cooled fast reactor design with lessons learned from the Fukushima-Daiichi accident
  - 11.4. Concluding remarks
- 12. Generation IV concepts: USSR and Russia
  - 12.1. Introduction
  - 12.2. History of the Soviet fast reactor program
  - 12.3. Sodium fast reactors
  - 12.4. Heavy liquid metal reactors
  - 12.5. Supercritical water reactor
  - 12.6. Conclusion
  - Nomenclature
- 13. Generation IV concepts in Korea
  - 13.1. Current status of nuclear power in Korea
  - 13.2. Plans for advanced nuclear reactors in Korea
  - 13.3. Current research and development on Generation IV reactor in Korea
  - Acronyms
- 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
  - Nomenclature
- 15. Generation IV concepts: India
  - 15.1. Introduction
  - 15.2. Advanced heavy water reactors
  - 15.3. High-temperature reactors
  - 15.4. Fast breeder reactor
  - 15.5. Molten salt reactors
  - 15.6. Conclusions
  - Nomenclature
Part Three. Related topics to Generation IV nuclear reactor concepts
- Preface to Part Three
- 16. The safety of advanced reactors
  - 16.1. Basic safety principles
  - 16.2. Safety and reliability goals
  - 16.3. Safety objectives and the classification of advanced reactor types
  - 16.4. Generic safety objectives and safety barriers
  - 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
  - 16.10. Literature review
  - 16.11. Modeling natural-circulation loops
  - 16.12. Conclusions
  - Nomenclature
  - Greek symbols
  - Nondimensional numbers
  - Subscripts
  - Acronyms and abbreviations
- 17. Nonproliferation for advanced reactors: Political and social aspects
  - 17.1. Introduction
  - 17.2. Nuclear history and basic science
  - 17.3. A look at the future
  - 17.4. The wider context
  - 17.5. Fuel cycles: sustainable recycling of used fuel compared to retrievable storage
  - Appendix 1: Euratom
  - Appendix 2: The 1997 IAEA additional protocol at a glance
  - Acronym
- 18. Thermal aspects of conventional and alternative fuels
  - 18.1. Introduction
  - 18.2. Metallic fuels
  - 18.3. Ceramic fuels
  - 18.4. Hydride fuels
  - 18.5. Composite fuels
- 19. Hydrogen cogeneration with Generation IV nuclear power plants
  - 19.1. Introduction
  - 19.2. Hydrogen review
  - 19.3. Hydrogen production methods
  - 19.4. Thermochemical cycles for hydrogen production
  - 19.5. Hydrogen cogeneration with Generation IV reactors
  - 19.6. Conclusions
- 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
Appendix A1. Additional materials (schematics, layouts, T–s diagrams, basic parameters, and photos) on thermal and nuclear power plants
Appendix A2. Comparison of thermophysical properties of reactor coolants
Appendix A3. Thermophysical properties of fluids at subcritical and critical/supercritical conditions
Appendix A4. Heat transfer and pressure drop in forced convection to fluids at supercritical pressures
Appendix A5. World experience in nuclear steam reheat
Appendix A6. Comparison of thermophysical properties of selected gases at atmospheric pressure
Appendix A7. Supplementary tables
Appendix A8. Unit conversion
Index

Igor Pioro

Professor Igor Pioro – Ph.D. (1983); Doctor of Technical Sciences (1992); Professional Engineer (Ontario, Canada) (2008); Foreign Fellow of the National Academy of Sciences of Ukraine (2021); Fellow of the American Society of Mechanical Engineers (ASME) (2012), Canadian Society of Mechanical Engineers (CSME) (2015), and Engineering Institute of Canada (EIC) (2013); life-time member of the American Nuclear Society (ANS) (2004), and member of Canadian NS (CNS) (2010); is an internationally recognized scientist within the areas of nuclear engineering (thermalhydraulics of nuclear reactors, Generation-IV nuclear-reactor concepts, etc.) and thermal sciences / engineering (boiling, forced convection including supercritical pressures, etc.) He is author/co-author of more than 534 publications 12 technical books, 48 chapters in encyclopedias, handbooks and books, 101 papers in refereed journals, 300 papers in refereed proceedings of international and national conferences and symposiums, 26 patents and inventions, and 47 major technical reports. Dr. Pioro graduated from the National Technical University of Ukraine "Kiev Polytechnic Institute" with M.A.Sc. in Thermal Physics in 1979. After that, he worked on various positions including an engineer, senior scientist, deputy director, professor, director of a graduate program in nuclear engineering, and associate dean. Currently, he is associated with the Department of Energy and Nuclear Engineering Faculty of Engineering and Applied Science, University of Ontario Institute of Technology (brand name: Ontario Tech University) (Oshawa, Ontario, Canada). Dr. Pioro is a Founding Editor – Editor-in-Chief of the ASME Journal of Nuclear Engineering and Radiation Science (J. NERS). He was a Chair of the Executive Committee of the Nuclear Engineering Division (NED) of the ASME (2011-2012) and a Chair of the International Conference On Nuclear Engineering (ICONE-20) (2011-2012). Professor Pioro has received many international and national awards and certificates of appreciation including Certificate of Appreciation for establishing the ASME J. NERS and in recognition of outstanding and distinguished service as an Editor of the Journal 2014 – 2020 (ASME); Harold A. Smith Outstanding Contribution Award from CNS (2017), Medal 60th Anniversary of NED (Nuclear Engineering Division of ASME) (2016); Service Recognition Award from the ASME (2014); Honorary Doctor’s Degree from National Technical University of Ukraine “Kiev Polytechnic Institute” (2013); The CNS Education and Communication Award (2011); UOIT Research Excellence Award (2011); ICONE Award from ASME (2009); Medal of the National Academy of Sciences of Ukraine for the best scientific work of a young scientist (1990); Badge "Inventor of the USSR" for implementation of inventions into industry (1990); etc.

Affiliations and expertise

Faculty of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, Canada

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Handbook of Generation IV Nuclear Reactors

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