Handbook of Generation IV Nuclear Reactors: A Guidebook

Cover
Handbook of Generation IV Nuclear Reactors
Copyright
Contributors
Foreword
Preface
Introduction
Current status of electricity generation in the world
Electricity generation in the world
Largest power plants of the world, industrial electricity-generating sources, and their pros and cons
Largest power plants of the world
Industrial electricity-generating sources
Pros and cons of various electricity-generating sources
Non-renewable-energy power plants
Thermal
Coal-fired thermal power plants
Gas-fired thermal power plants
Oil-shale-fired thermal power plants
Peat-fired thermal power plants
Mazut/heavy-oil-fired power plants
Internal-combustion engines power plants
Recovered energy generation
Nuclear
Renewable-energy power plants
Hydro
Conventional hydro-power plants
Run-of-the-river hydro-power plants
Pumped-Storage Hydro-Electric Power Plant (PSHEPP)
Wind
On-shore wind farms (power plants)
Off-shore wind farms (power plants)
Solar
Concentrated-solar thermal power plant with a tower and sun-tracking heliostats
Flat-panel PV and concentrated PV solar-power plants
Geothermal
Geothermal plants
Tidal
Tidal plants
Wave
Wave power plants
Actual examples of operating power grid with non-renewable and renewable sources
Conclusions
Acknowledgments
References
Further reading
Current status and future trends in the worldnuclear-power industry
Conclusions
References
Further reading
Generation IV International Forum (GIF)
Origins of GIF
Gen-IV goals
Selection of Gen-IV systems
Six Gen-IV nuclear-energy systems
Methodology working groups, task forces, cross-cutting items and the Senior Industrial Advisory Panel
Summary
Acknowledgments
References
Further reading
Very High Temperature Reactor
Development history and current status
Technology overview
Reactor design types
Design features
Safety
Fuel cycle
Multipurpose
Detailed technical description
Fuel design
Fuel burnup
Uranium fuel
Plutonium fuel
Reactor design
Prismatic core reactor design
Pebble bed core reactor design
Reactor safety
Plant design
Plant operations
Startup, rated operation, and shutdown
Dynamic operation
Applications and economics
Power generation
Cogeneration
Hydrogen cogeneration
Desalination cogeneration
Industrial application
Economics
Cost of electricity generation
Capital cost
Operating cost
Fuel cost
Power generation cost
Cost of hydrogen production
Cost of desalination cogeneration
Summary
References
Gas-cooled Fast Reactors (GFRs)
Rationale and generational R&D bridge
Gas-cooled Fast Reactor technology
Evolution of Generation-IV GFRs into small modular reactor and micro reactors
Conclusions
References
Sodium-cooled Fast Reactors (SFRs)
Introduction
Development history
System characteristics
Design features with sodium properties
Core configurations
Plant system
Loop type and pool type
Consistency with fuel cycle system (fuel cycle technology)
Safety issues
Safety design criteria and safety design guidelines
Safety characteristics and safety design
Reactor shutdown
Decay heat removal
Design measure against sodium chemical reactions
Containment measures
Future trends and key challenges
References
Lead-cooled Fast Reactors (LFRs)
Overview and motivation for lead-cooled fast reactor systems
Basic design choices
Lead versus LBE
Design choices for reactors with lead as the coolant
Primary system concepts: Evolution and challenges
Early conceptual designs derived from sodium-cooled fast reactor concepts
Primary system development and evolution
Safety principles
Fuel technology and fuel cycles for the lead-cooled fast reactor
Fuel assembly characteristics
Fuel cycle for the lead-cooled fast reactor
Summary of advantages and key challenges of the lead-cooled fast reactor
Advantages of the lead-cooled fast reactor
Key challenges of the lead-cooled fast reactor
Overview of Generation-IV lead-cooled fast reactor designs
Reference Generation IV systems
The European Lead-cooled Fast Reactor (ELFR)
The BREST-OD-300 reactor
The Small Secure Transportable Autonomous Reactor (SSTAR)
Additional Generation IV systems under study or development, and new directions
The Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED)
Westinghouse LFR
Newcleo LFR-AS-200
LFR-TL-X
The South Korean URANUS-40 system
The Chinese CLEAR-I reactor
The Pb-Bi-cooled direct contact boiling Water Fast Reactor (PBWFR)
The SVBR-100
SwEdish Advanced Lead Reactor (SEALER)
Sources of further information
Acknowledgments
References
Homogeneous Molten Salt Reactors (MSRs): The Molten Salt Fast Reactor (MSFR) concept
Introduction
The MSFR concept
Core and system description
Transient calculations
Fuel salt chemistry and material issues
Overview of the processing schemes
Impact of the salt composition on the corrosion of the structural materials
MSFR fuel cycle scenarios
Safety methodology and risk analysis
Liquid-fuel reactor specificities and decay heat removal
Safety approach
Safety evaluation
Conclusions and recommendations of the MSFR safety evaluation
Concept viability: Issues and demonstration steps
Identified limits
Demonstration steps
Other R&D activities on molten salt systems
Conclusion and perspectives
Acknowledgments
References
Sources for further information
SuperCritical Water-cooled Reactors (SCWRs)
Introduction
Types of supercritical water-cooled reactor concepts and main system parameters
Example of a pressure vessel concept
Example of a pressure tube concept
Fuel cycle technology
Fuel-assembly concept
High-Performance Light Water Reactor (HPLWR) fuel-assembly concept
Fast reactor fuel-assembly concept
Canadian SCWR fuel-assembly concept
Safety system concept
Safety system in a pressure vessel-type supercritical water-cooled reactor concept
Safety system in the Canadian SCWR concept
Containment pool
Automatic depressurization system
Gravity-driven core flooding system
Isolation condensers
Reserve water pool
Atmospheric air heat exchangers
Passive moderator cooling system
Dynamics and control
Start-up
Start-up system in a pressure tube-type supercritical water-cooled reactor concept
Stability
Advantages and disadvantages of supercritical water-cooled reactor concepts
Key challenges
Fuel qualification test
References
Generation IV: United States
Generation-IV program evolution in the United States
Energy market in the United States and Generation-IV systems
Electrical grid integration of Generation-IV nuclear energy systems in the United States
Industry and utilities interests in Generation-IV nuclear energy systems in the United States
Evolution of Generation-IV nuclear energy systems into small modular reactors and micro reactors
Deployment perspectives for Generation-IV systems in the United States and deployment schedule
Conclusions
References
Further reading
Generation IV: European Union: Breakthrough technologies to improve sustainability, safety & reliability, soc ...
Introduction: ``EU Energy Union´´ (2015) and ``EU Green Deal´´ (2020)-Going climate neutral by 2050-Euratom contrib ...
Total of 106 nuclear power reactors in the EU (= 26% of gross electricity production)
District heating and industrial heat applications world-wide
Good health and well-being (SDG 3-2030 Agenda, United Nations/UN/2015)
EUs ambition to become the worlds 1st major economy to go climate neutral by 2050
Energy transition toward climate neutrality: EUs support for ``green´´ technologies
EURATOM: Research safety of nuclear installations; health and safety (radiation protection); safeguards; radwaste ...
EURATOM-Brief history (21st century challenges) and links with IAEA and OECD/NEA
EURATOM legal framework-The most stringent safety requirements in the world
EURATOM-Science, technology and innovation (several ambitious Framework Programmes since 1994)
EURATOM-Dissemination of knowledge-``European Nuclear Education Network´´
Generation-IV: Breakthrough developments in sustainability, safety and performance through multilateral collaborati ...
Generation-IV International Forum (GIF): USA, Canada, France, Japan, South Africa, South Korea, Switzerland, Eura ...
Innovation in nuclear fission from Generation I to IV (Euratom contribution)
GIF Technology Roadmap (viability, performance, demonstration)-toward industrial deployment by 2045
GIF Roadmaps 2002 and 2013-viability, performance and demonstration phases
IAEA programme INPRO (International Project on Innovative Nuclear Reactors and Fuel Cycles)
GIF interaction with industry: The ``Senior Industrial Advisory Panel´´ (SIAP)
GIF interaction with regulators: NRC (USA), IRSN (FR) and MDEP (OECD/NEA)
Eight high-level goals for generation-IV nuclear energy systems and associated world-wide GIF R&D collaborative effort
Sustainability (efficient resource utilization and minimization of radioactive waste)
Safety (maximum safety performance through design, technology, regulation and culture) & Reliability
Economics (competitiveness w.r.t. other energy sources) and social aspects (e.g., public engagement in decision m ...
Proliferation resistance and physical protection (Non-Proliferation Treaty, IAEA 1970)
Euratom research and training actions in innovative reactor systems and EU ``Sustainable Nuclear Energy Technology ...
EURATOM actions that are considered as contributing to the six GIF reactor systems
European Sustainable Nuclear Fission Industrial Initiative (ESNII) and Nuclear Cogeneration Industrial Initiative ...
Experimental research reactors in the EU and small modular reactors
Experimental research reactors (training, materials testing, isotope production)
SMR technology is a great opportunity for the nuclear industry and could lead to a nuclear renaissance
Conclusion: The Euratom research and training program-Maintaining EU leadership in nuclear fission developments
ESFR SMART: A European Sodium Fast Reactor concept including the European feedback experience and the new saf ...
The ESFR-SMART project
Sodium fast reactors history in Europe
Safety improvement: Objectives and methodology
Some examples of safety improvement approach in the ESFR SMART
Reactivity control
New core concept with reduced sodium void effect
Passive-control rods
Ultra-sonic measurements for knowledge of the core geometry
Practically eliminated situations
Containment
Reactor pit taking over the functions of the safety vessel
In-vessel core catcher
Massive metallic roof
Leak tightness of roof penetrations
Decay-heat removal
Sodium fire
Sodium/water reaction
Severe-accident mitigation
Dosimetry and releases
Simplicity and human factor
Description of ESFR SMART primary system including these new options
General-plant characteristics
Core
Main vessel
Inner vessel
Reactor roof
Reactor pit
DHRS-3
Primary sodium confinement
Core support structure and connection to pump
Core catcher
Primary pump
Intermediate heat exchanger
Decay-heat removal concept
Polar table or dome
Description of ESFR SMART secondary loops
General description of ESFR SMART secondary loop
Secondary pump
Steam generator
DHRS-1 system
Piping
Safety analysis of the secondary loop
General layout of the plant
Handling systems
Spent-fuel handling
Fresh-fuel handling
Handling of components
Conclusions on safety improvements
R&D needs for the ESFR SMART options
Conclusion
Acknowledgment
References
Generation-IV Sodium-cooled Fast Reactor (SFR) concepts in Japan
Introduction
JSFR design and its key innovative technologies
General design features of JSFR
Key innovative technologies in the Japan sodium-cooled fast reactor design
High burn-up core
Safety enhancement
Compact reactor system
Two-loop cooling system
Integrated intermediate heat exchanger/pump component
Reliable steam generator
Natural-circulation decay heat removal system
Simplified fuel handling system
Steel plate-reinforced concrete containment vessel
Advanced seismic isolation system
Update of the Japan sodium-cooled fast reactor design with lessons learned from the Fukushima Daiichi accident
Concluding remarks
References
Generation-IV concepts in Korea
Current status of nuclear power in Korea
Plans for advanced nuclear reactors in Korea
Sodium-cooled fast reactor
Very-High-Temperature gas-cooled Reactor (VHTR)
Current research and development on Generation-IV reactor in Korea
Sodium-cooled Fast Reactor (SFR)
Development of a 150 MWel prototype Sodium-cooled Fast Reactor (SFR)
Top-tier design requirements
Core design
Fuel design
Fluid system design
Mechanical structure design
Research and development activities
Large-scale sodium thermal-hydraulic test program
Metal fuel development
Reactor physics experiment
Very-high-temperature reactor
Design and analysis codes
TRISO fuel technology
High-temperature materials
Hydrogen production
Lead fast reactor
Molten Salt Reactor (MSR)
Appendix: Paper list related to PEACER (including P-demo and Pyroprocess), PASCAR, URANUS, and other SNU-NUTRECK activities
References
Further reading
Generation-IV concepts: China
Current status of nuclear power in China
Plans for advanced nuclear reactors in China
Current research and development on Generation-IV reactors in China
SFR research and development
Research before CEFR construction
China SFR development strategy
CEFR
CDFR
Post-CDFR
Very-high-temperature reactor research and development
Early development of the HTGR program in China
HTR-10 test module project
Conceptual design and objectives of HTR-10
HTR-10 engineering experiments
Experiences learned by constructing HTR-10
HTR-PM project
The overall HTR-PM project
Design of HTR-PM
SCWR research and development
SCWR-M conceptual design
The CSR1000 concept design
MSR research and development
Thermal-hydraulic modeling and safety analysis
Neutronic modeling
Thermo-hydraulics and neutronics coupling analysis
Molten salt test loops
Material and salts research
Conceptual design of FuSTAR
LFR research and development
CLEAR-0
Clear-I
Clear-II
Clear-III
LESMOR
References
Generation-IV concepts: India
Introduction
Advanced Heavy Water Reactors (AHWRs)
Design features of AHWR-300
Enhanced safety features
Inherent safety features
Passive safety systems
Features to deal with severe accidents and Fukushima types of scenarios
Safety goals
Proliferation resistance
Physical protection
Improved economics
Research and development activities
High-Temperature Reactors (HTRs)
General description of compact High-Temperature Reactors (HTRs)
Reactor physics design
Thermal hydraulics design
Fuel development
Materials development
Inherent safety features and passive heat removal systems
Research and development activities
Innovative High-Temperature Reactor (IHTR)
Fast Breeder Reactor (FBR)
Fast reactor program in India
Fast breeder test reactors and their current status
The Prototype Fast Breeder Reactor (PFBR) and its current status
Motivation for improvements for future fast breeder reactors beyond the prototype Fast Breeder Reactor (FBR-600)
Conceptual design features of FBR-600
Enhanced safety features
Research and development status
Molten Salt Reactors (MSRs)
Conceptual designs of IMSBR
Design challenges
Research and development activities
Conclusions
Reference
Bibliography
The safety and risk assessment of Advanced Reactors (ARs)
Basic safety principles
Safety and reliability goals
Subsidiary safety requirements and licensing review
The safety focus for advanced concepts
Emerging and new safety design criteria
Safety design criteria: partial listing with edited NRC review comments (Sofu, 2014; Nakai, 2013; information courtesy o
The safety goal and objective of ``practical elimination´´
Safety objectives and the classification of advanced reactor types
Generic safety objectives and safety barriers
Safety goals and objectives
Risk informing safety requirements by learning from prior events
Major technical safety issues
Safety design approaches to achieving GIF goals (adapted from Kelly, 2014)
Multiple modules and plant risk
The role of Safety R&D for ARs
Risk informing advanced reactor safety: Quantifying the probability and uncertainty of core damage due to loss of p ...
Introduction to RIDM
Addressing limitations of RIDM methodology and safety-related societal judgments
Using prior data to inform RIDM: Applicability and exchangeability
Active and passive safety, power, and cooling systems
Risk informing the probability of extended power loss and core damage
RIDM: Specific worked dynamic example and core damage uncertainty estimate
Comparing and defining quantitative estimates of RIDM uncertainty
Natural circulation loop and parallel channel thermal-hydraulics
Introduction
Natural circulation flows
Literature review
The early investigations
Three Mile Island issues
BWR stability in the time and frequency domains
Numerical methods and artifacts
AR and Gen-IV passive residual heat-removal systems
Coupled natural circulation loops
Supercritical fluid states and NCLs
Computational Fluid Dynamics (CFD)
Nanofluids
Sodium and liquid metal reactors
Parallel channels
Conclusions
References
Further reading
Non-proliferation for Advanced Reactors (ARs): Political and Social aspects
Introduction
Non-proliferation: Past influence and future directions
Past dreams and present realities of the politics of power
The genesis of the Non-Proliferation Treaty (NPT) and its bargain
Effects of the treaty
Shortcomings of the treaty
Attempts to improve the treaty system
Nuclear history and basic science
Commercial nuclear power
Present situation and issues on research and sustainability
Research for advanced reactors
Commercial fuel supply
Alternate fuel cycles for advanced reactors
Enrichment
Reprocessing and recycling
Future policy implications of nuclear fuel cycles
A look at the future
Alternate fuel cycles
Advanced reactors and the NPT
The wider context
Fuel cycles: Sustainable recycling of used fuel compared to retrievable storage
Introduction: The cost of not burying the past
Economic and social aspects of recycling
The cost savings of the future
Waste to energy: Burning the benefits
Overcoming the ostrich syndrome
Annex 1: EURATOM
Annex 2: The 1997 IAEA additional protocol at a glance
The additional protocol
References
Thermal aspects of conventional and alternative nuclear fuels
Introduction
Metallic fuels
Ceramic fuels
Oxide fuels
UO2
ThO2
Mixed oxide fuels
UO2+PuO2
ThO2+UO2 and ThO2+PuO2
Carbide fuels
UC
UC2
Nitride fuels
UN
Hydride fuels
U-ZrH1.6
UTh4Zr10Hx
Composite fuels
UO2-SiC
UO2-C
UO2-BeO
Analysis results
Discussions
References
Hydrogen production pathways for Generation-IV reactors
Introduction
Coupling hydrogen and Generation-IV reactor technologies
Biomass and fossil-based technologies
Steam methane reforming
Gasification
Pyrolysis
Electrolysis
PEM electrolysis
Alkaline electrolysis
High-temperature electrolysis
Pure and hybrid thermochemical cycles
Sulfur-Iodine (S-I) pure thermochemical cycle
Hybrid Sulfur (HyS) thermochemical cycle
Copper-Chlorine (Cu-Cl) hybrid thermochemical cycle
Nuclear hydrogen production toward climate change mitigation
Conclusion
References
Systems of Advanced Small Modular Reactors (ASMRs)
Introduction
Early designs of small modular reactors
Nuclear reactors
Reactor coolant system components
Fuels
Containment
Emergency core cooling system
Economic and financing evaluation
Security of small modular reactors
Flexibility of small modular reactors
Conclusions and future trends
Acknowledgment
References
Current status of SMRs and S&MRs development in the world
Small Modular and Small- & Medium-size Reactors (SMRs and S&MRs)
Preconceptual
Conceptual
Basic
Developmental
Preliminary
Final/Certified
Construction
SMRs and S&MRs by type
SMRs and S&MRs by countries
Russian KLT-40S and RITM-200M SMRs
Special considerations on SMRs and future development and implementation
Safety and licensing requirements
Pathways to success
Conclusions
References
Alternative power cycles for Generation-IV reactors
Alternative power cycles for Generation-IV reactors
Basic-cycle options
Cycles for gas-cooled reactors (VHTRs and GFRs) and SFR
Cycles comparison
Conclusions
References
Closed Brayton-cycle configurations for Gas-cooled Fast Reactors
(GFRs) and Very-High-Temperature Reactors (VHTRs)
Introduction
Gas turbine as a power-conversion machine
Operational experience of helium gas turbines
Oberhausen II EVO helium turbine (1974)
Historical problems with helium gas turbines
Recent helium gas turbine tests
Nuclear power plant closed cycles
Brayton cycles
Helium as a coolant
Generation-IV closed-cycle configurations for GFRs and VHTRs
Component definition for design point performance
Axial compressor
Axial turbine
Recuperator
Precooler and intercooler
Reactor
Cycle performance
Cycle performance design considerations
DP performance comparison
Impact of temperature and pressure ratio on cycle efficiency and plant capacity
Component efficiencies
Compressor inlet temperature
Pressure losses
Working fluid for NPP cycle performance, and operation
Advances in working fluid studies
Working fluid cycle configuration and performance
Case studies on working fluid cycle configuration and performance
CO2 versus supercritical CO2 as working fluid
Helium-nitrogen binary mixture as working fluid
Thermophysical properties of reactor coolants and working fluids of power cycles
Nuclear power plant controls and operations
Inventory Control Strategy (ICS)
Bypass Control Strategy (BCS)
Reactor delivery temperature control (HST)
Combined control strategy
Risk assessment
Technology risk
Material technology for high temperature and pressure
Working fluid gas management
Technology maturity level for components based on selected working fluid
Working fluid TRL for turbomachinery components
Working fluid TRL for heat exchangers
Financial risk
Cost of working fluid
Sensitivity to pressure ratio and mass flow rate
Legislation
Other design considerations
Long-term off-design point performance and operation
Short-term off-design point performance and operation
Reactor technology status
Future trends
Smaller high-pressure ratio cycle configuration
Higher core outlet temperature >1000C
Improved helium compressors
Conclusion
References
Further reading
Regulatory and licensing challenges with Generation-IV nuclear energy systems
Introduction
The regulatory status
Regulatory requirements
Regulatory challenges for advanced reactors
Case study-Canadian perspectives on the design of Pressure Retaining Systems and Components (PRSCs) in small modula ...
Review framework for FA #1 on defense-in-depth
Operational limits and conditions
Defense-in-depth
Element #1: Leak detection system
Element #2: Aging Management Plan (AMP)
Element #3: Materials
Element #4: Leak rate
Element #5: Loadings
Element #6: Analysis
Element #7: Redundancy, diversity, and separation philosophy
Review framework for FA #2 on classification of systems, structures and components
Code classification requirements
Review framework for FA #10 on safety analysis
Accident prevention and plant safety characteristics
Postulated Initiating Events (PIEs)
Plant states
Quantitative application of the safety goals and PSA
Review framework for FA #11 on pressure boundary design
Design rules and limitations
Seismic classification, category, and qualification
Pressure boundary program and CSA standards
Non-design basis loading condition
Helical coil steam generator
Probabilistic fracture mechanics
Aging and Wear
In-service testing, maintenance, repair, inspection, and monitoring
Review framework for FA #16 on vendor Research and Development program
Conclusions
References
Further reading
ITER, the way to fusion energy
Nuclear fusion
The history of the ITER project
A gigantic fusion machine
A pharaonic worksite
Organizing a huge logistics
Delays and budget increases
The management challenge
Nuclear licensing
Safety and waste management
Natural hazards
The impact of ITER on the economy
Will fusion become commercial?
DEMO and the projects after ITER
Alternative technologies
The fusion era
References
Further reading
Additional materials (layouts, T-s diagrams, basic parameters, photos, etc.) on thermal and nuclear power plants
Introduction
Fossil-fuel thermal power plants
Combined-Cycle Power Plants (CCPPs)
Coal-fired thermal power plants
Current nuclear-power reactors and NPPs
Introduction
Current status, advantages and challenges of nuclear-power reactors
Technical considerations of various types of nuclear-power reactors and plants
Classifications of nuclear-power reactors and plants
By neutron spectrum
By reactor coolant and moderator
By pressure-boundary design
By power cycle
Comparison of various nuclear-power reactors by major features
Modern nuclear-power reactors and plants: Design features
Advanced Pressurized Water Reactors (APWRs) and PWRs
EPR (EDF, France)
APR-1400 (Doosan, South Korea)
AP-1000 (Westinghouse, USA)
PWR (USA)
Advanced PWR (MHI, Japan)
PWRs (Russia)
Boiling Water Reactors (BWRs) and Advanced BWRs (ABWRs)
BWR (USA)
ABWRs (Japan)
Advanced boiling water reactor (ABWR) (Hitachi-GE Nuclear Energy, Ltd.)
Application of ``Evolutional Designs´´:
ABWR Safety Features:
Additional features
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ABWR counter-measures against Fukushima Daiichi NPP severe accident (based on information from Hitachi-GE Nuclear Energy)
Pressurized Heavy Water Reactors (PHWRs)
History and global context of PHWRs
Details on present and advanced PHWRs
Summary of PHWR status and issues going forward
Extending life of existing units
Upgrading existing designs
Advanced PHWR concepts and fuel cycles
Conclusions
Additional PHWR information sources
Canada
Operators:
International members:
India
Argentina
Education
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Advanced Gas-cooled Reactors (AGRs) and Gas-Cooled Reactors (AGRs) and gas-cooled reactors
AGRs carbon-dioxide cooled
GCRs helium-cooled
Light-water-cooled Graphite-moderated Reactors (LGRs): RBMKs and EGPs
Sodium-cooled Fast Reactor (SFR): BN-600 (Generation-III), BN-800 (Generation-III+), and BN-1200 (Generation-IV)
Basis for Fast Reactors (FRs)
BN-600
The BN-600 reactor NPP has the following distinctive features:
Safety of BN-600 is based on:
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BN-800
Reactor vessel
Reactor core
Refueling system
In vessel refueling equipment performs the following:
The ex-vessel refueling system performs the following:
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Steam generator (SG)
Safety of the BN-800 reactor
The BN-800 design is based on progressive solutions to enhanced safety:
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Prospects of the fast-reactor technology-BN-1200
Objectives of BN-1200 development
Safety of the BN-1200 reactor
Conclusions
Acknowledgment
References
Comparison of thermophysical properties of reactor coolants
Introduction
Generations II, III and III+ reactor coolants
Generations-IV reactor coolants
Reactor coolants by type
Fluid coolants
Gas coolants
Liquid-metal coolants
Molten-salt coolants
Thermophysical properties of Generation III, III+, and IV reactor coolants
Heat transfer coefficients in nuclear-power rectors
Conclusions
References
Index