Physics of nuclear reactors

Front-Matter_2021_Physics-of-Nuclear-Reactors
	Front Matter
Copyright_2021_Physics-of-Nuclear-Reactors
	Copyright
Contributors_2021_Physics-of-Nuclear-Reactors
	Contributors
Foreword_2021_Physics-of-Nuclear-Reactors
	Foreword
Preface_2021_Physics-of-Nuclear-Reactors
	Preface
Acknowledgments_2021_Physics-of-Nuclear-Reactors
	Acknowledgments
Chapter-1---Introduction_2021_Physics-of-Nuclear-Reactors
	Introduction
		Introduction
		The atom and its nucleus
			The Bohr model of the atom
			The atomic energy levels
			Ionization
		The nucleus
			Nuclear dimensions and density
			Nuclear charge distribution
			Representation of a nucleus
			Isotopes, isotones, isobars
		Nuclide number density
		Nuclear stability
		Binding energy
			Binding energy per nucleon (BE/A)
			Binding energy of the last nucleon
			The drip lines
		Radioactivity
			Alpha decay
			Beta decay
			Positron decay or β+ decay
			Electron capture
			Gamma decay
			Rate of radioactive decay
			Mean life
			Half-life
			Units to express radioactivity emission
		Nuclear models
			The liquid drop model
			The shell model
		Nuclear energy levels
			Bound and excited energy levels
		Nuclear reactions
			Q-value of nuclear reactions
		Neutron interactions
		Neutron elastic scattering
			Kinematics of neutron elastic scattering
			The moderator
		Neutron inelastic scattering
			Discrete and continuum inelastic scattering
			Kinematics of neutron inelastic scattering
		Radiative capture
		(n, x) reaction
		Fission
		Probabilities of nuclear reactions
			Reaction rate and microscopic cross-section
			Flux, macroscopic cross-section, mean-free-path
			Total, capture, and absorption cross-sections
		Variation of cross-section with energy
		Compound nucleus reaction versus direct reaction
			Resonances
			Resolved and unresolved resonances
		Cross-section data representation and resonance formalism
		Fission mechanism
			Chance fissions
			Fission cross-sections
			Fission products
			Asymmetric fission
			Instability of the fission fragments
		Fission neutrons
			Prompt and delayed
			Fission neutron energies
		Energy released in fission
		Neutron chain reaction and generations
			Fission chain reaction
			Neutron generations and multiplication
		Concept of a nuclear reactor
			Criticality, critical size, and critical mass
			Basic design considerations
		Thermal reactor and fast reactor
			Energy region and fissile enrichment
			Energy distribution of neutrons in a reactor
		Spatial dependence of the neutrons in a reactor
		Basic physics of reactor design
		The reactor-power
		Summary
		Exercise problems
		References
		Further reading
Chapter-2---Nuclear-data_2021_Physics-of-Nuclear-Reactors
	Nuclear data
		Introduction: Nuclear data and its importance
			Application areas of nuclear data
		Nuclear data production
			Measurement of cross-sections
			Prediction of cross-sections
			Measurement of total, partial, and differential cross-sections
			Total cross-section estimation
			Partial cross-section estimation
			Use of cross-section standards
			EXFOR and CINDA
			Reaction model principles and cross-section predictions
			The spherical optical model
			The compound nucleus model
			Nuclear level densities
			The components of the potential
			Reaction model codes
			Validation of predicted data
			Error and co-variance data
		Nuclear data evaluation
			Evaluated nuclear data files
			The ENDF format
			Point-data representation in ENDF
			Resonance parameter data in ENDF
			The SLBW formalism
			The MLBW formalism
			The RM formalism
			Other RRR formalisms
			The unresolved resonance region
		Nuclear data processing
			Linearization
			Resonance reconstruction
			Doppler broadening
			Calculation of Doppler broadened resonance cross-sections
			Cross-section multigrouping
			Weighting function and self-shielding
			Problem-independent multigroup cross-section set
			Obtaining effective cross-sections in the resonance region
			Self-shielding in the URR
			Effect of Doppler-broadening on the reaction rate
			Group cross-sections in the resonance region for thermal neutron reactors
			Group-to-group transfer cross-sections
			Thermal scattering cross-sections
			Production cross-sections for particles other than neutrons
			Transport cross-sections
			Multigrouping the cross-sections from a linearized grid
			Cross-sections for shielding applications
			Displacement cross-sections
			Nuclear data processing codes
			Multigroup data validation
		Summary
		Exercise problems
		References
		Further reading
Chapter-3---Types-of-nuclear-reactors_2021_Physics-of-Nuclear-Reactors
	Types of nuclear reactors
		Introduction
		Classifications of nuclear reactors
			Based on materials used in reactors
			Other classifications
		Thermal neutron reactors
			Graphite moderated and water cooled reactors
			RBMK reactors
			Gas-cooled graphite-moderated reactors
			Boiling water reactors
			Pressurized water reactors
				Russian pressurized water reactors
			Pressured heavy water reactors
		Fast neutron reactors
			Breeding
			Sodium-cooled fast neutron reactors
		Future energy systems
		Fusion reactor systems
			Lawson criteria
			Inertial confinement
			Magnetic confinement
			Early fusion devices
		Fission-fusion hybrids
		Particle accelerators [11]
			LINACs and tandem accelerators
			Cyclotrons
			Synchrotron radiation sources
			Large hadron collider at CERN [13]
		Accelerator-driven subcritical systems (ADSS)
			Operational aspects of accelerator-driven system and its interface with reactor
			Overview of ADSS projects around the world
			The CERN energy amplifier
			The Los Alamos waste transmuter
			Other concepts
		One way coupled booster reactor concept in BARC
		More topics of interest
		Exercise problems
		References
		Further reading
Chapter-4---Homogeneous-reactor-and-neutron-diffu_2021_Physics-of-Nuclear-Re
	Homogeneous reactor and neutron diffusion equation
		Introduction
		Neutron density and flux
		Neutron continuity equation
		Transport cross-section
		Fick's law of neutron diffusion
			Physical interpretation
			Fick's law limitations
		Diffusion equation
			Boundary conditions for the steady-state diffusion equation
		Neutron diffusion in nonmultiplying media
			A point source in an infinite medium
			Line source in an infinite media
			Plane source in a finite medium (bare slab)
			Physical meaning of diffusion length
		Neutron diffusion in a multiplying media
			Time-dependent flux in a slab reactor
			Finite bare cylindrical reactor
		Reflected infinite slab reactor
			Reflector savings
		Neutron life cycle in a thermal reactor
		Slowing down equation
			Moderation of neutrons in a medium having A>1
			Slowing down in an infinite medium of moderator and resonant absorber
		Neutron thermalization
			Monoatomic gas as moderator
			Bound moderator atoms in liquids and solids
				Water
				Heavy water
				Graphite
			Maxwellian distribution of neutron flux
			Reaction rate in a Maxwellian spectrum
			Comparison of PWR spectrum and FBR spectrum
		Two-group diffusion theory
		Multigroup diffusion equation
		Numerical solution of the diffusion equation in 1-D slab geometry
			Center mesh finite differencing scheme
				Boundary conditions
			Corner mesh finite differencing scheme
		Summary and more topics of interest
		Exercise problems
		References
		Further reading
Chapter-5---Methods-of-solving-neutron-transport_2021_Physics-of-Nuclear-Rea
	Methods of solving neutron transport equation
		Introduction
		Assumptions in the neutron transport theory
		Neutron-nucleus interaction cross-sections
			Differential cross-section
		The neutron transport equation
			Formulation of the neutron transport problem
			Interface conditions
			Boundary conditions
			Time-independent transport equation: Fixed source problem
			Time-independent transport equation: k-eigenvalue problem
			The eigenvalues of neutron transport equations and their relations
		Solution to the neutron transport equation
			Multigroup approximation: Energy discretization
			The PN method
			P1 approximation
			The discrete ordinate method
		Numerical solution to the neutron transport equation
			The spatial discretization and sweeping scheme
			Limit on the size of spatial meshes and negative flux issue
			Source iteration method for multidimensions
			Discrete ordinate quadrature sets
				The double PN quadratures
			Discrete ordinate quadrature sets for multidimensional problems
				Level symmetric quadrature set
			The Ray effect
		The synthetic acceleration schemes
			The corrective equation
			The diffusion synthetic acceleration
			The transport synthetic acceleration
		The integral form of transport equation
		Solution to the integral transport equation
			Collision probability method
				Collision probability for slab geometry
				Collision probability for annular geometry
				Boundary condition
			Interface current method
				Discretized flux equation
				Properties of the collision probability matrices
				Calculation of collision probabilities in two-dimensional geometry
					Region-to-region collision probabilities
					Region-to-surface escape probability
					Surface-to-surface transmission probability
				Computation of collision probability integrals
				Use of boundary condition and solution to Collision Probability equations
		Method of characteristics approach
			MOC solution to one-dimensional slab geometry
			MOC solution for two-dimensional geometry
			Neutron tracking and quadrature weights
				Tracking of neutron trajectories
				Quadrature weights for polar and azimuthal angles
			Solution to the multigroup MOC equation
		Monte Carlo neutron transport
			Validity of Monte Carlo estimates
				Chebyshev inequality
				Law of large numbers
				The central limit theorem
			Random variable
				Uniform random variable
				Sampling from continuous distribution
			Simulating neutron transport with Monte Carlo method
				Simulating a neutron's history
				Sampling of source neutrons
				Sampling for neutron direction cosine from isotropic distribution
				Sampling for type of collision
			Tally computation
				Collision estimator
		More topics of interest
		Exercise problems
		References
Chapter-6---Fuel-burnup--fuel-management--and-fuel_2021_Physics-of-Nuclear-R
	Fuel burnup, fuel management, and fuel cycle physics
		Introduction
		Fuel burnup and its effects
			Energy production
			Changes in fuel composition and fission product buildup
				Units for measuring fuel burnup
			Fission product buildup
			Reactivity changes, effect of burnable poison, and absorber burnup
			Changes in kinetic parameters
		Bateman equation and solutions
			Solution to the Bateman equations
				Analytical methods
				Matrix exponential method
				Eigen-value methods
				Numerical recipes for solution of the Bateman equation
			Input data bases and reactor-specific cross-section sets
			Decay heat estimations
			Dose rates and radiotoxicities
		In-core fuel management
			Principles of continuous and batch refueling, in-core burnup and discharge burnup
				Out-to-in fueling
				In-to-out fueling
				Scattered fuel loading/checkerboard scheme
			Refueling aspects of BWR
			Refueling aspects of PHWRs
			Fuel management in PWRs
			Fuel management in FBRs
			Fuel doubling time
		Fuel management algorithms
		Long life cores
		Fuel utilization
			Components of a fuel cycle
			U-Pu fuel cycle
			Thorium fuel cycle
			FBR fuel cycle and closed fuel cycle
				System doubling time
		More topics of interest
		Exercise problems
		Additional exercises
		References
Chapter-7---Nuclear-reactor-kinetics_2021_Physics-of-Nuclear-Reactors
	Nuclear reactor kinetics
		Introduction
		Definition of important kinetics parameters
		Prompt and delayed neutrons
			Delayed neutrons
			Delayed photo neutrons
		The point kinetics equations
			Derivation of point kinetic equations
			The solution of point kinetics equations
				Step change in reactivity
				Relation between reactor period and reactivity
					Importance of delayed neutrons
				Prompt jump/drop
				Reactivity unit as Inhour
				One group of delayed neutrons
			Ramp reactivity change
				Periodic change in reactivity
			Inverse point kinetics
		Kinetics of a subcritical reactor
			Subcritical reactor equilibrium power
			Power behavior with neutron source addition or withdrawal
			Power at criticality with ramp reactivity addition
			Instantaneous reactor period
			Safe approach to criticality
				Inverse count rate plot
				The maximum permissible reactivity addition rates during startup
		Numerical solution of point kinetics equations
			Simple example of determination of phi
			Solution of Eq. (7.77)
			Solution of Eq. (7.77) with A in two parts
		Reactivity feedback in nuclear reactors
			Factors causing reactivity change
			Thermal reactor feedback reactivity
			Fast reactor reactivity feedback
			Reactivity coefficients
				Static coefficients
				Dynamic reactivity coefficient
		Long irradiation reactivity effects in reactors
			Xenon poisoning effects
				Equilibrium xenon concentration
				Xenon poisoning after shutdown
			The buildup of samarium poison
			Other fission product poisons
		Reactor stability analysis
			Linear stability analysis
				Stability of zero power reactor
				Stability of reactor with feedback
			Frequency domain methods of stability analysis
				Zero power transfer function (ZPTF)
				Feedback transfer function (FBTF)
				Dynamic power coefficient of reactivity
				Response of reactor for sinusoidal reactivity input
				Nyquist stability criterion
					Contour mapping
				Information about relative stability
				Nyquist stability: ZPTF
				Nyquist stability: FBTF
			Nonlinear stability analysis
		Neutronic coupling in a reactor
			Size effect
			Shape effects
			Safety concerns
		Xenon oscillations
			Affected systems
			Controlling spatial oscillations
			Example: PWR axial control
		Space time kinetics
			Necessity of space-dependent modeling
			Evolution of space time kinetics methodology
				Solution of multigroup diffusion equations
			Flux factorization method
				Amplitude function
				Shape function
			Coupled neutron kinetics and thermal hydraulics [20]
			Case study [19-21]
		Summary
		Annexure
			Formal statements of stability and Lyapunov's direct method for global stability
				Lyapunov stability
				Asymptotic stability
				Exponential stability and rate of convergence
				Bounded-input-bounded-output stability
				Global stability of nonlinear systems
					Lyapunov's direct method for nonlinear stability
					Lyapunov's indirect theorem
		Exercise problems
		References
Chapter-8---Nuclear-reactor-safety_2021_Physics-of-Nuclear-Reactors
	Nuclear reactor safety
		Introduction
			Broad safety objectives
			Defense in depth
			Safety during reactor operation
			The three Cs of nuclear safety
			Risk evaluation
			Safety assessment (regulatory authorities)
		Fundamental safety principles and safety framework [2, 9-11]
			Fundamental safety principles
			Safety framework
		Safety requirements and graded approach in safety
			Plant states
			Acceptance criteria
		Safety features in a NPP
			Inherent safety
			Reactor control system
			Reactor protection system
			Emergency core cooling system
			Reactor containment
			Active and passive safety features
			Analysis for safety assessment
		Deterministic safety assessment [12-22]
			The objective
			Evolution of DSA
			A case study of DSA (postulated rod withdrawal accident in a PWR) [18-22]
		Reliability analysis of safety systems [4, 23-30]
			Approach and aspects of reliability analysis
			Design and system function
			System boundary and system failure modes
			Common cause failures
				Beta (β) factor model
				Alpha factor model
			Human reliability analysis
				Human cognitive reliability model
			Data collection
			Reliability quantification
			Other issues
			Summary
		Probabilistic safety assessment: Level-1, Level-2, and Level-3. Case studies [31-54]
			Introduction approach to PSA
			Basis and approach
			Level-1 probabilistic safety assessment
				Management and organization
				Collection of information on design and operation of the plant
				Identification of sources of radioactivity and accident initiators
				Success criteria formulation and supporting analysis
				Accident sequence analysis
				Systems analysis
				Human reliability analysis
				Dependent failure analysis
				Data assessment and parameter estimation
				Model integration and core damage frequency quantification
				Sensitivity and uncertainty analysis
				Interpretation of results
				Quality assurance aspects of PSA
				Documentation of the analysis: Display and interpretation of results
				Computer programs available to perform Level-1 PSA
			Level-2 probabilistic safety assessment
				Interface between the Level-1 and Level-2 PSA
				Containment analysis and accident progression
				Quantification of the containment event tree and categorization of the endpoints into release categories
				Radiological source term analysis for the release categories
				Uncertainty analysis and sensitivity analysis
				Documentation of the analysis: Display and interpretation of results
				Codes available to perform Level-2 PSA
			Level-3 probabilistic safety assessment
				Interface between Level-2 analysis and Level-3 analysis
				Radionuclide release characterization
				Meteorological data and other information sources
				Environmental transport and deposition
				Radiological exposure and dose assessment
				Protective action (counter measure) modeling
				Economic consequences
				Uncertainty analysis and sensitivity analysis
				Documentation of the analysis: presentation and interpretation of results
				Programs available to perform Level-3 PSA
			Use and development of probabilistic safety assessment: Case studies
			Conclusion
		Uncertainty analysis in reliability and risk assessment [55-60]
			Introduction
			Types of uncertainty
				Input data (parametric) uncertainties
				Model uncertainty
			Uncertainty propagation [59, 60]
			Uncertainty methodology in PSA
				Monte-Carlo technique
				Discretization
				Method of moments
				Method of adjoint operator
			Importance and sensitivity analysis
				Importance measures
				Sensitivity analysis
		Major nuclear reactor accidents
			The three Mile Island-2 accident [2, 3]
			The Chernobyl accident [2, 3, 61, 62]
				Cause of accident
				The accident propagation sequence
					The event: April 25, 1986
					The event: April 26, 1986
					Actions initiated
			The Fukushima accident [3, 63]
		Brief history of the nuclear safety
		Summary
		Exercise problems
		References
Chapter-9---Design-methods-and-computer-code_2021_Physics-of-Nuclear-Reactor
	Design methods and computer codes
		Introduction
		Methods of neutronics analysis in thermal reactors
			Neutron cross-section library for lattice and core calculations
				Resonance cross-section treatment
					Narrow resonance approximation
					Intermediate resonance approximation
			Heterogeneous system
			Fuel assembly analysis methods
				Supercell methods
				Direct transport methods
				Improvements in homogenization
			Description of lattice analysis codes
				Indian lattice analysis codes
				International lattice analysis codes
			Methods for 3-D whole-core analysis
				Methods using assembly or pin homogenization
					Finite difference method
					Nodal method
				Methods without homogenization or one-step methods
			Description of whole-core analysis codes
				Codes using assembly homogenization or pin-cell homogenization (India)
				A few examples of codes using fuel assembly homogenization or pin-cell homogenization (world)
				Codes using single-step methods
		Methods of neutronics analyses in fast reactors
			Self-shielding calculations using subgroup/probability table method
				Theory of probability table method
				Effective cross sections using subgroup parameters
				Heterogeneity correction to cross sections using subgroup/probability table method
			ABBN-90 group constants system
			ERANOS system
			Indian codes and methodology for fast reactor analysis
				Steady-state deterministic neutronics analysis using FARCOB system
				Steady-state stochastic neutronics analysis using OpenMC
				Safety analysis methods and codes used
		More topics of interest
		References
Chapter-10---Experimental-and-operational-reacto_2021_Physics-of-Nuclear-Rea
	Experimental and operational reactor physics
		Introduction
		Neutron monitoring: Neutron detectors and instruments
			Modes of detector operation
				Pulse mode of operation
				Current mode of operation
			Detectors used for reactor start-up/power regulation
				Ion chambers
				Fission counter
				BF3 proportional counter
				10B-lined neutron detector
				Self-powered neutron detector
			In-core and out-core detectors
			Overlapping regimes for neutron monitoring during a reactor start-up
			Gamma sensitivity of neutron detectors
		Neutron flux measurement using activation method
			Characteristics of neutron spectrum in nuclear reactors
				Thermal neutron spectrum (Maxwellian spectrum)
				Intermediate neutron spectrum
				Fast flux-Fission neutron spectrum
			Foil activation methods
				Theory of activation method
					Self-shielding factor
					Flux depression factor:
				Activation detector properties
					Gold foil
					Copper foil
					Indium foil
					Manganese foil
				Thermal flux measurement and cadmium ratio
					Effective cadmium cut-off
				Measurement of fast neutron fluxes
			Measurement of Westcott spectrum parameters ``r´´ and ``T´´
			Neutron spectrum measurement
				One group case-total flux
				Two-group flux measurement
				Measurement of neutron spectrum in arbitrary number of groups
				Overview of SAND-II spectrum unfolding code
		Start-up or commissioning experiments in reactors
			Approach to criticality and subcritical multiplication measurements
			Nuclear instrumentation for start-up
			Neutron sources for reactor start-up
				Spontaneous fission neutrons
				Photo neutrons
				External neutron sources
			Neutron multiplication in a subcritical system
			Inverse multiplication (count rate) curves
			Reactivity addition rate and reactor period during first approach to criticality
		Reactivity measurements
			Asymptotic period method
			Rod drop method
			Subcritical counts method
			Inverse kinetics method
			Measurement of integral and differential worth of control rods
		Low power physics experiments in thermal reactors
			Calibration of the reactivity worth of regulation and protection devices
			Dynamic testing and measurement of worth of reactor protection systems
			Temperature coefficient measurement
			Absolute power calibration and thermal power measurement
			Measurement of xenon load
		Reactor start-up in sodium cooled fast reactors
			Initial fuel loading and FAC
			Control rod worth estimation
			Isothermal temperature coefficient measurement
			Power and flow coefficient measurements
			Power raising operation
		Failed fuel detection
			Failed fuel detection in thermal reactor systems
			Failed fuel detection in fast reactor systems
				Fast Breeder Test Reactor experiment of fuel pin clad rupture detection
		Regulatory aspects and reactor experimentation and operation
		Some critical/subcritical experimental facilities
			Examples of thermal reactor critical facilities
				ZERLINA
				PURNIMA I, II, and III
				KAMINI
				Advanced Heavy Water Reactor Critical Facility (AHWR-CF)
			Subcritical facilities with source
				MUSE
				YALINA
				BRAHMMA
			Some fast critical and experimental systems
				The zero power reactor (ZPR) and zero power physics reactor (ZPPR) facilities
				The Russian BFS critical facilities
				The MASURCA critical facility
				The experimental breeder reactor-I (EBR-I)
				The experimental breeder reactor-II (EBR-II)
				The fast breeder test reactor
		More topics of interest
		Exercise problems
		References
Chapter-11---Radiation-safety-and-radiation-shiel_2021_Physics-of-Nuclear-Re
	Radiation safety and radiation shielding design
		Introduction
		Basic radiation physics
			Fundamental definitions and concepts
			Alpha particles
				Bragg curve
			Beta particles
				Range of β particles
			Photons
				Photoelectric interactions
				Compton scattering
			Pair production
			Photon attenuation in materials
		Radiation dosimetry
			Radiation dose and units
			Radiation dose equivalent
			Radiation exposure
			External and internal exposures
			Radiation dose calculations
				Gamma radiation dose calculations by point kernel methods
				Point source in a vacuum
				Point source with a shield
				Extended sources
					Example: Line source
					Example: Disc source
				Buildup factor
				Dose rates with buildup factors
				Taylor's form
			Neutron radiation attenuation
				Boltzmann transport equation
				Discrete ordinates method
				Monte Carlo method
				Cross-sections for shielding applications
				Neutron irradiation effects in metals
				Displacement per atom
		Gamma shields
			Lead
			Bismuth
			Tungsten
			Iron and steel
			Concrete
			Half-value and tenth-value layers
		Neutron shields
			Hydrogenous materials
			Boron and its compounds
			Lithium
		Reactor sources of radiation
			Reactor core and vessel
			Reactor coolant system
			Steam and turbine system
			Radioactive waste system
		Radioactive sources in fuel cycle facilities
		Radiation dose limits for exposures
			Occupational dose limits
			Dose limits for members of the public
			Supervised areas
			Controlled areas
			Zone classification
		Shield design for reactors
			Interaction between shielding design and core design
				Low neutron leakage core design in LWR and FR
				Core monitoring and shield design in a pool type fast reactor
			Examples of shield design
				PHWR: TAPS 3 and 4 reactor
				Fast reactor: PFBR
				Secondary sodium activity computation in PFBR
		Complementary shielding
			Line of sight
				Cylindrical duct
				Rectangular duct
			Transport calculations
				Example: Streaming problem in AHWR top-shield
				Example: Apsara model for streaming through gaps in the transfer arm in PFBR
		Shield design methods in fuel cycle facilities
			Reprocessing plant
			Fuel fabrication
			Fuel assembly
			Waste disposal
				Example: Shield requirement estimation for a typical fuel pin storage room
		Summary and more topics of interest
		Exercise problems
		References
		Further reading
Chapter-12---Nuclear-reactors-of-the-future_2021_Physics-of-Nuclear-Reactors
	Nuclear reactors of the future
		Introduction
		Generation IV reactors
			Gen IV reactor types
				Very high temperature reactor
				Sodium-cooled fast reactor
				Lead-cooled fast reactors
				Gas-cooled fast reactor
				Molten salt reactor
				Super critical water-cooled reactor
				The Gen IV parameters
			Reactor physics challenges
			Indian high temperature reactors
				Compact high temperature reactor
				Innovative high temperature reactor
			Conclusions
		Small and modular reactors
			Introduction
			General features
				Size based
				Design based
			Classification of SMRs
				Temperature and power range
				LWR/PHWR water reactors
					A1. PHWR-220 (NPCIL, India)
					A2. AHWR-300 (BARC, India)
				Fast neutron spectrum reactor
					A1. GE-Hitachi (PRISM)
					A2. SVBR - 100
					A3. BREST-OD-300
				High temperature reactors
					HTR-PM (Tsinghua University, China)
				Molten salt reactors
					Integral molten salt reactor (Terrestrial Energy, Canada)
			The impact of SMRs
			Conclusions
		Traveling wave reactors
			Introduction
			The fuel burnup wave
				The genesis
				The mechanism
				Characteristics
			Requirements for sustained fuel burnup wave
				The necessary conditions
					A1. Feoktistov approach
					A2. The CANDLE fuel burnup
					A3. The van Dam approach
					A4. UC Berkeley approach
				The neutronics
				The constraints
			Mathematical methodology
				Mathematical approach
					A1. Steady state
					A2. Transient state
				General results
			Different types of B&B reactors
				Edward Teller
				HTGR TWR
				The CANDLE reactor
				The TerraPower standing wave reactor
				UC Berkeley SWR
			Conclusions
		Review questions
		References
		Further reading
Index_2021_Physics-of-Nuclear-Reactors
	Index
		A
		B
		C
		D
		E
		F
		G
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