course 
code 
teacher 
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ss 
ws cr. 
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Compulsory courses 
Nuclear Reactor Physics  17FAR 
Fejt, Frýbort, Frýbortová, Sklenka 
2+2 z,zk 
  
5 
 
Course:  Nuclear Reactor Physics  17FAR  doc. Ing. Sklenka Ľubomír Ph.D.  2+2 Z,ZK    5    Abstract:  Subject deals with nuclear reactor physics in lower advanced level  is consequential to introductory course read in bachelor degree course (17ZAF). Lectures on theoretical basic of neutron transport, advanced diffusion, critical equation are given to students. Also practical issues of reactor physics are mentioned.  Outline:  Multiplication factor in homogenous and heterogeneous
Scope: 2 lectures
Detailed derivation of four factor formula and six factor formula, differences between homogenous and heterogeneous systems, moderator to fuel ratio, factor?s derivation for heterogeneous system, practical use.
2. Neutron slowing down
Scope: 2 lectures
Detailed derivation and description of slowing down models, resonance theory, lethargy, parameters and factors calculations
3. Critical equation
Scope: 1 lecture
Derivation and solution of critical equation, solution of critical states and critical masses for variation geometries and boundary conditions
4. Neutron transport equation
Scope: 4 lectures
Classical heuristic derivation of neutron transport equation, statistical kinetics equation derivation, integrodifferential, integral, and other forms, solution and applicability of neutron transport equation
5. Slowing down kernels
Scope: 2 lectures
Derivation of neutron diffusion equation in general form, application of slowing down kernels into diffusion and transport equation, solution of diffusion equation with kernels, application to subcritical systems with external neutron source
6. Perturbation theory and adjoined equations
Scope: 1 lecture
Perturbation theory and its application in nuclear reactor physics, derivation and solution of adjoined equation, adjoined diffusion equation, determination of reactivity coefficients
7. Practical approach to nuclear reactor physics
Scope: 2 lectures
Lectures given by reactor physics experts working in industry  ŠKODA JS, a.s., ÚJV Řež, a.s., ČEZ, a.s. (EDU, ETE)
 Outline (exercises):  Multiplication factor on moderator to fuel dependency, slowing down factors determination, critical states solution, transport equation forms derivation, slowing down kernels application to general diffusion theory, subcritical system with external neutron source solution, adjoined diffusion equation solution
Credit thesis  derivation, application, calculation, and computer visualization of given reactor physics problem
 Goals:  Knowledge of relations in nuclear reactor physics
Skills and ability to calculate simple reactor physics problems, derivate basic equation and know their applicability
 Requirements:  17ZAF  Key words:  nuclear reactor, nuclear reactor physics, transport equation, diffusion equation, critical equation  References  Key references:
Stacey, W. M.: Nuclear Reactor Physics, WILEYVCH Verlag GmbH & Co. KGaA, Weinheim, 2007
Reuss, P.: Neutron Physics, EdP Sciences, 2009
Recommended references:
Galanin, A.D.: Teorie tepelných jaderných reaktorů, SNTL, 1959
Ganapol P.: Analytical Benchmarking in Neutron Transport Theory, Univ. of Arizona, 2009


Core Physics and Fuel Management  17PRF 
Sklenka 
  
2+0 z,zk 
 
3 
Course:  Core Physics and Fuel Management  17PRF  doc. Ing. Sklenka Ľubomír Ph.D.    2+0 Z,ZK    3  Abstract:  The course is focused on inner nuclear fuel cycle of the nuclear power plants, particularly PWR used and / or planned in the Czech Republic. The first part of the course consists of introduction to the core physics, e.g. fuel changes during the cycle, burnup, changes of keff during the cycle, xenon poisonings and xenon oscillations, samarium, etc. The second part of the course consists of NPP fuel cycle, fuel burnup and fuel management, e.g. fuel handling, fuel management, reactor operation, burnup, fuel loading, fuel reloading, loading pattern, legislative requirements for the core, core loading and fuel handling, fuel cycle of WWERs PWR, Fuel cycle of Dukovany & Temelín NPP, fuel cycle of western PWRs, BWR fuel cycle, CANDU fuel cycle. At the end of the course basic information about MOX fuel is mentioned. Note: Frontend & backend of the nuclear fuel cycle of the nuclear power plants is the part of 17JPC  Nuclear fuel cycle course  Outline:  1. Introduction
Duration: 1 lecture
Topic:
Introduction, familiarization with the structure of lectures and seminars, graduation requirements, definitions of basic concepts, introduction to the nuclear fuel cycle, frontend of the nuclear fuel cycle, inner nuclear fuel cycle and backend of the nuclear fuel cycle
2. Longterm reactor kinetics
Duration: 2 lectures
Topics:
Simple model of longterm reactor kinetics of the uraniumplutonium fuel cycle, fuel changes during the cycle, burnup, and changes of keff during the cycle
Fission products, assignment of the first part of the students' seminar work, deep burnup, longterm reactor kinetics of the thorium fuel cycle
3. Middleterm reactor kinetics
Duration: 2 lectures
Topics:
Xenon and samarium in the core during the cycle, simple model of middleterm reactor kinetics, xenon poisoning, assignment of the second part of the students' seminar work,
Xenon oscillations, castdown operation  thermal & power reactivity coefficients at the end of cycle, burnup absorbers
3. Linear reactivity model
Duration: 1 lecture
Topic:
Linear reactivity model, nonuniformity of power distribution in the core, neutron leakage from the core, leak reactivity, nonlinear reactivity model
4. NPP fuel cycle, fuel burnup and fuel management
Duration: 5 lectures
Topics:
Nuclear fuel cycle, inner nuclear fuel cycle, core physics, fuel handling, fuel management, reactor operation, burnup, fuel loading, fuel reloading, loading pattern, legislative requirements for the core, core loading and fuel handling
Inner fuel cycle strategy, core modelling methods, optimization of the core loading
Computer codes for core calculation, MOBYDICK computer code, practical use of the MOBYDICK code for WWER440 core calculation
Fuel cycle of WWERs PWR, Fuel cycle of Dukovany & Temelín NPP
Fuel cycle of western PWRs, BWR fuel cycle, CANDU fuel cycle
5. MOX fuel
Duration: 1 lecture
Topic:
Uranium  plutonium fuel cycle & reprocessing the spent nuclear fuel, MOX fuel, physics differences of MOX fuel, reactor operation with MOX fuel, use of MOX fuel in the world, perspective of MOX fuel in next generation of NPP
6. Students' seminar works evaluation
Duration: 1 lecture
Topic:
Students' seminar works, evaluation and discussion with the lecturer
 Outline (exercises):    Goals:  An overview of inner nuclear fuel cycle, core physics & fuel management of the nuclear power plants.
Application of acquired knowledge to solve problems, qualification and quantification of the effects of various physical quantities and phenomena on the operation of nuclear reactors and nuclear safety  Requirements:  17ZAF  Key words:  Nuclear reactor, nuclear fuel cycle, inner nuclear fuel cycle, core physics, fuel handling, fuel management, reactor operation, burnup, xenon, samarium, fuel loading, fuel reloading, loading pattern, Dukovany and Temelín NPP fuel cycle, PWR fuel cycle, BWR fuel cycle, CANDU fuel cycle, MOX fuel  References  Key references:
Stacey, W. M.: Nuclear Reactor Physics, Chapter 5 Nuclear Reactor Dynamics &
Chapter 6  Fuel Burnup, WILEYVCH Verlag GmbH & Co. KGaA, Weinheim, 2007
John R. Lamarsh: Introduction to Nuclear Engineering, 3rd Ed., Prentice Hall, 2001
Recommended references:
Core Management and Fuel Handling for Nuclear Power Plants, IAEA Safety Guide, NSG2.5, IAEA, Vienna, 2002
Design of the Reactor Core for Nuclear Power Plants, IAEA Safety Guide, NSG1.12, IAEA, Vienna, 2005
Media and tools:
Computer laboratory  2x2h, audiovisual technique, fuel cycle films on DVD 

Reactor Dynamics  17DYR 
Heřmanský, Huml 
  
2+2 z,zk 
 
4 
Course:  Reactor Dynamics  17DYR  prof. Ing. Heřmanský Bedřich CSc.    2+2 Z,ZK    4  Abstract:  Kinetics of reactors, delayed neutrons, prompt neutron mean lifetime, reactor period. Dynamics of a zero reactor  the formulation of shortterm kinetic equations and parameters of delayed neutrons, simplified solutions. Transfer function of zero reactor. Coefficients of reactivity for different reactor configurations, temperature coefficients, thermal feedback, stability of reactors, linear and nonlinear kinetics. Heat transfer in reactors, reactor dynamics. Mathematical model of power reactor with thermal feedback, Simplified models of the reactor dynamics, computer models of the reactor dynamics
 Outline:  1st Dynamics of zero reactor (shortterm kinetics)
Hours: 6 lectures
Topics of the lectures:
Reactor Kinetics Equations
Integration of the lectures to study and interaction with other lectures, the aim of lectures. Diffusion equation. Onegroup approximation. Kinetics equations without delayed neutrons, delayed neutrons, the reactor period and the effect of delayed neutrons. Parameters of delayed neutrons. Kinetics equations with delayed neutrons (production and destruction formulation), the initial conditions.
Integral form of kinetic equations
Laplace integral transformation. Transfer functions and dynamic response basics. Derivation of kinetic equations in integral form. Roots and constants of G0 function for the six groups of delayed neutrons case. "Best" parameters of delayed neutrons, energy spectrum correction.
Analytical solutions of kinetic equations
Response to an impulse change in reactivity (impulse response). Response to a step change in reactivity (transient response). Asymptotic period of a reactor. Special cases of a step change in reactivity. One group of delayed neutrons approximation. Response to linear changes in reactivity.
A simplified form of kinetic equations
Constant production of delayed neutrons. Prompt jump approximation: formulation of the equation in onegroup approximation, simplified response to step, harmonic and linear changes in reactivity. Numerical solutions of kinetic equations.
Transfer function of a zero reactor
Linearized model of a zero reactor. G0 function as transfer function of linearized zero reactor. Responses of linearized model to step and harmonic changes in reactivity.
Frequency response of zero reactor
Oscillatory experiments. Frequency response of linearized model of zero reactor. Bode plot. Logarithmic frequency response. Frequency response of zero reactor simplified models. Stability of zero reactor.
2nd Effect of temperature changes on the reactivity
Hours: 1 lecture
Topic of the lecture:
Dynamical systems with feedback. The stabilizing effect of the negative feedback. Reactivity temperature coefficients (RTC). Large reactor RTC. Doppler effect. Pressurized water reactors RTC (fuel RTC, coolant RTC). Effect of the boron concentration on the temperature feedback. Reactor's reactivity coefficients.
3rd Mathematical model of a power reactor
Hours: 3 lectures
Lectures:
Heat transfer in nuclear power reactors
Basic equations. Quasistationary approximation. Fuel and coolant equations in quasistationary approach (lumped parameters). Adiabatic model of the core heating. Differential approach (distributed parameters).
Mathematical model of the reactor with temperature feedback
Mathematical model and mathematical simulation of transients. Onechannel twocomponent (or threecomponent) model of the pressurized water reactor core. Classification of mathematical models according to the diffusion equations solution and thermalhydraulic processes.
Simplified models of the reactor dynamics
Nonlinear models: the NordheimFuchs model, oscillating reactor, reactor oscillation damping, model immediately jump. " Integral models: the adiabatic model, the integral model with heat losses. Linearized models with lumped parameters.
5th Transient heat transfer in the core  distributed parameters
Hours: 2 lectures
Lectures:
Analytical solutions of the transient heat transfer equation
Basic heat transfer equations, initial and boundary conditions. Solution via Laplace transformation. Formulation of heat transfers in fuel rods. First and second order approximations of the fuel rod transfer functions. Temperature delay of fuel rods. Impulse and step response.
Analytical solutions of the transient heat transfer equation of fuel channel
Basic fuel and coolant equations, initial and boundary conditions. Solution via Laplace transformation. Formulation of heat transfers in fuel channel. First and second order approximations of the fuel rod transfer functions. Temperature delay of fuel channel. Impulse and step response.
 Outline (exercises):  During the seminars the problems of the above chapters are solved and the basic formulas are derived. Furthermore, numerical simulations are carried out for some transients. The seminars include measurements of fundamental dynamic processes at the VR 1 reactor.  Goals:  Knowledge: Thorough knowledge of the kinetics of the reactor and the effects of temperature changes on reactor dynamics. Orientation in models of reactor dynamics. Ideas of the nuclear reactors thermal hydraulic analysis issue.
Skills: Understanding of the physical nature of processes taking place in different situations in a nuclear reactor. Application of acquired knowledge in the lectures Safety of Nuclear Power Plants.
 Requirements:  17ZAF, 17JARE, 17TER
 Key words:  Reactor kinetics, delayed neutrons, dynamics of zero reactor, parameters of delayed neutrons, transfer function, coefficients of reactivity, temperature coefficients, reactor dynamics, Laplace integral transformation, impulse response, step response, frequency response  References  Key references:
Heřmanský B.: "Nuclear reactor dynamics", Ministerstvo školství, Praha 1987, (In Czech)
Recommended references:
Kropš S.: "Temelin Low Power Tests", NUSIM 2001, České Budějovice, 2001.


Reactor Thermomechanics  17TERR 
Bílý, Heřmanský 
2+2 z,zk 
  
4 
 
Course:  Reactor Thermomechanics  17TERR  prof. Ing. Heřmanský Bedřich CSc.  2+2 Z,ZK    4    Abstract:  Heat generation in nuclear reactors  distribution and time evolution, residual heat generation. Steadystate and transient heat conduction in fuel elements, heat conduction in cladding, heat transfer in fuelcladding gap. Convection heat transfer in nuclear reactors and boiling crisis of the first kind. Temperature distribution in fuel channel in steadystate and transient conditions. Core hydrodynamics. Hot channel theory. Steady state thermohydraulic calculation of nuclear reactor.  Outline:  1. Heat generation in reactors
Scope: 4 lectures
Energy released by fission
Role of the course within studyprogram, relationship to other courses, goals of the course. Energy released in fission process and recoverable energy. Heat generation in core. Power peaking factors. Heat generation function. Linear heat generation.
Heat generation in cylindrical reactor
Uniform grid. Onegroup equation of the reactor. Heat generation in bare cylindrical reactor. Effect of reflector on spatial distribution of heat generation. Equivalent reactor. Influence of control rods. Influence of voids and gaps. Radial distribution of heat generation in campaign refueling strategy.
Xenon effect and chemical reactions
Xenon effect on spatial power distribution. Initial equation, heat generation distribution at a change of reactor operation regime, xenon oscillations, oscillations in VVER reactors. Chemical reaction of cladding with steam: chemical reaction kinetics, reaction energy, time evolution of cladding oxidation.
Residual heat generation in reactors
Fission reaction after shut down, radioactive fission products decay, radioactive decay of transuranic elements. Importance of residual heat generation for safety of nuclear reactors. Calculation code ORIGEN.
2. Heat transfer in fuel elements
Scope: 4 lectures
Heat conduction in fuel elements
Heat conduction equation for cylindrical geometry. Integral thermal conductivity. Thermophysical properties of the fuel. MATPRO code. Heat conduction in cylindrical fuel rod: simplified approach, effect of radiusdependent heat generation, central gap effect, heat conduction in hollow rod with double sided heat removal.
Heat transfer in fuelcladding gap
Heat transfer coefficient in fuelcladding gap: heat conduction in gas filling, contact thermal conductivity, radiation heat transfer. Thermophysical parameters affecting state of the gap. Schlykov similarity approach. Heat transfer coefficient of the VVER fuel  fresh fuel, burnedup fuel. Calculation models for fuelcladding gap heat transfer.
Heat transfer in core
Theory of similarity, dimensionless numbers. Singlephase flow: convective and conductive heat transfer. Forced convection, natural convection. Spacer grid effect on heat transfer coefficient. Heat transfer coefficient of VVER reactors.
Twophase flow and boiling crisis
Nucleate boiling regime. Nucleate boiling, film boiling. Boiling crisis of the first kind: physical principle, calculation correlations. Transition boiling, stable film boiling. Fuel performance under extreme conditions, results of PCM tests.
3. Steadystate temperature distribution in fuel channel
Scope: 2 lectures
Heat transfer in fuel channels
Energy equation of coolant flow. Temperature distribution in coolant, cladding, and fuel pellet. Homogenous model of fuel rod. Theory of similarity  dimensionless temperatures.
Steadystate temperature distribution in fuel channel
Axial temperature distribution of the coolant in the fuel channel with sine heat generation. Temperatures at fuel pin surface. Temperatures in fuel rod axes. Maximal temperatures at fuel rod surface and in fuel rod axis. Ring fuel rod. Similarity of temperature fields. Boiling in fuel channel.
4. Reactor hydrodynamics
Scope: 1 lecture
Pressure drop: Bernoulli formula, friction of coolant, local hydraulic resistances, acceleration pressure drop, gravity pressure drop. Total pressure drop and coolant distribution in the core. VVER reactors hydrodynamics. Hydraulics characteristics of core, reactor and primary loop. Pumps characteristics.
5.Thermohydraulic reactor analysis
Scope: 1 lecture
Hot channel theory: principle, hot channel factors, temperatures in hot channel. Deterministic and statistic approach. Thermohydraulic analysis of the reactor in steadystate conditions. Limiting criteria on maximal allowable thermal power of the reactor. Maximal allowable power (Operational limits and conditions). Integral computational codes (RELAP, ATHLET)
 Outline (exercises):  Content of exercises supports lectures with concrete calculations.
1. Heat generation in reactors
Energy released by fission, heat generation in cylindrical reactor, chemical reaction of cladding with steam
Scope: 6 tutorials
2. Heat transfer in fuel elements
heat conduction in fuel elements, heat transfer in fuelcladding gap
Scope: 4 tutorials
3. Steadystate temperature distribution in fuel channel
Heat transfer in fuel channels, steadystate temperature distribution in fuel channel
Scope: 3 tutorials
 Goals:  Detailed knowledge of physical aspects affecting spatial heat distribution in nuclear reactors. Orientation in basic laws of heat transfer in reactor core. Conception of nuclear reactor thermohydraulic analysis issues.
Application of basic courses (17ZAF, 17THN1, 17THN2) on nuclear power reactors. Orientation in given issues, use of obtained knowledge in further courses (17DYR, 17JBEZ).  Requirements:  17ZAF, 17JARE, 17THN1,2  Key words:  nuclear reactors heat generation, power peaking factor, linear heat generation, xenon oscillation, residual heat, heat conduction, MATPRO, heat convection, core hydrodynamics, hot channel theory, thermohydraulic calculation, forced convection, natural convection, nucleate boiling, film boiling, boiling crisis, hydraulic characteristics  References  Key references:
Heřmanský B.: "Thermomechanics of nuclear reactors", Academia, Praha 1986, (in Czech)
Recommended references:
Tong, L.S., Weisman, J.: Thermal Analysis of Pressurized Water Reactors, American Nuclear Society, Illinois USA, 1996, ISBN: 0894480383


Experimental Reactor Physics  17ERF 
Rataj, Sklenka 
  
4 kz 
 
4 
Course:  Experimental Reactor Physics  17ERF  Ing. Rataj Jan Ph.D.    4 KZ    4  Abstract:  The lectures are focused on experimental methods used for determination of neutronphysical and basic operational parameters of on nuclear reactors. The lectures deal with research nuclear reactors, their classification and utilisation in the field of experimental reactor physics, experimental methods focused on reactivity measurement, determination of control rod characteristics in the nuclear reactor, dynamics study of nuclear reactor, realisation of the critical experiment. Within the last lectures is prepared basic critical experiment at VR1 reactor.
The lectures are supplemented with experimental practices at the training reactor VR1: reactivity measurement, control rod calibration, dynamics study of nuclear reactor, prediction of unknown critical state. The main part of practices is focused on realization of basic critical experiment at VR1 reactor.
 Outline:  1. Research nuclear reactors
Range: 1 lecture,
Topic of lecture: Experimental methods of nuclear reactor physics. Classification of nuclear facilities and research reactors. Classification, parameters and utilisation of research reactors in the field of experimental reactor physics.
2. Reactivity measurement in nuclear reactors
Range: 1 lecture,
Topic of lecture: Static and dynamics techniques for reactivity determination. Source multiplication method. Sourcejerk method. Roddrop method. Positive period method. Inverse kinetic method.
3. Determination of control rod characteristics in the nuclear reactor
Range: 1 lecture,
Topic of lecture: Integral and differential control rod worth. Control rod worth. Methods focused on determination of control rod characteristics: inverse rate method and source multiplication method. Control rods intercalibration.
4. Studium dynamiky jaderného reaktoru
Range: 1 lecture,
Topic of lecture: Kinetics and dynamics of nuclear reactor. Study of zero power reactor behaviour. Nuclear reactor in subcritical, critical and supercritical state with and without external neutron source. Study of pulse, transient and frequency characteristics of nuclear reactor. Reactivity feedbacks and their influence on nuclear reactor control and operation.
5. Preparation of basic critical experiment at VR1 reactor
Range: 2 lectures,
Topic of lecture:
Basic requirements for core configurations in VR1 reactor. Design of VR1 reactor core configuration. Neutronphysical characteristics of VR1 reactor core and their determination. Legislative requirements of State office for nuclear safety for basic critical experiment realisation.
Procedure of basic critical experiment preparation and realisation. Schedule of basic critical experiment, its structure and content. Critical experiment in nuclear reactor. Critical state prediction by inverse rate method. Approaching unknown critical state in nuclear reactor.
 Outline (exercises):  Practices will be running at training reactor VR1
1. Reactivity measurement at VR1 reactor
Range: 1 practice
Topic of practice: Sourcejerk, roddrop and positive period method application on reactivity measurement at VR1 reactor.
2. Control rod calibration in VR1 reactor
Range: 1 practice
Topic of practice: Determination of control rod worth and its calibration curve in VR1 reactor by inverse rate method and source multiplication method. Control rods interintercalibration in VR1 reactor.
3. Study of nuclear reactor dynamics
Range: 1 practice
Topic of practice: Study of VR1 reactor behaviour in subcritical, critical and supercritical state with and without external neutron source. Study of VR1 reactor response to the pulse, transient and periodic reactivity changes. Study of void reactivity coefficient influence on undermoderated and overmoderated nuclear reactor.
4. Approaching the critical state at VR1 reactor
Range: 1 practice
Topic of practice: Prediction of unknown critical state at VR1 reactor by inverse rate method. Approaching the critical state at VR1 reactor by gradual change of control rod position.
5. Basic critical experiment at VR1 reactor
Range: 3 practices
Topic of practices:
Calculating preparation of the basic critical experiment.
Elaboration of the basic critical experiment schedule.
Realisation of basic critical experiment at VR1 reactor.
 Goals:  detailed knowledge in the field of experimental reactor physics, knowledge and assumption of method focused on determination of basic neutronphysical and operational parameters in nuclear reactors.
orientation in the given problems, application of gained knowledge in the fields of science, research and other experimental subject matter, ability of preparation and realisation of experimental works, processing of experimental values and its analysis and interpretation
 Requirements:  17ZAF, 17ENF  Key words:  experimental reactor physics, research reactor, reactivity measurement, control rods calibration, zero power reactor, nuclear reactor dynamics, critical experiment
 References  Key references:
Weston M. Stacey: Nuclear Reactor Physics, John Wiley and Sons, Inc., New York 2001, ISBN 0471391271
Recommended references:
Lewis E.,E.: Fundamentals of Nuclear Reactor Physics, Elsevier Inc., USA 2008, ISBN: 9780123706317
Media and tools:
training reactor VR1


Nuclear Fuel Cycle  17JPC 
Sklenka, Starý 
  
2+0 kz 
 
2 
Course:  Nuclear Fuel Cycle  17JPC  doc. Ing. Sklenka Ľubomír Ph.D. / Ing. Starý Radovan  2+0 KZ    2    Abstract:  The course is focused on frontend & backend of the nuclear fuel cycle of the nuclear power plants, particularly PWR used and / or planned in the Czech Republic. The first part of the course consists of introduction to frontend of the nuclear fuel cycle. After the first division and definitions of various types of fuel cycles, the lectures are pointed to various uranium and thorium sources, their mining, mechanical and chemical processing to the shape of yellow cake. The next step there are very briefly described types of purifications, conversions, enrichment and fabrication of nuclear fuel. The second part of the course consists of introduction to backend of the nuclear fuel cycle, namely spent nuclear fuel, spent nuclear fuel inventory, wet and dry spent fuel storage, interim spent fuel storage and final disposal of spent nuclear fuel. At the end of the course basic information about thorium fuel cycle is mentioned. Note: Inner nuclear fuel cycle is the part of 17PRF  Core physics and fuel management course.  Outline:  1. Introduction
Duration: 1 lecture
Topic:
Fuel cycle definition, description of fuel cycles and fuel cycle nodes, division of various fuel cycles
2. Reserves and uranium mining in environment
Duration: 2 lectures
Topic:
Uranium reserves on the Earth, their amount and distribution on the Earth, uranium mining in each regions, mining history, types of mining (opencast mines, mines, ISL), the biggest uranium mines of the World, uranium mining in the Czech Republic (history, description of mining fields, present time).
3. Mechanical and chemical processing of ore
Duration: 1 lecture
Topic:
Mechanical processing of ore (granulation, milling), leaching (acid, carbonate), miscellaneous methods of uranium separation from leachates (sorption, solventextraction, etc.), production and composition of yellow cake.
4. Purification and conversion to UF6
Duration: 1 lecture
Topic:
Nucleargrade specification, miscellaneous purification methods of yellow cake to the nucleargrade material (solvent extraction with TBP, etc.) processing of UF6 for enrichment.
5. Enrichment
Duration: 2 lectures
Topic:
Term definition (depleted uranium, highly enriched uranium, ...), enrichment history, theory of enrichment (enrichment cascade, separation work, ?) description and characteristic of each enrichment method: electromagnetic separation, gaseous diffusion, thermal liquid diffusion, gas centrifuge separation, aerodynamic separation, AVLIS.
6. Fuel fabrication
Duration: 1 lecture
Topic:
Conversion of UF6 to UO2, features of powder UO2, processing of fuel pellets, fabrication of fuel rods and fuel assemblies, construction fuel assemblies for: VVER, PWR, BWR, and CANDU.
7. Backend of the nuclear fuel cycle
Duration: 2 lectures
Topic:
Backend of the nuclear fuel cycle, oncethrough nuclear fuel cycle, closed nuclear fuel cycle, reprocessing of the spent nuclear fuel, legislative requirements for spent nuclear fuel and nuclear installation contain spent fuel
Spent nuclear fuel, spent nuclear fuel inventory, computer codes SCALE & ORIGEN for inventory calculation, practical use of the ORIGEN code for calculation of spent fuel inventory from WWER reactors
8. Spent nuclear fuel storage and final disposal
Duration: 2 lectures
Topic:
Basic requirements for spent fuel storage, interim spent fuel storage, various types of spent fuel storage, wet and dry spent fuel storage, and final disposal of spent nuclear fuel
Dry spent fuel storage, storage and transportation casks, physics and technology aspects of cask storage, safety of cask storage, CASTOR cask for spent fuel from Dukovany & Temelín NPP
9. Thorium fuel cycle
Duration: 1 lecture
Topic:
Thorium fuel cycle, thorium fuel, physics differences of thorium fuel, reactor operation with thorium fuel, use of thorium fuel in the world, perspective of thorium fuel in next generation of NPP
 Outline (exercises):  Seminars doesn't included to the course  Goals:  An overview of both frontend & backend of the nuclear fuel cycle of the nuclear power plants.
Application of acquired knowledge to solve problems, qualification and quantification of the effects of various physical quantities and phenomena on the operation of nuclear reactors and nuclear safety.  Requirements:  17ZAF  Key words:  Nuclear reactor, nuclear fuel cycle, frontend of the nuclear fuel cycle, backend of the nuclear fuel cycle, uranium reserves, thorium reserves, uranium mining, ISL, solvent extraction, yellow cake, enrichment, fabrication, spent nuclear fuel, spent nuclear fuel inventory, SCALE computer code, ORIGEN computer code, spent fuel storage, interim spent fuel storage, wet and dry spent fuel storage, transportation cask, CASTOR cask, final disposal of spent nuclear fuel, thorium fuel cycle  References  Key references:
Stacey, W. M.: Nuclear Reactor Physics, Chapter 5 Nuclear Reactor Dynamics &
Chapter 6  Fuel Burnup, WILEYVCH Verlag GmbH & Co. KGaA, Weinheim, 2007
John R. Lamarsh: Introduction to Nuclear Engineering, 3rd Ed., Prentice Hall, 2001
Recommended references:
Operation and Maintenance of Spent Fuel Storage and Transportation Casks/Containers, IAEATECDOC1532, IAEA, Vienna, 2007
Design of Fuel Handling and Storage Systems in Nuclear Power Plants Safety Guide, IAEA Safety Guide, NSG1.4, IAEA, Vienna, 2003
Media and tools:
Audiovisual technique, films on DVD 

Thermohydraulic Design of Nuclear Devices 4  17THNJ4 
Kobylka 
3+0 z,zk 
  
4 
 
Course:  Thermohydraulic Design of Nuclear Devices 4  17THNJ4  Ing. Kobylka Dušan Ph.D.  3+0 Z,ZK    4    Abstract:  This course is set to improve the basic knowledge of students about problems of thermohydraulics. The students will learn more about flow of compressible fluids (gases, steam, ..), twophase flow (important for emergency analyses of nuclear devices, description of power loaded parts of PWR or design of BWR), about subchannel analysis of fuel assemblies and about specific modes of heat transfer (liquid metals, molten salts and gases), which can bee used for designs of GEN IV reactors. It also includes extended commentary of turbulent flow and models, which were develop for its description.  Outline:  1. Compressible fluids flow
Time range: 2 lectures
Speed of sound, use of 1st thermodynamic law for open thermodynamic system, conversion of enthalpy to kinetic energy, flow through gap, sound velocity, Laval nozzle (principle and calculation), steadystate shock, steam flow (real gas).
2. Twophase flow
Time range: 7 lectures
Fundamentals of twophase flow and its miscellaneous patterns, diagrams of flow, principles of twophase flow description, definition of basic quantities and their calculation (void fraction, etc. ), twophase pressure drops, basic twophase flow modeling: single fluid models, two fluid model, four equation models, nonequilibrium models, sonic velocity and critical flow; flow instability: types of instability and their nature (flow pattern instability, Ledineggs instability, dynamic instability, thermal oscillation, etc.), analysis of selected instabilities; liquidvapor separation, ..
3. Turbulent flow
Time range: 2 lectures
Description of turbulent flow, review of turbulent flow models and their principles, detailed descriptions and principles of turbulence models: Kepsilon, Komega, RNG, model with Reynolds stress, LES models, comparison of models and their use.
4. Primary circuit thermohydraulic
Time range: 1 lecture:
Hydraulic characteristic of primary circuits particular components: reactor, primary pipes, modes of main circulating pump, steam generator. Principles of CFD, short list of CFD codes and their description, problematic of interpretation of results. Principle of subchannel analysis and their use for fuel assemblies calculations, s transcription of basic flow equations for subchannel analysis, follow up equations fior subchannel analysis (turbulent mixing, pressure losses, etc.), application of subchannel analysis in computer codes and their list and features, integral and systems codes.
5. Specific modes of conduction
Time range: 1 lecture
Convection in liquid metals: thermophysical properties of the most used liquid metals (sodium, solution of PbBi), differences of properties from normal coolants and their influence on calculations, forced and natural convection at internal flow in pipes or triangular channel in lattice of fast reactors and on plane wall. Convection in gases: thermophysical properties of the helium (coolant in VHTR), differences of properties from normal coolants and their influence on calculations, influence of high velocities, forced and natural convection at internal flow in pipes and on plane wall, principles of thermohydraulics design of layer with spherical fuel elements. Convection in molten salts: thermophysical properties of fluoride molten salts, their differences from normal coolants and dependence on solutions compositions, forced and natural convection at internal flow in pipes and on plane wall.
 Outline (exercises):  Lectures are in selected chapters completed with calculations of practical examples: conduction in liquid metals, two phase flow, pressure drops in primary circuits, CFD calculation.  Goals:  Knowledge: detailed knowledge of selected parts of fluid mechanics, thermodynamics and heat transfer (see list of lectures) which can be used in thermohydraulic designs generally or in specific design of primary circuit and others devices in nuclear power plants.
Abilities: orientation in given issue, use gained knowledge in other courses which are engaged in thermomechanics and designs of devices in nuclear power plants. On base of given knowledge students will be able to understand and analyse behavior and control of nuclear power plant as a complex.  Requirements:  THNJ1, THNJ2, THNJ3 or equivalent  Key words:  liquid metals, high temperature reactors, molten fluoride salts, sound velocity, Laval nozzle, steam flow, two phase flow, driftflux model, dynamic instability, turbulent flow, models of turbulent flow, Kepsilon, , RNG, LES, subchannel analyse, CFD.  References  1. Tong, L.S., Weisman, J.: Thermal Analysis of Pressurized Water Reactors, American Nuclear Society, Illinois USA, 1996, ISBN: 0894480383
2. Weseeling, P.: Principles of Computational Fluid Dynamics, Springer, 2000
3. Wilcox, D. C.: Turbulence modeling for CFD, DCW Industries, California, 2002
4. Tang Y.S., Coffield R.D., Markley R.A.: Thermal Analysis of Liquid Metal Fast Breeder Reactors, American Nuclear Society, Madison USA, 1978, ISBN: 0894480111
5. 2. Mareš R.  Šifner O.  Kadrnožka J.: Tabulky vlastností vody a vodní páry podle průmyslové formulace IAPWSIF97, VUTIUM , 1999, ISBN 8021413166
6. Lahey R.T., Moody F.J.: The ThermalHydraulics of a Boiling Water Nuclear Reactors, Američan Nuclear Society, La Grange Park, 1993, ISBN 0894480375 

Machines and Equipment of Nuclear Power Plants  17SAZ 
Kobylka 
2+1 z,zk 
  
3 
 
Course:  Machines and Equipment of Nuclear Power Plants  17SAZ  Ing. Kobylka Dušan Ph.D.    2+1 Z,ZK    3  Abstract:  The course familiarizes students with basic machine devices of nuclear power plants, which are important for their operation, as are: pressurizer system, pumps and blowers, steam and gas turbines, heat exchangers (condensers, steam generators, reheaters, feed water heaters, etc.) and pipes and valves. Informations about devices are given primarily in descriptive level. It means that students are familiarized with different designs, used materials, manufacturing and operational experiences and parameters of real devices from power plants. Students also receive basic outline of fundamental theory about calculations of devices.  Outline:  1. Pressurizer Systems of Primary Circuit
Time range: 1 lecture
The pressurizer function in the primary circuit, physical description of pressurizer, model and calculation of pressurizer, operational states of pressurizer, types of pressurizers and their construction, connection and construction of pressurizer subsystems (electrical heaters, letdown condenser, etc.).
2. Pumps
Time range: 2 lectures
Pumps classification, operation principle and main features of miscellaneous pumps types, typical construction features of pumps, description of the most important pumps in nuclear power plant: primary circuit main recirculation pumps, feed pumps, condensate pumps, others pumps, pumps for liquid metals, components and a subsystems of pumps.
3. Steam turbines
Time range: 2 lectures
Operation principle and basic construction, steam turbines classification, description of turbine construction, fundamental calculations, description of turbine components (seals, bearings, etc.), saturated steam turbines: unique features, humidity separation and reheating of steam, control of steam, turbines, detailed description of the 220 MW and 1000MW turbines.
4. Gas turbines and blowers for gas cooled reactors
Time range: 1 lecture
Blowers operation principle, fundamental calculations, description of blower construction, basic features of blowers, gas turbines and their unique features, description of components and subsystems of gas turbines and blowers (seals, bearings, etc.).
5. Condensers systems and bypass valves
Time range: 1 lecture
Condensers of steam turbines (thermal calculations, description, construction), cooling system of nuclear power plants (types of cooling, construction, cooling towers, etc.), bypass valves (their function and description).
6. Heat exchangers
Time range: 1 lecture
Classification, construction, principle of thermal and hydraulic calculations, feed water regeneration systems, description of regenerative heat exchangers, construction of regenerative heat exchangers, deaerator.
7. Pipes and fittings in nuclear power plant
Time range: 1 lecture
Standardization, classification, unique pipes in nuclear power plant, fitting types a description of typical fitting in nuclear power plant.
8. Steam generators
Time range: 2 lectures
Position of steam generators in heat schemes of nuclear power plants, heat calculation of steam generator, steam generator of the VVER 440 a VVER 1000, vertical steam generators, steam generators in nuclear power plants with gas cooled and fast reactors, hydraulic calculation of steam generators.
9. Complete disposition of nuclear power plant
Time range: 1 lecture
Bubble condenser, containments, air engineering, dieselgenerators, layout of nuclear power plants and its parts.
 Outline (exercises):    Goals:  Knowledge: description and construction of the most important machine devices of nuclear power plant, fundamental calculations of selected machine devices of nuclear power plant.
Abilities: orientation in field of machine devices of nuclear power plant
 Requirements:  THNJ1, THNJ2, THNJ3  Key words:  pressurizer, pump, steam turbine, blower, condenser, bypass valve, regenerative heat exchanger, deaerator, steam generator, bubble condenser, containment  References  Key references:
Hejzlar R.: Machines and Equipment of Nuclear Power Plants, Vol. 1, Nakladatelství ČVUT, Praha, 2000, (in Czech)
Hejzlar R.: Machines and Equipment of Nuclear Power Plants, Vol. 2, Nakladatelství ČVUT, Praha, 2000, (in Czech)
Recommended references:
Tong, L.S., Weisman, J.: Thermal Analysis of Pressurized Water Reactors, American Nuclear Society, Illinois USA, 1996, ISBN: 0894480383


Excursion Abroad  17EXZ 
Frýbort 
  
1 týden z 
 
2 
Course:  Excursion Abroad  17EXZ  Ing. Frýbort Jan Ph.D.          Abstract:  Within the course the students take a weekly excursion at workplaces and institutions related to nuclear energy and research and development in this field. The dominant part of the excursion takes place in Slovakia. Traditionally, students visit the FEI STU Bratislava, VÚJE Trnava, the Slovak nuclear power plants Mochovce and Jaslovské Bohunice, the selected hydroelectric power station, the International Atomic Agency in Vienna and the TRIGA reactor at ATI in Vienna.  Outline:  Traditionally, students visit the FEI STU Bratislava, VÚJE Trnava, the Slovak nuclear power plants Mochovce and Jaslovské Bohunice, the selected hydroelectric power station, the International Atomic Agency in Vienna and the TRIGA reactor at ATI in Vienna.  Outline (exercises):   Goals:  Knowledge: Broadening of awareness of activities of the close university STU Bratislava, getting acquaint with technology complexes of nuclear power plants, illustration of IAEA activities.
Abilities: Individual participation on the excursion, preparation of detailed travel report.  Requirements:  Only for students of Nuclear Engeneering  Key words:  nuclear facility, excursion,  References  

Research Project 1  17VUJR12 
Frýbort 
0+6 z 
0+8 kz 
6 
8 
Course:  Research Project 1  17VUJR1  Ing. Frýbort Jan Ph.D.  0+6 Z    6    Abstract:  The course is concerned on the officially assigned topic of the research project and its final presentation and defense. The guarantor of the project is the supervisor, who assigns literature, checks the progress of work and its defensibility and operatively solves problems of the project. Students independently solves the project. The project assignment, which usually follows the bachelor's project, is agreed by the Head of Department. Consultation hours are meant to provide contact with the supervisor and shall be handled according to the needs. Hence, there is no scheduling for the course.
 Outline:    Outline (exercises):    Goals:  Knowledge: a particular field depending on a given project topic
Abilities: working unaided on a given task, understanding the problem, producing an original specialist text
 Requirements:  closed bachelor study
 Key words:    References  according to given topic

Course:  Research Project 2  17VUJR2  Ing. Frýbort Jan Ph.D.    0+8 KZ    8  Abstract:  The course is concerned on the officially assigned topic of the research project and its final presentation and defense. The guarantor of the project is the supervisor, who assigns literature, checks the progress of work and its defensibility and operatively solves problems of the project. Students independently solves the project. The project assignment, which usually follows the bachelor's project, is agreed by the Head of Department. Consultation hours are meant to provide contact with the supervisor and shall be handled according to the needs. Hence, there is no scheduling for the course.  Outline:    Outline (exercises):    Goals:  Knowledge: a particular field depending on a given project topic
Abilities: working unaided on a given task, understanding the problem, producing an original specialist text
 Requirements:  closed bachelor study  Key words:    References  according to given topic 
 Optional courses 
Computer Control of Experiments  17PRE 
Kropík 
2+1 z,zk 
  
3 
 
Course:  Computer Control of Experiments  17PRE  doc. Ing. Kropík Martin CSc.  2+1 Z,ZK    3    Abstract:  Lectures provide information about standard interfaces of personal computers  parallel, serial, USB, LAN and special interface cards; about standalone equipment that communicate with computers via serial lines or GPIB (IEEE488) interface, further about measuring systems with VME, VXI and LXI interfaces, discuss their advantages and disadvantages. Next, lectures deal with programming of measuring systems  special dedicated software, problems of use of high programming languages and especially use of graphical oriented development tools (Agilent VEE ane LabView); data acquisition and evaluation. Finally, students prepare individual software project for data acquisition and evaluation.  Outline:  1. Standalone equipment, PC cards for measurement and bus based measuring systems (VME, VXI, LXI). Examples or measuring instruments, their features and capabilities of computer control
2. Parallel, serial, USB, LAN a Firewire interfaces for communication among PC and instruments, examples and demonstration
3. GPIB (IEEE488.2) interface, systems based on VXI bus with practical demonstration
4. Examples of measuring instruments and their computer control by standard communication programs and dedicated software
5. Graphical oriented development tool Agilent VEE 1; basics of developing environment, programming in VEE, interface for inputs and outputs
5. Graphical oriented development tool Agilent VEE 2; control of instruments, I/O drivers, work with files
5. Graphical oriented development tool Agilent VEE 3; work with variables, extended function for evaluation of experimental data, hierarchical structure of programs
8. Graphical oriented development tool LabView 1; basics of developing environment National Instruments LabView, software production in LabView, differences in comparison to Agilent VEE
9. Graphical oriented development tool LabView 1, control of instruments, data acquisition and evaluation
10. Demonstration of system for validation of software for VR 1 training reactor safety and control system controlled by software on basis of Agilent VEE
11.13. Individual students work on given software project under lecturer?s guidance  Outline (exercises):  Students gradually train work with measuring instruments, development tools for software, and finally, they develop individual software project for control of experiment, data acquisition and evaluation  Goals:  Knowledge: detailed knowledge of available instruments for control of experiments, measurement of electrical values and data acquisition; programming in graphical oriented development systems intended for control of experiments, data acquisition and their evaluation.
Abilities: orientation in matter of computer control of experiments, ability to practically use gained knowledge in own experimental work.  Requirements:  17ZEL  Key words:  graphical oriented development tools Agilent VEE and LabView, data acquisition and evaluation, interface, systems with USB, GPIB, LAN and VXI busses  References  Key references:
Agilent VEE Pro User?s Guide, Agilent Technologies, 2005
Getting Started with LabVIEW, National Instruments, 2009
Recommended references:
Robert Helsel: Visual Programming with HP VEE, Prentice Hall, 1997
Advanced Programming Techniques, Agilent Technologies, 2000
Hewlett Packard/Agilent Instruments Documentation
Media and tools:
electronic laboratory of Department of nuclear reactors, graphical oriented development tools Agilent VEE and LabView 

Stochastic Methods in Reactor Physics  17SMRF 
Huml 
2+2 kz 
  
4 
 
Course:  Stochastic Methods in Reactor Physics  17SMRF  Ing. Huml Ondřej Ph.D.  2+2 KZ    4    Abstract:  Course is intended to nuclear data processing for mathematical modeling in nuclear reactor physics, to analytical and numerical solution of various deterministic methods in reactor systems, statistic methods in nuclear reactor physics and to nuclear reactor burnup modeling.
Stress is put on practical examples, exercises and individual students? work on solving of given exercises. After passing the course, the attendees obtain not only theoretical knowledge, but also practical experience with various methods and approaches to modeling of neutronphysical characteristics of nuclear facilities and their application in real reactor systems.
 Outline:  1. Statistical methods of mathematical modeling in nuclear reactor physics
Rozsah: 8 přednášek
Témata přednášek:
utilization of Monte Carlo methods for solution of engineering issues ? principle of Monte Carlo method, random quantities, mathematical statistics and precision, normal distribution,
transformation to arbitrary distribution (Gaussian, Poisson, etc.), random and pseudorandom numbers and their testing, utilization on Monte Carlo method for solution of simple physical problem
application of Monte Carlo method in neutronics calculation of rector systems ? elementary principles of particle transport in a medium (transport and free path, absorption, fission, scattering), neutrons, charged particles
MCNP code and its application for neutronics calculation of reactor systems ? principle of MCNP run, algorithm development of physical problem and its transformation to MCNP environment, input definition, output files processing
solution of criticality problems ? calculations of multiplication coefficient keff for various reactor systems using MCNP code, precision of calculation and accuracy intervals, neutron sources definition for criticality calculations, material composition definition for various reactor systems
complex geometry structures ? definition of complex geometry structures in MCNP code, repeated structures, square and triangular lattices, modular approach to complex geometry description
preprocessors and postprocessors for input and output simplification ? Sabrina and MCNPVised codes for simplification of MCNP input geometry generation, MONACO code for verified inputs generation for VR1 reactor and for output files processing
solution of problems for neutron flux determination ? calculations of neutron flux densities and particle fluencies in reactor systems, fluxes and currents in simple and complex geometry structures, possibilities of calculated values processing by TECPLOT code
calculation optimization ? optimization approaches for fastening MCNP calculations, symmetric and nonsymmetric problems and various interfaces definition, Russian roulette and other computertime saving methods
2. Mathematical modeling of burnup in nuclear reactor systems
Rozsah: 4 přednášky
Témata přednášek:
Simple burnup models for reactor systems ? solution of shortterm and longterm kinetics using MATLAB code
Burnup modeling by diffusion and transport methods ? WIMS code application for burnup calculations, burnup problem definition for elementary cell, complex geometry
SCALE calculation system for nuclear reactor burnup modeling ? application of SCALE code for nuclear reactor neutronphysical characteristics calculations, overview of basic modules, description and characteristics of KENO, TWOONEDAT and ORIGEN modules, application of ORIGEN module for burnup calculations, problem definition, input data processing, geometry definition, research reactors fuel burnup calculation, fuel burnup in pressurized water reactors
HELIOS code ? Application of HELIOS code for nuclear reactor neutronphysical characteristics calculations, description and characteristics of HELIOS code and applied mathematical model, description of input and output files for fuel burnup calculation in pressurized water and boiling water reactor systems
3. Technical visit of nuclear reactor neutronics calculation department
Rozsah: 1 přednáška
Témata přednášek:
Technical visit of nuclear reactor neutronics calculation department in Nuclear Research Institute in Rez or department for reactor calculations in Skoda Nuclear Machinery in Pilsen
 Outline (exercises):  possibilities of calculated (or experimental) data processing, large data volumes processing, TECPLOT, ORIGIN and ROOT codes applications, presentation of outputs
JANIS code and nuclear data libraries processing
basics of MCNP code use, geometry and materials definition, simple criticality problems and multiplication coefficient estimates
complex geometry structure problem, problem with repeated geometry structures
application of Sabrina, MCNPVised and MONACO codes
two problems on neutron flux densities and particle fluencies calculation, processing and analysis of outputs by TECPLOT code
two problems on calculation optimization ? symmetry and Russian roulette
application of MATLAB code for solving simple nuclear reactor burnup problems
application of WIMS code for burnup modeling, research reactor burnup problem
basics of SCALE system use, modules, problem definition, geometry and material description, simple research reactor fuel burnup problem
fuel burnup calculation for pressurized water reactor VVER440 using ORIGEN module
seminar work ? generation of own model of given reactor problem, its solution and outputs processing using TECPLOT code.
presentation of students? seminar works
 Goals:  detailed knowledge of mathematical modeling in nuclear reactor physics, statistic methods in nuclear reactor physics and nuclear reactor fuel burnup modeling
orientation in the field, application of gained knowledge in other courses in the field of theoretical reactor physics
 Requirements:  17FAR  Fyzika jaderných reaktorů  nutná podmínka
18MOCA  Metoda Monte Carlo  doporučený předmět
 Key words:  nuclear reactor, reactor physics, nuclear data, statistical methods, burnup, neutron transport, Monte Carlo method  References  Stacey, W. M.: Nuclear Reactor Physics, WILEYVCH Verlag GmbH & Co. KGaA, Weinheim, 2007
Christian P. Robert, George Casella: Monte Carlo Statistical Methods (Springer Texts in Statistics), Springer, 2005
Jerome Spanier: Monte Carlo principles and neutron transport problems, AddisonWesley Pub. Co, 1969
James E. Gentle: Random Number Generation and Monte Carlo Methods (Statistics and Computing), Springer, 2004


Deterministic Methods in Reactor Physics  17DERF 
Frýbort 
  
2+2 kz 
 
4 
Course:  Deterministic Methods in Reactor Physics  17DERF  Ing. Frýbort Jan Ph.D.          Abstract:  Course is intended to nuclear data processing for mathematical modeling in nuclear reactor physics, to analytical and numerical solution of various deterministic methods in reactor systems, statistic methods in nuclear reactor physics and to nuclear reactor burnup modeling.
Stress is put on practical examples, exercises and individual students? work on solving given exercises. After passing the course the attendees obtain not only theoretical knowledge, but also practical experience with various methods and approaches to modeling of neutronphysical characteristics of nuclear facilities and their application on real reactor systems.
 Outline:  1. Introduction to mathematical modeling in nuclear reactor physics
Rozsah: 2 přednášky
Témata přednášek:
introductory lecture 
introductory lecture ? introduction to subject, role of the course within the studyprogram, relationship to other courses, goals of the course seminar work assignation, basic approaches to neutronics calculations of reactor systems, analytical and numerical solution of diffusion and transport equations, statistical methods, methodology of mathematical modeling in nuclear reactor physics ? analysis of problem to solve, selection of method for solution, physical model, mathematical model, algoritmization
outputs processing and analysis, their comparison with experiment, validation of mathematical model ? methods for outputs processing and analysis, general processing of computer codes output files, work with codes devoted to data analysis (TECPLOT, ORIGIN, ROOT), calculation uncertainties analysis, methods for mathematical model validation, importance of benchmark tests for reactor system mathematical modeling.
2. Nuclear data for mathematical modeling in nuclear reactor physics
Rozsah: 3 přednášky
Témata přednášek:
nuclear data for mathematical modeling in nuclear reactor physics ? introduction to cross sections theory, cross section experimental determination, cross section determination through calculation (codes GNASH, TALYS, etc.), evaluated nuclear data libraries (JEFF, JENDL, ENDF/B), other nuclear data libraries (ENDSF, EXFOR), general overview and division of libraries as data sources
processing of nuclear data libraries ? codes for searching and visualization of libraries? data, especially of cross sections (JEFPC, JANIS, NDX), data sources available via internet, codes for specialized processing of nuclear data, especially for processing from general data format to formats utilized by computer codes (stress put on data for MCNP code), NJOY code (basic code for cross section processing and adjustments)
nuclear data processing ? PREPRO code (alternative to NJOY, less general code, specialized mainly to MCNP), familiarization with CALENDF and TRANSX codes (for generation of group data for specialized codes), generation of activation data for SAND and UMG codes and basics of these codes utilization
3. Deterministic methods of mathematical modeling in nuclear reactor physics ? analytical solutions
Rozsah: 2 přednášky
Témata přednášek:
analytical methods for reactor physics equations solution ? utilization of analytical solution of nuclear reactor physics in praxis, derivation of particular usable equations
nuclear reactor physics equations analytical solution using MAPLE and MATLAB codes ?
analytical solutions (codes) history and their utilization in reactor physics, basics of MAPLE code and its possibilities for solution of particle transport in nuclear reactors, description of mathematical apparatus
4. Determinisctic methods of mathematical modeling in nuclear reactor physics ? numerical solutions
Rozsah: 6 přednášek
Témata přednášek:
overview of numerical methods for solution of particle transport in nuclear reactors ? general introduction to utilization of numerical methods in mathematical modeling, overview of numerical methods with respect to their application for particle transport in nuclear reactors, definition of initial and boundary conditions of particle transport numerical solution
numerical solution of diffusion equation ? introduction to numerical solution of diffusion equation, overview and selection of appropriate methods for diffusion equation numerical solution, initial and boundary conditions specification and selection, description of diffusion equation solution via selected numerical methods, solution outputs analysis, and analysis of their accuracy
numerical solution of transport equation ? introduction to numerical solution of transport equation, overview and selection of appropriate methods for numerical solution of transport equation, initial and boundary conditions specification and selection, description of transport equation solution via selected numerical methods solution outputs analysis, and analysis of their accuracy
numerical solution of particle transport in nuclear reactors using MATLAB code ? introduction to basics of MATLAB code and its possibilities for particle transport solution in nuclear reactors, MATLAB code mathematical apparatus description for numerical solution of particle transport, definition, and setting of solution conditions in MATLAB code, analysis and solutions? outputs processing in MATLAB code
calculation codes based on numerical methods of particle transport solution ? neutronphysical characteristics calculation of reactor systems ? Overview of computer codes utilizing numerical methods to neutronphysical characteristics of reactor systems. Description and characteristics of computer codes WIMS, TWODANTSYS.DANTYS, and CITATION. Input files generation for these codes for neutronphysical characteristics of reactor systems calculations. Output files description and analysis.
Computer codes based on numerical methods of particle transport solution ? nuclear facility shielding calculation ? Overview of computer codes using numerical methods for nuclear facility shielding calculations. Description and characteristics of computer codes ANISNORNL, TORTDORT. Input files generation for these codes for calculation of nuclear facilities shielding parameters. Output files description and analysis.
 Outline (exercises):  Témata cvičení:
possibilities of calculated (or experimental) data processing, large data volumes processing, TECPLOT, ORIGIN and ROOT codes applications, presentation of outputs
JANIS code and nuclear data libraries processing
NJOY code, data processing from general format to format used by MCNP, group data generation for user specified group boundaries, exporting data from NJOY to highquality PS figures for outputs publication or presentation
use of MATLAB and MAPLE codes for analytical solutions of reactor physics equations and outputs plotting by TECPLOT code
Diffusion equation solution by MATLAB code for selected reactor system geometries, solution outputs analysis and graphical processing
Use of WIMS code including given problem solution from the field of reactor physics
Use of TWODANTSYS.DANTYS code including given problem solution from the field of reactor physics
Use of CITATION code including given problem solution from the field of reactor physics
Use of ANISNORNL code including calculation of nuclear facility shielding parameters and characteristics
use of TORTDORT code including calculation of nuclear facility shielding parameters and characteristics
Seminar work ? generation of own model of given reactor problem, its solution by one of above mention code and output analysis and processing using TECPLOT code
Presentation of students? seminar works from the field of numerical solution of deterministic method
 Goals:  detailed knowledge of mathematical modeling in nuclear reactor physics, analytical and numerical solutions of deterministic methods in reactor systems.
Orientation in the field, application of gained knowledge in other courses from the field of theoretical reactor physics.
 Requirements:   Key words:  nuclear data, deterministic calculations, depletion calculations, transport equation, nuclear data uncertainty, calculation uncertainty, fuel depletion, macroscopic data, fullcore calculations  References  John R. Lamarsh: Introduction to Nuclear Engineering, 3rd Ed., Prentice Hall, 2001
Elmer E. Lewis: Fundamentals of Nuclear Reactor Physics, Academic Press, Amsterdam, 2008
Stacey, W. M.: Nuclear Reactor Physics, WILEYVCH Verlag GmbH & Co. KGaA, Weinheim, 2007
Weston M. Stacey, jr.: Variational Methods in Nuclear Reactor Physics, Academic Press, New York, 1974
Joe D. Hoffman: Numerical Methods for Engineers and Scientists, Second Edition, Marcel Dekkor Press, New York, 2001


Digital Safety Systems of Nuclear Reactors  17CIBS 
Kropík 
2+0 z,zk 
  
2 
 
Course:  Digital Safety Systems of Nuclear Reactors  17CIBS  doc. Ing. Kropík Martin CSc.    2+0 Z,ZK    2  Abstract:  Lectures deal with use of computers in safety systems of nuclear reactor, with requirements on their hardware and software. Attention is devoted to software life cycle, to software requirements, design, coding, integration of HW/SW, verification/validation, maintenance and configuration management of software. Requirements and limitation of programming languages by software coding are discussed. Problematic of programmable logical devices (CPLD, FPGA) for use in safety and control systems of nuclear devices was introduces into lectures. Subject is also completed by demonstration of validation of operational power measuring and independent power protection systems of VR 1 reactor I&C  Outline:  1. Computers in systems important to nuclear safety and requirements on hardware, preparation of requirements on functionality of computer based systems important to nuclear safety, requirements on computer hardware, redundancy, memory content check, testing, inputs/outputs, performance, qualification of on shelf hardware for systems important to nuclear safety
2. Requirements on software for safety systems 1, IEC60880, life cycle  requirements, specification, design, coding, verification, integration HW/SW, validation, operation and maintenance, quality assurance, configuration management, verification methods, testing, documentation, IEC62138  SW for category B systems according to IEC61226  e.g. control systems
3. Requirements on software for safety systems 1; use previously developed software, common cause failures, diversity, formal methods, integrated tools for software production
4. Coding of software 1; methods of coding for high quality software, basic attributes,  reliability (predictability of memory use, timing, flow control), robustness (diversity, exceptions handling, input and output tests), maintenance (readability, data abstraction, modularity, portability) and method for their achievement
5. Coding of software 2, programming languages and their use for safety systems of nuclear reactors, required features and limitation in their use for systems important to nuclear safety with respect to attributes mentioned in previous paragraph, us of Pascal and C languages
6. Upgrade of safety and control system (I&C) of VR 1 training reactor, preparation of hardware and software requirements, software production, quality assurance, practical examples
7. Configuration management at VR 1 training reactor, parameter setting for systems of operational power measurement, independent power protection, control system and human machine interface, used methodology and tools
8. Excursion at VR1 training reactor, demonstration of upgraded computer based safety and control system (I&C), exhibition of operation, of safety functions and system configuration management
9. Validation of systems important to nuclear safety 1; valdation methodology, simulation of input signals, tests of system response on them, available hardware and software tools for validation, computer controlled generators and signal sources, graphical oriented programming tools Agilent VEE and LabView
10.Validation of systems important to nuclear safety 2  demonstration of validation, validation of upgraded operational power measuring and independent power protection systems, testing of interfaces, testing of operational and safety functions using system based on IEEE488.2, VXI and programming tool Agilent VEE
11. Computer based safety and control systems in nuclear power plants 1; commercial computer based systems for nuclear power plants  Siemens Teleperm XS and software tool SPACE used e.g. in nuclear power plant Mochovce or new built power plants EPR, DSS Spinline used in upgraded I&C systems of nuclear power plant Dukovany, Westinghouse Eagle system in nuclear power plant Temelin
12. Safety and control systems of nuclear power plants Dukovany and Temelin, systems structure, used technology, quality assurance, redundancy, diversity, safety functions
13. Programmable logical devices (CPLD and FPGA) in safety and control system, reasons of use, advantages, disadvantages, circuits design, VHDL language, quality, qualification and testing
 Outline (exercises):  Excursion at VR 1 training reactor (paragraph 8.), demonstration of systems validation (paragraph 10), discussion on required literature  Goals:  Knowledge: problems of computer based safety system of nuclear reactors, differences in comparison to hardwired systems, requirements on hardware and software, systems testing, configuration management
Abilities: orientation in matter of computer based safety systems, use of gained knowledge in other subjects of reactor physics, nuclear power plants and in operator?s course during further education
 Requirements:  17ZAF, 17BES  Key words:  nuclear safety, computer based safety systems of nuclear reactors, quality assurance, software life cycle, coding, configuration management  References  Key references:
Nuclear power plants  Instrumentation and control systems important to safety  Software aspects for computerbased systems performing category A functions, IEC60880, 2006
Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems, NUREG/CR6463, 1996
Recommended references:
Nuclear power plants  Instrumentation and control important for safety  Software aspects for computerbased systems performing category B or C functions, IEC62138, 2004
Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations, IEEE7.4.3.22010
Media and tools:
training reactor VR 1 laboratory, electronic laboratory of Department of nuclear reactors with system for validation of computer based systems


Energy Sector and Energy Sources  17EEZ 
Tichý, Kobylka 
  
2+1 z,zk 
 
3 
Course:  Energy Sector and Energy Sources  17EEZ  Ing. Kobylka Dušan Ph.D. / Ing. Tichý Miloš CSc.    2+1 Z,ZK    3  Abstract:  The main purpose of this course is to transmit to students the basic information about energy sector as the part of economics, about its wide range, all important parts and about patterns of energy sector function. The course is  from the beginning  structured logically from definition of term "energetics? through the power consumption, power sources on Earth, fuel mining and its influence on our environment, to the transformation of fuel power to nobler types of power. This course describes power plants from the view as a device being used for the power transformation mostly from the view of their features for connection to energy network, how they influence the environment and national economy, etc. It points also to various types of nuclear reactors and their connection with the fuel cycles. It contains also power network features, their managing and structures, description of power networks in Europe and in the Czech Republic. The final part of this course is pointed to energetics of the Czech Republic and the State energy policy.
 Outline:  1. Definition of "energy sector", its division and energy consumption
Scope: 1 lecture
Limitation of energy sector, division of energy sector to parts, power engineering history and energy consumption in the World, energy sources: fossil fuel (solid, liquid and gaseous), renewable energy sources and their basic features.
2. Sources and fuels mining on Earth
Scope: 2 lectures
Reserves of basic fuels (solid fossil fuels, liquid fossil fuels, gaseous fossil fuels, nuclear fuels) on the Earth, their deposits an present mining, mining history, flow of energy raw materials in the World (transport, import, export), basic influence of mining on environment, forecasting.
3. Energy consumption, electricity
Scope: 2 lectures
Production  consumption equality, primary energy consumption in the World according to regions, energy consumption in the World according to fuels, energy consumption per capita, nonuniformity of consumption in the World, development of consumption in history and forecasting, influence of consumption on life quality, energy consumption in economy, fuels in economy, production and consumption of electricity, import and export of electricity, daily load curve, electricity accumulation.
4. Nuclear power in the World and basic features of nuclear power plant
Scope: 2 lectures
Nuclear power in the World, amount of operated nuclear reactors in various countries, forecasting, basic nuclear reactors types (PWR, BWR, CANDU, gas cooled reactors, RBMK, fast reactors, Generation IV) and their contribution in energy sector, connection of different reactor types to fuel cycles.
5. Power plants based on renewable energy sources
Scope: 2 lectures
Hydroelectric power plants (division of hydroelectric power plants, description of hydroelectric power plants, turbine types, basic features of hydroelectric power plants), wind power plants (principle, rotors, efficiency, weather map, basic features of wind power plants), solar energy (types of use, division of wind power plants, photovoltaics, efficiency, basic features of wind power plants)
6. Fossilfuel power plant and biomass
Scope: 2 lectures
Basic description of fossilfuel power plant, boilers types and their principles (stockerfired, fluid, drybottom, pulverized fuel, cyclone, ?), basic description of power plant components (coal feeding, boiler, filters, flue gas desulphurization, ?), basic features of power plants (efficiency, operational experiences, economy), influence on environment (gaseous emission and their reduction, solid wastes, ?), liquid fuel boiler, gaseous fuel boiler, internal combustion turbines and motors in power engineering, biomass boilers.
7. Energy supply system, hydrogen power engineering
Scope: 1 lecture
Electricity supply system: transmission network system, types of networks according to voltage, components of networks (wires, towers, etc.), transmission network system in Europe and their connection, European electricity business, transmission network system in the Czech republic, basic description of gas supply system and oil supply system, features of hydrogen power engineering (principle, hydrogen production and use)
8. Energy sector in the Czech Republic and the State energy policy
Scope: 2 lectures
Energy consumption in the Czech Republic, fuels and their contribution in energy supply, renewable energy sources, the most important power plants in the Czech Republic (EDU, ETE, Prunéřov, Mělník, Dlouhé stráně, Orlík), forecasting  the State energy policy.
9. Students reports
Scope: 1 lecture
Presentation of student?s reports prepared according to given topics.
 Outline (exercises):  Principles of energy systems designs, calculations of energy systems and their optimalizations by various aspects, versions comparison, use of code DESAE for calculations and optimalizations of energy systems.  Goals:  Basic knowledge of power engineering, energy sources and fuels, energy transformations a their influence on environment, knowledge of basic power plants descriptions and their features, description of energy sector in the Czech Republic and its forecasting (the State energy policy), calculations and optimalizations of energy systems.
Orientation in issue, ability of logical thinking in the power engineering, calculations and optimalizations of energy systems by using DESAE code.
 Requirements:  17THN1, 17JARE  Key words:  power engineering, energy sector, electricity, energy sources, coal, oil, natural gas, fuel mining, nuclear power plant, fossilfuel power plant, boiler, renewable energy sources, hydroelectric power plant, photovoltaics, transmission network system, energy sector in the Czech Republic, the State energy policy, DESAE.  References  Key references:
BP: BP Statistical Review of World Energy, London, 2009
The Ministry of Industry and Trade: The energy vision of the Czech Republic, Nakladatelstvi Arch, Praha 2005, ISBN: 8086165981
Recommended references:
WWW sites of Energy regulatory office: http://www.eru.cz/
Media and tools:
PC classroom, computer code DESAE


Selected Parts of Legislation  17VPL 
Bílková, Fuchsová 
  
2+0 z 
 
2 
Course:  Selected Parts of Legislation  17VPL  RNDr. Bílková Hana / Ing. Fuchsová Dagmar    2 Z    2  Abstract:  Lectures are focused on valid legislation of the Czech Republic for peaceful utilisation of nuclear energy and ionising radiation, i.e. above all on the Atomic Act and its implementing regulations. Attention is paid to Atomic Act structure, basic terms and legislation requirements for various control domain such as nuclear safety, radiation protection, emergency preparedness, etc.  Outline:  1. Legislation for peaceful utilisation of nuclear energy and ionising radiation, Atomic Act
Time range: 2 lectures
State Office for Nuclear Safety  history, status, competence, structure, international cooperation, acts concerning SUJB activities, view of Atomic Act implementing regulations, structure and content of the Atomic Act, basic terms, general conditions for performance of practices according to the Atomic Act, licences for particular practices, conditions for issuing the licence, licence application, obligations of licensees, radioactive waste management, supervising activities.
2. Quality assurance
Time range: 1 lecture
Decree on Quality assurance system in performing and ensuring activities related to the Utilisation of Nuclear Energy and Radiation Activities, and on Quality Assurance of Selected Equipment with Regard to their Ranking into Safety Classes, introduction and range of quality assurance, requirements for quality assurance system, requirements for documentation, persons in the quality assurance system, processes and activities, quality assurance program, selected equipment.
3.Radiation protection
Time range: 3 lectures
Goals and principles of radiation protection, Decree on radiation protection  ionising radiation source classification, workplace categorisation, categorisation of exposed workers, limit system and exposure reduction, optimisation of radiation protection, supervised and controlled area, methods of ionising radiation source management,
general conditions of safe operation, discharge of radionuclides into the environment, quantities, parameters and facts impacting on radiation protection , medical exposure, exposure to natural sources, radioactive waste management.
4.Radiation monitoring network
Time range: 1 lecture
Decree on function and organisation of the national radiation monitoring network, basic terms, monitoring network function, monitoring network organisation, early warning network, monitoring in the emergency, monitoring network performance.
5.Life cycle of nuclear facilities
Time range: 1 lecture
Definition of nuclear facility and other terms, list of nuclear facilities in the Czech Republic, siting of a nuclear facility, construction of a nuclear facility, particular stages of nuclear facility commissioning, operation of nuclear facility, particular stages of decommissioning, licence for particular practices, relevant legislation.
6.Nuclear safety
Time range: 1 lecture
Decree on requirements on nuclear installation for assurance of nuclear safety, radiation protection and emergency preparedness, Decree on nuclear safety and radiation protection assurance during commissioning and operating of nuclear facilities, basic terms, possible states of nuclear installations, basic requirements on nuclear installations to assurance of nuclear safety, fuel handling and storing.
7.Emergency preparedness
Time range: 1 lecture
Decree on details for emergency preparedness assurance at nuclear installations and workplaces with ionising radiation sources, basic terms concerning emergency preparedness, classification degrees for extraordinary events, assurance of emergency preparedness by licensee, identification of an extraordinary event occurrence, extraordinary event announcement, employees and other person exposure limitation, emergency preparedness verification, onsite emergency plan, national emergency response system, Emergency response centre of State office for nuclear safety.
8.Physical protection
Time range: 1 lecture
Decree on physical protection of nuclear materials and nuclear facilities and their classification, basic terms concerning physical protection, categorisation of nuclear materials and parts of nuclear facilities for purpose of physical protection, designation of guarded, protected and inner area, access of persons and vehicles, administrative and technical measures, physical protection of nuclear material in transport, documentation approved by the State Office for Nuclear Safety.  Outline (exercises):    Goals:  Knowledge: legislation of the Czech Republic for peaceful utilisation of nuclear energy and ionising radiation, basic terms used in that legislation, basic legislation requirements.
Abilities: be clever at legislation for peaceful utilisation of nuclear energy and ionising radiation, application of legislation provisions in the profession.  Requirements:    Key words:  State Office for Nuclear Safety, Atomic Act, Atomic Act implementing regulations, nuclear safety, radiation protection, physical protection, emergency preparedness, radiation monitoring network, quality assurance  References  Key references:
Act No. 18/1997 Coll. on Peaceful Utilisation of Nuclear Energy and Ionising Radiation (the Atomic Act), as amended
Recommended references:
Atomic Act implementing regulations
Annual Reports of State Office for Nuclear Safety 

Economic Evaluation of Nuclear Power Plants  17EHJE 
Starý 
2+0 zk 
  
2 
 
Course:  Economic Evaluation of Nuclear Power Plants  17EHJE  Ing. Starý Radovan  2 ZK    2    Abstract:  The course focuses on the economic evaluation of Nuclear power plants. Introductory lectures are concerned with an introduction to economy and the basic component parts of microeconomics. Lectures continued with insight into the business and managerial economics, explanation of the concepts of incomes, expenses, etc. and their applications in electrical energy resources evaluation. Second part of lectures is focused on evaluation of nuclear power plants  the fuel cycle and operations of NPP  Outline:  1. Introduction to Economics
Scope: 1 lecture
Lecture content:
Introductory lecture  introduction to the problems, explanation of basic terms. Economics and business administration, the basic concepts of microeconomics and macroeconomics.
2. Introduction to Microeconomics
Scope: 2 lectures
Lecture content:
Market and basic market elements, supply and demand, balance of economy, total,
average, fixed, variable and marginal costs and revenues, consumer behavior, theory of the firm, price formation, perfect and imperfect competition, examples of imperfect competition, monopoly, oligopoly, price and it´s regulation.
3. Introduction to Business Administration
Scope: 2 lectures
Lecture content:
Company's assets and capital structure, revenues, costs and profit of the enterprise, profit and relationships between the company's basic economic values, breakeven point analysis, operating and financial leverage effect.
4. Economics of Electricity Production
Scope: 3 lectures
Lecture content:
Supply and demand for electricity, typical annual and daily consumption chart, fixed and variable costs of different power sources, construction of the supply function of electric energy and its shifts, examples according to selected market, capacity factors of different power sources
5. Externalities in power generation
Scope: 1 lecture
Lecture content:
The concept of externality, negative and positive externalities, externalities in energy, nuclear power externalities, comparisons between sources
6. Economics of nuclear energy,
Scope: 2 lectures
Lecture content:
Capital, O&M and fuel costs of nuclear power plants and their comparison with other sources, evaluation of nuclear fuel cycle  mining and milling, conversion, enrichment, fabrication, evaluation of fuel reprocessing and permanent storage, calculation of LCOE
7. Evaluating methods of investments
Scope: 2 lectures
Lecture content:
Marketing analyzes, weighted average cost of capital  corporate discount rates, static and dynamic methods of evaluating, PBP, NPT, IRR, examples
8. The theory of innovation
Scope: 1 lecture
Lecture content:
Innovative processes, Schumpeter innovation waves, innovation impulses, innovation lifecycle  Outline (exercises):  There are no exercises, only a discussion about practical examples by lecture topics.  Goals:  Knowledge:
Understanding the basic topics of microeconomics and business administration. Orientation in the differences between the evaluation of different sources of electricity. Understending of power market. Understanding the economic eveluation of fuel cycle and NPP operation.
Abilities:
Orientation in the given issue, understanding the economic differences in the production of el. energy from different sources.
 Requirements:    Key words:  Electricity, pool, exchange, supply, demand, market, consumption, production, costs, base load, peak load, price, fuel cycle  References  Key refernces:
SCHILLER, Bradley R. Mikroekonomie dnes. Brno : Computer Press, 2004. 412 s. ISBN 802510169X.
STOFT, Steven. Power System Economics : Designing Markets for Electricity. 1st edition. [s.l.] : WileyIEEE Press, 2002. 496 s. ISBN 0471150401.
NORD POOL. Elspot Market Data [online]. Oslo, .
Recommended references:
KIRSCHEN, Daniel S., STRBAC, Goran. Fundamentals of Power System Economics. 1st edition. Chichester England : John Wiley & Sons Ltd., 2004. 296 s. ISBN 0470845724.
Media and tools:
Computer room 

Computer Science for Modern Physicists  17IMF 
Havlůj 
0+3 kz 
  
3 
 
Course:  Computer Science for Modern Physicists  17IMF  Ing. Havlůj František  0+3 KZ    3    Abstract:  Although the computers became an everyday and inherent part of the science and engineering, use of them is often reduced to ?office? tasks and to use of specialized computing tools. Surprisingly few researchers are able to use their computers for automated data processing in order to boost their efficiency. The subject in a form of an interactive seminar gets the students acquainted with the basic automation principles, mainly in data processing, but also in automated preparation of input decks for computing applications or in generation of charts and reports and in results presentation. Every lesson starts with a short lecture and a definition of a selected automation problem, which in turn the students try to solver under the teacher?s guidance. The most effort is put into individual, independent work and into preparation of the students for practical use of the lessons learned.  Outline:   Outline (exercises):  1. Automation basics and scripting (3 lessons)
Topics: Automation principles and methods. Introduction to Ruby language. Basic data formats and data file reading. Chart generation using gnuplot. Creation of text tables. Text processing. 2. Interaction with computing tools (2 lessons) Topics: Extraction of data from scientific application output files. Templating of input decks. Complex task  evaluation of critical positions of control rods using MCNP.
3. Automated document generation (3 lessons)
Topics: LaTeX text processor basics. ERb templating systems. Automated processing of PDF documents, including charts and tables.
4. Interactive documents (3 lessons)
Topics: Using HTML language for automated document generation. Style sheet basics with CSS. Javascript basics and the jQuery library. Complex task  interactive database of reactor records.
5. Advanced software tools for engineers (1 lesson)
Topics: Text encodings. Source code version control tools. Introduction to web applications.  Goals:  Knowledge: The students are well oriented in the practice of automated data processing and are acquainted with a core portfolio of relevant tools. They have a good overview of the automation processes and of the major methods used.
Skills: The students are by themselves able to design and implement basic automation processes, ranging from simple text and numeric processing to automated generation of documents, both in textual (using LaTeX text processor) and interactive (HTMLbased) form. They are able to use the Ruby scripting language at the level needed for the daytoday engineering scientific tasks and they are acquainted with a wide range of reference resources in order to be able study further whenever needed.  Requirements:  Introduction to Nuclear Reactor Physics 1  Key words:  scripting, Ruby language, automated data processing, gnuplot, LaTeX, HTML/CSS/JS, document generation  References  Key references:
Dave Thomas, with Chad Fowler and Andy Hunt, Programming Ruby 1.9 & 2.0 (4th edition): The Pragmatic Programmers' Guide, Pragmatic Programmers, 2013
Andrew Hunt, David Thomas, The Pragmatic Programmer: From Journeyman to Master, AddisonWesley, 1999
Recommended references: Philipp K. Janert, Gnuplot in Action  Understanding Data with Graphs, Manning Publications, 2009 Brian Marick, Everyday Scripting with Ruby: For Teams, Testers, and You, Pragmatic Programmers, 2007 

Materials Science for Reactors  14NMR 
Čech, Haušild 
  
2+0 zk 
 
2 
Course:  Materials Science for Reactors  14NMR  prof. Dr. Ing. Haušild Petr          Abstract:  Materials for classical and fusion reactors  Outline:  1. Radiation damage, effects of radiation on materials; interaction of radiation with crystal lattice, influence of irradiation temperature; mechanical properties irradiated materials.
2. Material conception of nuclear reactors primary demand on materials and welds of pressure vessel; summary of used materials; degradation mechanisms; short  term plus long  term characteristics; specificity of materials and construction of reactor VVER and PWR type; surveillance program of irradiated specimens; neutron dosimetry for surveillance program; verification of irradiation temperature ; non  destructive testing.
3. Zr alloys: production, types, using, PWR, VVER; characteristics of Zrcoating in normal service conditions (rust, hydrogen absorption), abnormal service conditions (boiling, short  term overheating) and accidents (RIA, LOCA); above  project accident pair with melting core.
4. Nuclear fusion: interaction of plasma with materials, transitional matters; demand on materials for inner component; specific materials for inner component; first wall, envelope, cooling system; wolfram, beryllium carbon composites; joining materials; plasma spraying (principle, using); specific materials for others application; vacuum vessel; sperconductive coils; materials for electric insulation; special materials under development.
 Outline (exercises):  Metallography
Tensile test  Goals:  Knowledge: Material conception of classical and fusion nuclear reactors
Skills: Orientation in nuclear material topics  Requirements:  14NMA  Key words:  Interaction of radiation with crystal lattice, Radiation damage, Materials for nuclear pressure vessels, Zr alloys, Interaction of plasma with materials, Materials for nuclear fusion.  References  Key references:
[1] G.S. Was, Fundamentals of Radiation Materials Science Metals and Alloys, SpringerVerlag 2007.
Recommended references:
[1] J. Koutsky and J. Kocik , Radiation Damage of Structural Materials. Material Science Monographs 79, Elsevier 1994.

 