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Curricula and Syllabi of FNSPE CTU in Prague

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Aktualizace dat: 28.8.2019

česky

Master degree programmeNuclear Engineering
Year 1
course code teacher ws ss ws cr. ss cr.

Compulsory courses

Nuclear Reactor Physics17FAR Fejt, Frýbort, Frýbortová, Sklenka 2+2 z,zk - - 5 -
Course:Nuclear Reactor Physics17FARdoc. 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, integro-differential, 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
ReferencesKey references:
Stacey, W. M.: Nuclear Reactor Physics, WILEY-VCH 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 Management17PRF Sklenka - - 2+0 z,zk - 3
Course:Core Physics and Fuel Management17PRFdoc. 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, burn-up, 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 burn-up and fuel management, e.g. fuel handling, fuel management, reactor operation, burn-up, 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: Front-end & back-end 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, front-end of the nuclear fuel cycle, inner nuclear fuel cycle and back-end of the nuclear fuel cycle

2. Long-term reactor kinetics
Duration: 2 lectures
Topics:
Simple model of long-term reactor kinetics of the uranium-plutonium fuel cycle, fuel changes during the cycle, burn-up, and changes of keff during the cycle

Fission products, assignment of the first part of the students' seminar work, deep burn-up, long-term reactor kinetics of the thorium fuel cycle

3. Middle-term reactor kinetics
Duration: 2 lectures
Topics:
Xenon and samarium in the core during the cycle, simple model of middle-term reactor kinetics, xenon poisoning, assignment of the second part of the students' seminar work,

Xenon oscillations, cast-down operation - thermal & power reactivity coefficients at the end of cycle, burn-up absorbers

3. Linear reactivity model
Duration: 1 lecture
Topic:
Linear reactivity model, non-uniformity of power distribution in the core, neutron leakage from the core, leak reactivity, non-linear reactivity model

4. NPP fuel cycle, fuel burn-up and fuel management
Duration: 5 lectures
Topics:
Nuclear fuel cycle, inner nuclear fuel cycle, core physics, fuel handling, fuel management, reactor operation, burn-up, 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, MOBY-DICK computer code, practical use of the MOBY-DICK code for WWER-440 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, burn-up, 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
ReferencesKey references:
Stacey, W. M.: Nuclear Reactor Physics, Chapter 5 Nuclear Reactor Dynamics &
Chapter 6 - Fuel Burnup, WILEY-VCH 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, NS-G-2.5, IAEA, Vienna, 2002
Design of the Reactor Core for Nuclear Power Plants, IAEA Safety Guide, NS-G-1.12, IAEA, Vienna, 2005

Media and tools:
Computer laboratory - 2x2h, audiovisual technique, fuel cycle films on DVD

Reactor Dynamics17DYR Heřmanský, Huml - - 2+2 z,zk - 4
Course:Reactor Dynamics17DYRprof. 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 short-term 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 (short-term 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. One-group 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 one-group 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. One-channel two-component (or three-component) model of the pressurized water reactor core. Classification of mathematical models according to the diffusion equations solution and thermal-hydraulic processes.

Simplified models of the reactor dynamics
Nonlinear models: the Nordheim-Fuchs 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
ReferencesKey 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 Thermomechanics17TERR Bílý, Heřmanský 2+2 z,zk - - 4 -
Course:Reactor Thermomechanics17TERRprof. Ing. Heřmanský Bedřich CSc.2+2 Z,ZK-4-
Abstract:Heat generation in nuclear reactors - distribution and time evolution, residual heat generation. Steady-state and transient heat conduction in fuel elements, heat conduction in cladding, heat transfer in fuel-cladding gap. Convection heat transfer in nuclear reactors and boiling crisis of the first kind. Temperature distribution in fuel channel in steady-state 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 study-program, 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. One-group 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 radius-dependent heat generation, central gap effect, heat conduction in hollow rod with double sided heat removal.

Heat transfer in fuel-cladding gap

Heat transfer coefficient in fuel-cladding 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, burned-up fuel. Calculation models for fuel-cladding gap heat transfer.

Heat transfer in core

Theory of similarity, dimensionless numbers. Single-phase flow: convective and conductive heat transfer. Forced convection, natural convection. Spacer grid effect on heat transfer coefficient. Heat transfer coefficient of VVER reactors.

Two-phase 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. Steady-state 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.

Steady-state 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 steady-state 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 fuel-cladding gap
Scope: 4 tutorials

3. Steady-state temperature distribution in fuel channel

Heat transfer in fuel channels, steady-state 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
ReferencesKey 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: 0-89448-038-3


Experimental Reactor Physics17ERF Rataj, Sklenka - - 4 kz - 4
Course:Experimental Reactor Physics17ERFIng. Rataj Jan Ph.D.-4 KZ-4
Abstract:The lectures are focused on experimental methods used for determination of neutron-physical 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 VR-1 reactor.
The lectures are supplemented with experimental practices at the training reactor VR-1: 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 VR-1 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. Source-jerk method. Rod-drop 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 inter-calibration.

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 VR-1 reactor
Range: 2 lectures,
Topic of lecture:
Basic requirements for core configurations in VR-1 reactor. Design of VR-1 reactor core configuration. Neutron-physical characteristics of VR-1 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 VR-1

1. Reactivity measurement at VR-1 reactor
Range: 1 practice
Topic of practice: Source-jerk, rod-drop and positive period method application on reactivity measurement at VR-1 reactor.

2. Control rod calibration in VR-1 reactor
Range: 1 practice
Topic of practice: Determination of control rod worth and its calibration curve in VR-1 reactor by inverse rate method and source multiplication method. Control rods inter-intercalibration in VR-1 reactor.

3. Study of nuclear reactor dynamics
Range: 1 practice
Topic of practice: Study of VR-1 reactor behaviour in subcritical, critical and supercritical state with and without external neutron source. Study of VR-1 reactor response to the pulse, transient and periodic reactivity changes. Study of void reactivity coefficient influence on under-moderated and over-moderated nuclear reactor.

4. Approaching the critical state at VR-1 reactor
Range: 1 practice
Topic of practice: Prediction of unknown critical state at VR-1 reactor by inverse rate method. Approaching the critical state at VR-1 reactor by gradual change of control rod position.

5. Basic critical experiment at VR-1 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 VR-1 reactor.


Goals:detailed knowledge in the field of experimental reactor physics, knowledge and assumption of method focused on determination of basic neutron-physical 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

ReferencesKey references:
Weston M. Stacey: Nuclear Reactor Physics, John Wiley and Sons, Inc., New York 2001, ISBN 0-471-39127-1

Recommended references:
Lewis E.,E.: Fundamentals of Nuclear Reactor Physics, Elsevier Inc., USA 2008, ISBN: 978-0-12-370631-7

Media and tools:
training reactor VR-1


Nuclear Fuel Cycle17JPC Sklenka, Starý - - 2+0 kz - 2
Course:Nuclear Fuel Cycle17JPCdoc. Ing. Sklenka Ľubomír Ph.D. / Ing. Starý Radovan2+0 KZ-2-
Abstract:The course is focused on front-end & back-end 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 front-end 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 back-end 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 (open-cast 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, solvent-extraction, etc.), production and composition of yellow cake.

4. Purification and conversion to UF6
Duration: 1 lecture
Topic:
Nuclear-grade specification, miscellaneous purification methods of yellow cake to the nuclear-grade 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. Back-end of the nuclear fuel cycle
Duration: 2 lectures
Topic:
Back-end of the nuclear fuel cycle, once-through 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 front-end & back-end 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, front-end of the nuclear fuel cycle, back-end 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
ReferencesKey references:
Stacey, W. M.: Nuclear Reactor Physics, Chapter 5 Nuclear Reactor Dynamics &
Chapter 6 - Fuel Burnup, WILEY-VCH 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, IAEA-TECDOC-1532, IAEA, Vienna, 2007
Design of Fuel Handling and Storage Systems in Nuclear Power Plants Safety Guide, IAEA Safety Guide, NS-G-1.4, IAEA, Vienna, 2003

Media and tools:
Audiovisual technique, films on DVD

Thermohydraulic Design of Nuclear Devices 417THNJ4 Kobylka 3+0 z,zk - - 4 -
Course:Thermohydraulic Design of Nuclear Devices 417THNJ4Ing. 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, ..), two-phase flow (important for emergency analyses of nuclear devices, description of power loaded parts of PWR or design of BWR), about sub-channel 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), steady-state shock, steam flow (real gas).

2. Two-phase flow
Time range: 7 lectures
Fundamentals of two-phase flow and its miscellaneous patterns, diagrams of flow, principles of two-phase flow description, definition of basic quantities and their calculation (void fraction, etc. ), two-phase pressure drops, basic two-phase 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; liquid-vapor 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: K-epsilon, K-omega, 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 sub-channel analysis and their use for fuel assemblies calculations, s transcription of basic flow equations for sub-channel analysis, follow up equations fior sub-channel analysis (turbulent mixing, pressure losses, etc.), application of sub-channel 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 Pb-Bi), 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, drift-flux model, dynamic instability, turbulent flow, models of turbulent flow, K-epsilon, , RNG, LES, sub-channel analyse, CFD.
References1. Tong, L.S., Weisman, J.: Thermal Analysis of Pressurized Water Reactors, American Nuclear Society, Illinois USA, 1996, ISBN: 0-89448-038-3
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: 0-89448-011-1
5. 2. Mareš R. - Šifner O. - Kadrnožka J.: Tabulky vlastností vody a vodní páry podle průmyslové formulace IAPWS-IF97, VUTIUM , 1999, ISBN 80-214-1316-6
6. Lahey R.T., Moody F.J.: The Thermal-Hydraulics of a Boiling Water Nuclear Reactors, Američan Nuclear Society, La Grange Park, 1993, ISBN 0-89448-037-5

Machines and Equipment of Nuclear Power Plants17SAZ Kobylka 2+1 z,zk - - 3 -
Course:Machines and Equipment of Nuclear Power Plants17SAZIng. 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, diesel-generators, 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
ReferencesKey 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: 0-89448-038-3


Excursion Abroad17EXZ Frýbort - - 1 týden z - 2
Course:Excursion Abroad17EXZIng. 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 117VUJR12 Frýbort 0+6 z 0+8 kz 6 8
Course:Research Project 117VUJR1Ing. 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:-
Referencesaccording to given topic


Course:Research Project 217VUJR2Ing. 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:-
Referencesaccording to given topic

Optional courses

Computer Control of Experiments17PRE Kropík 2+1 z,zk - - 3 -
Course:Computer Control of Experiments17PREdoc. 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
ReferencesKey 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 Physics17SMRF Huml 2+2 kz - - 4 -
Course:Stochastic Methods in Reactor Physics17SMRFIng. 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 burn-up 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 neutron-physical 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

pre-processors and post-processors for input and output simplification ? Sabrina and MCNPVised codes for simplification of MCNP input geometry generation, MONACO code for verified inputs generation for VR-1 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 non-symmetric problems and various interfaces definition, Russian roulette and other computer-time saving methods

2. Mathematical modeling of burn-up in nuclear reactor systems
Rozsah: 4 přednášky
Témata přednášek:
Simple burn-up models for reactor systems ? solution of short-term and long-term kinetics using MATLAB code

Burn-up modeling by diffusion and transport methods ? WIMS code application for burn-up calculations, burn-up problem definition for elementary cell, complex geometry

SCALE calculation system for nuclear reactor burn-up modeling ? application of SCALE code for nuclear reactor neutron-physical characteristics calculations, overview of basic modules, description and characteristics of KENO, TWOONEDAT and ORIGEN modules, application of ORIGEN module for burn-up calculations, problem definition, input data processing, geometry definition, research reactors fuel burn-up calculation, fuel burn-up in pressurized water reactors

HELIOS code ? Application of HELIOS code for nuclear reactor neutron-physical characteristics calculations, description and characteristics of HELIOS code and applied mathematical model, description of input and output files for fuel burn-up 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 burn-up problems

application of WIMS code for burn-up modeling, research reactor burn-up problem

basics of SCALE system use, modules, problem definition, geometry and material description, simple research reactor fuel burn-up problem

fuel burn-up calculation for pressurized water reactor VVER-440 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 burn-up 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, burn-up, neutron transport, Monte Carlo method
ReferencesStacey, W. M.: Nuclear Reactor Physics, WILEY-VCH 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, Addison-Wesley Pub. Co, 1969

James E. Gentle: Random Number Generation and Monte Carlo Methods (Statistics and Computing), Springer, 2004

Deterministic Methods in Reactor Physics17DERF Frýbort - - 2+2 kz - 4
Course:Deterministic Methods in Reactor Physics17DERFIng. 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 burn-up 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 neutron-physical 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 study-program, 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 (JEF-PC, 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 ? neutron-physical characteristics calculation of reactor systems ? Overview of computer codes utilizing numerical methods to neutron-physical characteristics of reactor systems. Description and characteristics of computer codes WIMS, TWODANT-SYS.DANTYS, and CITATION. Input files generation for these codes for neutron-physical 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 ANISN-ORNL, TORT-DORT. 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 high-quality 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 TWODANT-SYS.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 ANISN-ORNL code including calculation of nuclear facility shielding parameters and characteristics

use of TORT-DORT 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, full-core calculations
ReferencesJohn 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, WILEY-VCH 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 Reactors17CIBS Kropík 2+0 z,zk - - 2 -
Course:Digital Safety Systems of Nuclear Reactors17CIBSdoc. 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 VR-1 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
ReferencesKey references:
Nuclear power plants - Instrumentation and control systems important to safety - Software aspects for computer-based systems performing category A functions, IEC60880, 2006
Review Guidelines on Software Languages for Use in Nuclear Power Plant Safety Systems, NUREG/CR-6463, 1996

Recommended references:
Nuclear power plants - Instrumentation and control important for safety - Software aspects for computer-based systems performing category B or C functions, IEC62138, 2004
Standard Criteria for Digital Computers in Safety Systems of Nuclear Power Generating Stations, IEEE-7.4.3.2-2010

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 Sources17EEZ Tichý, Kobylka - - 2+1 z,zk - 3
Course:Energy Sector and Energy Sources17EEZIng. 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, non-uniformity 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. Fossil-fuel power plant and biomass
Scope: 2 lectures
Basic description of fossil-fuel power plant, boilers types and their principles (stocker-fired, fluid, dry-bottom, 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, fossil-fuel power plant, boiler, renewable energy sources, hydroelectric power plant, photovoltaics, transmission network system, energy sector in the Czech Republic, the State energy policy, DESAE.
ReferencesKey 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: 80-86165-98-1
Recommended references:
WWW sites of Energy regulatory office: http://www.eru.cz/

Media and tools:
PC classroom, computer code DESAE

Selected Parts of Legislation17VPL Bílková, Fuchsová - - 2+0 z - 2
Course:Selected Parts of Legislation17VPLRNDr. 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 co-operation, 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, on-site 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
ReferencesKey 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 Plants17EHJE Starý 2+0 zk - - 2 -
Course:Economic Evaluation of Nuclear Power Plants17EHJEIng. Starý Radovan2 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, break-even 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
ReferencesKey refernces:
SCHILLER, Bradley R. Mikroekonomie dnes. Brno : Computer Press, 2004. 412 s. ISBN 80-251-0169-X.
STOFT, Steven. Power System Economics : Designing Markets for Electricity. 1st edition. [s.l.] : Wiley-IEEE Press, 2002. 496 s. ISBN 0-471-15040-1.
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 0-470-84572-4.

Media and tools:
Computer room

Computer Science for Modern Physicists17IMF Havlůj 0+3 kz - - 3 -
Course:Computer Science for Modern Physicists17IMFIng. Havlůj František0+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 (HTML-based) form. They are able to use the Ruby scripting language at the level needed for the day-to-day 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
ReferencesKey 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, Addison-Wesley, 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 Reactors14NMR Čech, Haušild - - 2+0 zk - 2
Course:Materials Science for Reactors14NMRprof. 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 Zr-coating 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.
ReferencesKey references:
[1] G.S. Was, Fundamentals of Radiation Materials Science Metals and Alloys, Springer-Verlag 2007.
Recommended references:
[1] J. Koutsky and J. Kocik , Radiation Damage of Structural Materials. Material Science Monographs 79, Elsevier 1994.