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CZECH TECHNICAL UNIVERSITY IN PRAGUE
STUDY PLANS
2018/2019

Deterministic Methods in Reactor Physics

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Code Completion Credits Range Language
17DERF KZ 4 2+2 Czech
Lecturer:
Jan Frýbort (guarantor)
Tutor:
Jan Frýbort (guarantor)
Supervisor:
Department of Nuclear Reactors
Synopsis:

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.

Requirements:
Syllabus of lectures:

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.

Syllabus of tutorials:

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

Study Objective:

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.

Study materials:

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, 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

Note:
Time-table for winter semester 2018/2019:
Time-table is not available yet
Time-table for summer semester 2018/2019:
Time-table is not available yet
The course is a part of the following study plans:
Data valid to 2019-06-16
For updated information see http://bilakniha.cvut.cz/en/predmet3653206.html