Stochastic Methods in Reactor Physics

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Code Completion Credits Range Language
17SMRF KZ 4 2+2 Czech
Garant předmětu:
Ondřej Huml
Ondřej Huml
Ondřej Huml
Department of Nuclear Reactors

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.


17FAR - Fyzika jaderných reaktorů - nutná podmínka

18MOCA - Metoda Monte Carlo - doporučený předmět

Syllabus of lectures:

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

Syllabus of tutorials:

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

Study Objective:

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

Study materials:

Stacey, 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

Time-table for winter semester 2023/2024:
Time-table is not available yet
Time-table for summer semester 2023/2024:
Time-table is not available yet
The course is a part of the following study plans:
Data valid to 2023-08-30
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