Stochastic Methods in Reactor Physics
Code  Completion  Credits  Range  Language 

17SMRF  KZ  4  2+2  Czech 
 Garant předmětu:
 Lecturer:
 Tutor:
 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 burnup modeling.
Stress is put on practical examples, exercises and individual students? work on solving of given exercises. After passing the course, the attendees obtain not only theoretical knowledge, but also practical experience with various methods and approaches to modeling of neutronphysical characteristics of nuclear facilities and their application in real reactor systems.
 Requirements:

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
preprocessors and postprocessors for input and output simplification ? Sabrina and MCNPVised codes for simplification of MCNP input geometry generation, MONACO code for verified inputs generation for VR1 reactor and for output files processing
solution of problems for neutron flux determination ? calculations of neutron flux densities and particle fluencies in reactor systems, fluxes and currents in simple and complex geometry structures, possibilities of calculated values processing by TECPLOT code
calculation optimization ? optimization approaches for fastening MCNP calculations, symmetric and nonsymmetric problems and various interfaces definition, Russian roulette and other computertime saving methods
2. Mathematical modeling of burnup in nuclear reactor systems
Rozsah: 4 přednášky
Témata přednášek:
Simple burnup models for reactor systems ? solution of shortterm and longterm kinetics using MATLAB code
Burnup modeling by diffusion and transport methods ? WIMS code application for burnup calculations, burnup problem definition for elementary cell, complex geometry
SCALE calculation system for nuclear reactor burnup modeling ? application of SCALE code for nuclear reactor neutronphysical characteristics calculations, overview of basic modules, description and characteristics of KENO, TWOONEDAT and ORIGEN modules, application of ORIGEN module for burnup calculations, problem definition, input data processing, geometry definition, research reactors fuel burnup calculation, fuel burnup in pressurized water reactors
HELIOS code ? Application of HELIOS code for nuclear reactor neutronphysical characteristics calculations, description and characteristics of HELIOS code and applied mathematical model, description of input and output files for fuel burnup calculation in pressurized water and boiling water reactor systems
3. Technical visit of nuclear reactor neutronics calculation department
Rozsah: 1 přednáška
Témata přednášek:
Technical visit of nuclear reactor neutronics calculation department in Nuclear Research Institute in Rez or department for reactor calculations in Skoda Nuclear Machinery in Pilsen
 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 burnup problems
application of WIMS code for burnup modeling, research reactor burnup problem
basics of SCALE system use, modules, problem definition, geometry and material description, simple research reactor fuel burnup problem
fuel burnup calculation for pressurized water reactor VVER440 using ORIGEN module
seminar work ? generation of own model of given reactor problem, its solution and outputs processing using TECPLOT code.
presentation of students? seminar works
 Study Objective:

detailed knowledge of mathematical modeling in nuclear reactor physics, statistic methods in nuclear reactor physics and nuclear reactor fuel burnup modeling
orientation in the field, application of gained knowledge in other courses in the field of theoretical reactor physics
 Study materials:

Stacey, W. M.: Nuclear Reactor Physics, WILEYVCH Verlag GmbH & Co. KGaA, Weinheim, 2007
Christian P. Robert, George Casella: Monte Carlo Statistical Methods (Springer Texts in Statistics), Springer, 2005
Jerome Spanier: Monte Carlo principles and neutron transport problems, AddisonWesley Pub. Co, 1969
James E. Gentle: Random Number Generation and Monte Carlo Methods (Statistics and Computing), Springer, 2004
 Note:
 Further information:
 No timetable has been prepared for this course
 The course is a part of the following study plans:

 Jaderné inženýrství  Jaderné reaktory (compulsory elective course)