The MCU-TR Code

Alekseev N.I., Bolshagin S.N., Bykov V.P., Gomin E.A., Gorodkov S.S., Gurevich M.I., Kalugin M.A., Kulakov A.S., Marin S.V., Oleynik D.S., Prjanichnikov A.V., Shkarovsky D.A., Yudkevich M.S.

Software registration certificate N2011619384.

Passport of the code N336.




The code is designed as a machine-independent. It was tested on personal and multiprocessor computers.


The code MCU-TR with its nuclear data bank MDBTR50 is designed for precise modelling of the transport of neutrons and photons in the core of transport reactors by means of analog and weight (non-analog) Monte Carlo methods based on evaluated nuclear data for nuclear reactors, taking into account changes in the isotopic composition of reactor materials during the its operation.

Main areas of application:

- Precision neutron-physical calculations of multiplying systems with complex spatial and energy distribution;

- Verification of the engineering codes of physic calculations of nuclear reactors.

A code allows carrying out benchmark calculations of neutron-physical characteristics of the two- and three-dimensional fragments of cores of transport reactors, as well as full-scaled cores of these reactors. The technology of simulation of fast neutrons near the vessel to assess the fast-neutron flux at the surface and in depth of the reactor vessel, as well as simulation of fast and thermal neutrons in the active volume of ionization chamber for the reactor vessel. Technology has been developed for calculating the change in the isotopic composition of reactor materials and calculating the radiation characteristics of irradiated nuclear fuel.


MCU-TR simulation of interaction of neutrons with the nuclei of the medium is based on information stored in the files of evaluated nuclear data with an accurate description of the energy dependence of the laws of interaction of neutrons with nuclei. Herewith:

- In the fast energy region at simulation of elastic and inelastic neutron the scattering laws of the interactions defined in files of evaluated nuclear data are used without any simplification;

- In the region of unresolved resonances the code uses subgroup representation of the interaction of neutrons with nuclei, taking into account the temperature dependence of subgroup parameters;

- In the resolved resonance region a description of the continuous dependence of the resonance cross sections is implemented, taking into account the Doppler effect, i.e., the calculation of cross sections produced for each neutron energy at a given temperature environment without pre-tabulation and subsequent interpolation of these sections;

- In the thermalization region the calculation is made for the given temperature of the moderator, using cross sections derived from the generalized phonon spectra, and not counted for some pre-set temperature scattering laws.

In modelling the interaction of photons with matter and photons' birth in the reactions of neutrons with nuclei not only the group representation of cross-sections of the reactions, but also pointwise representation is used, including all information available in the files of evaluated nuclear data.

MDBTR50 data bank contains information for 375 isotopes.

The geometry of the system can be an arbitrary collection of bodies bounded by second-order surfaces and/or planes. Additionally, there is the Woodcock method that will accurately simulate the helical shape of the fuel elements.

There is two- and three-dimensional visualization of the entire geometry of the system and its individual structural elements.

A set of calculated functionals enables verification of engineering codes (spectral) and reactor codes, namely, it is possible to calculate not only the effective neutron multiplication factor and the nuclear reaction rates, but also all of the constants, including the diffusion coefficients, which are used in engineering codes to calculate reactor cores.

The code provides an opportunity of calculating the distribution of energy over fuel assemblies and/or fuel pins in the volume of the reactor core .

There is the opportunity to solve the steady homogeneous and inhomogeneous transport equations. The latter is necessary to model the reactor start-ups with a given external source of neutrons.

It is possible to carry out the calculation of a start-up state at permanently poisoned by Xe for a given power.

When calculating the change in the isotopic composition of materials during the campaign all isotopes of minor actinides, nuclear fission activation products of structural materials are taken into account, moreover, it is possible to calculate the radiation characteristics of spent nuclear fuel.

MCU-TR code is written in standard language Fortran-93/95/2003 with dynamical memory allocation.

The code can operate on PCs with WINDOWS operating system and multi-core processors or multiprocessor computers equipped with MPI (Message Passing Interface) software interface.

MCU-TR is based on the modules of the MCU-5 package.


The code has no fundamental limitations on the complexity of simulated systems.


The computing time depends on the isotopic composition of reactor materials, the complexity of the geometry of the simulated system and the necessary statistical accuracy of estimated functional.

The computing time depends on the user's qualification in the choice of physical model, the initial data to describe the geometry and the choice of time step calculation of burnup.


A working version of the code for each specific computer and installed operating system is generated from the modules of the MCU-5 package using its preprocessor. Source text of modules is written in the input language preprocessor, which generates text files in the Fortran language. Parameters defining the amount of memory needed to solve a certain class of problems, as well as functional features of the code may be changed by the user at the stage of generation.

As compared with previous versions of the family MCU there is the possibility of parallel computing using multiprocessor systems, the possibility of modelling photon transport, the use of dynamic memory, which makes it much more economical in respect of memory consumption.

Greatly expanded the nuclear data bank.

In addition, the code provides the following features:

- Calculation of the scattering laws for the desired temperature during data input for calculated options for solving the neutron transport equation.

- Automatic change of any parameter in the calculation of burnup.

- Calculation of several states with an automatic change in the source data.

- Automatic withdrawal of the problem at a given value of the effective neutron multiplication factor Keff.


MCU Office, which is a graphical interface for MCU. Its main functions are as follows:

- 2D and 3D visualization of input data for geometric module NCG and the generation of error messages, if any;

- A possibility of editing the original data file for the MCU;

- Viewing files created by the MCU in the calculation;

- Run a task in a separate window;

In addition, the user is presented with some additional features:

- Generation of a working version of MCU;

- Comparison of two images of a calculated system;

- Display in colour the reaction rates obtained in the calculation;

- Display in text form and colour the reaction rates obtained in the process of calculation, as well as the number of materials, areas, facilities, and other information;

- Etc.


The code is deposited in OFAP NR 18 November 2010 (inventory number 713 OFAP YAR).

There is documentation[1].


1. N.I. Alekseev, et. al. Development and verification of software tools to perform precise calculations of the neutron-physical characteristics of the transport reactors cores based on the Monte Carlo method. Report RRC KI Inv. № 36-10/35-09, Moscow, 2009.


The texts of the code MCU-TR with the library MDBTR50 and description of the code take upto 2 GB hard drive, the recommended minimum amount of RAM is 1 GB per processor core.



Source text file modules are written in the input language preprocessor of the MCU package, which generates text files in Fortran-90/95.

Graphical visualization is written in C++.


MS-DOS (version higher than 3.3); WINDOWS 95, 98, NT, 2000, XP, Server 2003, Vista, 7.