E. A. Gomin

The status of MCU-5

The software developed within the framework of the MCU project which implements the Monte Carlo method for solving the transport equations for various kinds of radiation is widely used for calculating nuclear reactors of different types. The final version of MCU-5 is described. The developed libraries of constants are briefly described, and the potentialities of the MCU-5 package modules and the executable codes compiled from them are characterized. Examples of important problems of reactor physics solved with the codes from the MCU family are presented.

Evaluated benchmark experiments at critical assemblies simulating features of an HTHR at the ASTRA facility

The design of the ASTRA facility and critical assemblies that simulate physics features of modular high-temperature reactors (HTHR-Ms) with a graphite moderator and reflectors loaded with fuel particles having multilayer ceramic coatings is described in detail. Geometrical dimensions of the main elements and regions of the critical assemblies, composition of the materials used, and experimental results for various configurations of the critical assemblies are presented. A detailed computational benchmark model allowing for the structural and compositional features of the critical assembly configurations in question is developed on the basis of all the above data. The results are to be used for verification of the neutronics codes used for calculations of high-temperature helium-cooled reactors.

Simulation of Nuclear Reactor Kinetics by the Monte Carlo Method

The KIR computer code intended for calculations of nuclear reactor kinetics using the Monte Carlo method is described. The algorithm implemented in the code is described in detail. Some results of test calculations are given.

Calculation of the Neutron Importance Function and the Delayed Neutron Effective Fraction by the Monte Carlo Method

An algorithm for calculation of the neutron importance function and the delayed neutron effective fraction by the Monte Carlo method implemented in the KIR program is presented. The results of calculation of the delayed neutron effective fraction in some critical experiments are given in comparison with the experimental results.

Calculation of Prompt Fission Neutron Lifetimes by the Monte Carlo Method

An algorithm for calculation of prompt fission neutron lifetimes in a nuclear reactor by the Monte Carlo method is described. Evaluation of the importance function is carried out with solution of the neutron transport equation without solving the adjoint equation. The results of the prompt neutron lifetime calculations performed within some critical experiments are presented and compared with the experimental results.

Experiments on small-size fast critical fuel assemblies at the AKSAMIT facility and their use for development of computational models

Small-size fast critical assemblies with highly enriched fuel at the AKSAMIT facility are described in detail. Computational models of the critical assemblies at room temperature are given. The calculation results for the critical parameters are compared with the experimental data. A good agreement between the calculations and the experimental data is shown. The physical models developed for the critical assemblies, as well as the experimental results, can be applied to verify various codes intended for calculation of the neutronic characteristics of small-size fast nuclear reactors. For these experiments, the results computed using the codes of the MCU family show a high quality of the neutron data and of the physical models used.

Status of PERST-5 package

The methods and algorithms used in the PERST-5 package are described. This package is part of the MCU-5 code and is intended for neutron-physical calculation of the cells and parts of nuclear reactors using a generalized method of first collision probabilities.

Calculated analysis of experiments in fast neutron reactors

In this paper, the results of computational simulation of experiments with the MK-I core of the JOYO fast neutron sodium-cooled reactor are presented. The MCU-KS code based on the Monte Carlo method was used for calculations. The research was aimed at additional verification of the MCU-KS code for systems with a fast neutron spectrum.