The article presents the results of a mathematical Monte-Carlo simulation of the passive mode of an active well coincidence counter (AWCC) in the Serpent software environment. The developed model was experimentally tested on different types of neutron sources. Estimates show that the results of the numerical simulation in the Serpent software environment can be used to refine and expand the range of effective mass measurements for 240Pu by adjusting the calibration coefficients.

This paper analyzes the methods and technologies for assessing the method of formation, composition, characteristics and features of corium, which is a mixture of nuclear and structural materials of the nuclear reactor core, formed as a result of an accident accompanied by partial or complete core melting. The study is based on data from the study of corium formed as a result of the accident at the Fukushima Daiichi nuclear power plant, which are in the public domain and are the result of the work of many scientific organizations around the world. Corium research is one of the main issues in the framework of improving nuclear safety in the future and is one of the objectives of the successful procedure for eliminating the consequences of the accident at the Fukushima Daiichi nuclear power plant. Without a detailed analysis of the neutronic, materials science, gravimetric and other characteristics of the corium, as well as the creation of a complex model of the corium that combines these data, it is impossible to organize an efficient and safe process for removing nuclear materials from the damaged units of the Fukushima Daiichi nuclear power plant. The objective of this work is to combine the existing research results into a data set that allows modeling of the corium using neutronic calculation codes and includes such data as the size, density and morphology of corium samples and their approximate nuclide composition. Such modeling allows not only to perform tasks related to increasing the level of safety in the implementation of the procedure for eliminating the consequences of the accident at the Fukushima Daiichi nuclear power plant, but also to serve as an international benchmark for modeling a mixture containing nuclear materials.

Virtual reality (VR) technology is now being adopted in many industries, including entertainment, medicine, science, and engineering. In the nuclear field, the primary purposes of VR are: reducing radiation dose rates, security of nuclear facilities, visualization of physical processes, and training of personnel. Additionally, VR is a much cheaper alternative to expensive and license-requiring experimental nuclear facilities. This work focuses on reconstructing the workroom with the Uranium-Water Subcritical Assembly (UWSA) located at the National Research Nuclear University MEPhI to determine the optimal uranium–water ratio associated with this assembly in virtual reality. The creation of the virtual analog using Unreal Engine 4 was introduced to integrate the physical model into the virtual environment. The neutronic model of the UWSA was obtained by the MCU code. A similar model was generated by the Serpent code for verification purposes. Additional functions such as neutron flux visualization, radiation dose rate distribution visualization, and dose accumulation mechanics were introduced into the project to improve the quality of education. Visualization of both neutron flux in the assembly and gamma radiation distribution in the workroom was performed using particle systems and volumetric fog based on calculated and experimental data. Operating experience feedback was introduced to prevent or minimize difficulties that may occur in the future by learning from events that have already occurred.

Fuel debris removal is the most challenging part of damaged nuclear power station decommissioning. It is important to carry out nuclear safety calculations accurately and quickly enough. Here, it was clarified that modern codes based on the Monte Carlo method were capable of performing neutronic analysis with the same accuracy and without significant differences in the results. The benchmark calculations were performed using three codes: MVP, Serpent, and MCU. In this study, the comparison focused on multiplication factor, neutron fluxes and reaction rates relative difference, and calculation time of many fuel debris particles system. Then the calculation results were used when codes comparing. It was shown that the calculation results showed good agreement between all codes. It was assumed that minor differences in the thermal range of neutron fluxes can be caused by different thermal neutrons scattering treatment for all codes. The study also showed that solving such problems requires significant computing power and time. It has been proven that the statistical geometry model in the MVP and the explicit stochastic geometry model in the Serpent have the possibility to provide solutions with the same accuracy, but much faster.

Abstract This work aimed at generating homogenized group constants using the Serpent code and then using the CriMR diffusion code to model the large SFR OECD 3600 MWth MOX core. The results were compared with a full core reference Monte Carlo solution by Serpent. Reactivity feedback parameters were also considered. Generating the group constants from separate fuel assemblies allows for simultaneously carrying out calculations and then using the results as input in diffusion codes rather than waiting so long for a 3D full core Monte Carlo calculation to be completed. From the results of the integral parameters we see a close agreement in the calculation codes. The differences can be attributed to the errors that could arise from generating the constants from individual sub-assemblies. The differences in the underlying physics and approximations used in development of the codes could also be a factor. Another way the errors could be reduced is by checking to see that the sub-assembly configurations used in the non-multiplying zones are as close as possible to the real layout in a full 3D core.

Accident at Fukushima Daiichi nuclear power plant led to increase of importance of safety justification for processes at post-accident facilities in nuclear industry. One of such processes is extraction of corium from reactors cavity. Recriticality of this process is defined by potential unacceptable accident. This paper introduces supporting code for neutron fluxes and reaction rates visualization in systems with complex geometry that can be used in modeling of corium removing works. Visualization code is based on Unreal Engine 4 game engine. Code allows observing neutronic functionals distribution in three dimensions. The reseach and provided implementation details help to understand the physical processes that take place as the accidents occur during corium removing works.

After the accident at the Fukushima Daiichi NPP, the attention of the scientific community is riveted on how the consequences are being eliminated. Removing corium – a lava-like resolidified mixture of nuclear fuel with other structural elements of the reactor – remains the most difficult task, the solution of which can take several decades. It is extremely important to exclude the occurrence of any emergency processes during the removal of corium. The purpose of this work was to solve a coordinated hydrodynamic and neutronic problem characterized by a large number of randomly oriented and irregularly located corium particles in water as part of the development of a benchmark for this class of problems. Monte Carlo-based precision codes were used to perform a neutronic analysis. The positions of corium particles were determined from the numerical simulation results. The analysis results obtained using the codes involved showed good agreement for all the states considered. It was shown that the modern neutronic codes based on the Monte Carlo method successfully cope with the geometric formation and solution of the problem with a nontrivial distribution of corium particles in water. The results of the study can be used to justify the safety of corium handling procedures, including its extraction from a damaged power unit.

As part of the tasks to improve the nuclear safety of nuclear power plants, a new program code was developed. The CORIUMSITY program code developed, considered in this work, is intended to analyze the scenario in which an accident at a nuclear power plant is simulated with the melting of the core and the formation of the so-called “corium” - a mixture of nuclear and structural materials of the nuclear reactor core, formed as a result of thermal and mechanical impact during an accident. The CORIUMSITY program code, is intended to analyze several scenarios of different accidents, include an accident with reactor core melting. The functions of this code can help in solving many urgent nuclear safety problems. One of the main methods of operation of the CORIUMSITY code algorithms is the matrix exponential method, which consists in using a matrix function of a square matrix, in which as values are used indicators corresponding to nuclides from the CORIUMSITY code database. The program implements an iterative Euler method for solving the system of levels of nuclear fuel burnup. The CORIUMSITY code was verified with benchmark data to assess the accuracy of the calculation.

The article describes the new software product developed at MEPhI. It represents a virtual reality simulation of an experiment on a subcritical uranium-graphite assembly. This practical work plays an important role in the training of young specialists studying the physics of nuclear reactors. However not all students have access to real experimental facilities, this fact makes it necessary to complement real experiment with simulation in virtual reality that allows to accurately reproduce the actions that the student performs during the real practical work. This approach let to increase the efficiency of the educational process and even expand the capabilities of real experimental assembly by visualizing physical processes during its operation.

Using virtual reality technology - a modern trend. The nuclear industry is no exception. This article provides an overview of mathematical models used to create virtual analogue of critical assembly Godiva in virtual reality. Godiva - there is a simple example that allows to hone techniques for creating more complex virtual analogues of nuclear reactors and nuclear facilities. Mathematical models include stationary and dynamic ones. The stationary model is based on data from calculations carried out using Monte Carlo programs such as MCU, Serpent and Geant4. An approach is also described that makes it possible to calculate the reverse multiplication from the values of the effective multiplication factor for various states of the subcritical assembly. The dynamic model allows one to calculate the neutron-physical characteristics of the supercritical assembly during fast processes such as a neutron burst. In conclusion, there are other examples of virtual analogs created using similar approaches.

The paper presents algorithm of solving neutron diffusion equation in three-dimensional space based on multiple application of alternating direction implicit method. The point of the method is to reduce solving of multidiagonal systems of equations to solving of few three diagonal systems in iterative processes. Diffusion code CriMR that uses this algorithm implemented for triangular mesh is described. Results of initial tests of the code are also presented (comparison of parameters of three different homogenized cores of BN-600 calculated by CriMR with ones of MCU-PTR).

. Minimization of the dose load on personnel of nuclear facilities is an important task. This work focuses on the possibilities of reducing the accumulated dose rate using augmented reality technology. The concept of a system for visualizing the distribution of the radiation background in a workroom located in the educational laboratory of the Department of Theoretical and Experimental Physics of Nuclear Reactors at National Research Nuclear University MEPhI is proposed. The workroom contains subcritical experimental facilities with sources of ionizing radiation. The system allows avoiding places of increased radiation, thereby minimizing the dose load. The concept is implemented in the form of an Android application made on Unreal Engine 4. The application creates a multi-colored fog, the color of which at each point depends on the dose rate of gamma radiation.