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.

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.

In this paper, the idea of modifying the benchmark by increasing effective multiplication factor value in the considered system without increasing the number of particles under consideration and fuel enrichment is discussed. The technology of data transmission, processing and comparison of the results of neutron-physical calculations using three modern codes developed in different countries and implementing the Monte Carlo method has been worked out. For the corium and water mixture parameters search, the method conventionally called the gradient descent method is used. The search method for areas of critical state corium slurry in water is described. This method can be used to justify nuclear safety in the corium extraction and transportation processes. The methodology is based on the combined use of 1-D and 3-D criticality calculations capabilities of the SCALE 6.2 program pack. Fall of corium particles in water simulation benchmark version is formulated. The possibility of using the algorithm to find the parameters of a corium and water mixture is demonstrated. This benchmark includes the critical state of corium slurry in water and assumes the use of regular structures in the formation of geometric models. The proposed version of the critical benchmark for the corium particles in water state contains 55% fuel.

Among the various international policies for the management of spent nuclear fuel and radioactive waste (RW), an important issue is the need to reprocess this waste. The aim of this work is to analyze current modern methods for transmutation of minor actinides (MA) in the reactor cores of thermal, fast (with homogeneous and heterogeneous placement) and in accelerated driven systems in order to identify factors limiting the efficiency of transmutation. This work is an analytical review and it defines the main methods for removing the limitations of the effectiveness of the transmutation process. On the basis of MA formation sources data and MA decay chains, namely americium, neptunium and curium, as well as transmutation methods in the latest reactor systems, factors limiting the efficiency of the transmutation process and methods of reducing their influence are investigated. The main factors are: an increased level of gassing during transmutation, industrial restrictions and the influence of the transmutation process on the processes occurring in the core during transient processes. The main current problem of transmutation is the high activity in terms of heat release level and neutron emission of irradiated assemblies due to the presence of curium isotopes.

This paper describes the development of full-scale models of the BR-1200 reactor for the MCU-FR code with a homogeneous and heterogeneous description of fuel assemblies’ geometry. The correctness of the control rods efficiency calculation in a homogeneous model is analyzed. The control rods requiring heterogeneous modeling are defined.

In this work was carried out the simulation in the SCALE6 code of an experiment on the BN-600 reactor on irradiating of fuel assemblies, containing samples of mixed nitride uranium-plutonium fuel. A comparison of the results for SCALE6 on the results of other codes is presented. The results of an estimation of uncertainties in the calculated data connected with uncertainties of an irradiation of an experimental sample and used neutron cross-sections library. Discussion of possible differences between analytical and experimental results is given.

The object of research in this work is the core of a high-power fast neutron reactor with lead coolant (hereinafter referred to as the BR-1200). The purpose of this work is to determine the effectiveness of the CPS in the core of such a reactor in various states that differ in the type and number of CPS organs, which are simultaneously in the core. An additional goal is to determine the effectiveness of the method of partial heteronization of the fuel assembly structure in the model. In addition, the task is posed to compare the results with the data obtained using other neutron-physical codes and libraries of neutron constants. The calculations were carried out in the KENOVI module of the SCALE code. Neutron constant libraries such as the ENDF / B7 point library and the v7-238 group library were used. This paper describes the creation of full-scale models of the BR-1200 reactor with a homogeneous and heterogeneous description of the fuel assembly structure. The article describes the simulation of fuel assemblies with a homogeneous description of the fuel structure and a heterogeneous description of the structure of absorbing rods. The calculation of such parameters as the effective multiplication factor and the efficiency of CPS organs is carried out.