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A simple theoretical model is proposed in this paper based on an analysis of the experimental data available in the available literature. It describes the rate of mass loss during the dissociation of uranium nitride in a vacuum and in a gas atmosphere due to the evaporation of substances from the surface of materials. The proposed model can be used to assess the degree of dissociation of uranium nitride, which is one of the main causes of fuel destruction, since congruent melting is observed only at high nitrogen pressures.

When fuel rods melt during a severe accident, the movement of the structural and fuel materials along fuel assemblies can be spatially non-uniform, so that a three-dimensional thermohydraulic model is implemented in the EVKLID/V2 code as part of the HYDRA-IBRAE/LM module. The transport of the same sets of components and their mixtures as in the one-dimensional version of the HYDRA-IBRAE/LM module can be calculated in the model: liquid sodium, sodium vapor or vapor-gas mixture of sodium vapor and non-condensable gases, solid lead, liquid lead, solid uranium dioxide, liquid uranium dioxide, hard stainless steel, liquid stainless steel, steel vapor. The three-dimensional model is implemented in a cylindrical coordinate system, which makes it easier to include the geometric dimensions and parameters of a particular fuel assembly (number and diameter of fuel rods, lattice pitch, and others) and the core. A description is given of the basic system of equations describing the motion of the components of the destroyed core in the three-dimensional r–z–φ geometry, and its numerical realization. Examples of test calculations showing the serviceability of the model are presented.

The article describes the basic models laid down in the second version of the EUCLID/V2 integrated code developed for carrying out end-to-end analysis of severe accidents in liquid metal cooled reactors. Brief information about the basic analogs of the code is given. Unlike the first version of the code, its second version includes additional tools for analyzing design-basis and beyond-design-basis accidents involving fuel pin, fuel assembly, and reactor core failures. To this end, the code is supplemented with additional modules using which it is possible to calculate fuel rod tightness failure as a consequence of its melting, escape of fission products into the coolant, their transport over the circuit, and release into the nuclear power plant rooms. The code also incorporates modules for calculating the core failure processes. Special attention is paid to the physical models for calculating the core materials' melting processes, motion of the produced melt, its interaction with the coolant and with other materials, and propagation of fission materials. For calculating the core failure processes, a multicomponent 3D model has been implemented. The methods used for calculating heat transfer and friction between the components are based on well-proven analytical and empirical relations for determining the heat transfer and friction coefficients. The coefficients presented in the article also depend on the obtained multicomponent flow motion regime and the type of components (metal and ceramics). The algorithms governing joint operation of the thermomechanical, thermal-hydraulic, neutronics, and the fuel rod thermal failure module are described. Emphasis is placed on data exchange methods in the course of an accident in the reactor. The approaches used for calculating the transport of fission products in the coolant and in the NPP rooms are presented.

The article presents the results obtained from verification of the EUCLID/V2 coupled code developed at the Nuclear Safety Institute of the Russian Academy of Sciences, which is intended for analysis of accident conditions in liquid metal cooled fast reactors. The additional capabilities available in the code in comparison with its first version include, in particular, analysis of individual equipment components in the 3D approximation, consideration of the transport of fission products and corrosion products in the coolant and in the nuclear power plant buildings, and also analysis of severe accidents in a fast reactor. The article presents the code verification results and assessment of its applicability for analysis of accidents involving destruction of fuel pins and the reactor core. The verification was carried out against the data obtained at experimental facilities and from analytic tests. Information about the key experiments used to validate the code is briefly outlined. In particular, data of experiments carried at the Oak Ridge, Argonne, and Sandia National Laboratory, the United States; at the National Nuclear Center of the Kazakhstan Republic; and on the test bench at the Nizhny Novgorod State Technical University (NSTU) in Russia are used. Modules of the coupled code EUCLID/V2 integrated code verification matrix are given. The errors of calculating the parameters most important for analysis of an accident’s consequences evaluated using the EUCLID/V2 code are proven with due regard to the computation and experimental results. The ranges of parameters within which the code has been verified are determined. The uncertainty and sensitivity of computation results are also analyzed based on the results from simulating certain experiments. Factors having the main influence on the computation results are determined. It is shown that the computation results are consistent with the experimental results subject to the input data uncertainty.

Due to the revival of interest to the development of fast reactors cooled by liquid metals, the problem of carrying out theoretical research in support of their safety is actual. A detailed calculation of all stages of the accident from the beginning to the end requires knowledge of the laws for modeling physical processes occurring in the reactor in an emergency. The most serious are accidents with the destruction of the core. Simulation of severe accident in nuclear reactor is the key element in safety analysis of nuclear power plants. Destruction of fuel rods is one of the most important processes that should be calculated during core degradation. For different type of fuels the mechanism of the degradation are different too. For example, oxide and metallic fuels usually melt congruently at high temperature, but nitride fuel dissociates. The main objective of the proposed research is developing of models and numerical algorithms for calculation fuel rods destruction with oxide, metallic and nitride fuels. The models of the destruction processes and some calculation results are presented in the paper. The processes are investigated for the first phase of severe accidents covering the period from the onset of fuel-rod melting to the melt escape from the core center.

The results of verification of the IBRAE-developed SAFR/V1 module, describing in the EVKLID/V2 integral code the thermal breakdown of fuel pins, on the basis of data obtained using experimental facilities. The experiments performed on the TREAT and DEH facilities at the Argonne National Laboratory and on a bench at the Nizhny Novgorod State Technical University were picked for the verification process. The computational error was evaluated for individual parameters on the basis of the verification results. The impact of the uncertainties in the initial data on the computational results was analyzed.

The basic approaches used in the SAFR/V1 module of the integral code EVKLID/V2 to simulate the movement of the melt formed upon melting of fuel elements are presented. The system of mass, energy, and momentum conservation equations used to simulate the movement of melt is presented. Special attention is devoted to methods of numerical approximation of the equations as well as to the solution of problems involving smearing of the solution at the melt boundary. The realized methods of stimulating the motion of melt have been verified on the basis of tests with known analytical solutions.

For the confirmation of the claimed design properties of a reactor plant with a heavy liquid-metal coolant, computational and theoretical studies should be performed in order to justify its safety. As one of the basic scenarios of an accident, the leakage of water into the liquid metal is considered in the case of steam generator tube decompression. The most important in the analysis of such a kind of accidents are questions concerning the motion and heat exchange of steam bubbles in the steam generator and the probability of blocking the flow area owing to freezing coolant, since the temperature of boiling feedwater in the steam generator can become lower than the melting point of lead. In this paper, we present main approaches and relationships used for the simulation of the motion of gas bubbles and heat transfer between bubbles and liquid metal flow. A brief description of the HYDRA-IBRAE/LM computational code that can be used to analyze emergency situations in a liquid metal-cooled reactor facility is also presented. It should be noted that the existing experimental data on the motion and heat transfer of gas bubbles in a heavy liquid metal are insufficient. For this reason, in order to verify the HYDRA-IBRAE/LM code models, experiments have been performed on the cooling of liquid lead by argon and on the motion of gas bubbles in the Rose’s alloy. In particular, a change in the temperature of the coolant over time has been studied, and the void fraction of gas at different flow rates of gas has been measured. A detailed description of the experiments and a comparison of the results of the calculations with the experimental data are presented. The analysis of uncertainties made it possible to reveal the main factors that exert the greatest effect on the results of calculations. The numerical analysis has shown that the models incorporated into the HYDRA-IBRAE/LM code allow one to describe to a sufficient degree of confidence the process of cooling liquid lead melt when argon bubbles pass through it, simulating the flow of water into the liquid metal in the course of steam generator tube rupture.

The approaches used in the SAFR/V1 module modeling the thermal destruction of fuel pins in fast reactors for calculating in integrated code EVKLID/V2 the melting of fuel pins and solidification of the formed melt are described. The enthalpy formulation of the heat-conduction equation is used. The numerical scheme used for the heat-conduction equation in performing the calculations is presented. The realized boundary conditions as well as a model of the gas gap in a fuel pin are described. The module’s capability of modeling heat propagation, including in the presence of a phase transition, is verified on the basis of analytical tests.

The presented work is dedicated to the development of approaches to simulate cladding melt relocation along the surface of the fuel pin. Development of the approaches is based on the results of the experiments carried out at the NSI RAS and IT SB RAS. Features of the melt relocation are studied in the experiments. It is demonstrated that the laminar film flow regime in the heated part of the fuel simulator is the main flow regime. Model of the melt relocation is constructed. This model is the part of the SAFR module of the EUCLID/V2 coupled code. It is shown that the proposed approaches allow simulating the melt relocation with good accuracy.

The results of thermohydraulic code HYDRA-IBRAE/LM/V1 verification on experimental data obtained using the BN-600 reactor in different years are described. The particulars of the computational scheme of the reactor using HYDRA-IBRAE/LM/V1 for modeling the BN-600 operating regime are presented. In preparing the scheme, special attention was devoted to accurate modeling of the reactor cool-down regimes on natural circulation as being most important from the safety validation standpoint. The computational results obtained with the uncertainty of the initial data taken into account are presented for such a cool-down regime.

The basic relations used in the HYDRA-IBRAE/LM/V1 code for calculating the flow and heat-exchange of sodium coolant in fuel-rod assemblies in regimes with and without boiling and in the presence of a crisis of heat exchange are presented. The experimental data suitable for verifying the thermohydraulic code for calculating friction in one- and two-phase regimes as well as heat exchange in regimes with and without boiling are picked on the basis of a review of the literature. The computational error of different parameters is obtained on the basis of the verification results. It is shown that the thermohydraulic code HYDRA-IBRAE/LV/V1 can be used to calculate correctly the primary processes in design basis and beyond-design basis accidents in sodium-cooled reactor facilities.

Heat-exchanger models in the SOKRAT-BN code which are used to calculate problems with boiling sodium in channels with different geometry are presented. The results of modeling of different experiments on boiling of liquid-metal coolant are presented. Good agreement is obtained between the SOKRAT-BN calculations and experiments performed with stationary and nonstationary boiling of sodium. It is shown that the thermohydraulic processes occurring in reactor facilities during design-basis and beyond design basis accidents can be calculated correctly using the thermohydraulic module of the SOKRAT-BN computer code.

The investigation of the melt flow of a liquid metal along the surface of a heated rod is carried up. In the experiments, the metal from the upper volume was drained over the surface of a quartz tube, heated from the inside. This pipe is used to simulate a fuel column. A high-speed video of the process was produced. Data are obtained on the features of the flow of a metal melt. The stages of formation of droplets are shown. The data on the shape and velocity of the droplet movement are given. It is shown that in the air atmosphere around the test section when the first drop passes on the surface of quartz, a trace is formed, along which the metal moves in the future. Direct and curvilinear rivulet flows on the rod's surface are also observed.

The paper contains the results of numerical simulation of stainless steel melt motions on the surface of uranium dioxide. The investigations are performed for purposes of understanding of the fuel rod behavior during the core disruptive accident in the fast reactors. The systems of mass, energy and momentum conservation equations are solved to simulate melt motion on the surface of the fuel pin. Heat transfer and friction between melt and pin's surface and melt and coolant flow are taken into consideration. The dependences of mass of the melt and the features of the melt motion on coolant velocity and contact angle between melt and surface of the fuel rod are presented.

A new model which describes the dynamics of a vertically stratified flow correctly within the limits of a single-pressure two-fluid model has been developed. The model is based on the modification of finite-differences of convective terms and pressure gradients taking into account a distinct interface. We propose to use the vapor quality as a criterion for the onset of dryout. The choice of the criterion is based on the analysis of experimental and theoretical studies. To determine the boundary vapor quality we used the correlation xcr = 1.26·G0.2, which was found from experimental data fit. A review of articles has shown that for today it is impossible to predict correct superheat value. Therefore the superheat value was determined as a parameter of the model from the experimental data of a particular simulated experiment. Thus a boiling up regime was selected. The model described in this paper allows us to calculate the boiling up of sodium under the superheat conditions as well as problems of the evolution of the vapor volume. The verification of the models was done by using the SOCRAT-BN code [1]. SOCRAT-BN is a coupled code which consists of modules for calculation of damage and melting of a reactor’s core, thermohydraulic processes and neutron physics. The models of vertical stratification, dryout and slug boiling of superheated sodium are described in details in this paper. Also we present the results of verification for the models within analytic tests and experimental data.

SOCRAT-BN is a software package to simulate design and severe accidents of sodium-cooled fast reactors. The package consists of modules for calculating damage to the reactor’s core, thermohydraulic processes and neutron physics. The thermohydraulic module has been developed to calculate one- and two-phase flows in channels with different geometry and bundles. The module is based on a two-fluid model for equal pressures of phases. In this paper we present an explanation of the deciding constitutive models for equations used in the system. Validation of the module was performed on the experimental data for one- and two-fluid flows in complex geometry channels and on calculation of running a first loop of the reactor BN-600 in nominal mode.

The results of testing the thermohydraulic module of the SOKRAT-BN computing code for analyzing accidents with boiling of sodium coolant in fast reactors are presented. The computational results are compared with experimental data. It is shown that the thermohydraulic module of the SOKRAT-BN code models stationary sodium boiling well. Using as a basis the results obtained by modeling sodium boiling in a vertical heated channel, a system of closure relations for calculating two-phase sodium flow regimes, including the interphase velocity, was modified and checked. Modeling sodium boiling in a vertical annular channel also showed that the closure relations incorporated in the thermohydraulic module of the SOKRAT-BN code are suitable for calculating heat-exchange with a wall.