This article is devoted to an overview of the conducted work and the current status of decommissioning of the world’s first BN-350 industrial fast neutron reactor. The reactor was put into operation on 16 July 1973 in Aktau. In 1999, the government of Kazakhstan decided to shut down the reactor, and from that moment to the present, it has been in the decommissioning stage. All work on decommissioning the reactor facility was grouped into five stages. The first stage was completed in 2010 when the spent fuel of the BN-350 reactor was placed for long-term storage. The second stage is nearing completion. Research is currently underway to develop technologies for processing radioactive sodium. The goal of the third and fourth stages of the BN-350 reactor decommissioning is the comprehensive processing of liquid and solid radioactive waste. Now such waste is stored in special storage directly on the territory of the nuclear power plant.

This paper presents the results of a calculation code approach providing a solution to the point kinetics problem for the IVG.1M research reactor of the National Nuclear Center of the Republic of Kazakhstan and allowing the simulation of dynamic processes going on during reactor start-ups, including changes in the thermal state of all its elements, reactor regulator displacement, accumulation of absorbers in the fuel, and the beryllium reflector. A mathematical description of the IVG.1M point kinetics model is presented, which provides a calculation of the reactor neutron parameters, taking into account the dependence of reactivity effects on the temperature, changes in the isotopic composition of materials, and thermal expansion of core structural elements. An array of data values was formed of reactivity added by separate elements of the core when changing their thermal state and other reactor parameters, as well as an array of data with the parameters of heat exchange of coolant-based reactor structural elements. These are used in the process of solving the point kinetics problem to directly replace formal parameters, eliminating the need to calculate the values of these parameters at each calculation step. Preliminary calculations to form an array of values of reactivity effects was applied to the reactor by separate structural elements when their temperature changes were performed using the IVG.1M precision reactor calculation model. The model was validated by the reactor parameters in the critical state. Preliminary calculations to form an array of data with the parameters of heat exchange of coolant-based reactor structural elements were performed in ANSYS Fluent software using the calculation model that describes the IVG.1M reactor fuel element in detail. Validation of the developed calculation code based on the results of two start-ups of the IVG.1M reactor was performed and its applicability for the analysis of transient and emergency modes of reactor operation and evaluation of its safe operation limits was confirmed.

The nuclear power industry preferably aims to a closed nuclear fuel cycle and a transition to a two‐component structure based on the joint operation of thermal and fast neutron reactors. This approach makes it possible to significantly reduce the volume of radioactive waste and increase the efficiency of using uranium fuel. Sodium fast neutron reactors currently demonstrate the highest technical readiness for commercial operation. This status has been achieved through the accumulation of operational experience from research reactors to pilot demonstration projects. Many of these reactors are currently in the process of decommissioning. Decommissioning of sodium fast neutron reactor facilities is associated with serious intellectual, material, and technological difficulties. There is still no overall experience in the world of completing the decommissioning of such facilities. Nevertheless, it is believed that the experience and competencies accumulated in these activities are key to forming a strategy for decommissioning of the future nuclear power plants with fast neutron reactors as an important final stage of the life cycle of such nuclear facilities. This is very important because decommissioning is a complex process that requires careful planning and coordination to ensure the safety of personnel, the public, and the environment in the future. In this regard, this review is devoted to an overview of the current progress in decommissioning various experimental sodium fast neutron reactor facilities. The accumulated experience, existing problems, and prospects for further work have been analyzed.

Research of prototype fuel elements and fuel assemblies in in-pile tests are an integral part in substantiating the safety and operability of installations being designed, and are also necessary for validating the calculation codes. Despite the long service life and a large number of studies conducted in the IGR reactor, the capabilities of this reactor under conditions of implementing long-term experiments at minimum power are still not fully clear. This paper presents a substantiation of the possibility of implementing an experiment with power reactor fuel during the implementation of the long-term operation mode of the IGR reactor. Neutronic and thermal-hydraulic calculations were performed to determine the energy release in the fuel and structural materials of the experimental device, the time to achieve steady-state heat exchange between the fuel element and the cooling gas at a steady-state fuel element power level, and the temperature distribution in the test section at specified energy parameters.