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 computational and physical studies on the production of corium and its retention in an MR’s melt trap of the Lava-B facility. A feature of the Lava-B facility used in the IAE NNC RK to study the processes occurring during a severe accident at a nuclear reactor, is the separation of the stages of the reactor core corium formation and its interaction with structural materials. The melting of materials takes place in an induction furnace with a hot crucible, after which it moves to a melt receiver (MR) in which the test object is located. In the case of studies of processes occurring outside the reactor vessel, this is a special trap, which is placed in the inductor to simulate decay heat. However, based on the conservative computational estimates, it was found that the inductor power in the MR can be sufficient to directly produce, melt, and, subsequently, maintain the corium in the liquid phase. In this regard, in order to optimize the experiments under controlled conditions, the authors came up with the idea to experimentally test the possibility of producing corium by induction heating directly in the MR’s melt trap. In addition, according to the authors, this method would obviate the problem of corium contact with the carbon environment of the melting furnace of the Lava-B facility. Previously, burden heating simulating corium was modeled on the computer using available parameters of the MR’s induction heater. Based on the numerical experiment, the conditions for physical modeling of the corium production in the MR’s melt trap were established. An analysis of the physical modeling showed that during the burden heating in the melt trap, its metal components became liquid, thus, forming a melt pool. However, in terms of this design of the trap, there were problems associated with the complete melting of all corium components, as well as with the integrity of the experimental device when forming the corium pool and during the actual physical modeling.

This paper presents the results of an experimental study of the interaction of a candidate sacrificial material (SM) for a light water reactor melt trap with corium at the Lava-B test-bench. The candidate sacrificial material is a combination of aluminum oxide and a lead layer. The idea of using such a combination of SM is based on the fact that when the lead layer interacts with corium, there will be an increase in the intensity of heat removal from the corium, as well as the chemical interaction between the corium and SM due to the high heat-conducting properties of lead. This approach will improve the efficiency of corium localization in the melt trap compared to the current set of sacrificial material. Experiments have shown active melting and boiling of lead during its interaction with corium. This is confirmed both by the readings of thermocouples and by the X-ray diffraction phase analysis of the deposit material formed on the walls of the melt receiver (MR) of the Lava-B bench, sampled after the experiment. The experiment results show that the lead layer reduces the rate of increase in the temperature of the corium and increases the rate of erosion of the ceramic part of the SM. With these circumstances, it is possible to conclude that the use of aluminum oxide with a lead layer is promising in practice.

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.