The BN-350 is the first industrial fast neutron reactor in the history of nuclear energy. It is currently undergoing decommissioning. One of the key challenges of decommissioning is managing the solid radioactive waste that has accumulated throughout the reactor’s operational life. At the moment, the accumulated solid radioactive waste is stored in a storage facility within the BN-350 reactor complex. An analysis showed that more than ~7262 tons with 5.17 × 10¹⁴ Bq activity of various types of solid radioactive waste have been accumulated over the reactor operation. They are mainly represented by materials with low activity. At the same time, the main share of activity is comprised of highly active waste with a total mass of ~170 tons and an activity of 4.73 × 10¹⁴ Bq. A solid radioactive waste management strategy has been developed. It includes all stages from collection and classification to transportation and long-term storage. Modern technologies now offer new possibilities. Some radioactive waste can be processed and reused in other economic sectors. In particular, recycling metals and alloys can reduce the volume of solid radioactive waste. It can also return valuable materials to industrial use.

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.

In 2022, the physical start-up stage of the IVG.1M research reactor was successfully completed, initiated by reducing the fuel enrichment of ²³⁵U. This phase included the loading of nuclear fuel into the reactor core and conducting experiments to determine the neutron-physical characteristics of the reactor. Prior to the physical start-up, preliminary calculations were performed using the computational code MCNP6 and a full-scale model of the IVG.1M reactor with low-enriched uranium fuel (LEU). During the start-up series, the reactivity worth curves of the reactor control and protection system’s operating and compensating elements were determined. Additionally, experiments were performed to measure the reactivity effects of technological channel draining and to obtain activation reaction rates in the central experimental channel using nickel and gold activation indicators. The results of determining the neutron-physical characteristics of the IVG.1M reactor have confirmed the operability of the reactor core with LEU fuel.

The paper considers safety ensuring issues during the work at the plant intended for researching the technology of irradiated beryllium purification using chlorination method. Presented technology assumes handling beryllium and its compounds, chlorine and hydrogen chloride in compliance with the following safety aspects: radiation and toxic safety. Based on the results of the calculation of an expected dose and taking into account the requirements for the attended personnel, the allowable duration of maintenance the reaction chamber with the loaded irradiated beryllium sample was determined. The amount of radioactive beryllium in each cleaning cycle was determined from the point of view of the reasonable minimization (based on the safety rules) of the possible internal exposure of personnel due to the intake of gaseous tritium. The paper describes a unique way to measure the concentration of free chlorine by optical method and methods to protect the harmful effects of chlorine gas and hydrogen. The results of this study demonstrate possibility of safe technology implementation during preparation and conduction of experiments for the treatment of irradiated beryllium.