On the Development of Tolerant Fuel --- a New Generation Fuel

The concept of tolerant fuel is considered as applied to water-cooled power reactors. The concept is based on eliminating the steam-zirconium reaction. For this, two work areas, i.e., using the physical and the thermodynamic barriers, were considered. Physical barrier presupposes exclusion of contact between water and zirconium, and the thermodynamic barrier (the most radical method) envisages replacement of the alloy containing zirconium with other materials inert to water when exposed to high temperature in the reactor core (∼ 1200 °C). Consequences of the most devastating accidents at the nuclear power plants in the world were discussed: Three Mile Island, Chernobyl and Fukushima. The latest accident in Japan brought to the fore the concept of tolerant nuclear fuel, i.e., being resistant to accidents. Work orientation in creating the tolerant fuel is indicated. Main attention is paid to materials and technologies applied to tolerant fuel. General requirements to safety analysis of the reactor facility fuel system currently developed in the Russian Federation and abroad, as well as current safety criteria for fuel elements, under design-based accidents are presented. Procedure for calculating justification of the safety criteria fulfillment for fuel elements under design-basis accidents is briefly considered. Main characteristics of the new generation materials under development for reactor cores as applied to tolerant fuel are presented. Based on comparing the proposed materials as the tolerant fuel for the fuel element claddings, composite materials based on the heat-resistant SiC/SiC ceramic system could be recommended, and as far as fuel materials are concerned --- materials with increased density, uranium capacity and thermal conductivity values, i.e., nitride fuel and fuel made of uranium silicide

Radiation-hygienic support of work with nitride fuel for fast reactors: problems, achievements and prospects. Initial position

The article is devoted to the substantiation of the need to implement a special program of radiation-hygienic support of work with nitride fuel for fast neutron reactors. It is shown that at the current pace of implementation of the project direction "Breakthrough", in conditions when achievements in scientific research lead to a revision of design and technological solutions, it is possible to manage radiation and hygiene support only in a mode that provides a quick response to changes in the real production and environmental situation.

Thermophysical Properties of Mixtures of Thorium and Uranium Nitride

The miscibility, lattice parameter, and thermophysical properties of (Th0.2U0.8)N and (Th0.5U0.5)N have been investigated. It is shown that additions of thorium nitride (ThN) to uranium nitride (UN) increases the thermophysical performance of the mixed nitride fuel form in comparison to reference UN. In the more dilute limit, additions of ThN serve as a burnable neutronic poison and reduces the change in keff over the lifecycle of the fuel. At higher concentrations, additions of ThN serve as a significant fertile source of 233U. Where appropriate, comparisons to previous work on UN + PuN mixtures are made, as this is a comparable fuel form for potential fast reactor concepts, and a suitable point of contrast in the possible design space afforded by mixed (ThxU1 − x)N fuel forms. The data from this work are the input parameters for finite element modeling of the temperature distribution in a compact reactor. The results of modeling and simulation of this core design are shown for the case of steady-state operation and during double, adjacent heat pipe failure.

Opening of Nitride Spent Nuclear Fuel

Nitride nuclear fuel (UN + 10-20% PuN) is considered a promising alternative to the widely used oxide nuclear fuel (UO2). Thermal conductivity and density of nitride fuel are ∼ 7 times and 1.3 times higher than that of oxide fuel, respectively. Nitride fuel demonstrates a good compatibility with the cladding of fuel rods made of stainless steel. Along with the development of new fuel, methods for its subsequent processing are being developed. Various options for the initial opening of nitride spent nuclear fuel (SNF) are considered in this article. The use of gaseous chlorine is technologically inconvenient and dangerous when working with radioactive substances. The electrochemical dissolution of nitride SNF cannot be realized due to the formation of a by-product - UNCl. Uranium nitride chloride is an insulator and it blocks the electrochemical process. It was found that the chlorination of nitride SNF with cadmium or lead chlorides makes it possible to carry out 100% UN → UCl3 conversion. The use of voloxidation (oxidation of nitride SNF to oxides) as the first stage of processing will make the entire technology universal, suitable for processing both nitride and oxide SNF. The choice of the method for opening SNF depends on the choice of the subsequent stages of its processing. Keywords: nitride spent nuclear fuel, SNF, chlorination, anodic dissolution, UNCl, “soft” chlorination, voloxidation, processing