Study on specific heat capacity and thermal conductivity of uranium nitride

In this study, we have proposed an analytical method for calculating the specific heat capacity of uranium nitride nuclear material. The specific heat capacity results have obtained by the use of the Debye-Einstein approximation. The thermal conductivity of nuclear material has been obtained by using the experimental data of thermal diffusivity and the calculation results of specific heat capacity. This method shows that our results are satisfactory for the wide range temperature variations. The proposed approach can be easily applied to determine the thermodynamic properties of the other nuclear materials.

Neutronic Performance of a Compact 10 MWe Nuclear Reactor with Low Enrichment (ThxU1−x)N Fuel

Recent interest in compact nuclear reactors for applications in space or in remote locations drives innovation in nuclear fuel design, especially non-oxide ceramic nuclear fuels. This work details neutronic modeling designed to support the development of a new nuclear fuel concept based on a mixture of thorium and uranium nitride. A Monte Carlo N-Particle Version 6.2 (MCNP-6) model of a compact 10 MWe reactor design which incorporates (ThxU1−x)N fuel is presented. In this context, a “compact” reactor is a completely assembled reactor which may be emptied of coolant and transported by specialized commercial vehicle, deployed by a C130J aircraft, or launched into space. Core geometry, reflector barrels, and the heat exchange zones are designed to support reduction of overall reactor volume of core components while maintaining criticality with a fixed total fuel mass of 4500 kg. Dense mixed nitrides of thorium nitride (ThN) additions in uranium nitride (UN) in 5 wt.% increments between $$0.05 \le x \le 0.5$$ 0.05 ≤ x ≤ 0.5 have been considered for calculation of $$k_{\infty }$$ k ∞ and $$k_{{{\text{effective}}}}$$ k effective . ThN additions in UN results in a slight increase in the magnitude of the temperature coefficient of reactivity, which is negative by design. The isotopic distribution of the principal actinide inventory as a function of burnup, time, and initial fuel composition is presented and discussed within the context of the proliferation risk of this core design.

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.

Computational Modeling Towards Accelerating Accident Tolerant Fuel Concepts and Determining In-Pile Fuel Behavior

To mitigate global warming, we need to develop carbon-free ways to generate power. Nuclear energy currently generates more carbon-free power in the United States than all other sources combined at 55%. To make nuclear as viable a power source as possible, we need to maximize power density and safety. Both of these can be improved with Accident Tolerant Fuel (ATF) materials. Uranium nitride (UN), a candidate ATF material, offers high fuel economy due to its uranium density and improved safety margins from thermal properties. However, its instability in the presence of water, a reactor coolant, must be addressed. This dissertation employs Density Functional Theory-based methods to investigate the atomistic and electronic mechanisms in UN corrosion initiation. To ensure accuracy in future UN models, the effects of magnetic treatments on UN surface stability and corrosion properties are also determined. The performance of advanced nuclear materials must be tested in research reactors before they can be implemented in power reactors. To get real-time temperature data from these tests, sensors are required that can survive the high temperatures and irradiation. To meet these needs, Idaho National Laboratory has been developing High Temperature Irradiation Resistant Thermocouples (HTIR-TCs). Towards increasing temperature resolution and in-pile lifetime, an ab initio method has been developed to predict HTIR-TC performance. The method considers the effects of composition and temperature on performance and has been validated against experiment. To predict the interaction of HTIR-TCs with research reactor coolant, corrosion and oxidation mechanisms have been investigated. By examining the diffusion behaviors of water and oxygen, recommendations are made for which thermoelement materials may be the most resistant to corrosion and/or oxidation.