Incorporation of uranium nitride fuel capability into the ENIGMA fuel performance code: Model development and validation

Abstract

Uranium dioxide (UO${}_2$) is the primary fuel form for nuclear reactors but its moderate uranium density and low thermal conductivity have prompted the exploration of alternative materials. Uranium nitride (UN) has emerged as a promising candidate for a variety of reactors, offering higher uranium density and thermal conductivity. This paper details the development and implementation of UN fuel capabilities within the ENIGMA fuel performance code for Light Water Reactor (LWR) applications. The new UN capability in ENIGMA includes correlations for theoretical density at room temperature, thermal conductivity, specific heat capacity, enthalpy, thermal expansion strain, Young’s modulus, Poisson’s ratio, thermal creep strain rate, irradiation creep strain rate and emissivity. Additionally, it incorporates models for densification, solid fission product swelling, fission gas bubble swelling, and fission gas release, along with a modified RADAR model for determining the pellet radial power profile and helium generation. Validation of the UN model was conducted using data from the L414 pin irradiation in the JOYO fast reactor in Japan. Further validation efforts are planned using datasets from JOYO and the Siloé thermal reactor in France. The paper also outlines areas of future work to address experimental data gaps and enhance model accuracy to cover a broader range of cladding materials, manufacturing parameters (including porosity volume fraction) and operating conditions (including fuel temperatures and burnups). Although focused on LWR applications, the work outlined supports the use of UN fuel across various reactor systems.