Recycling and burning minor actinides (MA, e.g., americium, neptunium) in mixed-oxide (MOX) nuclear fuel is a strategic option for fast reactor concepts of Generation IV, especially considering the current interest in the ultimate radioactive waste management and sustainability improvement by better use of natural resources. Among the fuel properties, thermal conductivity and melting temperature are pivotal since they determine, respectively, the fuel temperature profile and the fundamental safety limit on the margin to fuel melting, hence impacting on the overall fuel performance under irradiation and allowing the safe irradiation of the fuel pin. Nevertheless, the available literature about Am- or Np-containing MOX is currently scarce, both regarding experimental data and models. Moreover, state-of-the-art fuel performance codes (FPCs, e.g., TRANSURANUS) do not account for the effects of minor actinides on MOX fuel properties. This work presents original correlations for thermal conductivity and melting temperature of minor actinide-MOX fuels, i.e., (U, Pu, Am, Np)O2-x, derived based on the available literature and accessible data, which are herein extensively reviewed. The assessment of the novel correlations is first performed in a statistical way, evaluating the regressor p-values which indicate their significance with respect to the available experimental dataset used for the fitting procedure. Additionally, the novel correlations for MA-MOX are assessed against both measured and calculated data (from Molecular Dynamics simulations), yielding an accuracy in line with the already existing correlations and with the state-of-the-art experimental uncertainties. Finally, the potential integral impact of a homogeneous minor actinide content in the fuel is illustrated on the basis of a fuel pin fast-ramped up to fuel melting during the HEDL P-19 irradiation experiment.

Modelling of thermal conductivity and melting behaviour of minor actinide-MOX fuels and assessment against experimental and molecular dynamics data

A. Magni;L. Luzzi;D. Pizzocri;
2021

Abstract

Recycling and burning minor actinides (MA, e.g., americium, neptunium) in mixed-oxide (MOX) nuclear fuel is a strategic option for fast reactor concepts of Generation IV, especially considering the current interest in the ultimate radioactive waste management and sustainability improvement by better use of natural resources. Among the fuel properties, thermal conductivity and melting temperature are pivotal since they determine, respectively, the fuel temperature profile and the fundamental safety limit on the margin to fuel melting, hence impacting on the overall fuel performance under irradiation and allowing the safe irradiation of the fuel pin. Nevertheless, the available literature about Am- or Np-containing MOX is currently scarce, both regarding experimental data and models. Moreover, state-of-the-art fuel performance codes (FPCs, e.g., TRANSURANUS) do not account for the effects of minor actinides on MOX fuel properties. This work presents original correlations for thermal conductivity and melting temperature of minor actinide-MOX fuels, i.e., (U, Pu, Am, Np)O2-x, derived based on the available literature and accessible data, which are herein extensively reviewed. The assessment of the novel correlations is first performed in a statistical way, evaluating the regressor p-values which indicate their significance with respect to the available experimental dataset used for the fitting procedure. Additionally, the novel correlations for MA-MOX are assessed against both measured and calculated data (from Molecular Dynamics simulations), yielding an accuracy in line with the already existing correlations and with the state-of-the-art experimental uncertainties. Finally, the potential integral impact of a homogeneous minor actinide content in the fuel is illustrated on the basis of a fuel pin fast-ramped up to fuel melting during the HEDL P-19 irradiation experiment.
Nuclear fuel, Minor actinide-bearing MOX, Thermal conductivity, Melting temperature, Modelling and assessment.
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Utilizza questo identificativo per citare o creare un link a questo documento: http://hdl.handle.net/11311/1204278
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