Abstract
Thermal conductivity and melting temperature of nuclear fuel are essential for analysing its performance under irradiation, since they determine the fuel temperature profile and the melting safety margin, respectively. A starting literature review of data and correlations revealed that most models implemented in state-of-the-art fuel performance codes (FPCs) describe the evolution of thermal conductivity and melting temperature of Light Water Reactor (LWR) MOX (uranium-plutonium mixed oxide) fuels, in limited ranges of operation and without considering the complete set of fundamental dependencies (i.e., fuel temperature, burn-up, plutonium content, stoichiometry, and porosity). Since innovative Generation IV nuclear reactor concepts (e.g., ALFRED, ASTRID, MYRRHA) employ MOX fuel to be irradiated in Fast Reactor (FR) conditions, codes need to be extended and validated for application to design and safety analyses on fast reactor MOX fuel. The aim of this work is to overcome the current modelling and code limitations, providing fuel performance codes with suitable correlations to describe the evolution under irradiation of fast reactor MOX fuel thermal conductivity and melting temperature. The new correlations have been obtained by a statistically assessed fit of the most recent and reliable experimental data. The resulting laws are grounded on a physical basis and account for a wider set of effects on MOX thermal properties (fuel temperature, burn-up, deviation from stoichiometry, plutonium content, porosity), providing clear ranges of applicability for each parameter considered. As a first test series, the new correlations have been implemented in the TRANSURANUS fuel performance code, compared to state-of-the-art correlations, and assessed against integral data from the HEDL P-19 fast reactor irradiation experiment. The integral validation provides promising results, pointing out a satisfactory agreement with the experimental data, meaning that the new models can be efficiently applied in engineering fuel performance codes.
Highlights
The development of Generation IV reactor concepts, employing uranium-plutonium mixed oxide (MOX1) fuel irradiated under fast neutron flux, calls for dedicated and advanced fuel modelling with respect to the state of the art
The first step towards the proposal of a novel thermal conductivity correlation for fast reactor MOX is the selection of the best set of experimental data, i.e., composed of measurements recently obtained from up-to-date techniques and covering wide ranges of the considered dependencies, as explained in what follows
K0(T, x, [Pu], p) − kin f where k0(T, x, [Pu], p) is the fresh MOX thermal conductivity calculated with Eq 2, bu is the fuel burn-up in GWd/tHM, kinf is the asymptotic thermal conductivity at high burn-up based on the only two available sets of experimental data on irradiated fast reactor MOX
Summary
The development of Generation IV reactor concepts, employing uranium-plutonium mixed oxide (MOX1) fuel irradiated under fast neutron flux, calls for dedicated and advanced fuel modelling with respect to the state of the art (target of the INSPYRE H2020 Project [1]). The aforementioned experimental works have been exploited to build models of MOX fuel thermal conductivity for fast reactor applications, developed e.g., by Bonnerot [15] and Philipponneau [9], based on the comprehensive review of thermal properties of oxide fuels carried out by Martin [16] They showed that in the wide temperature range between 800 and 3100 K the MOX thermal conductivity is always lower with respect to UO2, with only slight effects of plutonium concentration (below 12 wt.%) in the entire temperature range, and of O/M ratio at high temperatures. An extensive review of experimental and modelling works dedicated to MOX melting temperature has been recently performed by Calabrese et al [41], highlighting the importance of the oxygento-metal ratio for the melting behaviour of MOX fuel, especially in the high Pu concentration domain, and the prominent role played by the plutonium content in MOX fuel for fast breeder reactor applications.
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