For the past 5 decades, active development of alternative, so-called high density low enriched fuels for research and test reactors have been ongoing. This has led to the conversion of many research reactors (RR) from high-enriched uranium (HEU) based fuels to low enriched uranium (LEU) based fuels, thus lowering the risk of proliferation or illicit use of these sensitive materials. In particular, the qualification of the uranium silicide (U3Si2) based dispersion fuel system has led to a number of succesful conversions. Because one class of RRs, the so-called high-performance research reactors (HPRR), are currently still incapable of converting to LEU based fuels without very significant loss in performance, fuel development is still ongoing. 3 major development paths are being followed : higher loaded U3Si2 dispersion fuel, U-Mo alloy dispersion fuel and U-Mo alloy monolithic fuel. Several irradiation programs are ongoing and planned for the future to qualify these fuels for the conversion of the HPRR in Europe, the US and potentially the rest of the world. SCK-CEN has played a key role in these developments over the last 20 years [1, 2], particularly in the dispersion fuels where the fuel compound is dispersed in an aluminum matrix. Towards the future, SCK-CEN will also get more involved in the monolithic fuel developments.
RR fuels used historically were known to have an athermal behaviour in normal operating conditions (typically coolant temperatures of <50°C and fuel temperatures of <200°C), which means that the fuel performance is not influenced by the temperature reached by the fuel as long as the fuels are operated within the thermohydraulic limits of the reactor in which they are used. This means that RR operators are not required to operate a fuel performance code to determine the domain in which the fuel is allowed to be operated in function of burnup, corrosion layer buildup, thermal conductivity, etc, which is different from power reactor operators. The limits are determined by the ability of the reactor's thermohydraulics to evacuate the heat of the fuel and coolant boiling is the limiting phenomenon. The fuels never reach the operating temperatures at which thermally activated phenomena have a dominating impact on the fuel behaviour during normal operation and use of the fuel. However, recent results have indicated the possibility that U-Mo based and potentially U3Si2 dispersion fuels may show thermally activated phenomena which occur at temperatures which can be reached by the fuel during normal operation . The most important material property evolution leading to this activation is the degradation of the thermal conductivity with the accumulation of burnup in the fuel, leading to higher fuel internal temperatures while the cladding temperature remains low, even if the thickness of the fuel layer is only ~0.5mm. Recent results of measurements on U-Mo alloy dispersion fuel  do indicate a dramatic decrease in thermal conductivity with burnup in that system, but a more systematic approach would need to be applied in order to arrive at conclusive results. The reduction in thermal conductivity in these fuel systems is mainly linked to buildup of low conductivity interaction phases in the fuel, outer corrosion layers on the aluminium cladding and buildup of defects in the fuel particles (eg. (nano)bubbles, precipitates, etc.). It is currently not known which phenomena are dominating, but the competition between reduction in fuel power due to burnup accumulation (less uranium available) leading to lower temperatures and the degrading thermal conductivity leading to higher fuel temperatures, may allow the fuel temperature to reach values leading to activation of thermal phenomena late in life and at high burnup, even if the fuel power at that moment is low compared to the power reached when the fuel was still fresh. Such a hypothesis would be able to explain some of the recent observations in the post-irradiation examinations (PIE) of the last irradiation campaigns .
The online measurement of fuel temperature during irradiation is virtually impossible for research reactor fuel plates and thermal conductivity is a difficult-to-access parameter. As a result, only very limited data is available to predict the evolution of the fuel thermal conductivity with burnup for research reactor fuels. Typically, thermal conductivity measurements are performed using relatively complex laser flash methods in the nuclear fuel R&D. This is mainly linked to the high temperature needs of the power reactor fuel community, while the research reactor fuels operate at temperatures well below 300°C, for which other methods are likely more suitable. Outside of the nuclear field, other methods for thermal conductivity determination have been developed, which may be more simple for operation in a hot cell environment and thus allow more frequent and systematic measurement sets and more suitable for the lower temperature requirements of the RR fuel development field. At SCK-CEN, the Maldonado method  is under evaluation for application on irradiated research reactor fuel samples and thanks to the recently acquired availability of the focused ion beam (FIB) technique on radioactive materials, also a microscopic method for measurement of thermal conductivity  is being developed. If successful, the use of the Maldonado method offers a completely novel and innovative solution for cheap and flexible measurements on research reactor fuel samples in hot cell environment as a highly valuable alternative to the cumbersome and expensive laser flash method. The FIB based analyses allow a decoupling of the evolution of the different components in the dispersion fuel system, allowing a more detailed understanding of the phenomena at play and a more predictive, mechanistic modelling. SCK-CEN is particularly well placed to acquire a systematic dataset on thermal conductivity degradation for different fuel systems with various irradiation histories, thanks to the significant library of fuel samples available from previous irradation campaigns (FUTURE, E-FUTURE, COBRA, SEMPER FIDELIS, SELENIUM, etc.).
This PhD topic is targeting the development of a systematic dataset of thermal conductivity evolution in the U-Mo and U3Si2 based fuel systems, with the final goal to derive empirical relations for the evolution and assess the underlying mechanisms. The candidate will focus on the further optimisation and validation of the Maldonado method and the microscopic FIB method for thermal conductivity measurements on irradiated RR fuels. Using different available methods, the systematic measurement of thermal conductivity evolution with burnup in systematic series of samples available at SCK-CEN, will then result in a highly valuable dataset for modeling of research reactor fuel behaviour. Results can be directly compared to results obtained by laser flash or any other method in literature or performed during the PhD through collaborations. Using the determined thermal conductivities in fuel performance models and calculating irradiation temperatures is possible through collaborations with groups involved in these domains (Argonne National Lab and CEA) and if the candidate develops an interest in these theoretical models, he/she can be invited to help develop them. SCK-CEN has a large series of different research reactor fuels with different irradiation histories at its disposal from which a systematic set of samples can be developed, but transportation of this highly radioactive material is expensive and difficult. Thanks to its extensive collaborations, SCK-CEN has direct access to specialists from JRC Karlsruhe, the University of Munich, Argonne National Lab, Idaho National Lab and Pacific Northwest National Lab, who are active in thermal conductivity assessment in nuclear fuels using different approaches. The PhD should eventually bring an answer to the hypothesis of the importance of thermally activated phenomena in the different research reactor fuel systems, focusing on the U3Si2 and U-Mo alloy dispersion fuels.
 A. Leenaers, PhD thesis, UGENT/SCKCEN, 2014, ISBN-9789076971223, https://biblio.ugent.be/publication/5664867/file/5664868
 S. Van den Berghe, P. Lemoine, Nucl. Eng. Technol. 46 (2) (2014) 125-146.
 Y. Bei, D. Salvato, et al., unpublished results
 T.K. Huber et al., J. Nucl. Mater., 503 (2018), pp. 304-313
 T.K. Huber, PhD thesis, TU Munich, 2016, https://mediatum.ub.tum.de/doc/1286734/1286734.pdf
 A. Leenaers et al., Presentation on SEMPER FI PIE results, RERTR-2019 conference, September 2019, Zagreb, Croatia
 O. Maldonado, Cryogenics 1992 Vol 32, No 10
 Y. Miao et al., J. of Nucl. Mater. 527 (2019), in print