Optoelectronic developments for remote-handled maintenance tasks in ITER

Remotely handled maintenance tools operated in the future International Thermonuclear Experimental Reactor (ITER) will face a harsh radiation environment, with total dose level requirements of several MGy. Optical fiber data communication has been considered as an alternative to conventional electronic transmission between the control room and remote handled maintenance equipment, mainly owing to its insentivity to electro-magnetic interference and to its wavelength encoded multiplexing capabilities. In this paper we summarise main results obtained at SCK•CEN over the past years towards the development of radiation tolerant fibre-optic communication links and report on the radiation tolerance of various individual optical components such as optical fibres, laser diodes and photodetectors, as well as their associated electronic driver circuits

Improved models of melting temperature and thermal conductivity for mixed oxide fuels doped with low minor actinide contents

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 to improve the sustainability of nuclear energy by reducing ultimate radioactive waste and improving the exploitation of fuel resources. Thermal conductivity and melting temperature are fundamental properties of nuclear fuels, since they determine the fuel temperature profile and the melting safety margin, respectively and affect the overall fuel performance under irradiation. The available literature on thermal properties of Am or Np- containing MOX, both experimental data and models, is currently scarce. Moreover, state-of-the-art fuel performance codes (FPCs), e.g., GERMINAL and TRANSURANUS, do not account for the effects of minor actinides on MOX fuel properties. This deliverable presents the development and validation of original correlations for the thermal conductivity and melting temperature of minor actinide-bearing MOX (U,Pu,Am,Np)O2-x based on available literature data. These correlations are derived by extending those obtained in the project for U-Pu MOX fuels with the inclusion of the effect of Am and Np content, while preserving the physically- grounded formulation depending on the most relevant parameters. Ways to improve these correlations further in the future are also discussed

IAEA FUMAC BENCHMARK ON THE HALDEN, STUDISVIK AND QUENCH-L1 LOCA TESTS

The International Atomic Energy Agency (IAEA) sponsored the Coordinated Research Project (CRP) on Fuel Modeling under Accident Conditions (FUMAC) to coordinate and support research on nuclear fuel modelling under accident conditions in member countries. The focus of the FUMAC CRP (2015- 2018) has been on loss-of-coolant accidents (LOCA). Various institutions performed fuel performance simulations of selected experiments using different fuel performance codes (e.g., FRAPCONFRAPTRAN, TRANSURANUS, ALCYONE, DIONISIO, SOCRAT, FTPAC, BISON, RAPTA) and system codes (e.g SOCRATE, ATHLET). One of the results of the FUMAC CRP is a comprehensive code-to-code benchmark of selected results, and a comparison of simulations with experimental data as well. This paper represents an overview of the current state-of-the-art of nuclear fuel simulation capabilities for LOCAs and paves the way to further analyses and future developments. More precisely, we discuss the results of the simulation of a subset of the experiments considered in the FUMAC CRP, i.e., (i) the Halden LOCA tests (IFA-650.9/10/11, but only IFA-650.10 is in detail presented in this paper), (ii) the Studsvik LOCA test NRC-192, and (iii) rod 4 of the KIT QUENCH-L1 bundle test. These experiments, briefly presented in the paper, cover a wide range of conditions relevant for LOCA scenarios from different sources. The presented benchmark results are considered in more detail at the end of the LOCA transient (e.g., time of failure, cladding outer diameter, cladding oxidation thickness…). The experimental data are always included in the comparisons, when available. The results are also critically discussed, with the aim of identifying modelling developments required for the improvement of LOCA analyses. Finally, the outcome is complemented with an uncertainty and sensitivity analysis in a separate paper in this conference

Towards grain-scale modelling of the release of radioactive fission gas from oxide fuel. Part I: SCIANTIX

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-of-the-art semi-empirical approach (ANS 5.4-2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusion-decay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as inter- related phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS